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| {{#Wiki_filter:Krishna P. Singh Technology Campus, 1 Holtec Blvd., Camden, NJ 08104 HOLTEC Telephone (856) 797-0900 DECOMMISSIONING Fax (856) 797-0909 INTERNATIONAL HDI PNP 2023-030 10 CFR 50.90 December 14, 2023 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Palisades Nuclear Plant NRC Docket No. 50-255 Renewed Facility Operating License No. DPR-20 | | {{#Wiki_filter:}} |
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| ==Subject:==
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| License Amendment Request to Revise Renewed Facility Operating License and Permanently Defueled Technical Specifications to Support Resumption of Power Operations In accordance with Title 10 of the Code of Federal Regulations, Part 50, Section 90 (10 CFR 50.90), Application for amendment of license, construction permit, or early site permit, Holtec Decommissioning International, LLC (HDI) on behalf of Holtec Palisades LLC, hereby requests U. S. Nuclear Regulatory Commission (NRC) review and approval of a proposed license amendment request (LAR) to revise the Palisades Nuclear Plant (PNP) Renewed Facility Operating License (RFOL) DPR-20. The proposed LAR would revise the RFOL, Appendix A, Permanently Defueled Technical Specifications (PDTS), and Appendix B Environmental Protection Plan (EPP) to reflect the resumption of power operations at PNP.
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| In Reference 1, Entergy Nuclear Operations, Inc. notified the NRC that it had permanently ceased operations and permanently removed fuel from the reactor vessel at PNP. Upon docketing the 10 CFR 50.82, Termination of license, paragraph a, subparagraph 1, 10 CFR 50.82(a)(1) certifications 10 CFR 50.82(a)(2) no longer authorizes operation of the reactor, or emplacement or retention of fuel into the reactor vessel. However, shortly after PNP transitioned to a decommissioning facility, Holtec Palisades LLC assumed ownership of PNP (Reference 2) and given the support from the Governor of the State of Michigan, HDI commenced a project to return PNP to a power operations plant. The regulatory part of this project as described in an HDI {{letter dated|date=March 13, 2023|text=letter dated March 13, 2023}} (Reference 3), has identified the regulatory path to reinstate the power operations licensing basis (POLB) to resume power operations through a series of licensing submittals referred to as a regulatory framework. HDI intends to submit additional licensing actions over the next several months to reinstate the Plant POLB. In this submittal HDI proposes to restore the PNP power operations technical specifications (POTS) which is a part of this regulatory framework.
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| The PNP repower regulatory framework consists of this LAR, a LAR to revise PDTS administrative requirements, an emergency plan LAR, an exemption to 10 CFR 50.82(a)(2)
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| (Reference 4), and a license transfer order for PNP operating authority (Reference 5). As discussed in Reference 4, to coordinate implementing this requested amendment, after receipt
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| HDI PNP 2023-030 Page 2 of 3 of NRC approvals needed to return PNP to power operations, HDI is proposing to submit a notification of transition to power operations to the NRC that will docket HDIs satisfaction of the implementation conditions for license transfer, 10 CFR 50.82(a)(2) exemption, and license amendments. Upon docketing this notification letter, PNP will transition from a facility in decommissioning back to a power operations plant.
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| HDI is currently targeting the implementation of the POTS in the third quarter of 2025. To support this schedule, HDI respectfully requests that the NRC review the enclosed LAR on a schedule that that will permit approval of the proposed LAR by January 31, 2025, and that the proposed amendment become effective upon docketing the transition notification letter, with a 30-day implementation period.
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| The proposed changes to the PNP PDTS are in accordance with 10 CFR 50.36, Technical specifications, 10 CFR 50.36(c)(1) through 10 CFR 50.36(c)(5). The proposed changes to the PNP EPP are in accordance with 10 CFR 50.36b, Environmental conditions, paragraph b, 10 CFR 50.36b(b).
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| The enclosure to this letter provides a detailed description and evaluation of the proposed changes for PNP. Attachment 1 to the enclosure contains a mark-up of the current RFOL, PDTS, and EPP pages. Attachment 2 to the enclosure contains the retyped RFOL, TS, and EPP containing the proposed changes. Attachment 3 to the enclosure contains the proposed changes to the TS Bases. The proposed TS Bases changes are provided for information and will be incorporated in accordance with the TS Bases Control Program upon implementation of the approved amendment.
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| The proposed changes have been evaluated in accordance with 10 CFR 50.91(a), Notice for public comment, subparagraph (1), using the standards in 10 CFR 50.92, Issuance of amendment, paragraph (c), and it has been determined that the changes involve no significant hazards consideration. The basis for this determination is included in the enclosure.
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| In accordance with 10 CFR 50.91(b), State consultation, HDI is notifying the State of Michigan of this proposed LAR by transmitting a copy of this letter, with its enclosure, to the designated State of Michigan official.
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| If you have any questions regarding this submittal, please contact Jim Miksa, regulatory assurance engineer, at (269) 764-2945.
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| This letter contains no new regulatory commitments and no revisions to existing regulatory commitments.
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| I declare under penalty of perjury that the foregoing is true and correct. Executed on December 14, 2023.
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| Respectfully, Jean A. Fleming Digitally signed by Jean A. Fleming Date: 2023.12.14 09:26:30 -05'00' Jean A. Fleming Vice President, of Licensing, Regulatory Affairs & PSA Holtec International
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| HDI PNP 2023-030 Page 3 of 3
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| ==References:==
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| : 1. Entergy Nuclear Operations, Inc. letter to U.S. Nuclear Regulatory Commission, "Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel," dated June 13, 2022 (ADAMS Accession No. ML22164A067)
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| : 2. U.S. Nuclear Regulatory Commission letter to Holtec International, Palisades Nuclear Plant and Big Rock Point Plant - Issuance of Amendment Nos. 129 and 273 re: Order Approving Transfer of Licenses and Conforming Administrative License Amendments, dated June 28, 2022 (ADAMS Accession No. ML22173A173)
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| : 3. Holtec Decommissioning International, LLC letter to U.S. Nuclear Regulatory Commission, "Regulatory Path to Reauthorize Power Operations at the Palisades Nuclear Plant" dated March 13, 2023 (ADAMS Accession No. ML23072A404)
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| : 4. Holtec Decommissioning International, LLC letter to U.S. Nuclear Regulatory Commission, "Request for Exemption from Certain Termination of License Requirements if 10 CFR 50.82" dated September 28, 2023 (ADAMS Accession No. ML23271A140)
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| : 5. Holtec Decommissioning International, LLC letter to U.S. Nuclear Regulatory Commission, "Application for Order Consenting to Transfer of Control of License and Approving Conforming License Amendments" dated December 6, 2023 (ADAMS Accession Nos. ML23340A161, ML23340A162)
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| ==Enclosure:==
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| Evaluation of the Proposed Changes Enclosure Attachments:
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| : 1. Proposed Changes (mark-up) to Palisades Plant Renewed Facility Operating License DPR-20, Appendix A Permanently Defueled Technical Specifications, and Appendix B Environmental Protection Plan Pages
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| : 2. Page Change Instructions and Retyped Pages for the Palisades Plant Renewed Facility Operating License DPR-20, Appendix A Technical Specifications, and Appendix B Environmental Protection Plan
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| : 3. Proposed Technical Specifications Bases Changes (for information only) cc: NRC Region III Regional Administrator NRC Decommissioning Inspector - PNP NRC Project Manager PNP Designated Michigan State Official
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| HDI PNP 2023-030 Enclosure Page 1 of 97 Enclosure HDI PNP 2023-030 Evaluation of the Proposed Changes
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| HDI PNP 2023-030 Enclosure Page 2 of 97 TABLE OF CONTENTS 1.0
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| ==SUMMARY==
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| DESCRIPTION ............................................................................................... 3 2.0 DETAILED DESCRIPTION ................................................................................................ 3 2.1 Reason for Proposed Change ..................................................................................... 3 2.2 Description of Proposed Change ................................................................................ 4
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| ==3.0 TECHNICAL EVALUATION==
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| .............................................................................................. 4 3.1 Accident and Transient Analyses Applicable to the Proposed Change ................. 5 3.2 Evaluation of the Proposed Change ........................................................................... 8 3.2.1 Proposed Changes to the PNP Renewed Facility Operating License.................. 8 3.1.2 Proposed Changes to the Permanently Defueled Technical Specifications ..... 17 3.2.3 Proposed Changes to RFOL Appendix B, Environmental Protection Plan ....... 84 3.2.4 Proposed Changes to the PNP Technical Specification Bases.......................... 89
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| ==4.0 REGULATORY EVALUATION==
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| ........................................................................................ 89 4.1 Applicable Regulatory Requirements ....................................................................... 89 4.2 Precedent .................................................................................................................... 91 4.3 No Significant Hazards Consideration Determination ............................................ 91 4.4 Conclusion .................................................................................................................. 94 5.0 ENVIRONMENTAL EVALUATION ................................................................................. 94
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| ==6.0 REFERENCES==
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| ................................................................................................................. 96 7.0 ATTACHMENTS .............................................................................................................. 97
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| HDI PNP 2023-030 Enclosure Page 3 of 97 EVALUATION OF THE PROPOSED CHANGES 1.0
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| ==SUMMARY==
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| DESCRIPTION In accordance with Title 10 of the Code of Federal Regulations, Part 50, Section 90 (10 CFR 50.90), Application for amendment of license, construction permit, or early site permit, Holtec Decommissioning International, LLC (HDI) on behalf of Holtec Palisades, LLC hereby requests U. S. Nuclear Regulatory Commission (NRC) review and approval of a license amendment request (LAR) to revise the Palisades Nuclear Plant (PNP) Renewed Facility Operating License (RFOL) DPR-20. The proposed license amendment would revise the RFOL, Appendix A, Permanently Defueled Technical Specifications (PDTS), and Appendix B Environmental Protection Plan (EPP). The proposed changes reinstate PNP license requirements that were removed based on docketing the 10 CFR 50.82(a)(1) certifications of permanent cessation of power operations and permanent removal of fuel from the reactor vessel to support a return of the PNP plant to power operations. The requested changes involve no significant hazards consideration. HDI requests approval of the proposed LAR by January 31, 2025, and that the proposed amendment become effective upon docketing the transition notification letter, with a 30-day implementation period.
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| 2.0 DETAILED DESCRIPTION 2.1 Reason for Proposed Change In Reference 1, Entergy Nuclear Operations, Inc. notified the NRC that it decided to permanently cease operations at PNP no later than May 31, 2022. Certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel were submitted to the NRC in accordance with 10 CFR 50.82(a)(1)(i) and (ii), respectively, and were docketed (Reference 2). Upon docketing the 10 CFR 50.82(a)(1) certifications 10 CFR 50.82(a)(2) no longer authorizes operation of the reactor, or emplacement or retention of fuel into the reactor vessel. The regulatory framework for the reauthorization of power operations at PNP includes submitting a request for exemption from 10 CFR 50.82(a)(2) to remove the restriction that prohibits operation of the PNP reactor, or emplacement or retention of fuel into the PNP reactor vessel (Reference 3). This restriction, imposed by the voluntary docketing of the 10 CFR 50.82(a)(1) certifications, was used as the basis for licensing actions that allowed relaxation of power operation license requirements at PNP. Implementation of the NRC approved licensing actions included revising the PNP licensing basis to accurately reflect the status and reduced risk of a facility in decommissioning. No major decommissioning activities occurred at PNP to support this transition, and none have occurred since. There are no physical changes to facility design proposed or required to support this exemption. LAR along with the referenced exemption, an operating authority transfer order, and LARs to the RFOL administrative requirements and Emergency Plan are required to support reinstatement of the PNP power operations licensing basis that was in effect just prior to the 10 CFR 50.82(a)(1) certifications.
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| As described above, this LAR is necessary to reinstate the PNP RFOL, TS, and the EPP that were in effect just prior to the 10 CFR 50.82(a)(1) certifications to support returning PNP to a power operations licensing basis (POLB). To retain a clear connection between the RFOL decommissioning license amendments (References 4 and 5) and the license amendments to return PNP to power operations, HDI has elected to submit two separate LARs to revise the RFOL, PDTS and EPP. One is this LAR, which reinstates the TS needed for resumption of power operation. A second LAR reinstates TS for certain administrative requirements for plant
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| HDI PNP 2023-030 Enclosure Page 4 of 97 staff. Although both LARs must be approved and implemented prior to the resumption of power operation, they are not linked.
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| 2.2 Description of Proposed Change This LAR proposes to revise the PNP RFOL, the PDTS, and the EPP. The proposed changes are consistent with the previously approved PNP power operations technical specifications (POTS) that allowed emplacement of fuel into the reactor vessel and power operations at PNP.
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| The proposed changes would revise certain requirements contained within the PNP RFOL and PDTS to reinstate requirements that are necessary for power operation and revise or remove requirements that would no longer be applicable. The proposed EPP changes are editorial/administrative to more accurately reflect a power operations plant. The proposed changes to the PNP PDTS are in accordance with 10 CFR 50.36(c)(1) through 10 CFR 50.36(c)(5). The proposed changes to the PNP EPP are in accordance with 10 CFR 50.36b(b).
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| The Updated Final Safety Analysis Report (UFSAR), now titled the Defueled Safety Analysis Report (DSAR), will be updated, via the 10 CFR 50.59, Changes, tests and experiments, process to reflect the docketed version that was in effective prior to the 10 CFR 50.82(a)(1) certifications, PNP UFSAR Revision 35 (Reference 6). Any DSAR retained changes to UFSAR Revision 35 have been or will be evaluated via the 50.59 process against UFSAR Revision 35 to determine if NRC approval is required prior to exiting the period of decommissioning. This will include reinstatement of accident analyses and the safety reclassification of systems, structures, and components (SSCs), required to support the PNP power operations licensing basis (POLB). Changes made to the UFSAR after Revision 35 will be evaluated for retention, to the extent appropriate for an operating plant. The DSAR change back to the PNP POLB UFSAR will be accomplished under the 10 CFR 50.59 process and be implemented coincident with the associated license amendments.
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| HDI has submitted an application to the NRC for the transfer of the PNP operating authority from HDI to a new entity (Reference 11). Ownership of the PNP license will remain with Holtec Palisades LLC. This application is necessary because the order that transferred operating authority to HDI limits HDI to the performance of spent fuel management and decommissioning activities at PNP (Reference 7). The license transfer request changes are independent of the changes proposed in this enclosure and, as such, the changes proposed by the license transfer request are not included in the RFOL and PDTS markups of this LAR.
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| ==3.0 TECHNICAL EVALUATION==
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| This LAR proposes modifications to the PNP RFOL, the PDTS, and the EPP to support reinstatement of power operation at PNP.
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| The regulatory requirements related to the content of TS are promulgated in 10 CFR 50.36, Technical Specifications. As detailed in a subsequent section of this LAR, this regulation lists the criteria to define the scope of items that must be included in TS. The scope of systems structures and components (SSCs) and parameters that must have Limiting Conditions for Operation (LCO) included in the PNP TS to support the PNP POLB are those needed to address the reinstated UFSAR Revision 35 postulated design basis accidents (DBAs), so that the consequences of the DBAs are maintained within acceptable limits. PNP TS LCOs that were removed in Amendment 272 (Reference 8) will be reinstated to support the PNP POLB, UFSAR Revision 35 (Reference 6).
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| HDI PNP 2023-030 Enclosure Page 5 of 97 3.1 Accident and Transient Analyses Applicable to the Proposed Change As stated in Reference 8, Chapter 14 of the PNP UFSAR (Reference 6) and license Amendment No. 226 (Reference 9) describe the postulated DBA and transient scenarios applicable to PNP during power operations and when fuel is in the reactor vessel. These scenarios analyses demonstrate that the PNP plant design supports safe power operations and retention of fuel in the reactor vessel, and that radiological consequences from the postulated accident scenarios do not exceed the regulatory requirements of 10 CFR 50.67 or 10 CFR Part 100, as applicable. Two basic groups of events are pertinent to safety: abnormal operational transients and postulated DBAs. The analyses of the abnormal operational transients evaluate the ability of the plant protection features to ensure that during these transients no fuel damage occurs, and the primary coolant system (PCS) pressure limit is not exceeded. The safety design limits require that damage to the fuel be limited and that no nuclear system process barrier damage results from any abnormal operational occurrence.
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| Thus, analysis of this group of events evaluates the features that protect the first two radioactive material barriers. The radioactive material barriers are the fuel cladding, primary coolant system and the Containment. Analyses of the events in the second group, postulated DBAs, evaluate situations that require functioning of the engineered safeguards systems in order to protect the fission product barriers, including containment, in order to mitigate the offsite radiological consequences.
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| PNP UFSAR Revision 35, Chapter 14 (Reference 6) contains the DBA and transient scenarios applicable to PNP during plant operations. It also describes the analyses that were performed to demonstrate that the plant could be operated safely and that radiological consequences from postulated DBAs do not exceed the applicable limits. The most severe postulated accidents for nuclear power plants involve damage to the nuclear reactor core and the release of large quantities of fission products to the PCS. Many of the accident scenarios postulated in the UFSAR involve failures or malfunctions of systems which could affect the reactor core. Once the reactor is permitted to be refueled and resume power operation, these UFSAR events are applicable.
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| The analyses in UFSAR Chapter 14 provide results that are compared to regulatory acceptance criteria and are integral to the plant's design and licensing basis. These analyses demonstrate the integrity of the fission product barriers, the capability to shut down the reactor and maintain it in a safe shutdown condition, and the capability to prevent or mitigate the consequences of DBAs and transients. Certain systems, structures, and components (SSCs) are credited in these analyses for the purpose of mitigating the DBAs or transients. These SSCs are considered for inclusion in TS per the requirements of 10 CFR 50.36(c).
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| As noted above, the UFSAR will be reinstated to Revision 35 (Reference 6) which is the last docketed version of the UFSAR that was in effect prior to the 10 CFR 50.82(a)(1) certifications. This includes reinstatement of the DBA and transient scenarios in Chapter 14. These scenarios and analyses are reinstated as they existed in Revision 35 of the UFSAR. No changes to these analyses are made as part of the return to power for PNP. As a normal part of evaluating core designs and operational parameters, reviews of Chapter 14 sections are performed for each fuel reload cycle by the nuclear fuel vendor to determine if the proposed core design is still bounding or if it needs to be reanalyzed (disposition of events). The UFSAR is updated accordingly, as needed.
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| Therefore, future changes to Chapter 14 of the UFSAR may occur separately from the licensing actions necessary to resume power operation and will be evaluated under the 10 CFR 50.59 process against the PNP POLB.
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| HDI PNP 2023-030 Enclosure Page 6 of 97 In the PDTS LAR (Reference 4), PNP discussed four accident analyses that remained valid for a permanently shut down and defueled condition. They are the postulated cask drop accident, fuel handling incident in the fuel handling building, the liquid waste incident, and the waste gas incident. These four accident analyses will be retained and revised as necessary to the UFSAR Revision 35 versions that were in place prior to the PDTS amendment 272 (Reference 8).
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| In the PDTS LAR (Reference 4) the postulated cask drop accident was not reanalyzed for the permanently defueled condition. In lieu of revising the cask drop accident analysis, restrictions were placed on the movement of a fuel cask (License Condition 2.C.(5)) to support removal of TS 3.7.10, Control Room Ventilation (CRV) Filtration, TS 3.7.11, Control Room Ventilation (CRV) Cooling, and TS 3.7.12, Fuel Handling Area Ventilation System, thereby ensuring continued compliance with the radiological consequences described for a postulated cask drop during shutdown and defueled conditions. Therefore, since the postulated cask drop analysis described in UFSAR Revision 35 was not revised to support the PDTS LAR, it is proposed to remain unchanged as the analysis of record with the proposed deletion of License Condition 2.C.(5) supported by the reinstatement of TS 3.7.10, TS 3.7.11, and TS 3.7.12.
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| The fuel handling accident (FHA) was updated to add a sensitivity analysis to determine the fuel decay time at the time of a FHA (determined to be 17 days) that is required to no longer credit isolation of the Control Room (CR) ventilation system and radionuclide removal through the CR and SFP ventilation systems. This sensitivity analysis addressed only a fuel assembly drop in the spent fuel pool and did not revise the previously approved fuel handling accident in containment as described in UFSAR Revision 35. This previous version of the FHA as discussed in UFSAR Revision 35 will be reinstated as the analysis of record and includes a fuel assembly drop in containment as well as in the spent fuel pool. Additionally, with the reinstatement of TS 3.7.10, TS 3.7.11, and TS 3.7.12 the sensitivity analysis for fuel decay time is no longer necessary.
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| The accidental release of liquid waste was not reanalyzed for the permanently defueled condition. Accidental discharge of radioactive liquid from a volume control tank (VCT) rupture, primary makeup storage tank (T-90) failure, or utility water storage tank (T-91) failure to the circulating water canal will continue to be controlled by administrative process, system design, and system monitoring during normal operation. The same controls will remain in place in the operating condition as were previously credited both during operation and the permanently defueled condition. Therefore, no change is proposed to the accidental release of liquid waste incident, and it will reinstated as described in UFSAR Revision 35.
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| The analysis for accidental release of waste gas consists of two parts, waste gas decay tank failure and rupture of the volume control tank. For the permanently defueled condition the waste gas decay tank failure analysis was revised to support deletion of TS 3.7.10 and TS 3.7.11 by demonstrating that the dose consequences are bounded by the FHA. This LAR proposes to reinstate TS 3.7.10 and TS 3.7.11 to support the approved waste gas decay tank rupture analysis described in UFSAR Revision 35. The volume control tank rupture analysis that was no longer applicable in the permanently defueled condition will again be applicable in the power operations condition and will be reinstated without revision to the version described in UFSAR Revision 35. The reinstated analysis, without credit for release path filtration, demonstrates EAB, LPZ, and CR dose criteria are not exceeded. Therefore, reinstatement of the UFSAR Revision 35 waste gas incident supports power operations at PNP.
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| HDI PNP 2023-030 Enclosure Page 7 of 97 A list of the PNP UFSAR Chapter 14 safety analysis DBAs and transients is provided in Table 3-1. Analyses are reviewed on a cycle specific basis as appropriate and are addressed by the fuel vendor as described in UFSAR Section 14.1. Analyses of radiological impacts for certain events were reviewed and approved by the NRC in Amendment 226 (Reference 9). This amendment was the full implementation of an alternate source term in accordance with the guidance in Regulatory Guide 1.183 (Reference 10). The alternate source term amendment approved by the NRC will remain the analysis of record and is unchanged since approved by amendment 226.
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| No changes were made to these analyses during the transition to decommissioning.
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| Table 3 PNP DBAs and Transients UFSAR DBA or Transient Section 14.2 Uncontrolled Control Rod Withdrawal 14.3 Boron Dilution 14.4 Control Rod Drop 14.5 Core Barrel Failure 14.6 Control Rod Mis-operation 14.7 Decreased Reactor Coolant Flow 14.8 Start-up of an Inactive (Primary Coolant Pump) Loop 14.9 Excessive Feedwater Incident 14.10 Increase in Steam Flow (Excess Load) 14.11 Postulated Cask Drop Accidents 14.12 Loss of External Load 14.13 Loss of Normal Feedwater 14.14 Steam Line Rupture Incident 14.15 Steam Generator Tube Rupture with a Loss of Offsite Power 14.16 Control Rod Ejection 14.17 Loss of Coolant Accident 14.18 Containment Pressure and Temperature Analysis 14.19 Fuel Handling Incident 14.20 Liquid Waste Incident 14.21 Waste Gas Incident 14.22 Maximum Hypothetical Accident Radiological Consequences of Failure of Small Lines Carrying Primary Coolant 14.23 Outside Containment 14.24 Control Room Radiological Habitability These DBAs and incidents are once again applicable to PNP in a power operating condition. They are the same ones previously described in UFSAR, Revision 35.
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| When reinstated, these analyses will reflect the licensing basis as it existed just prior to the 10 CFR 50.82(a)(1) certifications to support returning PNP to a POLB. These accident analyses form the basis for the TS that existed prior to the defueled period, and they now form the basis for the TS proposed in this LAR.
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| HDI PNP 2023-030 Enclosure Page 8 of 97 3.2 Evaluation of the Proposed Change HDI proposes to modify the PNP RFOL, PDTS, and EPP as shown below. Each section that is proposed to be changed is identified, the proposed changes are shown, and the basis for each change is given. Changes to the RFOL are listed first followed by changes to the PDTS and EPP. If appropriate, proposed deletions are shown with strikethrough and additions are shown in bold italics. TS sections that were deleted in their entirety with issuance of Amendment 272, PDTS, (Reference 8) are proposed for reinstatement without change to create TS sections consistent with power operation. The addition of these complete TS sections is described below. to this enclosure contains a mark-up of the current RFOL, PDTS, and EPP pages.
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| The proposed changes to the PDTS are considered a major rewrite with the addition of numerous TS. TS sections that are added in their entirety are listed by number and title, however the inserted TS pages are not included with the marked-up pages in Attachment 1. In addition, the following editorial changes are not shown in the marked-up RFOL and PDTS in because they do not affect the technical content of the RFOL or the PDTS:
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| Reformatting (margins, font, tabs, line spacing, etc.) content to create a continuous electronic file, Renumbering of pages, where appropriate, and Removal of historical amendment numbers. of this enclosure provides the re-typed and added pages to reflect the proposed changes in their entirety.
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| The markups of the TS Bases, including TS Bases sections that will be added in their entirety are provided in Attachment 3. They are provided for information only. Upon approval of this amendment, changes to the TS Bases will be incorporated in accordance with PNP TS 5.5.12, Technical Specifications (TS) Bases Control Program.
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| 3.2.1 Proposed Changes to the PNP Renewed Facility Operating License Changes proposed to the RFOL are necessary to reinstate the license conditions that were removed by license amendment 272 (Reference 8) due to docketing the 10 CFR 50.82 decommissioning certifications as conditioned by the exemption to 10 CFR 50.82(a)(2)
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| (Reference 3). A License Condition removed in Amendment 272 because it was identified as superseded (i.e., original License Condition 1.B) will not be reinstated. License Conditions removed in Amendment 272 because they were identified as historical Conditions will not be reinstated (i.e., original License Conditions 2.C(4), 2.C(7), 2.H, and 2.I). See Reference 4 for the discussion and approval of the superseded and historical designations. No changes are proposed to the previously approved reinstated license conditions.
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| As previously discussed, HDI has submitted an application to the NRC for the transfer of the PNP operating authority from HDI to a new entity (Reference 11). Ownership of the PNP license will remain with Holtec Palisades LLC. Note, references to "HDI" are replaced by bracketed Palisades Energy, LLC, or Palisades Energy (e.g. [Palisades Energy]) to reflect the change in operating authority per license transfer application conforming amendments (Reference 11). They will be changed coincident with the implementation of the operating authority transfer through issuance of conforming amendments. They are not addressed in this submittal.
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| HDI PNP 2023-030 Enclosure Page 9 of 97 License Condition 2.B.(1)
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| Current License Condition 2.B.(1) Proposed License Condition 2.B.(1)
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| Pursuant to Section 104b of the Act, as Pursuant to Section 104b of the Act, as amended, and 10 CFR Part 50, "Licensing of amended, and 10 CFR Part 50, "Licensing of Production and Utilization Facilities", (a) Production and Utilization Facilities", (a)
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| Holtec Palisades to possess and use, and (b) Holtec Palisades to possess and use, and (b)
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| HDI to possess and use the facility at the HDI [Palisades Energy] to possess, and use designated location in Van Buren County, and operate, the facility as a utilization Michigan, in accordance with the procedures facility at the designated location in Van and limitation set forth in this license; Buren County, Michigan, in accordance with the procedures and limitation set forth in this license; Basis This License Condition is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, this license condition is reinstated to acknowledge that the plant will resume power operation as a utilization facility defined by 10 CFR Part 50.
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| License Condition 2.B.(2)
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| Current License Condition 2.B.{2} Proposed License Condition 2.B.{2}
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| HDI, pursuant to the Act and 10 CFR Parts HDI [Palisades Energy], pursuant to the Act 40 and 70, to possess source and special and 10 CFR Parts 40 and 70, to receive, nuclear material that was used as reactor possess, and use source and special fuel, in accordance with the limitations for nuclear material that was used as reactor storage as described in the Updated Final fuel, in accordance with the limitations for Safety Analysis Report, as supplemented storage and amounts required for reactor and amended; operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; Basis This License Condition is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, this license condition is reinstated to permit authorization for receipt and use of special nuclear material (SNM) as reactor fuel, add the reference to use of the SNM for reactor operations, and allow possession of SNM as reactor fuel in the amount required for reactor operation. With the resumption of power operation, PNP has a need to receive SNM in the form of reactor fuel and use SNM as reactor fuel for reactor operations.
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| HDI PNP 2023-030 Enclosure Page 10 of 97 License Condition 2.B.(3)
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| Current License Condition 2.B.(3) Proposed License Condition 2.B.(3)
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| HDI pursuant to the Act and 10 CFR Parts HDI [Palisades Energy] pursuant to the Act 30, 40 and 70, to receive, possess, and and 10 CFR Parts 30, 40 and 70, to receive, use byproduct, source, and special nuclear possess, and use byproduct, source, and material as sealed sources that were used special nuclear material as sealed sources for reactor startup, sealed sources that that were used for reactor startup, sealed were used for reactor instrumentation and sources that were used for reactor are used in the calibration of radiation instrumentation, and are used in the monitoring equipment, and that were used calibration of radiation monitoring equipment as fission detectors in amounts as calibration, and that were used as fission required; detectors in amounts as required; Basis This License Condition is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, this license condition is reinstated to authorize the receipt and use of byproduct, source, and SNM as sealed neutron sources for reactor startup, reactor instrumentation, and fission detectors. This reinstatement is consistent with the fact that PNP will become a power operation plant and use of this byproduct material is necessary for plant functions as described.
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| License Condition 2.B.(5)
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| Current License Condition 2.B.(5) Proposed License Condition 2.B.(5)
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| HDI pursuant to the Act and 10 CFR Parts 30, HDI [Palisades Energy] pursuant to the Act 40 and 70, to possess, but not separate, such and 10 CFR Parts 30, 40 and 70, to possess, byproduct and special nuclear materials that but not separate, such byproduct and special were produced by the operations of the facility. nuclear materials as may be that were produced by the operations of the facility.
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| Basis This License Condition is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, this license condition is reinstated to reflect that PNP will become a power operation plant and produce byproduct and special nuclear materials in the course of plant operation.
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| HDI PNP 2023-030 Enclosure Page 11 of 97 License Condition 2.C.(1)
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| Current License Condition 2.C.(1) Proposed License Condition 2.C.(1)
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| [deleted] [deleted] [Palisades Energy] is authorized to operate the facility at steady state reactor core power levels not in excess of 2565.4 Megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.
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| Basis This License Condition is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, this License Condition is reinstated in its entirety to reflect the operating condition and technical specifications for an operating plant. The maximum power level is defined in this License Condition and is consistent with that used to support the accident analyses evaluated in the UFSAR.
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| License Condition 2.C.(2)
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| Current License Condition 2.C.(2) Proposed License Condition 2.C.(2)
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| The Technical Specifications contained in The Technical Specifications contained in Appendix A, as revised through Amendment Appendix A, as revised through Amendment No. 273, and the Environmental Protection No. 273XXX, and the Environmental Plan contained in Appendix B are hereby Protection Plan contained in Appendix B are incorporated in the license. HDI shall maintain hereby incorporated in the license. HDI the facility in accordance with the Technical [Palisades Energy] shall maintain operate Specifications and the Environmental the facility in accordance with the Technical Protection Plan. Specifications and the Environmental Protection Plan.
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| Basis This License Condition is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, this License Condition is reinstated to reflect that PNP will become a power operation plant. The current amendment number, 273, is deleted and replaced with XXX which acts as a placeholder for the new amendment number associated with the approval of this LAR.
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| HDI PNP 2023-030 Enclosure Page 12 of 97 License Condition 2.C.(3)
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| Current License Condition 2.C.(3) Proposed License Condition 2.C.(3)
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| [deleted] [deleted]
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| Fire Protection
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| [Palisades Energy] shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment requests
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| .[for entirety of text proposed for addition, see Attachment 1] 2. The licensee shall implement the modifications to its facility, as described in Table S-2, Plant Modifications Committed, of Entergy Nuclear Operations, Inc. (ENO) letter PNP 2019-028 dated May 28, 2019, to complete the transition to full compliance with 10 CFR 50.48(c) before the end of the refueling outage following the fourth full operating cycle after NRC approval. .[for entirety of text proposed for addition, see Attachment 1] will be completed once the related modifications are installed and validated in the PRA model.
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| Basis This License Condition is proposed for reinstatement in its entirety, with editorial revision, to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, this License Condition is reinstated to provide requirements for implementation of a fire protection program that complies with 10 CFR 50.48(a) and 10 CFR 50.48(c), including requirements for risk informed changes that may be made without prior NRC approval, other changes that may be made without prior NRC approval, and transition License Conditions. See Attachment 1 for the complete License Condition. The revision to the earlier version of the License Condition is to spell out the first occurrence of ENO as Entergy Nuclear Operations, Inc. (ENO) when referring to letters previously issued.
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| 10 CFR 50.48(a) and 10 CFR 50.48(c) apply to holders of operating licenses issued under Part 50. The conditions specified in 2.C.(3) include consideration of risk metrics for core damage frequency and large early release frequency, which are associated with power operation. This License Condition, which is based on maintaining an operational fire protection program in accordance with 10 CFR 50.48, with the ability to achieve and maintain safe shut down of the reactor in the event of a fire, will become applicable at PNP with the resumption of power operation. Therefore, License Condition 2.C.(3) is reinstated to reflect power operation of the facility.
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| HDI PNP 2023-030 Enclosure Page 13 of 97 License Condition 2.C.(5)
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| Current License Condition 2.C.(5) Proposed License Condition 2.C.(5)
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| Movement of a fuel cask in or over the spent Movement of a fuel cask in or over the spent fuel pool is prohibited when irradiated fuel fuel pool is prohibited when irradiated fuel assemblies decayed less than 90 days are in assemblies decayed less than 90 days are the spent fuel pool. in the spent fuel pool.
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| [deleted]
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| Basis This License Condition was added to prohibit movement of a fuel cask in or over the spent fuel pool when irradiated fuel assemblies with less than 90 days decay time are in the spent fuel pool. Previously, TS 3.7.10, Control Room Ventilation (CRV) Filtration, and TS 3.7.11, Control Room Ventilation (CRV) Cooling, were required during movement of a fuel cask in or over the SFP; TS 3.7.12, Fuel Handling Area Ventilation System, was required during movement of a fuel cask in or over the SFP when fuel assemblies with less than 90 days decay time are in the fuel handling building. These TS were deleted in Amendment 272; therefore, this License Condition was needed to control the timing of cask movement in or over the spent fuel pool.
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| TS 3.7.10, TS 3.7.11, and TS 3.7.12 are being reinstated in the TS as described below. They will contain the limits on cask movement in or over the spent fuel pool. This License Condition is no longer needed to control cask movement. Reinstatement of these TS will ensure an appropriate spent fuel assembly decay time requirement is maintained and the associated analyses presented in UFSAR, Revision 35, Table 14.1-6 remains bounding.
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| License Condition 2.C.(8)
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| Current License Condition 2.C.(8) Proposed License Condition 2.C.(8)
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| [deleted] [deleted]
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| Amendment 257 authorizes the implementation of 10 CFR 50.61a in lieu of 10 CFR 50.61.
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| Basis This License Condition is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, this License Condition is reinstated in its entirety to reflect the regulatory requirements for an operating plant. 10 CFR 50.61a, Fracture toughness requirements for protection against pressurized thermal shock events, applies to pressurized water reactors for which an operating license has been issued.
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| HDI PNP 2023-030 Enclosure Page 14 of 97 License Condition 2.D Current License Condition 2.D Proposed License Condition 2.D
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| [deleted] [deleted]
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| The facility has been granted certain exemptions from Appendix J to 10 CFR Part 50, Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors. This section contains leakage test requirements, scheduled and acceptance criteria for tests of the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. These exemptions were granted in a {{letter dated|date=December 6, 1989|text=letter dated December 6, 1989}}.
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| These exemptions granted pursuant to 10 CFR 50.12, are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. With these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.
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| Basis This License Condition is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, this License Condition is reinstated in its entirety to reflect the operating condition of the facility. During power operation, the containment plays a role in mitigating the consequences of the DBAs discussed in the UFSAR, Revision 35.
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| 10 CFR Part 50, Appendix J provides leakage test requirements for the containment. The exemptions previously approved by the NRC for use at PNP related to 10 CFR 50, Appendix J are provided in this License Condition. Providing this License Condition is necessary to resume power operations.
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| HDI PNP 2023-030 Enclosure Page 15 of 97 License Condition 2.J Current License Condition 2.J Proposed License Condition 2.J
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| [deleted] [deleted]
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| All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of American Society for Testing and Materials (ASTM) E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal scheduled, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC, as required by 10 CFR Part 50, Appendix H.
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| Basis This License Condition is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, this License Condition is reinstated to address the requirements of 10 CFR 50, Appendix H.
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| 10 CFR 50 Appendix H requires that the design of the reactor vessel surveillance capsule program and withdrawal schedule must meet the requirements in the version of ASTM Standard Practice E 185 that is current on the issue date of the American Society of Mechanical Engineers (ASME) Code to which the reactor pressure vessel (RPV) was purchased. The rule also requires the licensee to perform capsule testing and to report the test results in accordance with the requirements in ASTM Standard Practice E 185-82 to the extent practicable for the configuration of the test specimens in the RPV surveillance capsules.
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| The requirements in Appendix H are applicable to nuclear plants that are in power operations in the reactor critical operating mode because: (a) this is the plant operating mode that produces high energy neutrons as a result of the reactor's nuclear fission process; and (b) the requirements are set in place to provide assurance that the RPV will maintain adequate levels of fracture toughness throughout the operating life of the reactor.
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| Continued implementation of the applicable surveillance capsule testing and reporting requirements are necessary for PNP because the plant will resume operation, thereby exposing the reactor vessel to high energy neutrons and subjected to high thermal stress environments, as induced by operating the reactor coolant system at an elevated temperature. Any corresponding commitments in the UFSAR, Revision 35, will also be reinstated under the provisions of 10 CFR 50.59.
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| HDI PNP 2023-030 Enclosure Page 16 of 97 License Condition 2.K Current License Condition 2.K Proposed License Condition 2.K This license is effective as of the date of This license is effective as of the date of issuance issuance and until the Commission and until the Commission notifies the licensee in notifies the licensee in writing that the writing that the license is terminated shall expire license is terminated. at midnight March 24, 2031.
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| Basis This License Condition is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, this License Condition currently conforms with 10 CFR 50.51, Continuation of license, in that the license authorizes ownership and possession by HDI until the Commission notifies the licensee in writing that the license is terminated. While appropriate for a licensee that has permanently ceased operation, 10 CFR 50.51, Continuation of license requires that an operating reactor have a license issued for a fixed period of time. Therefore, this change is necessary to resume power operations.
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| ATTACHMENTS AND DATE OF ISSUANCE Current Attachments and Date of Issuance Proposed Attachments and Date of Issuance Appendix A - Permanently Defueled Appendix A - Permanently Defueled Technical Specifications Technical Specifications Basis The title of Appendix A is updated to reflect that the Permanently Defueled Technical Specifications will be retitled as the Technical Specifications. This is an administrative change.
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| HDI PNP 2023-030 Enclosure Page 17 of 97 3.1.2 Proposed Changes to the Permanently Defueled Technical Specifications APPENDIX A TITLE PAGE Current Title Proposed Title PALISADES PLANT PALISADES PLANT RENEWED FACILITY OPERATING LICENSE RENEWED FACILITY OPERATING LICENSE DPR-20 DPR-20 APPENDIX A APPENDIX A PERMANENTLY DEFUELED TECHNICAL PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS SPECIFICATIONS Basis The title is modified to remove Permanently Defueled before Technical Specifications to reflect that PNP will be operating and not permanently defueled. This is an administrative change.
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| APPENDIX A Table of Contents Current Table of Contents deleted 1.0 USE AND APPLICATION 5.7 High Radiation Area Basis The Table of Contents is proposed to be removed from the PDTS and placed under licensee control. Placing the TOC under licensee control eliminates the regulatory burden of submitting TOC pages for NRC review and allows timely administrative corrections and improvements to the TOC without NRC review and approval. The TOC does not meet the criteria specified in 10 CFR 50.36 requiring its inclusion within a plant's TS. The TOC references the page number of each Specification in the TS and does not contain any technical information required by 10 CFR 50.36. Since the TOC does not include information required to be in the TS by 10 CFR 50.36, inclusion of a TOC within the TS is optional. Removal of the TOC from the TS is an administrative change and is acceptable. The TS TOC will be maintained, revised, and distributed in accordance with administrative procedures. Holders of copies of the TS, including the NRC, will continue to receive periodic updates of the TOC pages.
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| HDI PNP 2023-030 Enclosure Page 18 of 97 TS SECTION 1.1, DEFINITIONS TS 1.1, Definitions, provides defined terms that are applicable throughout the TS and TS Bases.
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| Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.
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| AVERAGE shall be the average (weighted in proportion to the DISINTEGRATION ENERGY concentration of each radionuclide in the primary coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total noniodine activity in the coolant.
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| AXIAL OFFSET (AO) AO shall be the power generated in the lower half of the core less the power generated in the upper half of the core, divided by the sum of the power generated in the lower and upper halves of the core (determined using the incore monitoring system).
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| AXIAL SHAPE INDEX (ASI) ASI shall be the power generated in the lower half of the core less the power generated in the upper half of the core, divided by the sum of the power generated in the lower and upper halves of the core (determined using the excore monitoring system).
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| CERTIFIED FUEL A CERTIFIED FUEL HANDLER is an individual who HANDLER complies with provisions of the CERTIFIED FUEL HANDLER training and retraining program required by Specification 5.3.2.
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| CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST.
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| Calibration of instrument channels with Resistance Temperature Detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel.
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| HDI PNP 2023-030 Enclosure Page 19 of 97 Whenever a RTD or thermocouple sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.
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| The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps.
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| CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
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| CHANNEL FUNCTIONAL A CHANNEL FUNCTIONAL TEST shall be:
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| TEST
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| : a. Analog and bistable channels - the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, of all devices in the channel required for channel OPERABILITY;
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| : b. Digital channels - the use of diagnostic programs to test (continued) digital hardware and the injection of simulated process data into the channel to verify OPERABILITY, of all devices in the channel required for channel OPERABILITY.
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| The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested.
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| CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or control rods within the reactor vessel with the vessel head removed and fuel in the vessel.
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| Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
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| CORE OPERATING LIMITS The COLR is the plant specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation
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| HDI PNP 2023-030 Enclosure Page 20 of 97 within these limits is addressed in individual Specifications.
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| DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The dose conversion factors used for this calculation shall be those listed in Federal Guidance Report 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion, 1989; (Table 2.1, Exposure-to-Dose Conversion Factors for Inhalation).
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| INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).
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| LEAKAGE LEAKAGE shall be:
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| : a. Identified LEAKAGE
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| : 1. LEAKAGE, such as that from pump seals or valve packing (except Primary Coolant Pump seal water leakoff), that is captured and conducted to collection systems or a sump or collecting tank;
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| : 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; and
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| : 3. Primary Coolant System (PCS) LEAKAGE through a Steam Generator to the Secondary System (primary to secondary LEAKAGE).
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| : b. Unidentified LEAKAGE All LEAKAGE (except Primary Coolant Pump seal leakoff) that is not identified LEAKAGE;
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| : c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an PCS component body, pipe wall, or vessel wall.
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| MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average primary coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
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| HDI PNP 2023-030 Enclosure Page 21 of 97 NON-CERTIFIED A NON-CERTIFIED OPERATOR is a non-licensed OPERATOR operator who complies with the qualification requirements of Specification 5.3.1.
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| OPERABLE - A system, subsystem, train, component, or device OPERABILITY shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
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| PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:
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| : a. Described in Chapter 13, Initial Tests and Operation, of the FSAR;
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| : b. Authorized under the provisions of 10 CFR 50.59; or
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| : c. Otherwise approved by the Nuclear Regulatory Commission.
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| QUADRANT POWER TILT Tq shall be the maximum positive ratio of the power (Tq) generated in any quadrant minus the average quadrant power, to the average quadrant power.
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| RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to (RTP) the primary coolant of 2565.4 MWt.
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| REFUELING BORON REFUELING BORON CONCENTRATION shall be a CONCENTRATION Primary Coolant System boron concentration of >
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| 1720 ppm and sufficient to assure the reactor is subcritical by > 5% with all control rods withdrawn.
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| SHUTDOWN MARGIN SDM shall be the instantaneous amount of reactivity (SDM) by which the reactor is subcritical or would be subcritical from its present condition assuming: a. All full length control rods (shutdown and regulating) are fully inserted except for the single rod of highest reactivity worth, which is assumed to be fully withdrawn. However, with all full length control rods verified fully inserted by two independent means, it is not necessary to account for a stuck rod in the SOM calculation. With any full length control rods not capable of being fully inserted, the reactivity worth of these rods must be accounted for in the determination
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| HDI PNP 2023-030 Enclosure Page 22 of 97 of SDM; and b. There is no change in part length rod position.
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| STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
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| THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the primary coolant.
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| TOTAL RADIAL PEAKING FRT shall be the maximum ratio of the individual fuel FACTOR (FRT) pin power to the core average pin power integrated over the total core height, including tilt.
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| Table 1.1-1 MODES
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| % RATED AVERAGE PRIMARY REACTIVITY THERMAL MODE TITLE COOLANT CONDITION (keff) POWER(a) TEMPERATURE (F) 1 Power Operation 0.99 >5 NA 2 Startup 0.99 5 NA 3 Hot Standby < 0.99 NA 300 4 Hot Shutdown(b) < 0.99 NA 300 > Tave > 200 5 Cold Shutdown(b) < 0.99 NA 200 6 Refueling(c) NA NA NA (a) Excluding decay heat.
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| (b) All reactor vessel head closure bolts fully tensioned.
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| (c) One or more reactor vessel head closure bolts less than fully tensioned.
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| HDI PNP 2023-030 Enclosure Page 23 of 97 Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, these Definitions are proposed to be reinstated because they are necessary to support the TS added by this LAR. When combined with the Definition for ACTIONS in the current PDTS, they comprise all the Definitions that were in the TS prior to Amendment 272 (Reference 8). Additionally, these Definitions have not changed from the Definitions used prior to PNP Amendment 272. Each Definition is used elsewhere in the proposed TS. The Definitions will be added to Section 1.0 in alphabetical order with the existing Definitions. The existing Definitions remain in the TS and are not modified by this LAR.
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| TS SECTION 1.2, LOGICAL CONNECTORS TS 1.2, Logical Connectors, explains how the arrangement of these connectors constitutes logical conventions with specific meanings. The proposed modifications reflect the logical connectors necessary to support the TS that are reinstated with this LAR. Modifications to Section 1.2 in the PDTS are shown below. The addition of previously deleted text is described here and shown in Attachment 1 Current PURPOSE Proposed PURPOSE Logical connectors are used in Technical Logical connectors are used in Technical Specifications (TS) to discriminate between, Specifications (TS) to discriminate between, and yet connect, discrete Conditions, and yet connect, discrete Conditions, Required Actions, Completion Times, Required Actions, Completion Times, Surveillances, and Frequencies. The only Surveillances, and Frequencies. The only logical connector that appears in TS is AND. logical connectors that appears in TS is are The physical arrangement of this connector AND and OR. The physical arrangement of constitutes logical conventions with specific this these connectors constitutes logical meanings. conventions with specific meanings.
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| HDI PNP 2023-030 Enclosure Page 24 of 97 Current BACKGROUND Proposed BACKGROUND Levels of logic may be used to state Required Several levels Levels of logic may be used to Actions. These levels are identified by the state Required Actions. These levels are placement (or nesting) of the logical connectors identified by the placement (or nesting) of the and by the number assigned to each Required logical connectors and by the number Action. The first level of logic is identified by assigned to each Required Action. The first the first digit of the number assigned to a level of logic is identified by the first digit of the Required Action and the placement of the number assigned to a Required Action and the logical connector in the first level of nesting placement of the logical connector in the first (i.e., left justified with the number of the level of nesting (i.e., left justified with the Required Action). number of the Required Action). The successive levels of logic are identified by additional digits of the Required Action number and by successive indentions of the logical connectors.
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| When logical connectors are used to state a Condition, Completion Time, Surveillance, or Frequency, only the first level of logic is used, and the logical connector is left justified with the statement of the Condition, Completion Time, Surveillance, or Frequency.
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| Current EXAMPLES Proposed EXAMPLES The following example illustrates the use of The following examples illustrates the use logical connectors of logical connectors EXAMPLE 1.2-1 REQUIRED ACTION EXAMPLE 1.2-1 REQUIRED ACTION A.1 Suspend ... A.1 Verify Suspend ...
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| AND AND A.2 Initiate ... A.2 Restore Initiate ...
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| EXAMPLE 1.2-2 is proposed to be reinstated.
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| [for entirety of text proposed for reinstatement, see Attachment 1]
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| HDI PNP 2023-030 Enclosure Page 25 of 97 Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, this section will be reinstated. This section currently reflects the logical connector utilized in TS 3.7.14, TS 3.7.15, and TS 3.7.16. These are the only TS that utilize a logical connector in the PDTS. The Required Actions of Example 1.2-1 are currently Suspend and Initiate which more closely aligns with TS in the PDTS. They will be replaced with the standard Verify and Restore. Reinstatement of this TS necessary to support power operation will include TS with the logical connector OR and more complex nesting. Therefore, Example 1.2-2 is reinstated since it pertains to the logical connector OR and more complex nesting. See Attachment 1 for the wording of Example 1.2-2. These changes are administrative.
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| TS SECTION 1.3, COMPLETION TIMES TS 1.3, Completion Times, establishes the Completion Time convention, and provides guidance for its use. It is modified to reflect the power operation condition and the Completion Times that are proposed for the reinstated TS. Modifications to Section 1.3 are shown below. The addition of previously deleted text is described here and shown in Attachment 1.
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| Current BACKGROUND Proposed BACKGROUND Limiting Conditions for Operation (LCOs) Limiting Conditions for Operation (LCOs) specify specify minimum requirements for minimum requirements for ensuring safe ensuring storage and handling of spent operation of the plant storage and handling of fuel. spent nuclear fuel.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, the Background section of TS 1.3 is modified to reflect that PNP will be returned to power operation conditions. This change is administrative.
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| HDI PNP 2023-030 Enclosure Page 26 of 97 Current DESCRIPTION Proposed DESCRIPTION The Completion Time is the amount of time The Completion Time is the amount of time allowed for completing a Required Action. It allowed for completing a Required Action. It is is referenced to the discovery of a situation referenced to the discovery of a situation (e.g.,
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| (e.g., variable not within limits) that requires inoperable equipment or variable not within entering an ACTIONS Condition unless limits) that requires entering an ACTIONS otherwise specified, providing the facility is Condition unless otherwise specified, providing in a specified condition stated in the the plant facility is in a MODE or specified Applicability of the LCO. condition stated in the Applicability of the LCO.
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| The Completion time begins when a Certified Unless otherwise specified, the The Fuel Handler (CFH) on the shift crew with Completion time begins when a senior responsibility for facility operations makes the licensed operator Certified Fuel Handler (CFH) determination that an LCO is not met and an on the operating shift crew with responsibility ACTIONS Condition is entered. for plant operations makes the determination that an LCO is not met and an ACTIONS Required Actions must be completed prior to Condition is entered. The otherwise the expiration of the specified Completion specified.[for entirety of text proposed for Time. An ACTIONS Condition remains in reinstatement, see Attachment 1] in the effect and the Required Actions apply until Completion Time are satisfied.
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| the Condition no longer exists or the plant is not within the LCO Applicability. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the plant is not within the LCO Applicability.
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| If situations are discovered[for entirety of text proposed for reinstatement, see Attachment 1] Example 1.3-3 may not be extended.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, the Description section of TS 1.3 is modified to reflect that PNP will be returned to a power operation condition. As a result, the TS will contain operability requirements for numerous systems and Completion Time rules are expanded to reflect the various situations in the reinstated TS. In addition, reinstatement of the term plant better represents PNP in the operating condition. Senior licensed operator and operating shift crew are reinstated to reflect the required operating staff. The terms Certified Fuel Handler (CFH) is deleted as these personnel will no longer have an operating role once these TS are reinstated. A separate LAR for administrative TS changes will address the staffing changes necessary to support a power operations plant.
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| HDI PNP 2023-030 Enclosure Page 27 of 97 Current EXAMPLES Proposed EXAMPLE The following example illustrates the use The following examples illustrates the use of of Completion Times with different Completion Times with different types of Required Actions. Conditions and changing Conditions Required Actions.
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| See Attachment 1 for complete text of EXAMPLE 1.3-1. EXAMPLE 1.3-1 is modified to address Completion Times as utilized by the reinstated TS. See Attachment 1 for the proposed changes.
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| EXAMPLE 1.3-2 is proposed to be reinstated.
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| EXAMPLE 1.3-3 is proposed to be reinstated.
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| EXAMPLE 1.3-4 is proposed to be reinstated.
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| EXAMPLE 1.3-5 is proposed to be reinstated EXAMPLE 1.3-6 is proposed to be reinstated.
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| EXAMPLE 1.3-7 is proposed to be reinstated.
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| See Attachment 1 for the entirety of the text to be reinstated.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, this section is modified to reflect the use of Completion Times that are utilized in the reinstated TS. The reinstated examples reflect the variety of conditions that are contained in the reinstated TS. These changes to the Examples section of TS 1.3 are administrative changes.
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| TS SECTION 1.4, FREQUENCY TS 1.4, Frequency, defines the proper use and application of Frequency requirements. It is modified to reflect the reinstated TS which support the power operating condition for PNP.
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| Modifications to Section 1.4 in the PDTS are shown below. The addition of previously deleted text is described here and shown in Attachment 1 Current DESCRIPTION Proposed DESCRIPTION The specified Frequency is referred to The specified Frequency is referred to throughout this section and each of the throughout this section and each of the Specifications of Section 3.0, Surveillance Specifications of Section 3.0, Surveillance Requirement (SR) Applicability. The specified Requirement (SR) Applicability. The specified Frequency consists of the requirements of the Frequency consists of the requirements of the Frequency column of each SR. Frequency column of each SR, as well as certain Notes in the Surveillance column The use of "met and performed in these that modify performance requirements.
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| instances conveys specific meanings. A Surveillance is "met only when the acceptance Sometimes special situations dictate when
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| HDI PNP 2023-030 Enclosure Page 28 of 97 criteria are satisfied. Known failure of the the requirements of a Surveillance are to requirements of a Surveillance, even without a be met. [See Attachment 1 for the text to Surveillance specifically being performed, be reinstated.]
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| constitutes a Surveillance not "met. The use of "met and performed in these Performance refers only to the requirement to instances conveys specific meanings. A specifically determine the ability to meet the Surveillance is "met only when the acceptance criteria. acceptance criteria are satisfied. Known failure of the requirements of a Surveillance, even without a Surveillance specifically being performed, constitutes a Surveillance not "met. Performance refers only to the requirement to specifically determine the ability to meet the acceptance criteria
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| [See Attachment 1 for the entirety of the text to be reinstated.] The use of "met and performed in these instances conveys specific meanings. A Surveillance is "met only when the acceptance criteria are satisfied.
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| Known failure of the requirements of a Surveillance, even without a Surveillance specifically being performed, constitutes a Surveillance not "met. Some Surveillances contain notes that modify the Frequency of (continued) performance or the conditions during which the acceptance criteria must be satisfied.[See Attachment 1 for the entirety of the text to be reinstated.]
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| Examples 1.4-3, 1.4-4, 1.4-5, and 1.4-6 discuss these special situations.
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| HDI PNP 2023-030 Enclosure Page 29 of 97 Current EXAMPLES Proposed EXAMPLE The following examples illustrate the type of The following examples illustrate the Frequency statements that appear in the various ways that Frequencies are Technical Specifications (TS). specified. In these examples, the Applicability of the LCO (LCO not shown)
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| Example 1.4-1 is MODES 1, 2, and 3 illustrate the type of Frequency statement that appears in the Example 1.4-2 Technical Specifications (TS).
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| [for entirety of text for Example 1.4-1 is modified to address examples 1.4-1 and 1.4-2, Frequencies as utilized by the reinstated TS.
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| see Attachment 1] See Attachment 1 for proposed changes.
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| Example 1.4-2 is modified to address Frequencies as utilized by the reinstated TS.
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| See Attachment 1 for proposed changes Example 1.4-3 is proposed to be reinstated.
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| Example 1.4-4 is proposed to be reinstated.
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| Example 1.4-5 is proposed to be reinstated.
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| Example 1.4-6 is proposed to be reinstated.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, the Description section of TS 1.4 is modified to reflect the reinstatement of the TS supporting a power operation condition. The number and types of Surveillance Requirements that are included in the operational TS contain many types of Frequencies and examples are added back into the TS to reflect these Frequencies. These changes to the Definitions and examples section of TS 1.4 are administrative changes.
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| TS SECTION 2.0, SAFETY LIMITS (SLs)
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| TS Section 2.0 contains SLs that are necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity from the reactor core and the primary coolant system (PCS) in accordance with 10 CFR 50.36(c)(1).
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| The SLs established in TS Section 2.1 prevent overheating of the fuel and possible cladding perforation, which would result in the release of fission products to the reactor coolant and protects the integrity of the PCS from overpressurization, thereby preventing the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. SL violations in TS Section 2.2 are values of various parameters for which automatic protective action is needed during normal operations or anticipated transients to prevent exceeding an SL.
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| HDI PNP 2023-030 Enclosure Page 30 of 97 Current TS 2.0 Proposed TS 2.0 TS 2.0 (Deleted) TS 2.0 SAFETY LIMITS (SLs) (Deleted)
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| TS 2.1 SLs See Attachment 1 for the complete TS.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS Section 2.0 is proposed for reinstatement since the SLs will apply to PNP reactor power operations in accordance with 10 CFR 50.36(c)(1).
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| TS 2.1.1, Reactor Core SLs, prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. It is applicable in MODES 1 and 2. TS 2.1.1 applies to an operating reactor, and it will be reinstated.
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| TS 2.1.2, Primary Cooling System (PCS) Pressure SL, protects the PCS from over-pressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. It is applicable in MODES 1 through 6.
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| TS 2.1.2 applies to maintaining the PCS pressure and it will be reinstated.
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| In 2.2, SL Violations, TS 2.2.1 defines the action to take if SL 2.1.1 is not met. It requires the unit to be placed in MODE 3. TS 2.2.2 defines the action to take if SL 2.1.2 is not met. If the unit is in MODE 1 or 2, it requires the unit to be placed in MODE 3. If the unit is MODE 3, 4, 5, or 6, it requires compliance to be restored within five minutes. These TS will be reinstated.
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| These TS are required to comply with 10 CFR 50.36(c)(1).
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| TS SECTION 3.0, LIMITING CONDITIONS FOR OPERATION (LCO)
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| TS Section 3.0 contains the general requirements applicable to all Limiting Conditions for Operation (LCOs) and applies at all times unless otherwise stated in a TS. Proposed revisions to the PDTS (including those proposed for reinstatement) are described below. The corresponding TS Bases are also revised to reflect these changes.
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| A mark-up of this section is provided in Attachment 1.
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| Current LCO 3.0.1 Proposed LCO 3.0.1 LCOs shall be met during the specified LCOs shall be met during the MODES or conditions in the Applicability, except as other specified conditions in the Applicability, provided in LCO 3.0.2. except as provided in LCO 3.0.2, LCO 3.0.7, LCO 3.0.8, and LCO 3.0.9.
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| HDI PNP 2023-030 Enclosure Page 31 of 97 Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, the MODES as defined in Table 1.1-1 apply to operating or refueling conditions and are used throughout the reinstated TS. Thus, the reference to MODES is reinstated. In addition, the references to LCOs 3.0.7, LCO 3.0.8, and 3.0.9 are reinstated to reflect the proposed reinstatement of those LCOs as discussed below.
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| Current LCO 3.0.2 Proposed LCO 3.0.2 Upon discovery of a failure to meet an LCO, Upon discovery of a failure to meet an LCO, the Required Actions of the associated the Required Actions of the associated Conditions shall be met. Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, LCO 3.0.2 is modified by reinstating the references to LCOs 3.0.5 and 3.0.6. This change reflects the proposed reinstatement of those LCOs as discussed below.
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| Current LCO 3.0.3 Proposed LCO 3.0.3 Not included When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the plant shall be placed in a MODE or other specified condition in which the LCO is not applicable.LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4.
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| For the entire text, see Attachment 1.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, LCO 3.0.3 provides the actions that must be implemented when an LCO is not met.
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| It is only applicable in MODES 1 through 4. These operating MODES are included in the reinstated TS. Thus, LCO 3.0.3 is needed to address these operational conditions.
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| HDI PNP 2023-030 Enclosure Page 32 of 97 Current LCO 3.0.4 Proposed LCO 3.0.4 Not included When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the plant. For the entire text, see Attachment 1.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, LCO 3.0.4 provides limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. MODES are included in the reinstated TS. Thus, LCO 3.0.4 is needed to address these operational conditions.
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| Current LCO 3.0.5 Proposed LCO 3.0.5 Not included Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, LCO 3.0.5 provides the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The allowance of LCO 3.0.5 to not comply with the requirements of LCO 3.0.2 (i.e., to not comply with the Required Actions) to allow the performance of SRs on equipment declared inoperable or removed from service is necessary as part of the reinstated TS because some TS will include requirements to declare equipment inoperable or to remove it from service.
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| Thus, LCO 3.0.5 is needed to address these Required Actions.
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| HDI PNP 2023-030 Enclosure Page 33 of 97 Current LCO 3.0.6 Proposed LCO 3.0.6 Not included When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered.When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. For the entire text, see Attachment 1.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, LCO 3.0.6 addresses the actions required for a supported system when the support system LCO is not met. It is proposed to be reinstated since there are supported system LCOs in the reinstated TS. Thus, LCO 3.0.6 is needed to address these operational conditions.
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| Current LCO 3.0.7 Proposed LCO 3.0.7 Not included Special Test Exception (STE) LCOs in each applicable LCO section allow specified Technical Specifications (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged.
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| Compliance with STE LCOs is optional.
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| When an STE LCO is desired to be met but is not met, the ACTIONS of the STE LCO shall be met. When an STE LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with the other applicable Specifications.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, LCO 3.0.7 allows certain LCO exceptions when special tests are required to be performed at various times over the life of the plant. It is proposed for reinstatement since there are special test exceptions that will be reinstated in the TS. Thus, LCO 3.0.7 is needed to address these operational conditions.
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| HDI PNP 2023-030 Enclosure Page 34 of 97 Current LCO 3.0.8 Proposed LCO 3.0.8 Not included When one or more required snubbers are unable to perform their associated support function(s), any affected supported LCO(s) are not required to be declared not met solely for this reason if risk is assessed and managed, and:At the end of the specified period the required snubbers must be able to perform their associated support function(s),
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| or the affected supported system LCO(s) shall be declared not met. For the entire text, see Attachment 1.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, LCO 3.0.8 addresses the actions required when one or more required snubbers are unable to perform their associated support function(s). It is proposed to be reinstated because, in an operating plant, snubbers are required to perform their associated support functions. Thus, LCO 3.0.8 is needed to address these operational conditions.
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| Current LCO 3.0.9 Proposed LCO 3.0.9 Not included When one or more required barriers are unable to perform their related support function(s), any supported system LCO(s) are not required to be declared not met solely for this reason for up to 30 days provided that at least one train or subsystem of the supported system is OPERABLE and supported by barriers capable of providing their related support function(s), and risk is assessed and managed.At the end of the specified period, the required barriers must be able to perform their related support function(s) or the supported system LCO(s) shall be declared not met. For the entire text, see Attachment 1.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, LCO 3.0.9 addresses the actions required when one or more required barriers are unable to perform their related support function(s). It is proposed for reinstatement, because there are LCOs that require equipment to be operable or in operation in the TS. Barriers may be required to support certain TS functions. Thus, LCO 3.0.9 is needed to address these operational conditions.
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| HDI PNP 2023-030 Enclosure Page 35 of 97 TS SECTION 3.0, SURVEILLANCE REQUIREMENT (SR)
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| APPLICABILITY TS Section 3.0 contains the general requirements applicable to all SRs and applies at all times unless otherwise stated in a TS. Proposed revisions to these PDTS are described below. The corresponding TS Bases are also revised to reflect these changes.
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| A mark-up of this section is provided.
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| Current SR 3.0.1 Proposed SR 3.0.1 SRs shall be met during the specified SRs shall be met during the MODES or other conditions in the Applicability for individual specified conditions in the Applicability for LCOs, unless otherwise stated in the individual LCOs, unless otherwise stated in SR[for entirety of text, see Attachment the SR[for entirety of text, see 1]Surveillances do not have to be Attachment 1]Surveillances do not have to performed on variables outside specified be performed on inoperable equipment or limits. variables outside specified limits.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, this TS is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| SR 3.0.1 is modified by adding the reference to MODES. MODES are used in the reinstated TS.
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| MODES as defined in Table 1.1-1 are for operating or refueling conditions. This term applies to a plant with reinstated TS.
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| In addition, SR 3.0.1 reinstates the discussion regarding inoperable equipment. The reinstated LCOs include equipment operability requirements.
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| HDI PNP 2023-030 Enclosure Page 36 of 97 Current SR 3.0.2 Proposed SR 3.0.2 The specified Frequency for each SR is met if The specified Frequency for each SR is met if the Surveillance is performed within 1.25 the Surveillance is performed within 1.25 times the interval specified in the Frequency, times the interval specified in the Frequency, as measured from the previous performance. as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.
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| For Frequencies specified as once, the above interval extension does not apply.
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| If a Completion Time requires periodic performance on a once per . . . basis, the above Frequency extension applies to each performance after the initial performance.
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| Exceptions to this Specification are stated in the individual Specifications.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, this TS is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| SR 3.0.2 provides an allowance for extending the frequency for performance of a SR to 1.25 times the interval specified in the Frequency to facilitate scheduling or unforeseen problems that may prevent performance during normal intervals. It is proposed to reinstate the discussion of frequency requirements that will exist in the reinstated TS LCOs.
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| HDI PNP 2023-030 Enclosure Page 37 of 97 Current SR 3.0.4 Proposed SR 3.0.4 Entry into a specified condition in the Entry into a MODE or other specified condition Applicability of an LCO shall only be made in the Applicability of an LCO shall only be when the LCO's Surveillances have been met made when the LCO's Surveillances have within their specified Frequency, except as been met within their specified Frequency, provided by SR 3.0.3. except as provided by SR 3.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.
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| This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the plant.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, this TS is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| SR 3.0.4 is modified by adding the reference to MODES. MODES are used in the reinstated TS.
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| MODES as defined in Table 1.1-1 are for operating or refueling conditions. This term applies to a plant with reinstated TS.
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| In addition, SR 3.0.4 reinstates the provision that states that it shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. The reinstated TS contain Required Actions that would require an entry into another specified condition defined in the Applicability of a TS.
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| TS SECTION 3.1, REACTIVITY CONTROL SYSTEMS TS Section 3.1 contains requirements to assure and verify operability of reactivity control systems to ensure the reactor remains within the bounds of the PNP accident analyses.
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| TS Section 3.1 is proposed for reinstatement in its entirety. Thus, a mark-up of this TS section is not provided. See Attachment 2 for the complete text of the reinstated TS. The corresponding TS Bases are also reinstated to reflect these changes.
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| HDI PNP 2023-030 Enclosure Page 38 of 97 Proposed PNP TS Basis for Change TS 3.1.1, SHUTDOWN MARGIN (SDM) This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.1.1 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.1.1 ensures the SDM is maintained within the limits specified in the COLR. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown events and Anticipated Operational Occurrences (AOOs).
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| The SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.1.2, Reactivity Balance This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.1.2 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.1.2 ensures that core reactivity balance remains within +/- 1% of predicted values. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown events and Anticipated Operational Occurrences (AOOs).
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| The SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.1.3, Moderator Temperature This TS is proposed for reinstatement in its entirety Coefficient (MTC) to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to
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| HDI PNP 2023-030 Enclosure Page 39 of 97 10 CFR 50.82(a)(2), Reference 3, TS 3.1.3 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.1.3 establishes MTC limits during plant operation to ensure stable plant operation. The MTC relates a change in core reactivity to a change in primary coolant temperature.
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| The MTC satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.1.4, Control Rod Alignment This TS is proposed for reinstatement in its entirety, with revision, to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.1.4 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant, except for the NOTE in SR 3.1.4.3.
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| Maximum control rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM.
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| Control rod alignment satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2).
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| The NOTE in SR 3.1.4.3 excepted a control rod drive from the SR for a specific period of time. The cycle referenced in the NOTE has passed and all control rod drives are subject to the SR going forward. Therefore, the NOTE is no longer needed and is not reinstated. This NOTE is related to previous License Condition 2.C.(4) which was deleted as a historical License Condition in Reference 3. See Reference 3, section 4.2.11 for the discussion.
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| TS 3.1.5, Shutdown and Part-Length Control This TS is proposed for reinstatement in its entirety Rod Group Insertion Limits to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.1.5 is reinstated as it existed in the previously
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| HDI PNP 2023-030 Enclosure Page 40 of 97 approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.1.5 establishes limits on insertion of PNPs shutdown and part length control rods to ensure that core reactivity, ejected rod worth, and SDM are preserved. Limits on shutdown rod insertion have been established, and all rod positions are monitored and controlled during power operation.
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| The shutdown and part-length rod group insertion limits satisfy Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.1.6, Regulating Rod Group Position This TS is proposed for reinstatement in its entirety Limits to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.1.6 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.1.6 ensures regulating rod groups are limited to the withdrawal sequence, overlap, and insertion limits specified in the COLR. The insertion limits directly affect core power distributions, assumptions of available SDM, and initial reactivity insertion rate.
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| The regulating rod group position limits satisfy Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.1.7, Special Test Exceptions (STE) This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.1.7 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.1.7 permits suspension of existing LCOs to allow the performance of certain PHYSICS TESTS in MODE 2. These tests are conducted to determine control rod worths, SHUTDOWN MARGIN (SDM), and specific reactor core characteristics. It is acceptable to suspend certain
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| HDI PNP 2023-030 Enclosure Page 41 of 97 LCOs for PHYSICS TESTS because fuel damage criteria are not exceeded.
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| TS SECTION 3.2, POWER DISTRIBUTION LIMITS TS Section 3.2 contains power distribution limits that provide assurance that fuel design criteria are not exceeded, and the accident analysis assumptions remain valid.
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| TS Section 3.2 is proposed for reinstatement in its entirety. Thus, a mark-up of this TS section is not provided. See Attachment 2 for the complete text of the reinstated TS. The corresponding TS Bases are also reinstated to reflect these changes.
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| Proposed PNP TS Basis for Change TS 3.2.1, Linear Heat Rate (LHR) This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.2.1 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.2.1 ensures the LHR remains within the limits specified by the COLR. This LCO limits the amount of damage to the fuel cladding during an accident by ensuring that the plant is operating within acceptable bounding conditions at the onset of a transient.
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| LHR satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.2.2, Total Radial Peaking This TS is proposed for reinstatement in its entirety Factor (FRT) to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.2.2 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.2.2 ensures the FRT remains within the limits specified in the COLR. This LCO limits the amount of damage to the fuel cladding during an accident by ensuring that the plant is operating
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| HDI PNP 2023-030 Enclosure Page 42 of 97 within acceptable bounding conditions at the onset of a transient.
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| The Total Radial Peaking Factor satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.2.3, Quadrant Power Tilt (Tq) This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.2.3 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| The limitations on Tq, specified in TS 3.2.3, ensure that assumptions used in the analysis for establishing LHR limits and DNB margin remain valid during operation. This LCO limits the amount of damage to the fuel cladding during an accident by ensuring that the plant is operating within acceptable bounding conditions at the onset of a transient.
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| The Tq satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.2.4, Axial Shape Index (ASI) This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.2.4 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.2.4 ensures the ASI remains within the limits specified in the COLR. This LCO limits the amount of damage to the fuel cladding during an accident by ensuring that the plant is operating within acceptable bounding conditions at the onset of a transient.
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| The ASI satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| HDI PNP 2023-030 Enclosure Page 43 of 97 TS SECTION 3.3, INSTRUMENTATION TS Section 3.3 contains operability requirements for sensing and control instrumentation required for safe operation of the facility.
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| TS Section 3.3 is proposed for reinstatement in its entirety. Thus, a mark-up of this TS section is not provided. See Attachment 2 for the complete text of the reinstated TS. The corresponding TS Bases are also reinstated to reflect these changes.
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| Proposed PNP TS Basis for Change TS 3.3.1, Reactor Protection System (RPS) This TS is proposed for reinstatement in its entirety Instrumentation to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.3.1, including Tables 3.3.1-1 and 3.3.1-2, is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| The RPS initiates a reactor shutdown as required to mitigate design basis accidents and transients. By tripping the reactor, the RPS also assists the Engineered Safety Features (ESF) systems in mitigating accidents.
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| The RPS instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.3.2, Reactor Protective System (RPS) This TS is proposed for reinstatement in its entirety Logic and Trip Initiation to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.3.2 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| The RPS initiates a reactor shutdown as required to mitigate design basis accidents and transients.
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| By tripping the reactor, the RPS also assists the Engineered Safety Features (ESF) systems in mitigating accidents.
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| The RPS Logic and Trip Initiation satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| HDI PNP 2023-030 Enclosure Page 44 of 97 TS 3.3.3, Engineered Safety Features This TS is proposed for reinstatement in its entirety (ESF) Instrumentation to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.3.3, including Table 3.3.3-1, is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| The ESF Instrumentation initiates necessary safety systems, based upon the values of selected plant parameters, to protect against violating core design limits and the Primary Coolant System (PCS) pressure boundary and to mitigate accidents.
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| The ESF Instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.3.4, Engineered Safety Features This TS is proposed for reinstatement in its entirety (ESF) Logic and Manual Initiation to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.3.4, including Table 3.3.4-1, is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| The ESF Instrumentation initiates necessary safety systems, based upon the values of selected plant parameters, to protect against violating core design limits and the Primary Coolant System (PCS) pressure boundary and to mitigate accidents. ESF manual initiation permits the operator to manually actuate an ESF system when necessary.
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| The ESF satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| HDI PNP 2023-030 Enclosure Page 45 of 97 T 3.3.5, Diesel Generator (DG) - This TS is proposed for reinstatement in its entirety Undervoltage Start (UV Start) to that which was in effect prior to the Instrumentation 10CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 3.3.5 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| The DGs provide a source of emergency power to allow safe operation of the plant.
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| Undervoltage protection instrumentation will generate a DG start in the event a loss of voltage or degraded voltage condition occurs.
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| The DG - UV Start channels satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.3.6, Refueling Containment High This TS is proposed for reinstatement in its entirety Radiation (CHR) Instrumentation to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.3.6 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| CHR Instrumentation provides automatic containment isolation during refueling operations.
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| This ensures the radioactive materials are not released directly to the environment and significantly reduces the offsite doses from those calculated by the safety analyses.
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| The Refueling CHR Instrumentation satisfies the requirements of Criterion 4 of 10 CFR 50.36(c)(2).
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| TS 3.3.7, Post Accident Monitoring (PAM) This TS is proposed for reinstatement in its entirety Instrumentation to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.3.7, including Table 3.3.7-1, is reinstated as it existed in the previously approved TS prior to Amendment
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| HDI PNP 2023-030 Enclosure Page 46 of 97 272, to reflect the power operation condition of the plant.
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| The PAM instrumentation displays plant variables that provide information required by the operators during accident situations. This information provides the necessary support for the operator to take the manual actions, for which no automatic control is provided, that are required for safety systems to accomplish their safety Functions for Design Basis Events.
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| PAM instrumentation that satisfies the definition of Type A in Regulatory Guide 1.97 meets Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.3.8, Alternate Shutdown System This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.3.8, including Table 3.3.8-1, is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| The Alternate Shutdown System provides the control room operator with sufficient instrumentation and controls to maintain the plant in a safe shutdown condition from a location other than the control room.
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| The Alternate Shutdown System has been identified as an important contributor to the reduction of plant risk to accidents and, therefore, satisfies the requirements of Criterion 4 of 10 CFR 50.36(c)(2).
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| TS 3.3.9, Neutron Flux Monitoring Channels This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.3.9 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| HDI PNP 2023-030 Enclosure Page 47 of 97 The neutron flux monitoring channels are necessary to monitor core reactivity changes. By monitoring neutron flux, loss of SDM caused by boron dilution can be detected as an increase in flux.
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| The neutron flux monitoring channels satisfy Criterion 4 of 10 CFR 50.36(c)(2).
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| TS 3.3.10 Engineered Safeguards Room This TS is proposed for reinstatement in its entirety Ventilation (ESRV) Instrumentation to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.3.10 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| The ESRV instrumentation provides isolation of the engineered safeguards pump rooms in the event of high radiation in the pump rooms.
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| Typically, high radiation would only be expected due to excessive leakage during the recirculation phase of operation following a Loss of Coolant Accident (LOCA).
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| The ESRV Instrumentation satisfies the requirements of Criterion 3 of 10 CFR 50.36(c)(2).
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| TS SECTION 3.4, PRIMARY COOLANT SYSTEM (PCS)
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| TS Section 3.4 contains requirements that provide for appropriate control of process variables, design requirements, or operating restrictions needed for appropriate functional capability of PCS equipment required for safe operation of the facility.
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| TS Section 3.4 is proposed for reinstatement in its entirety. Thus, a mark-up of this TS section is not provided. See Attachment 2 for the complete text of the reinstated TS. The corresponding TS Bases are also reinstated to reflect these changes.
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| HDI PNP 2023-030 Enclosure Page 48 of 97 Proposed PNP TS Basis for Change TS 3.4.1, PCS Pressure, Temperature, and This TS is proposed for reinstatement in its entirety Flow Departure from Nucleate Boiling (DNB) to that which was in effect prior to the Limits 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.1 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| The requirements of TS 3.4.1 ensure that PCS pressure, temperature and flow rate remain within the limits specified in the COLR. The limits placed on DNB related parameters ensure that these parameters will not be less conservative than were assumed in the analyses.
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| The PCS DNB limits satisfy Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.4.2, PCS Minimum Temperature for This TS is proposed for reinstatement in its entirety Criticality to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.2 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.2 ensures the PCS temperature remains above the minimum temperature for reactor criticality to prevent operation in an unanalyzed condition.
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| The PCS minimum temperature for criticality satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.4.3, PCS Pressure and Temperature This TS is proposed for reinstatement in its entirety (P/T) Limits to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.3, including Figures 3.4.3-1 and 3.4.3-2, is reinstated as it existed in the previously approved TS prior to
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| HDI PNP 2023-030 Enclosure Page 49 of 97 Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.3 limits the pressure and temperature changes during PCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.
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| The PCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.4.4, PCS Loops - MODES 1 and 2 This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.4 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.4 ensures adequate PCS heat transfer capability during power operation. The primary function of the PCS is removal of the heat generated in the fuel due to the fission process and transfer of this heat, via the Steam Generators (SGs), to the secondary plant.
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| PCS Loops - MODES 1 and 2 satisfy Criteria 2 and 3 of 10 CFR 50.36(c)(2).
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| TS 3.4.5, PCS Loops - MODE 3 This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.5 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.5 ensures adequate PCS heat transfer capability during hot standby. The primary function of the primary coolant in MODE 3 is removal of decay heat and transfer of this heat, via the Steam Generators (SGs), to the secondary plant fluid. The secondary function of the primary coolant is to act as a carrier for soluble neutron poison, boric acid.
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| HDI PNP 2023-030 Enclosure Page 50 of 97 PCS Loops - MODE 3 satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.4.6, PCS Loops - MODE 4 This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.6 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.6 ensures adequate PCS heat transfer capability during hot shutdown. The intent of this LCO is to provide forced flow for decay heat removal and transport.
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| PCS Loops - MODE 4 satisfies Criterion 4 of 10 CFR 50.36(c)(2).
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| TS 3.4.7, PCS Loops - MODE 5, Loops This TS is proposed for reinstatement in its entirety Filled to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.7 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.7 ensures adequate PCS heat transfer capability during cold shutdown with the PCS piping filled. In MODE 5 with the PCS loops filled, the primary function of the primary coolant is the removal of decay heat and transfer this heat either to the Steam Generator (SG) secondary side coolant via natural circulation or the Shutdown Cooling (SDC) heat exchangers.
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| PCS Loops - MODE 5 (Loops Filled) satisfies Criterion 4 of 10 CFR 50.36(c)(2).
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| HDI PNP 2023-030 Enclosure Page 51 of 97 TS 3.4.8, PCS Loops - MODE 5, Loops This TS is proposed for reinstatement in its entirety Not Filled to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.8 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.8 ensures adequate PCS heat transfer capability during cold shutdown with the PCS piping not filled. In MODE 5 with loops not filled, only the SDC System can be used for coolant circulation.
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| PCS loops - MODE 5 (Loops Not Filled) satisfies Criterion 4 of 10 CFR 50.36(c)(2).
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| TS 3.4.9, Pressurizer This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.9 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.9 provides requirements for pressurizer water level, required pressurizer heater capacity, and heater power supply to ensure proper operation of the pressurizer. The pressurizer provides a point in the PCS where liquid and vapor are maintained in equilibrium under saturated conditions for pressure control purposes.
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| The pressurizer satisfies Criterion 2 (for pressurizer water level) and Criterion 4 (for pressurizer heaters) of 10 CFR 50.36(c)(2).
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| TS 3.4.10, Pressurizer Safety Valves This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.10, including Table 3.4.10-1, is reinstated as it existed
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| HDI PNP 2023-030 Enclosure Page 52 of 97 in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.10 ensures the pressurizer safety valves are capable of providing PCS overpressure protection. Operating in conjunction with the Reactor Protection System, these valves are used to ensure that the pressure Safety Limit is not exceeded for analyzed transients.
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| The pressurizer safety valves satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.4.11, Pressurizer Power Operated This TS is proposed for reinstatement in its entirety Relief Valves (PORVs) to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.11 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.11 ensures the PORVs are capable of providing PCS overpressure protection. The primary purpose of this LCO is to ensure that the PORV and the block valve are operating correctly so the potential for a LOCA through the PORV pathway is minimized, or if a LOCA were to occur through a failed open PORV, the block valve could be manually operated to isolate the path.
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| Pressurizer PORVs satisfy Criterion 4 of 10 CFR 50.36(c)(2).
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| HDI PNP 2023-030 Enclosure Page 53 of 97 TS 3.4.12, Low Temperature Overpressure This TS is proposed for reinstatement in its entirety Protection (LTOP) System to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.12, including Figure 3.4.12-1, is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.12 ensures the integrity of the primary coolant pressure boundary by maintaining the PCS pressure within allowable values (the Pressure and Temperature (P/T) limits of 10 CFR 50, Appendix G) at low temperatures.
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| The LTOP System satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.4.13, PCS Operational LEAKAGE This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.13 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.13 limits operation when PCS leakage is present. The purpose of the PCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from certain sources to amounts that do not compromise safety.
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| PCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| HDI PNP 2023-030 Enclosure Page 54 of 97 TS 3.4.14, PCS Pressure Isolation Valve This TS is proposed for reinstatement in its entirety (PIV) Leakage to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.14 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.14 ensures that PIV leakage or inadvertent valve positioning does not result in overpressure of low-pressure piping and components.
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| Although this specification provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low-pressure portions of connecting systems.
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| PCS PIV leakage satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.4.15, PCS Leakage Detection This TS is proposed for reinstatement in its entirety Instrumentation to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.15 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.15 ensures instrumentation is provided to detect and identify PCS leakage. Leakage detection instrumentation must have the capability to detect significant Primary Coolant Pressure Boundary (PCPB) degradation as soon after occurrence as practical to minimize the potential for propagation to a gross failure.
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| PCS leakage detection instrumentation satisfies Criterion 1 of 10 CFR 50.36(c)(2).
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| HDI PNP 2023-030 Enclosure Page 55 of 97 TS 3.4.16, PCS Specific Activity This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.16 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.16 limits the allowable concentration level of radionuclides in the primary coolant to minimize the offsite dose consequences in the event of a steam generator tube rupture or other accident.
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| PCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.4.17, Steam Generator (SG) Tube This TS is proposed for reinstatement in its entirety Integrity to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.4.17 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.4.17 ensures the primary containment pressure boundary (PCPB) function of the SG.
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| This Specification addresses only the PCPB integrity function of the SG.
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| Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
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| TS SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS (ECCS)
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| TS Section 3.5 contains requirements that provide for appropriate functional capability of ECCS equipment required for mitigation of DBAs or transients to protect the integrity of a fission product barrier.
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| TS Section 3.5 is proposed for reinstatement in its entirety. Thus, a mark-up of this TS section is not provided. See Attachment 2 for the complete text of the reinstated TS. The corresponding TS Bases are also reinstated to reflect these changes.
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| HDI PNP 2023-030 Enclosure Page 56 of 97 Proposed PNP TS Basis for Change TS 3.5.1, Safety Injection Tanks (SITs) This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.5.1 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.5.1 ensures the functions of the SITs are maintained. They are to supply water to the reactor vessel during the blowdown phase of a Loss of Coolant Accident (LOCA), to provide inventory to help accomplish the refill phase that follows thereafter, and to provide Primary Coolant System (PCS) makeup for a small break LOCA.
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| The SITs satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.5.2, ECCS - Operating This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.5.2 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.5.2 ensures the ECCS can provide core cooling and negative reactivity to protect the reactor core during accidents involving inventory loss.
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| ECCS - Operating satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.5.3, ECCS - Shutdown This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.5.3 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power
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| HDI PNP 2023-030 Enclosure Page 57 of 97 operation condition of the plant.
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| TS 3.5.3 ensures the ECCS can provide core cooling and negative reactivity to protect the reactor core during accidents involving inventory loss while in the hot shutdown condition.
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| ECCS - Shutdown satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.5.4, Safety Injection Refueling This TS is proposed for reinstatement in its entirety Water Tank (SIRWT) to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.5.4 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.5.4 ensures a source of borated water is available for containment spray system and engineered safeguards pump operation.
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| The SIRWT satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.5.5, Containment Sump Buffering This TS is proposed for reinstatement in its entirety Agent and Weight Requirements to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.5.5 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.5.5 ensures the buffering agent, i.e., sodium tetraborate, results in a post-LOCA sump water pH value consistent with accident analyses.
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| STB satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| HDI PNP 2023-030 Enclosure Page 58 of 97 TS SECTION 3.6, CONTAINMENT SYSTEMS TS Section 3.6 contains requirements that assure the integrity of the containment, depressurization and cooling systems, and containment isolation valves.
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| TS Section 3.6 is proposed for reinstatement in its entirety. Thus, a mark-up of this TS section is not provided. See Attachment 2 for the complete text of the reinstated TS. The corresponding TS Bases are also reinstated to reflect these changes.
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| Proposed PNP TS Basis for Change TS 3.6.1, Containment This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.6.1 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.6.1 ensures a containment configuration that is consistent with the safety analyses. Compliance with this LCO will ensure a containment configuration, including the equipment hatch, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analysis.
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| The containment satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.6.2, Containment Air Locks This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.6.2 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.6.2 ensures the containment air locks perform their design function as part of the containment pressure boundary. As part of the containment pressure boundary, the air lock safety function is related to limiting the containment leakage rate to 1.0 La.
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| HDI PNP 2023-030 Enclosure Page 59 of 97 The containment air locks satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.6.3, Containment Isolation Valves This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.6.3 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.6.3 ensures the containment isolation valves and devices are capable of providing containment isolation within the time limits assumed in the safety analyses.
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| The containment isolation valves satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.6.4, Containment Pressure This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.6.4 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.6.4 ensures containment pressure is limited during normal operation to preserve the initial conditions assumed in accident analyses.
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| Containment pressure satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.6.5, Containment Air Temperature This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.6.5 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.6.5 ensures containment pressure is limited during normal operation to preserve the initial
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| HDI PNP 2023-030 Enclosure Page 60 of 97 conditions assumed in accident analyses.
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| Containment pressure satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| TS 3.6.6, Containment Cooling Systems This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.6.6 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.6.6 ensures Containment Spray and Containment Air Cooler systems are capable of limiting post-accident pressure and temperature in containment to less than the design values.
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| The Containment Spray System and the Containment Cooling System satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS SECTION 3.7, PLANT SYSTEMS TS Section 3.7 provides requirements for the appropriate functional capability of plant equipment required for safe operation of the facility, including requirements that apply when the facility is in a defueled condition.
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| TS 3.7.1 through TS 3.7.13, and TS 3.7.17 are proposed for reinstatement in their entirety.
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| Thus, a mark-up of these TS is not provided. See Attachment 2 for the complete text of the reinstated TS.
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| TS 3.7.14 through TS 3.7.16 are modified to reinstate the reference to LCO 3.0.3. PNP will also use the Surveillance Frequency Control Program (SFCP), implemented in Amendment 271 (Reference 10), in the TS. Therefore, the Frequency of SR 3.7.14.1 and SR 3.7.15.1 will revert to the SFCP. The actual values of the Frequency do not change. The relocation of the Frequencies is administrative in nature.
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| Mark-ups of TS 3.7.14, TS 3.7.15, and TS 3.7.16 are provided in Attachment 1 in this enclosure.
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| The corresponding TS Bases are also reinstated and revised to reflect these changes.
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| Proposed PNP TS Basis for Change TS 3.7.1, Main Steam Safety Valves This TS is proposed for reinstatement in its entirety (MSSVs) to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of
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| HDI PNP 2023-030 Enclosure Page 61 of 97 the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.7.1, including Table 3.7.1-1, is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.7.1 ensures the MSSVs are capable of providing protection against over-pressurization of the secondary system and the primary coolant system.
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| The MSSVs satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.7.2, Main Steam Isolation Valves This TS is proposed for reinstatement in its entirety (MSIVs) to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.7.2 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.7.2 ensures the MSIVs are capable of isolating steam flow from the secondary side of the steam generators in the event of a high energy line break.
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| The MSIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.7.3, Main Feedwater Regulating Valves This TS is proposed for reinstatement in its entirety (MFRVs) and MFRV Bypass Valves to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.7.3 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.7.3 ensures the MFRVs and MFRV bypass valves provide steam generator level control during normal plant operation and provide isolation in the event of a high energy line break.
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| HDI PNP 2023-030 Enclosure Page 62 of 97 The MFRVs and MFRV bypass valves satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.7.4, Atmospheric Dump Valves This TS is proposed for reinstatement in its entirety (ADVs) to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.7.4 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.7.4 ensures the ADVs are capable of removing decay heat should the preferred heat sink not be available.
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| The ADVs satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.7.5, Auxiliary Feedwater (AFW) This TS is proposed for reinstatement in its entirety System to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.7.5 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.7.5 ensures the AFW System automatically supplies feedwater to the steam generators to remove decay heat from the Primary Coolant System upon the loss of normal feedwater supply.
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| The AFW System satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.7.6, Condensate Storage and Supply This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.7.6 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| HDI PNP 2023-030 Enclosure Page 63 of 97 TS 3.7.6 ensures a supply of a safety-grade source of water to the steam generators for removing decay and sensible heat from the Primary Coolant System (PCS).
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| The Condensate Storage and Supply satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.7.7, Component Cooling Water (CCW) This TS is proposed for reinstatement in its entirety System to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.7.7 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.7.7 ensures a heat sink for the removal of process and operating heat from safety related components during a DBA or transient.
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| The CCW System satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.7.8, Service Water System (SWS) This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.7.8 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.7.8 ensures a heat sink for the removal of process and operating heat from safety related components during a DBA or transient.
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| The SWS satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.7.9, Ultimate Heat Sink (UHS) This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.7.9 is
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| HDI PNP 2023-030 Enclosure Page 64 of 97 reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.7.9 ensures a heat sink is provided for process and operating heat from safety related components during a DBA or transient, as well as during normal operation.
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| The UHS satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.7.10, Control Room Ventilation This TS is proposed for reinstatement in its entirety (CRV) Filtration to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.7.10 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.7.10 ensures the CRV Filtration system provides a protected environment from which occupants can control the plant following an uncontrolled release of radioactivity.
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| The CRV Filtration system satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| 3.7.11, Control Room Ventilation This TS is proposed for reinstatement in its entirety (CRV) Cooling to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.7.11 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.7.11 ensures the CRV Cooling system provides temperature control for the control room during normal and emergency conditions.
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| The CRV Cooling system satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| HDI PNP 2023-030 Enclosure Page 65 of 97 3.7.12, Fuel Handling Area This TS is proposed for reinstatement in its entirety Ventilation System to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.7.12 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.7.12 ensures the Fuel Handling Area Ventilation System filters airborne radioactive particulates from the area of the spent fuel pool following a fuel handling accident or a fuel cask drop accident.
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| For the fuel handling accident, the Fuel Handling Area Ventilation System satisfies Criterion 4 of 10 CFR 50.36(c)(2). For the fuel cask drop accident, the Fuel Handling Area Ventilation System satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.7.13, Engineered Safeguards This TS is proposed for reinstatement in its entirety Room Ventilation (ESRV) Dampers to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.7.13 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| The ESRV dampers provide isolation of the engineered safeguards room in the event of a high radiation alarm.
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| The ESRV Dampers satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.7.14, Spent Fuel Pool (SFP) Water TS 3.7.14 is in the PDTS. Modifications to this TS Level are shown in Attachment 1. These modifications ensure that TS 3.7.14 is reinstated in its entirety, as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.7.14 the initial SFP water level assumed in the fuel handling accident analysis and cask drop analysis is maintained.
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| HDI PNP 2023-030 Enclosure Page 66 of 97 The TS section title is administratively changed from Facility Systems to Plant Systems to reflect that this TS section addresses operational facility requirements. In addition, the NOTE in the Actions (LCO 3.0.3 is not applicable) is proposed to be reinstated to conform to the reinstatement of TS LCO 3.0.3 as previously proposed.
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| PNP will use the SFCP, implemented as Amendment 271 (Reference 10), in the TS.
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| Therefore, the Frequency of SR 3.7.14.1 will revert to the SFCP. Consequently, SR 3.7.14.1 Frequency is changed from 7 days to In accordance with the Surveillance Frequency Control Program. The actual value of the Frequency does not change. The relocation of this Frequency is administrative in nature.
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| HDI PNP 2023-030 Enclosure Page 67 of 97 TS 3.7.15, Spent Fuel Pool (SFP) Boron TS 3.7.15 is in the PDTS. Modifications to this TS Concentration are shown in Attachment 1. These modifications ensure that TS 3.7.15 is reinstated in its entirety, as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.7.15 ensures the initial SFP boron concentration assumed in the fuel handling accident analysis in maintained.
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| The TS section title is administratively changed from Facility Systems to Plant Systems to reflect that this TS section addresses operational facility requirements. In addition, the NOTE in the Actions (LCO 3.0.3 is not applicable) is proposed to be reinstated to conform to the reinstatement of TS LCO 3.0.3 as previously proposed.
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| PNP will use the SFCP, implemented as Amendment 271 (Reference 10), in the TS.
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| Therefore, the Frequency of SR 3.7.15.1 will revert to the SFCP. Consequently, SR 3.7.15.1 Frequency is changed from 7 days to In accordance with the Surveillance Frequency Control Program. The actual value of the Frequency does not change. The relocation of this Frequency is administrative in nature.
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| TS 3.7.16, Spent Fuel Pool Storage TS 3.7.16, including Tables 3.7.16-1 through 3.7.16-5, is in the PDTS. Modifications to this TS are shown in Attachment 1. These modifications ensure that TS 3.7.16 is reinstated in its entirety, as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| The TS section title is administratively changed from Facility Systems to Plant Systems to reflect that this TS section addresses operational facility requirements. In addition, the NOTE in the Actions (LCO 3.0.3 is not applicable) is proposed to be reinstated to conform to the reinstatement of TS LCO 3.0.3 as previously proposed.
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| HDI PNP 2023-030 Enclosure Page 68 of 97 TS 3.7.17, Secondary Specific Activity This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.7.17 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.7.17 ensures steam generator tube out-leakage is identified. A limit on secondary coolant specific activity during power operation minimizes releases to the environment because of normal operation, anticipated operational occurrences, and accidents.
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| Secondary specific activity limits satisfy Criterion 2 of 10 CFR 50.36(c)(2).
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| TS SECTION 3.8, ELECTRICAL POWER SYSTEMS TS Section 3.8 contains operability requirements that provide for appropriate functional capability of plant electrical equipment required for safe operation of the facility.
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| TS Section 3.8 is proposed for reinstatement in its entirety. Thus, a mark-up of this TS section is not provided. See Attachment 2 for the complete text of the reinstated TS. The corresponding TS Bases are also reinstated to reflect these changes.
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| Proposed PNP TS Basis for Change TS 3.8.1, AC Sources - Operating This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.8.1 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.8.1 ensures independent and redundant sources of power to the ESF systems comprised of offsite power sources and onsite standby diesel generators.
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| The AC sources satisfy Criterion 3 of
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| HDI PNP 2023-030 Enclosure Page 69 of 97 10 CFR 50.36(c)(2).
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| TS 3.8.2, AC Sources - Shutdown This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.8.2 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.8.2 ensures independent and redundant sources of power to the ESF systems comprised of offsite power sources and onsite standby diesel generators.
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| The AC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.8.3, Diesel Fuel, Lube Oil, and Starting This TS is proposed for reinstatement in its entirety Air to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.8.3 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.8.3 ensures the diesel generators (DGs) are capable of performing their design function.
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| Since diesel fuel, lube oil, and starting air subsystems support the operation of the standby AC power sources, they satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.8.4, DC Sources - Operating This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.8.4 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| HDI PNP 2023-030 Enclosure Page 70 of 97 TS 3.8.4 ensures the DC electrical system supports the AC power system and selected safety related equipment.
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| The DC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.8.5, DC Sources - Shutdown This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 3.8.5 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.8.5 ensures the DC electrical system supports the AC power system and selected safety related equipment.
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| The DC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.8.6, Battery Cell Parameters This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.8.6, including Table 3.8.6-1, is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.8.6 ensures battery cell parameters remain within acceptable limits to ensure availability of the required DC power.
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| Battery cell parameters satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| HDI PNP 2023-030 Enclosure Page 71 of 97 TS 3.8.7, Inverters - Operating This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.8.7 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.8.7 ensures the inverters are capable of providing continuous AC power to the preferred AC buses.
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| Inverters are part of the distribution system and, as such, satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.8.8, Inverters - Shutdown This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.8.8 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.8.8 ensures the inverters are capable of providing continuous AC power to the preferred AC buses.
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| Inverters are part of the distribution system and, as such, satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.8.9, Distribution Systems - Operating This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.8.9 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.8.9 ensures an independent and redundant source of power to the ESF systems.
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| HDI PNP 2023-030 Enclosure Page 72 of 97 The distribution systems satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.8.10, Distribution Systems This TS is proposed for reinstatement in its entirety
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| - Shutdown to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.8.10 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.8.10 ensures an independent and redundant source of power to the ESF systems.
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| The distribution system satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS SECTION 3.9, REFUELING OPERATIONS TS Section 3.9 contains requirements that provide for appropriate functional capability of parameters and equipment that are required for mitigation of DBAs during refueling operations (moving irradiated fuel to or from the reactor core).
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| TS Section 3.9 is proposed for reinstatement in its entirety. Thus, a mark-up of this TS section is not provided. See Attachment 2 for the complete text of the reinstated TS. The corresponding TS Bases are also reinstated to reflect these changes.
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| Proposed PNP TS Basis for Change TS 3.9.1, Boron Concentration This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.9.1 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.9.1 ensures the boron concentration of the PCS and the refueling cavity during refueling activities remain within limits to ensure that the reactor remains subcritical during MODE 6.
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| Boron concentration satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| HDI PNP 2023-030 Enclosure Page 73 of 97 TS 3.9.2, Nuclear Instrumentation This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.9.2 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.9.2 ensures instrumentation is available to monitor the core reactivity condition during refueling activities.
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| Nuclear Instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| TS 3.9.3, Containment Penetrations This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.9.3 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.9.3 ensures that a release of fission product radioactivity from a fuel handling accident in containment will be mitigated.
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| The Containment Penetrations satisfy the requirements of Criterion 4 of 10 CFR 50.36(c)(2).
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| HDI PNP 2023-030 Enclosure Page 74 of 97 TS 3.9.4, Shutdown Cooling (SDC) and This TS is proposed for reinstatement in its entirety Coolant Circulation - High Water Level to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.9.4 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.9.4 ensures the removal of decay and sensible heat and the mixing of borated coolant during refueling activities.
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| SDC and Coolant Circulation - High Water Level satisfies Criterion 4 of 10 CFR 50.36(c)(2).
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| TS 3.9.5, Shutdown Cooling (SDC) and This TS is proposed for reinstatement in its entirety Coolant Circulation - Low Water Level to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.9.5 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.9.5 ensures the removal of decay and sensible heat and the mixing of borated coolant during refueling activities.
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| SDC and Coolant Circulation - Low Water Level satisfies Criterion 4 of 10 CFR 50.36(c)(2).
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| TS 3.9.6, Refueling Cavity Water Level This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, TS 3.9.6 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 3.9.6 ensures sufficient water level in the refueling cavity and spent fuel pool during refueling activities.
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| HDI PNP 2023-030 Enclosure Page 75 of 97 Refueling cavity water level satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| TS Section 4.0, DESIGN FEATURES TS Section 4.0, Design Features, provides information and design requirements associated with plant systems.
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| TS 4.2 and TS 4.3.1.4 apply in an operating condition and are proposed to be reinstated. See Attachment 1 for the marked-up pages.
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| Proposed PNP TS Basis for Change TS 4.2, Reactor Core This TS is proposed for reinstatement in its entirety to that which was in effect prior to 4.2.1 Fuel Assemblies the 10 CFR 50.82(a)(1) certifications to The reactor core shall contain 204 fuel restore the PNP power operations RFOL.
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| assemblies. Each assembly shall consist of a Upon recission of the 10 CFR 50.82(a)(1) matrix[for entirety of text proposed for certifications, as conditioned by the reinstatement, see Attachment 1]. exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 4.2 is reinstated as it 4.2.2 Control Rod Assemblies existed in the previously approved TS prior The reactor core shall contain 45 control rods. to Amendment 272, to reflect the power Four of these control rods may consist of operation condition of the plant.
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| part-length absorbers. The control material shall be silver-indium-cadmium, as approved TS 4.2 provides requirements for reactor by the NRC. fuel assemblies and control rods in the reactor core. In an operating plant, fuel assemblies and control rod requirements for the core are necessary. Therefore, this TS will be reinstated.
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| 4.3.1 Criticality This TS is proposed for reinstatement in its entirety to that which was in effect prior to 4.3.1.1.a New or irradiated fuel assemblies the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| 4.3.1.2.e New or irradiated fuel assemblies Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the 4.3.1.3.e New or irradiated fuel assemblies exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 4.3.1 is reinstated as it
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| [For entirety of TS 4.3.1 text, see Attachment 1] existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| These TS sections are proposed to be revised to reinstate references to new fuel storage. TS Sections 4.3.1.1, 4.3.1.2, and 4.3.1.3 provide requirements to ensure the new and irradiated fuel stored in racks in
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| HDI PNP 2023-030 Enclosure Page 76 of 97 Regions I and II of the spent fuel pool remains subcritical. As an operating plant, PNP will receive new fuel and these sections address storage of that new fuel.
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| TS 4.3.1.4 This TS is proposed for reinstatement in its entirety to that which was in effect prior to The new fuel storage racks are designed and the 10 CFR 50.82(a)(1) certifications to shall be maintained with[for entirety of text restore the PNP power operations RFOL.
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| proposed for reinstatement, see Attachment Upon recission of the 10 CFR 50.82(a)(1) 1]. certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 4.3.1.4, including Figure 4.3-1, is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 4.3.1.4 provides requirements for the fuel racks for new fuel assemblies. With the return to power operation, PNP will be authorized to receive new fuel assemblies for reactor reloads.
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| TS Section 5.0, Administrative Controls TS Section 5.0 establishes the requirements associated with site personnel responsibilities, the site organization, staffing, training, procedures, programs, reporting requirements, and high radiation areas. Portions of this section are proposed to be reinstated as administrative requirements are needed for power operation of PNP. Sections reinstated in their entirety are shown in Attachment 2. Sections that are marked up from the current PDTS are shown in Attachment 1.
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| Proposed PNP TS Basis for Change TS 5.5.2, Primary Coolant Sources Outside This TS is proposed for reinstatement in its Containment entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.5.2 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| This program provides controls to minimize primary coolant leakage to the
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| HDI PNP 2023-030 Enclosure Page 77 of 97 engineered safeguards rooms from portions of systems outside containment during mitigation of a DBA occurring in containment. Those systems include the containment spray system, the safety injection system, the shutdown cooling system, and the containment sump suction piping.
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| As discussed in Section 3.0, the UFSAR, Revision 35, Chapter 14 accidents and transients inside containment are applicable and therefore, recirculation of post-accident highly radioactive primary coolant outside containment could occur.
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| Therefore, this program is reinstated in the TS.
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| TS 5.5.5, Containment Structural Integrity This TS is proposed for reinstatement in its Surveillance Program entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.5.5 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| This program controls monitoring of several containment attributes to ensure containment structural integrity.
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| During power operations the containment is credited as part of the initial conditions of the applicable accident analyses and as part of the primary success path for mitigation of these events.
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| In addition, this program is invoked in TS 3.6.1, Containment which is being reinstated. Therefore, this program is also reinstated in the TS.
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| TS 5.5.6, Primary Coolant Pump Flywheel This TS is proposed for reinstatement in its Surveillance Program entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the
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| HDI PNP 2023-030 Enclosure Page 78 of 97 exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.5.6 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| With the return to power operation, the flywheel surveillance program provides the inspection frequencies and acceptance criteria for the reactor coolant pump flywheel inspection program. Therefore, this program is reinstated in the TS.
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| TS 5.5.8, Steam Generator (SG) Program This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.5.8 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| This TS requires that a steam generator program be established and implemented to ensure that steam generator tube integrity is maintained.
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| With the reinstatement of TS 3.4.17, Steam Generator Tube Integrity this program is needed as this TS invokes its use. Therefore, this program is also reinstated in the TS.
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| TS 5.5.9, Secondary Water Chemistry This TS is proposed for reinstatement in its Program entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.5.9 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| The purpose of the secondary water chemistry program is to maintain water chemistry to inhibit steam generator tube
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| HDI PNP 2023-030 Enclosure Page 79 of 97 degradation. This program is necessary for the operating plant. Therefore, this program is reinstated in the TS.
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| TS 5.5.10, Ventilation Filter Testing Program This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.5.10 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| This program implements certain testing procedures for the control room ventilation and fuel handling area ventilation systems.
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| With the reinstatement of TS 3.7.10, Control Room Ventilation (CRV)
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| Filtration and TS 3.7.12, Fuel Handling Area Ventilation this program is needed as these TS invoke its use. Therefore, this program is also reinstated in the TS.
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| TS 5.5.11, Fuel Oil Testing Program This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.5.11 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| This program implements required testing of both new fuel oil and stored fuel oil to ensure compliance with manufacturers specifications and applicable ASTM Standards.
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| With the reinstatement of TS 3.8.3, Diesel Fuel, Lube Oil and Starting Air, this program is needed as this TS invokes its use. Therefore, this program is also reinstated in the TS.
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| HDI PNP 2023-030 Enclosure Page 80 of 97 TS 5.5.13, Safety Functions Determination This TS is proposed for reinstatement in its Program (SFDP) entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.5.13 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| This program was established to ensure loss of safety function is detected and appropriate actions taken when a supported system LCO is not met solely due to a support system LCO not being met.
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| With the resumption of power operation and reinstatement of the TS, redundant systems are required to mitigate the UFSAR, Revision 35, Chapter 14 accidents and transients. Therefore, the requirements of the SFDP, which directs cross train checks of multiple and redundant safety systems, apply.
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| Additionally, the SFDP is invoked in LCO 3.0.6, which is being reinstated in its entirety as previously discussed.
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| Therefore, this program is also reinstated in the TS.
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| TS 5.5.14, Containment Leak Rate Testing This TS is proposed for reinstatement in its Program entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.5.14 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 5.5.14 requires that a program to address the leakage rate testing of the containment be established. With the resumption of power operation and the reinstatement of the TS, Containment
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| HDI PNP 2023-030 Enclosure Page 81 of 97 integrity is credited in the analysis of the UFSAR, Revision 35, Chapter 14 accidents and transients.
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| In addition, this program is invoked in TS 3.6.1, Containment and TS 3.6.2, Containment Air Locks which are being reinstated. Therefore, this program is also reinstated in the TS.
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| TS 5.5.16, Control Room Envelope This TS is proposed for reinstatement in its Habitability Program entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.5.16 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 5.5.16 establishes the administrative program for testing of the control room habitability systems to ensure operators can safely implement actions to control the reactor and mitigate accidents from within the control room envelope. These tests of control room habitability systems are necessary to support safe operation of the plant because reactor accidents challenging control room habitability are possible.
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| With the reinstatement of TS 3.7.10, Control Room Ventilation (CRV) Filtration this program is needed as this TS invokes its use. Therefore, this program is also reinstated in the TS.
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| TS. 5.5.17, Surveillance Frequency Control This TS is proposed for reinstatement in its Program entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.5.17 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| HDI PNP 2023-030 Enclosure Page 82 of 97 The SFCP at PNP was implemented via Amendment 271 (Reference 12). The SFCP requires that a program be established to ensure that SRs specified in the TS are performed at intervals sufficient to assure the associated LCOs are met.
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| PNP uses the SFCP in the TS as shown in Attachment 2. With the reinstatement of numerous TS invoking its use, this program is also reinstated.
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| TS 5.6.2, Radiological Environmental This TS is proposed for reinstatement in its Operating Report entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.6.2 is modified by replacing the reference to the facility with the plant to reflect the power operation condition as follows:
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| The Radiological Environmental Operating Report covering the operation of the plant facility during[for entirety of 5.6.2 text, see Attachment 1]
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| This is an administrative change.
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| TS 5.6.3, Radioactive Effluent Release This TS is proposed for reinstatement in its Report entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.6.3 is modified by replacing the references to the facility with the plant to reflect the power operation condition as follows:
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| The Radioactive Effluent Release Report covering operation of the plant facility in the previous year.[for entirety of 5.6.3 text, see Attachment 1]
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| This is an administrative change.
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| HDI PNP 2023-030 Enclosure Page 83 of 97 TS 5.6.5, CORE OPERATING LIMITS This TS is proposed for reinstatement in its REPORT (COLR) entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.6.5 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| The COLR provides the required documentation and analytical methods used to determine the reactor core operating limits. The COLR is the plant specific document that provides cycle specific parameter limits for the current reload cycle. The COLR will be reestablished and controlled by this TS.
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| Additionally, numerous reinstated TS reference the COLR for operating limits.
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| Therefore, the requirements for this report are also reinstated.
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| TS 5.6.6, Post Accident Monitoring Report This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.6.6 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| Since TS 3.3.7, Post Accident Monitoring (PAM) Instrumentation is reinstated, the post-accident monitoring instrument reporting requirements from TS 3.3.7, Conditions B and G, will also reinstated, including references to TS 5.6.6.
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| Therefore, this report is also reinstated.
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| HDI PNP 2023-030 Enclosure Page 84 of 97 TS 5.6.7, Containment Structural Integrity This TS is proposed for reinstatement in its Surveillance Report entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.6.7 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 5.5.5, Containment Structural Integrity Surveillance Program, is proposed to be reinstated in its entirety. It references this report. Therefore, this report is also reinstated.
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| TS 5.6.8, Steam Generator Tube Inspection This TS is proposed for reinstatement in its Report entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL.
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| Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, TS 5.6.8 is reinstated as it existed in the previously approved TS prior to Amendment 272, to reflect the power operation condition of the plant.
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| TS 5.5.8, Steam Generator (SG)
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| Program, which ensures that SG tube integrity is maintained, is proposed to be reinstated. It references this report.
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| Therefore, this report is also reinstated.
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| 3.2.3 Proposed Changes to RFOL Appendix B, Environmental Protection Plan All changes to the EPP proposed in this LAR are made solely to more accurately reflect the PNP plant after it resumes power operation and to ensure the terminology used in the EPP is consistent with that used in the plant license. The proposed changes do not alter the obligations in the environmental area, including, as appropriate, requirements for reporting and keeping records of environmental data, and any conditions and monitoring requirement for the protection of the nonaquatic environment. As such, the changes to the EPP proposed by this LAR are administrative changes only.
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| The EPP is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, the EPP is reinstated as it existed in the previously approved RFOL prior to
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| HDI PNP 2023-030 Enclosure Page 85 of 97 Amendment 272, to reflect the power operation condition of the plant.
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| Current Table of Contents Proposed Table of Contents 3.1 Facility Design and Operation 3.1 Plant Facility Design and Operation 5.4 Facility Reporting Requirements 5.4 Plant Facility Reporting Requirements Basis To reflect PNP as an operating plant, the EPP is administratively revised to replace the word Facility with Plant to be consistent with terminology for the TS proposed in this LAR.
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| Current Section 1.0 Proposed Section 1.0 1.0 Objectives of the Environmental 1.0 Objectives of the Environmental Protection Plan Protection Plan The Environmental Protection Plan (EPP) is The Environmental Protection Plan (EPP) is to to provide for protection of environmental provide for protection of environmental values values during handling and storage of spent during construction and operation handling fuel and maintenance of the nuclear facility. and storage of spent fuel and maintenance of The principal objectives of the EPP are as the nuclear facility. The principal objectives of follows: the EPP are as follows:
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| (1) Verify that the facility is maintained in an (1) Verify that the plant is operated facility is environmentally acceptable manner, as maintained in an environmentally established by the FES and other NRC acceptable manner, as established by the environmental impact assessments. FES and other NRC environmental impact assessments.
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| (3) Keep NRC informed of the environmental effects of handling and storage of spent (3) Keep NRC informed of the environmental fuel and maintenance of the facility and of effects of handling and storage of spent actions taken to control those effects. fuel and maintenance of the facility construction and operation and of actions taken to control those effects.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, the proposed administrative changes to Section 1.0 will replace a reference to handling and storage of spent fuel and maintenance with a reference to construction and operation, a reference to facility is maintained with plant is operated, and a reference to handling and storage of spent fuel and maintenance of the facility with
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| HDI PNP 2023-030 Enclosure Page 86 of 97 facility construction and operation. These proposed administrative changes more accurately reflect the revised purpose of the plant in the operating condition.
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| Current Section 2.1 Proposed Section 2.1 2.1 Aquatic Issues 2.1 Aquatic Issues The need for aquatic monitoring programs The need for aquatic monitoring programs to to confirm that thermal mixing occurs as confirm that thermal mixing occurs as predicted, that chlorine releases are predicted, that chlorine releases are controlled controlled within those discharge within those discharge concentrations concentrations evaluated, and that effects on evaluated, and that effects on aquatic biota aquatic biota and water quality due to facility and water quality due to plant facility operation are no greater than predicted operation are no greater than predicted Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, to reflect PNP as an operating plant, the EPP is administratively revised to replace the word Facility with Plant to be consistent with terminology for the TS proposed in this LAR.
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| Current Section 3.1 Proposed Section 3.1 3.1 Facility Design and Operation 3.1 Plant Facility Design and Operation The licensee may make changes in facility The licensee may make changes in station design or operation or perform tests or facility design or operation or perform tests or experimentsChanges in facility design or experimentsChanges in plant facility design operation or performance of tests or or operation or performance of tests or experiments experiments A proposed change, test, or experiment A proposed change, test, or experiment shall(2) a significant change in effluents [in shall(2) a significant change in effluents or accordance with 10 CFR Part 51.5(b)(2)] power level [in accordance with 10 CFR Part 51.5(b)(2)]
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| HDI PNP 2023-030 Enclosure Page 87 of 97 Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, to reflect PNP as an operating plant, the EPP is administratively revised to replace the word Facility with Plant or Station to be consistent with terminology for the TS proposed in this LAR.
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| The proposed change to Section 3.1 to reinstate the reference to power level [in accordance with 10 CFR Part 51.5(b)(2)] reflects the operating condition of the plant. With the return to power operation of PNP, references to power level changes are now applicable. This proposed administrative change more accurately reflect the revised purpose of the plant.
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| Current Section 3.3 Proposed Section 3.3 3.3 Changes Required for Compliance with 3.3 Changes Required for Compliance with Other Environmental Regulations Other Environmental Regulations Changes in Facility design or operation and Changes in plant facility design or operation and Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, to reflect PNP as an operating plant, the EPP is administratively revised to replace the word Facility with Plant to be consistent with terminology for the TS proposed in this LAR.
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| Current Section 4.1 Proposed Section 4.1 4.1 Unusual or Important Environmental 4.1 Unusual or Important Environmental Events Events Any occurrence of an unusual or important Any occurrence of an unusual or important event that indicates or could result in event that indicates or could result in significant significant environmental impact causally environmental impact causally related to plant related to the handling and storage of spent operation the handling and storage of spent fuel and maintenance of the facility shall be fuel and maintenance of the facility shall be recorded and recorded and Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, the proposed change to Section 4.1 replaces the handling and storage of spent fuel and maintenance of the facility with plant operation to more accurately reflect the revised purpose of the facility in the operating condition. This proposed administrative change more accurately reflect the revised purpose of the plant and is consistent
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| HDI PNP 2023-030 Enclosure Page 88 of 97 with the TS terminology proposed in this LAR.
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| Current Section 5.2 Proposed Section 5.2 5.2 Records Retention 5.2 Records Retention Records and logs relative to the environmental Records and logs relative to the environmental aspects of previous plant operation and the aspects of previous plant operation and the handling and storage of spent fuel and handling and storage of spent fuel and maintenance of the facility shall be made and maintenance of the facility shall be made and retained in a manner convenient for review and retained in a manner convenient for review and inspection. These records and logs shall be inspection. These records and logs shall be made available to the NRC on request. made available to the NRC on request.
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| Records of modifications to facility structures, Records of modifications to plant facility systems and components determined to structures, systems and components potentially affect the continued protection of determined to potentially affect the continued the environment shall be retained for the life protection of the environment shall be retained of the facility. All other records, data and logs for the life of the plant facility. All other records, relating to this EPP shall be retained for five data and logs relating to this EPP shall be years or, where applicable, in accordance with retained for five years or, where applicable, in the requirements of other agencies. accordance with the requirements of other agencies.
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2),
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| Reference 3, the proposed changes to Section 5.2 removes the reference to previous plant operation and the handling and storage of spent fuel and maintenance of the facility to more accurately reflect the revised purpose of the facility in an operating condition. This proposed administrative change more accurately reflect the revised purpose of the plant in an operating condition and is consistent with the TS terminology proposed in this LAR.
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| To reflect PNP as an operating plant, the EPP is administratively revised to replace the word Facility with Plant to be consistent with terminology for the TS proposed in this LAR.
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| Current Section 5.4 Current Section 5.4 5.4 Facility Reporting Requirements 5.4 Plant Facility Reporting Requirements Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, to reflect PNP as an operating plant, the EPP is administratively revised to replace the word Facility with Plant to be consistent with terminology for the TS proposed in this LAR.
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| HDI PNP 2023-030 Enclosure Page 89 of 97 Current Section 5.4.1 Proposed Section 5.4.1 5.4.1 Routine Reports 5.4.1 Routine Reports and an assessment of the observed impacts and an assessment of the observed impacts of the facility operation on the environment of the plant facility operation on the environment (b) A list of all changes in facility design or operation, tests, and experiments (b) A list of all changes in station facility design or operation, tests, and experiments Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, to reflect PNP as an operating plant, the EPP is administratively revised to replace the word Facility with Plant or Station to be consistent with terminology for the TS proposed in this LAR.
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| Current Section 5.4.2 Proposed Section 5.4.2 5.4.2 Nonroutine Reports 5.4.2 Nonroutine Reports The report shall (a) describe, analyze, The report shall (1) describe, analyze, and and evaluate the event, including extent evaluate the event, including extent and and magnitude of the impact and facility magnitude of the impact and plant facility operating characteristics, (2) operating characteristics, (2)
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| Basis This TS is proposed for reinstatement in its entirety to that which was in effect prior to the 10 CFR 50.82(a)(1) certifications to restore the PNP power operations RFOL. Upon recission of the 10 CFR 50.82(a)(1) certifications, as conditioned by the exemption to 10 CFR 50.82(a)(2), Reference 3, to reflect PNP as an operating plant, the EPP is administratively revised to replace the word Facility with Plant to be consistent with terminology for the TS proposed in this LAR.
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| 3.2.4 Proposed Changes to the PNP Technical Specification Bases Mark-ups of the TS Bases are provided in Attachment 3 for information only. Upon approval of this amendment, changes to the TS Bases will be incorporated in accordance with TS 5.5.12, Technical Specifications (TS) Bases Control Program.
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| ==4.0 REGULATORY EVALUATION==
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| 4.1 Applicable Regulatory Requirements 10 CFR 50.36, Technical Specifications In accordance with 10 CFR50.36, TS are required to include items in the following five
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| HDI PNP 2023-030 Enclosure Page 90 of 97 categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2)
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| LCOs; (3) SRs; (4) design features; and (5) administrative controls. However, the rule does not specify the particular requirements to be included in a plants TS.
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| This proposed amendment reinstates the portions of the previous PNP TS that are applicable to a power operation plant. It proposes changes to every section of the TS listed in 10 CFR 50.36(c)(1) through 10 CFR 50.36(c)(5). The safety limits and limiting safety system settings are reinstated in Section 2.0 as described previously. The LCOs and SRs are reinstated to support plant power operation. The four criteria listed in 10 CFR 50.36(c)(2) are used to determine which structures, systems and components are required to be in the TS. As described in the previous sections, each TS in Sections 3.1 through 3.9 list the 10 CFR 50.36(c)(2) criterion that applies to a particular TS. The design features are revised to support the receipt and handling of new fuel for power operations. The administrative controls are modified and added to support the necessary programs and reports for an operating reactor.
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| Therefore, the proposed PNP TS meet the criteria of 10 CFR 50.36 for an operating reactor.
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| 10 CFR 50.36b, Environmental Conditions 10 CFR 50.36b states that TS may include conditions to protect the environment during operation and decommissioning. These conditions are set out in an attachment to the license and are derived from information contained in the environmental report or the supplement to the environmental report. Obligations in the environmental area, including, as appropriate, requirements for reporting and keeping records of environmental data, and any conditions and monitoring requirement for the protection of the nonaquatic environment are included. The proposed modifications to the EPP are administrative and do not change the environmental obligations required by 10 CFR 50.36b.
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| 10 CFR 50.48(a) and (c), Fire Protection These regulations establish the fire protection requirements for operating power reactors.
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| 10 CFR 50.48(a) states: Each holder of an operating license issued under this partmust have a fire protection plan that satisfies Criterion 3 of appendix A to this part. 10 CFR 50.48(c) provides the requirements for use of National Fire Protection Association Standard NFPA 805 in creating a satisfactory fire protection plan. The reinstatement of License Condition 2.C.(3) addresses PNP compliance with these regulations.
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| 10 CFR 50.51, Continuation of License 10 CFR 50.51 states, in part: (a) Each license will be issued for a fixed period of time to be specified in the license. PNP complies with this regulation by the reinstatement of the RFOL expiration date in the RFOL.
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| 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants The General Design Criteria establish minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the Commission. The PNP design and licensing basis and its relationship to the General Design Criteria are described in the UFSAR, Revision 35 and other plant-specific licensing basis documents. UFSAR Revision 35 will be reinstated. This will include reinstatement of accident analyses and the safety reclassification of SSCs required to support the PNP POLB. Changes made to the UFSAR after Revision 35 will be evaluated for retention, to the extent appropriate for an operating plant.
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| HDI PNP 2023-030 Enclosure Page 91 of 97 10 CFR50, Appendix H, Reactor Vessel Material Surveillance Program Requirements 10 CFR 50 Appendix H requires that the design of the reactor vessel surveillance capsule program and withdrawal schedule must meet the requirements in the version of ASTM Standard Practice E 185 that is current on the issue date of the American Society of Mechanical Engineers (ASME) Code to which the reactor pressure vessel (RPV) was purchased. Reinstatement of License Condition 2.J addresses these requirements.
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| 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors This section contains leakage test requirements, scheduled and acceptance criteria for tests of the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. The manner in which PNP complies with this regulation, including exemptions, is reinstated in License Condition 2.D.
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| 4.2 Precedent No nuclear power plant licensee to date has requested reauthorization of power operation after docketing the 10 CFR 50.82(a)(1) certifications and before reaching the renewed facility license expiration date. There have been instances in which a licensee submitted to the NRC, and then subsequently withdrew, a certification of an intent to cease operations under 10 CFR 50.82(a)(1)(i). In those cases, the licensee had not submitted on the docket the certification of permanent cessation of operation and permanent removal of fuel from the reactor vessel.
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| While current regulations do not specify a particular mechanism for reauthorizing operation of a nuclear power plant after both certifications are submitted on the docket and before operating license expiration, there is no statute or regulation prohibiting such action.
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| Additionally, the NRC has considered the possibility of returning a plant to power operations as mentioned in Regulatory Guide 1.184, Decommissioning of Nuclear Power Reactors (Reference 13), and SECY-20-110, Denial of Petition for Rulemaking on Criteria to Return Retired Power Reactors to Operations (Reference 14). Thus, the NRC may address such requests under the existing regulatory frameworkincluding granting exemptions, where neededon a case-by-case basis. This proposed change to the RFOL, TS, and EPP supports the regulatory framework for reauthorization of power operations at PMP Precedent for Removal of Technical Specification Table of Contents Precedence for the removal of the Table of Contents from the Technical Specifications is found in approval of a request for Southern Nuclear Company in 2021. See Reference 15.
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| 4.3 No Significant Hazards Consideration Determination In accordance with 10 CFR 50.92, Issuance of amendment, Holtec Decommissioning International (HDI) has reviewed the proposed changes and concludes that the changes do not involve a significant hazards consideration since the proposed changes satisfy the criteria in 10 CFR 50.92(c). These criteria require that operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
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| HDI PNP 2023-030 Enclosure Page 92 of 97 The proposed license amendment would revise the Renewed Facility Operating License (RFOL), the Appendix A Permanently Defueled Technical Specifications (PDTS), and the Appendix B Environmental Protection Plan (EPP). The proposed changes are consistent with resumption of power operation of the reactor and emplacement and retention of fuel into the reactor vessel. The review of the proposed changes is based on the reinstatement of the plant operating licensing basis (POLB) as it was prior to the 10 CFR 50.82(a)(1) certifications. There are no physical changes to facility design proposed or required to support this amendment, and no changes proposed to the processes or procedures that were previously used during PNP power operations.
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| The proposed changes would revise certain requirements contained within the Palisades Nuclear Plant (PNP) RFOL, PDTS and EPP to add or revise license conditions or specifications that are necessary for power operation and revise or remove license conditions or specifications that would no longer be applicable. The proposed changes to the PNP RFOL and PDTS are in accordance with 10 CFR 50.36(c)(1) through 10 CFR 50.36(c)(5). The proposed changes to the PNP EPP are in accordance with 10 CFR 50.36b(b).
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| The discussion below addresses each 10 CFR 50.92(c) no significant hazards consideration criterion and demonstrates that the proposed amendment does not constitute a significant hazard.
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| : 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
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| Response: No.
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| The proposed changes to the PNP RFOL, PDTS and EPP do not impact the design basis function of plant structures, systems, or components (SSC). The proposed changes do not affect accident initiators or precursors, nor do they alter design assumptions that could increase the probability or consequences of previously evaluated accidents.
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| Chapter 14 of the PNP Updated Final Safety Analysis Report (UFSAR) Revision 35 (ADAMS Accession No. ML21125A285) describes the postulated design basis accidents (DBA) and transient scenarios applicable to PNP during power operations. The UFSAR will be reinstated to reflect the docketed version (Revision 35) that was in effect prior to docketing the 10 CFR 50.82(a) certifications of permanent cessation of power operations and permanent removal of fuel at PNP. This will include restoration of the UFSAR Revision 35 which includes previously evaluated accident analyses and safety classification of SSCs to support power operations at PNP. The proposed changes to the PDTS simply revise and/or add license conditions or specifications applicable to the PNP POLB as previously evaluated in UFSAR Revision 35. The proposed changes do not involve physical changes to the facility or in the procedures governing operation of the plant that were in effect prior to 10 CFR 50.82(a)(1) certifications.
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| The proposed addition / revision to TS definitions and rules of usage and application are those applicable to the reinstated PNP power operations technical specifications (TS) and have no impact on plant SSCs or the methods of operation of such SSCs.
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| HDI PNP 2023-030 Enclosure Page 93 of 97 The proposed reinstatement of PNP safety limits (SLs) and SL violations contain SLs that are necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity from the reactor core and the Primary Coolant System (PCS) pursuant to 10 CFR 50.36(c)(1). Since the proposed SLs are applicable to the power operations at PNP and provide protection of physical barriers to prevent uncontrolled radioactive release, they would not increase the probability or consequences of previously evaluated DBAs.
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| The reinstatement of TS Limiting Conditions for Operation (LCO) and Surveillance Requirements (SR) that are related to the operation of the nuclear reactor or to the prevention, diagnosis, or mitigation of reactor-related transients or accidents do not affect the applicable DBAs previously evaluated in the reinstated UFSAR. The safety functions involving core reactivity control, reactor heat removal, primary coolant system inventory control, and containment integrity are applicable at PNP as a power operation plant.
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| The proposed reinstatement of PNP design features contain features of the plant such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety pursuant to 10 CFR 50.36(c)(4).
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| Since the proposed design features are applicable to the power operations at PNP and provide protection of important design features, they would not increase the probability or consequences of previously evaluated DBAs.
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| The addition and modification of provisions of the administrative controls of the PDTS and the non-radiological environmental protection requirements in the EPP do not affect any accidents applicable during power operation of the plant.
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| The probability of occurrence of previously evaluated accidents in the UFSAR is not increased since reinstatement of the previously approved licensing basis, including the RFOL, PDTS and EPP, is bounded by the reinstated analyses.
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| Additionally, the proposed changes do not impact the function of plant structures, systems, or components.
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| Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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| : 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
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| Response: No.
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| The proposed changes to the PNP RFOL, the PDTS, and the EPP have no impact on plant structures, systems or components. The proposed changes do not involve installation of new equipment or modification of existing equipment that could create the possibility of a new or different kind of accident. Hence, the proposed changes do not result in a change to the way the facility or equipment is operated in a manner which could cause a new or different kind of accident initiator to be created.
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| The addition of TS that are related only to the operation of the nuclear reactor or only to the prevention, diagnosis, or mitigation of reactor-related transients or accidents, cannot result in different or more adverse failure modes or accidents than
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| HDI PNP 2023-030 Enclosure Page 94 of 97 previously evaluated because the plant will be operated within the previously approved POLB.
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| Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.
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| : 3. Does the proposed amendment involve a significant reduction in a margin of safety?
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| Response: No.
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| Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed amendment would modify the PNP RFOL and PDTS by adding the portions of the RFOL and TS that are credited in the accident analyses for the DBAs in the reinstated UFSAR. Postulated DBAs involving reactor operation are applicable because the plant will be in a power operation condition. These proposed changes impact operation of the facility and its response to transients or DBAs by reinstating requirements for equipment that is related to the operation of the nuclear reactor or to the prevention, diagnosis, or mitigation of reactor-related transients or accidents. The changes ensure that equipment required to respond to DBAs and transients described in the UFSAR remain capable of performing their safety function. No accident analyses or safety analyses acceptance criteria will be affected by the proposed changes.
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| Therefore, the proposed amendment does not involve a significant reduction in the margin of safety.
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| Based on the above, HDI concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of no significant hazards consideration is justified.
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| 4.4 Conclusion Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
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| 5.0 ENVIRONMENTAL EVALUATION This amendment request meets the eligibility criteria for categorical exclusion from environmental review set forth in 10 CFR 51.22, Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review, paragraph (c)(9). In support of this conclusion, as described in Reference 3, an independent environmental review of potentially new and significant information, and environmental issues not addressed in the October 2006 Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Supplement 27, Regarding Palisades Nuclear Plant was performed. The review concluded that the proposed
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| HDI PNP 2023-030 Enclosure Page 95 of 97 licensing actions environmental impacts are consistent with the findings in the PNP RFOL Supplemental Environmental Impact Statement (NUREG 1427, Supplement 27), and hence the NRC staff recommendation to the Commission is applicable to this activity. The 10 CFR 51.22(c)(9) criteria are met as follows:
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| (i) The amendment involves no significant hazard consideration.
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| As described in Section 4.3 of this evaluation, the proposed amendment involves no significant hazards consideration. There are no changes to the design configuration or operation of the plant as constructed. There are no relaxations in the criteria used to establish safety limits or safety system settings or TS LCOs that were in effect prior to the 10 CFR 50.82(a)(1) certifications.
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| (ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.
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| There are no design configuration or operational changes proposed or required to support the reinstatement of the POLB that would change the type or amount of any effluents previously considered in the provisional, full-term, or renewed facility operating license environmental impact statements that considered power operations impacts through March 24, 2031. Reference 3 provides additional information. There are no expected changes in the types, characteristics, or quantities of effluents discharged to the environment associated with the proposed license amendment. The license amendment will not cause any materials or chemicals to be introduced into the plant that could affect the characteristics or types of effluents released offsite. Resumed power operations will be conducted under existing environmental permits. In addition, the method of operation of waste processing systems will not be affected by the proposed license amendment. The proposed license amendment will not result in changes to the design basis requirements of SSCs that function to limit or monitor the release of effluents. All the SSCs associated with limiting the release of effluents will continue to be able to perform the necessary functions.
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| (iii) There is no significant increase in individual or cumulative occupational radiation exposure.
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| There are no design configuration or operational changes proposed or required to support reinstatement of the POLB that would change the cumulative public or occupational radiation exposure than previously considered in the provisional, full-term, or renewed facility operating license environmental impact statements that considered power operations impacts through March 24, 2031. Reference 3 provides additional information. Plant programs and processes to support an operating plant will be reinstated to ensure 10 CFR 20 limits are not exceeded for individual or cumulative occupational exposure. Since the proposed license amendment does not involve any physical change to the facility or in the procedures governing operation of the plant, the proposed license amendment does not involve a significant increase in individual or cumulative public or occupational radiation exposure.
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| Based on the above, HDI concludes that the proposed amendment meets the eligibility criteria for categorical exclusion as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
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| HDI PNP 2023-030 Enclosure Page 96 of 97
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| ==6.0 REFERENCES==
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| : 1. Entergy Nuclear Operations, Inc. letter to U. S. Nuclear Regulatory Commission, Supplement to Certification of Permanent Cessation of Power Operations, dated October 19, 2017 (ADAMS Accession No. ML17292A032)
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| : 2. Entergy Nuclear Operations, Inc. letter to U. S. Nuclear Regulatory Commission, Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel. Dated June 13, 2022 (ADAMS Accession No. ML22164A067)
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| : 3. Holtec Decommissioning International letter to U. S. Nuclear Regulatory Commission, Request for Exemption from Certain Termination of License Requirements of 10 CFR 50.82, dated September 28, 2023 (ADAMS Accession No. ML23271A140)
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| : 4. Entergy Nuclear Operations, Inc. letter to U. S. Nuclear Regulatory Commission, License Amendment Request to Revise Renewed Facility Operating License and Technical Specifications for Permanently Defueled Condition, dated June 1, 2021 (ADAMS Accession No. ML21152A108)
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| : 5. Entergy Nuclear Operations, Inc. letter to U.S. Nuclear Regulatory Commission, License Amendment Request - Administrative Controls for a Permanently Defueled Condition, dated July 27, 2017 (ADAMS Accession No. ML17208A428)
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| : 6. Entergy Nuclear Operations, Inc. letter to U.S. Nuclear Regulatory Commission, Final Safety Analysis Report Update - Revision 35, dated April 14, 2021, (ADAMS Accession No. ML21125A285)
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| : 7. U.S. Nuclear Regulatory Commission letter to Entergy Nuclear Operations, Inc.,
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| Palisades Nuclear Plant and Big Rock Point Plant - Issuance of Amendment Nos. 129 and 273 re: Order Approving Transfer of Licenses and Conforming Administrative License Amendments, dated June 28, 2022 (ADAMS Accession No. ML22173A173)
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| : 8. U. S. Nuclear Regulatory Commission letter to Entergy Nuclear Operations, Inc.,
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| Issuance of Amendment No. 272 re: Permanently Defueled Technical Specifications, dated May 13, 2022 (ADAMS Accession No. ML22039A198)
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| : 9. U. S. Nuclear Regulatory Commission letter to Entergy Nuclear Operations, Inc.,
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| Palisades Plant - Issuance of Amendment Re: Alternative Radiological Source Term, dated September 28, 2007 (ADAMS Accession No. ML072470676)
| |
| : 10. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 0, dated July 2000.
| |
| : 11. Holtec Decommissioning International letter to U. S. Nuclear Regulatory Commission, Application for Order Consenting to Transfer of Control of License and Approving Conforming License Amendments, dated December 6, 2023 (ADAMS Accession Nos. ML23340A161, ML23340A162)
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| | |
| HDI PNP 2023-030 Enclosure Page 97 of 97
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| : 12. U. S. Nuclear Regulatory Commission letter to Entergy Nuclear Operations, Inc.,
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| Palisades Plant - Issuance of Amendment No. 271 Regarding Adoption of TSTF-425, Relocate Surveillance Frequencies to License Control - RITSTF Initiative 5B, dated December 30, 2019 (ADAMS Accession No. ML19317D855)
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| : 13. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.184, Decommissioning of Nuclear Power Reactors, Revision 1, dated October 4, 2013 (ADAMS Accession No. ML13144A840)
| |
| : 14. U. S. Nuclear Regulatory Commission, SECY-20-110: Enclosure 1 - Federal Register Notice - Denial of Petition for Rulemaking on Criteria to Return Retired Nuclear Power Reactors to Operations (PRM 50-117; NRC 2019-0063), dated December 7, 2020 (ADAMS Accession No. ML20205L307)
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| : 15. U. S. Nuclear Regulatory Commission letter to Southern Nuclear Operating Co., Inc, Issuance of Amendments Regarding Removal of Table of Contents from the Technical Specifications, dated September 29, 2021 (ADAMS Accession No. ML21232A149) 7.0 ATTACHMENTS
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| : 1. Proposed Changes (mark-up) to Palisades Plant Renewed Facility Operating License DPR-20, Appendix A Permanently Defueled Technical Specifications, and Appendix B Environmental Protection Plan Pages
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| : 2. Page Change Instructions and Retyped Pages for the Palisades Plant Renewed Facility Operating License DPR-20, Appendix A Technical Specifications, and Appendix B Environmental Protection Plan
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| : 3. Proposed Technical Specifications Bases Changes (for information only)
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| | |
| Enclosure Attachment 1 to HDI PNP 2023-030 Proposed Changes (mark-up) to Palisades Plant Renewed Facility Operating License DPR-20, Appendix A Permanently Defueled Technical Specifications, and Appendix B Environmental Protection Plan Pages Note, references to "HDI" are replaced by bracketed Palisades Energy, LLC, or Palisades Energy (e.g. [Palisades Energy]) to reflect the change in operating authority per license transfer application conforming amendments.
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| 86 pages follow
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| | |
| 1 HOLTEC PALISADES, LLC HOLTEC DECOMMISSIONING INTERNATIONAL, LLC[PALISADES ENERGY, LLC]
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| DOCKET NO. 50-255 PALISADES NUCLEAR PLANT RENEWED FACILITY OPERATING LICENSE Renewed License No. DPR-20
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| : 1. The Nuclear Regulatory Commission (NRC or the Commission) having previously made the findings set forth in Operating License No. DPR-20, dated February 21, 1991, has now found that:
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| A. The application for Renewed Operating License No. DPR-20 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
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| and the Commissions rules and regulations set forth in 10 CFR Chapter I, and all required notifications to other agencies or bodies have been duly made; B. [deleted];
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| C. Actions have been identified and have been or will be taken with respect to:
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| (1) managing the effects of aging on the functionality of structures and components that have been identified to require review under 10 CFR 54.21(a)(1) during the period of extended operation, and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21(c), such that there is reasonable assurance that the activities authorized by this renewed operating license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3 for the facility, and that any changes made to the facilitys current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commissions regulations; D. The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; E. There is reasonable assurance: (i) that the activities authorized by this renewed operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations set forth in 10 CFR Chapter I; F. Holtec Palisades, LLC (Holtec Palisades) is financially qualified and [Palisades Energy, LLC (Palisades Energy)]Holtec Decommissioning International, LLC is financially and technically qualified to engage in the activities authorized by this renewed operating license in accordance with the Commissions regulations set forth Renewed License No. DPR-20 Amendment No. XXX
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| | |
| 2 in 10 CFR Chapter I; G. Holtec Palisades and HDI[Palisades Energy] have satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements" of the Commissions regulations; H. The issuance of this renewed operating license will not be inimical to the common defense and security or to the health and safety of the public; I. After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of this renewed Facility Operating License No. DPR-20, subject to the conditions for protection of the environment set forth herein, is in accordance with 10 CFR Part 51 (formerly Appendix D to Part 50), of the Commissions regulations and all applicable requirements have been satisfied; and J. The receipt, possession, and use of source, byproduct, and special nuclear material as authorized by this renewed operating license will be in accordance with 10 CFR Parts 30, 40, and 70.
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| : 2. Renewed Facility Operating License No. DPR-20 is hereby issued to Holtec Palisades and HDI[Palisades Energy] as follows:
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| A. This renewed license applies to the Palisades Plant, a pressurized light water moderated and cooled reactor and electrical generating equipment (the facility). The facility is located in Van Buren County, Michigan, and is described in the Palisades Plant Updated Final Safety Analysis Report, as supplemented and amended, and in the Palisades Plant Environmental Report, as supplemented and amended.
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| B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:
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| (1) Pursuant to Section 104b of the Act, as amended, and 10 CFR Part 50, Licensing of Production and Utilization Facilities, (a) Holtec Palisades to possess and use, and (b) HDI[Palisades Energy] to possess, and use, and operate, the facility as a utilization facility at the designated location in Van Buren County, Michigan, in accordance with the procedures and limitation set forth in this license; (2) HDI[Palisades Energy], pursuant to the Act and 10 CFR Parts 40 and 70, to receive, possess and use source, and special nuclear material that was used as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3) HDI[Palisades Energy], pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use byproduct, source, and special nuclear material as sealed sources that were used for reactor startup, sealed sources that were used for reactor instrumentation, and are used in the calibration of radiation monitoring equipment calibration, and that were used as fission detectors in amounts as Renewed License No. DPR-20 Amendment No. XXX
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| | |
| 3 required; (4) HDI [Palisades Energy], pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material for sample analysis or instrument calibration, or associated with radioactive apparatus or components; and (5) HDI[Palisades Energy], pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials that were as may be produced by the operations of the facility.
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| C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act; to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
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| (1) [deleted][Palisades Energy] is authorized to operate the facility at steady state reactor core power levels not in excess of 2565.4 Megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.
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| (2) The Technical Specifications contained in Appendix A, as revised through Amendment No. XXX273, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. [Palisades Energy]HDI shall maintain operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
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| (3) [deleted] Fire Protection
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| [Palisades Energy] shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment requests dated December 12, 2012, November 1, 2017, November 1, 2018, and March 8, 2019, as supplemented by letters dated February 21, 2013, September 30, 2013, October 24, 2013, December 2, 2013, April 2, 2014, May 7, 2014, June 17, 2014, August 14, 2014, November 4, 2014, December 18, 2014, and January 24, 2018, and May 28, 2019, as approved in the safety evaluations dated February 27, 2015, February 27, 2018, and August 20, 2019. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
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| Renewed License No. DPR-20 Amendment No. XXX
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| 4 (a) Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.
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| : 1. Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
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| : 2. Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1x10-7/year (yr) for CDF and less than 1x10-8/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
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| (b) Other Changes that May Be Made Without Prior NRC Approval
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| : 1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard.
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| The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.
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| Renewed License No. DPR-20 Amendment No. XXX
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| 5 The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is adequate for the hazard. Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:
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| Fire Alarm and Detection Systems (Section 3.8);
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| Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);
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| Gaseous Fire Suppression Systems (Section 3.10); and Passive Fire Protection Features (Section 3.11).
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| This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.
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| : 2. Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation dated February 27, 2015, to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.
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| (c) Transition License Conditions
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| : 1. Before achieving full compliance with 10 CFR 50.48(c), as specified by 2, below, risk-informed changes to the licensees fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2. above.
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| Renewed License No. DPR-20 Amendment No. XXX
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| 6
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| : 2. The licensee shall implement the modifications to its facility, as described in Table S-2, Plant Modifications Committed, of Entergy Nuclear Operations, Inc. (ENO) letter PNP 2019-028 dated May 28, 2019, to complete the transition to full compliance with 10 CFR 50.48(c) before the end of the refueling outage following the fourth full operating cycle after NRC approval. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.
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| : 3. The licensee shall implement the items listed in Table S-3, Implementation Items, of ENO letter PNP 2014-097 dated November 4, 2014, within six months after NRC approval, or six months after a refueling outage if in progress at the time of approval with the exception of Implementation Items 3 and 8 which will be completed once the related modifications are installed and validated in the PRA model.
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| (4) [deleted]
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| (5) Movement of a fuel cask in or over the spent fuel pool is prohibited when irradiated fuel assemblies decayed less than 90 days are in the spent fuel pool.[deleted]
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| (6) Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:
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| : a. Fire fighting response strategy with the following elements:
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| : 1. Pre-defined coordinated fire response strategy and guidance
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| : 2. Assessment of mutual aid fire fighting assets
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| : 3. Designated staging areas for equipment and materials
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| : 4. Command and control
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| : 5. Training of response personnel
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| : b. Operations to mitigate fuel damage considering the following:
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| : 1. Protection and use of personnel assets
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| : 2. Communications
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| : 3. Minimizing fire spread
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| : 4. Procedures for implementing integrated fire response strategy
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| : 5. Identification of readily-available pre-staged equipment
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| : 6. Training on integrated fire response strategy
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| : 7. Spent fuel pool mitigation measures
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| : c. Actions to minimize release to include consideration of:
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| : 1. Water spray scrubbing
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| : 2. Dose to onsite responders (7) [deleted]
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| Renewed License No. DPR-20 Amendment No. XXX
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| 7 (8) Amendment 257 authorizes the implementation of 10 CFR 50.61a in lieu of 10 CFR 50.61.[deleted]
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| D. The facility has been granted certain exemptions from Appendix J to 10 CFR Part 50, Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors.
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| This section contains leakage test requirements, scheduled and acceptance criteria for tests of the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. These exemptions were granted in a {{letter dated|date=December 6, 1989|text=letter dated December 6, 1989}}.
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| These exemptions granted pursuant to 10 CFR 50.12, are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. With these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.[deleted]
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| E. HDI[Palisades Energy] shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Palisades Nuclear Plant Physical Security Plan.
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| HDI[Palisades Energy] shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Palisades CSP was approved by License Amendment No. 243 as supplemented by changes approved by License Amendment Nos. 248, 253, 259, and 264.
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| F. [deleted]
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| G. Holtec Palisades and HDI[Palisades Energy] shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.
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| H. [deleted]
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| I. [deleted]
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| J. All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of American Society for Testing and Materials (ASTM) E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal scheduled, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage Renewed License No. DPR-20 Amendment No. XXX
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| | |
| 8 must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC, as required by 10 CFR Part 50, Appendix H.[deleted]
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| K. This license is effective as of the date of issuance and shall expire at midnight March 24, 2031until the Commission notifies the licensee in writing that the license is terminated.
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| FOR THE NUCLEAR REGULATORY COMMISSION
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| /RA/
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| J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:
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| : 1. Appendix A - Permanently Defueled Technical Specifications
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| : 2. Appendix B - Environmental Protection Plan Date of Issuance: January 17, 2007 Renewed License No. DPR-20 Amendment No. XXX
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| | |
| PALISADES PLANT RENEWED FACILITY OPERATING LICENSE DPR-20 APPENDIX A PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS As Amended through Amendment No. 273
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| | |
| Definitions 1.1 1.1 DEFINITIONS 1.0 USE AND APPLICATION 1.1 Definitions
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| -----------------------------------------------------------NOTE----------------------------------------------------------
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| The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.
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| Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.
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| AVERAGE DISINTEGRATION shall be the average (weighted in proportion to the ENERGY - concentration of each radionuclide in the primary coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total noniodine activity in the coolant.
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| AXIAL OFFSET (AO) AO shall be the power generated in the lower half of the core less the power generated in the upper half of the core, divided by the sum of the power generated in the lower and upper halves of the core (determined using the incore monitoring system).
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| AXIAL SHAPE INDEX (ASI) ASI shall be the power generated in the lower half of the core less the power generated in the upper half of the core, divided by the sum of the power generated in the lower and upper halves of the core (determined using the excore monitoring system).
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| CERTIFIED FUEL HANDLER A CERTIFIED FUEL HANDLER is an individual who complies with provisions of the CERTIFIED FUEL HANDLER training and retraining program required by Specification 5.3.2.
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| Palisades Nuclear Plant 1.1-1 Amendment No. 266, 272, XXX
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| | |
| Definitions 1.1 1.1 DEFINITIONS CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST.
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| Calibration of instrument channels with Resistance Temperature Detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel.
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| Whenever a RTD or thermocouple sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.
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| The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps.
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| CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
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| CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:
| |
| : a. Analog and bistable channels - the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, of all devices in the channel required for channel OPERABILITY; Palisades Nuclear Plant 1.1-3 Amendment No. XXX
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| | |
| Definitions 1.1 1.1 DEFINITIONS CHANNEL FUNCTIONAL TEST b. Digital channels - the use of diagnostic programs to test (continued) digital hardware and the injection of simulated process data into the channel to verify OPERABILITY, of all devices in the channel required for channel OPERABILITY.
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| The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested.
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| CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or control rods within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
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| CORE OPERATING LIMITS The COLR is the plant specific document that provides cycle REPORT (COLR) specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.
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| DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The dose conversion factors used for this calculation shall be those listed in Federal Guidance Report 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion, 1989; (Table 2.1, Exposure-to-Dose Conversion Factors for Inhalation).
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| INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).
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| Palisades Nuclear Plant 1.1-3 Amendment No. XXX
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| | |
| Definitions 1.1 1.1 DEFINITIONS LEAKAGE LEAKAGE shall be:
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| : a. Identified LEAKAGE
| |
| : 1. LEAKAGE, such as that from pump seals or valve packing (except Primary Coolant Pump seal water leakoff), that is captured and conducted to collection systems or a sump or collecting tank;
| |
| : 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known not to interfere with the operation of leakage detection systems and not to be pressure boundary LEAKAGE; and
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| : 3. Primary Coolant System (PCS) LEAKAGE through a Steam Generator to the Secondary System (primary to secondary LEAKAGE).
| |
| : b. Unidentified LEAKAGE All LEAKAGE (except Primary Coolant Pump seal leakoff) that is not identified LEAKAGE;
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| : c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in a PCS component body, pipe wall, or vessel wall.
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| MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average primary coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
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| NON-CERTIFIED OPERATOR A NON-CERTIFIED OPERATOR is a non-licensed operator who complies with the qualification requirements of Specification 5.3.1.
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| Palisades Nuclear Plant 1.1-4 Amendment No. 266, 272, XXX
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| | |
| Definitions 1.1 1.1 DEFINITIONS OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
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| PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:
| |
| : a. Described in Chapter 13, Initial Tests and Operation, of the FSAR;
| |
| : b. Authorized under the provisions of 10 CFR 50.59; or
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| : c. Otherwise approved by the Nuclear Regulatory Commission.
| |
| QUADRANT POWER TILT Tq shall be the maximum positive ratio of the power (Tq) generated in any quadrant minus the average quadrant power, to the average quadrant power.
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| RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) primary coolant of 2565.4 MWt.
| |
| REFUELING BORON REFUELING BORON CONCENTRATION shall be a Primary CONCENTRATION Coolant System boron concentration of 1720 ppm and sufficient to assure the reactor is subcritical by 5% with all control rods withdrawn.
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| Palisades Nuclear Plant 1.1-5 Amendment No. XXX
| |
| | |
| Definitions 1.1 1.1 DEFINITIONS SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
| |
| : a. All full length control rods (shutdown and regulating) are fully inserted except for the single rod of highest reactivity worth, which is assumed to be fully withdrawn.
| |
| However, with all full length control rods verified fully inserted by two independent means, it is not necessary to account for a stuck rod in the SDM calculation. With any full length control rods not capable of being fully inserted, the reactivity worth of these rods must be accounted for in the determination of SDM; and
| |
| : b. There is no change in part length rod position STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
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| THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the primary coolant.
| |
| TOTAL RADIAL FRT shall be the maximum ratio of the individual PEAKING FACTOR fuel pin power to the core average pin power integrated over (FRT) the total core height, including tilt.
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| Palisades Nuclear Plant 1.1-6 Amendment No. XXX
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| | |
| Definitions 1.1 1.1 DEFINITIONS Table 1.1-1 (page 1 of 1)
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| MODES
| |
| % RATED AVERAGE PRIMARY REACTIVITY THERMAL MODE TITLE COOLANT CONDITION (a)
| |
| POWER TEMPERATURE (keff)
| |
| (F) 1 Power Operation 0.99 >5 NA 2 Startup 0.99 5 NA 3 Hot Standby < 0.99 NA 300 4 Hot Shutdown(b) < 0.99 NA 300 > Tave > 200 5 Cold Shutdown(b) < 0.99 NA 200 6 Refueling(c) NA NA NA (a) Excluding decay heat.
| |
| (b) All reactor vessel head closure bolts fully tensioned.
| |
| (c) One or more reactor vessel head closure bolts less than fully tensioned.
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| Palisades Nuclear Plant 1.1-7 Amendment No. XXX
| |
| | |
| Logical Connectors 1.2 1.0 USE AND APPLICATION 1.2 Logical Connectors PURPOSE The purpose of this section is to explain the meaning of logical connectors.
| |
| Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies. The only logical connectors that appears in TS is are AND and OR. The physical arrangement of this these connectors constitutes logical conventions with specific meanings.
| |
| BACKGROUND Several Levels levels of logic may be used to state Required Actions.
| |
| These levels are identified by the placement (or nesting) of the logical connectors and by the number assigned to each Required Action. The first level of logic is identified by the first digit of the number assigned to a Required Action and the placement of the logical connector in the first level of nesting (i.e., left justified with the number of the Required Action).
| |
| The successive levels of logic are identified by additional digits of the Required Action number and by successive indentions of the logical connectors.
| |
| When logical connectors are used to state a Condition, Completion Time, Surveillance, or Frequency, only the first level of logic is used, and the logical connector is left justified with the statement of the Condition, Completion Time, Surveillance, or Frequency.
| |
| Palisades Nuclear Plant 1.2-1 Amendment No. 189, 272, XXX
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| | |
| Logical Connectors 1.2 1.2 Logical Connectors EXAMPLES The following examples illustrates the use of logical connectors.
| |
| EXAMPLE 1.2-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Suspend Verify . . .
| |
| AND A.2 Initiate Restore . . .
| |
| In this example the logical connector AND is used to indicate that when in Condition A, both Required Actions A.1 and A.2 must be completed.
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| Palisades Nuclear Plant 1.2-2 Amendment No. 189, 272, XXX
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| | |
| Logical Connectors 1.2 1.2 Logical Connectors EXAMPLES EXAMPLE 1.2-2 (continued)
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Trip . . .
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| OR A.2.1 Verify . . .
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| AND A.2.2.1 Reduce . . .
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| OR A.2.2.2 Perform . . .
| |
| OR A.3 Align . . .
| |
| This example represents a more complicated use of logical connectors.
| |
| Required Actions A.1, A.2, and A.3 are alternative choices, only one of which must be performed as indicated by the use of the logical connector OR and the left justified placement. Any one of these three Actions may be chosen.
| |
| If A.2 is chosen, then both A.2.1 and A.2.2 must be performed as indicated by the logical connector AND. Required Action A.2.2 is met by performing A.2.2.1 or A.2.2.2. The indented position of the logical connector OR indicates that A.2.2.1 and A.2.2.2 are alternative choices, only one of which must be performed.
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| Palisades Nuclear Plant 1.2-3 Amendment No. XXX
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| | |
| Completion Times 1.3 1.3 Completion Times 1.0 USE AND APPLICATION 1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.
| |
| BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe storage and handling of spent nuclear fueloperation of the plant. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s).
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| DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the discovery of a situation (e.g.,
| |
| inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the facility plant is in a MODE or specified condition stated in the Applicability of the LCO.
| |
| Unless otherwise specified, The the Completion Time begins when a Certified Fuel Handler (CFH)senior licensed operator on the operating shift crew with responsibility for plant operations makes the determination that an LCO is not met and an ACTIONS Condition is entered. The "otherwise specified" exceptions are varied, such as a Required Action Note or Surveillance Requirement Note that provides an alternative time to perform specific tasks, such as testing, without starting the Completion Time. While utilizing the Note, should a Condition be applicable for any reason not addressed by the Note, the Completion Time begins. Should the time allowance in the Note be exceeded, the Completion Time begins at that point. The exceptions may also be incorporated into the Completion Time. For example, LCO 3.8.1, "AC Sources - Operating,"
| |
| Required Action B.2, requires declaring required feature(s) supported by an inoperable diesel generator, inoperable when the redundant required feature(s) are inoperable. The Completion Time states, "4 hours from discovery of Condition B concurrent with inoperability of redundant required feature(s)." In this case the Completion Time does not begin until the conditions in the Completion Time are satisfied.
| |
| Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the plant is not within the LCO Applicability.
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| Palisades Nuclear Plant 1.3-1 Amendment No. 270, 272, XXX
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| | |
| Completion Times 1.3 1.3 Completion Times If situations are discovered that require entry into more than one Condition at a time within a single LCO (multiple Conditions), the Required Actions for each Condition must be performed within the associated Completion Time. When in multiple Conditions, separate Completion Times are tracked for each Condition starting from the discovery of the situation that required entry into the Condition, unless otherwise specified.
| |
| Once a Condition has been entered, subsequent trains, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition, unless specifically stated. The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition, unless otherwise specified.
| |
| However, when a subsequent train, subsystem, component, or variable expressed in the Condition is discovered to be inoperable or not within limits, the Completion Time(s) may be extended. To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability:
| |
| : a. Must exist concurrent with the first inoperability; and
| |
| : b. Must remain inoperable or not within limits after the first inoperability is resolved.
| |
| The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either:
| |
| : a. The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours; or
| |
| : b. The stated Completion Time as measured from discovery of the subsequent inoperability.
| |
| The above Completion Time extensions do not apply to those Specifications that have exceptions that allow completely separate re-entry into the Condition (for each train, subsystem, component, or variable expressed in the Condition) and separate tracking of Completion Times based on this re-entry. These exceptions are stated in individual Specifications.
| |
| The above Completion Time extension does not apply to a Completion Time with a modified "time zero." This modified "time zero" may be expressed as a repetitive time (i.e., "once per 8 hours," where the Palisades Nuclear Plant 1.3-1 Amendment No. 270, 272, XXX
| |
| | |
| Completion Times 1.3 1.3 Completion Times Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase "from discovery . . ." Example 1.3-3 illustrates one use of this type of Completion Time. The 10 day Completion Time specified for Conditions A and B in Example 1.3-3 may not be extended.
| |
| EXAMPLES The following examples illustrates the use of Completion Times with different types of Conditions and changing ConditionsRequired Actions.
| |
| EXAMPLE 1.3-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME AB. Required AB.1 Be in Immediately6 hours Action and MODE 3Suspend associated movement of fuel Completion assemblies in the Time not Spent Fuel Pool. Immediately36 hours metSpent Fuel Pool AND boron concentration AB.2 Be in not within MODE 5.Initiate limit. action to restore Spent Fuel Pool boron concentration to within limit Condition A B has two Required Actions. Each Required Action has its own separate Completion Time. Each Completion Time is referenced to the time that Condition A B is entered.
| |
| The Required Actions of Condition A B are to be in MODE 3 within 6 hours AND in MODE 5 within 36 hours. A total of 6 hours is allowed for reaching MODE 3 and a total of 36 hours (not 42 hours) is allowed for reaching MODE 5 from the time that Condition B was entered. If MODE 3 is reached within 3 hours, the time allowed for reaching MODE 5 is the next 33 hours because the total time allowed for reaching MODE 5 is 36 hours.
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| Palisades Nuclear Plant 1.3-2 Amendment No. 270, 272, XXX
| |
| | |
| Completion Times 1.3 1.3 Completion Times If Condition B is entered while in MODE 3, the time allowed for reaching MODE 5 is the next 36 hours.immediately suspend movement of fuel assemblies in the Spent Fuel Pool and initiate action to restore Spent Fuel Pool boron within limit.
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| Palisades Nuclear Plant 1.3-1 Amendment No. 270, 272, XXX
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| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-2 (continued)
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pump A.1 Restore pump to 7 days inoperable. OPERABLE status.
| |
| B. Required B.1 Be in MODE 3. 6 hours Action and associated AND Completion Time not met. B.2 Be in MODE 5. 36 hours When a pump is declared inoperable, Condition A is entered. If the pump is not restored to OPERABLE status within 7 days, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable pump is restored to OPERABLE status after Condition B is entered, Condition A and B are exited, and therefore, the Required Actions of Condition B may be terminated.
| |
| When a second pump is declared inoperable while the first pump is still inoperable, Condition A is not re-entered for the second pump.
| |
| LCO 3.0.3 is entered, since the ACTIONS do not include a Condition for more than one inoperable pump. The Completion Time clock for Condition A does not stop after LCO 3.0.3 is entered, but continues to be tracked from the time Condition A was initially entered.
| |
| While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has not expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition A.
| |
| While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition B.
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| Palisades Nuclear Plant 1.3-2 Amendment No. XXX
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| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-2 (continued)
| |
| The Completion Time for Condition B is tracked from the time the Condition A Completion Time expired.
| |
| On restoring one of the pumps to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first pump was declared inoperable. This Completion Time may be extended if the pump restored to OPERABLE status was the first inoperable pump. A 24 hour extension to the stated 7 days is allowed, provided this does not result in the second pump being inoperable for > 7 days.
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| Palisades Nuclear Plant 1.3-3 Amendment No. XXX
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| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued)
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Restore Function X 7 days Function X train to OPERABLE train status. AND inoperable.
| |
| 10 days from discovery of failure to meet the LCO B. One B.1 Restore Function Y 72 hours Function Y train to OPERABLE train status. AND inoperable.
| |
| 10 days from discovery of failure to meet the LCO C. One C.1 Restore Function X 12 hours Function X train to OPERABLE train status.
| |
| inoperable.
| |
| OR AND C.2 Restore Function Y One train to OPERABLE 12 hours Function Y status.
| |
| train inoperable.
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| Palisades Nuclear Plant 1.3-4 Amendment No. XXX
| |
| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued)
| |
| When one Function X train and one Function Y train are inoperable, Condition A and Condition B are concurrently applicable. The Completion Times for Condition A and Condition B are tracked separately for each train starting from the time each train was declared inoperable and the Condition was entered. A separate Completion Time is established for Condition C and tracked from the time the second train was declared inoperable (i.e., the time the situation described in Condition C was discovered).
| |
| If Required Action C.2 is completed within the specified Completion Time, Conditions B and C are exited. If the Completion Time for Required Action A.1 has not expired, operation may continue in accordance with Condition A. The remaining Completion Time in Condition A is measured from the time the affected train was declared inoperable (i.e., initial entry into Condition A).
| |
| The Completion Times of Conditions A and B are modified by a logical connector, with a separate 10 day Completion Time measured from the time it was discovered the LCO was not met. In this example, without the separate Completion Time, it would be possible to alternate between Conditions A, B, and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO. The separate Completion Time modified by the phrase "from discovery of failure to meet the LCO" is designed to prevent indefinite continued operation while not meeting the LCO. This Completion Time allows for an exception to the normal "time zero" for beginning the Completion Time "clock." In this instance, the Completion Time "time zero" is specified as commencing at the time the LCO was initially not met, instead of at the time the associated Condition was entered.
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| Palisades Nuclear Plant 1.3-5 Amendment No. XXX
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| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-4 (continued)
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve(s) to 4 hours valves OPERABLE status.
| |
| inoperable.
| |
| B. Required B.1 Be in MODE 3. 6 hours Action and associated AND Completion Time not met. B.2 Be in MODE 4. 30 hours A single Completion Time is used for any number of valves inoperable at the same time. The Completion Time associated with Condition A is based on the initial entry into Condition A and is not tracked on a per valve basis. Declaring subsequent valves inoperable, while Condition A is still in effect, does not trigger the tracking of separate Completion Times.
| |
| Once one of the valves has been restored to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first valve was declared inoperable. The Completion Time may be extended if the valve restored to OPERABLE status was the first inoperable valve. The Condition A Completion Time may be extended for up to 4 hours provided this does not result in any subsequent valve being inoperable for > 4 hours.
| |
| If the Completion Time of 4 hours (including the extension) expires while one or more valves are still inoperable, Condition B is entered.
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| Palisades Nuclear Plant 1.3-6 Amendment No. XXX
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| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-5 (continued)
| |
| ACTIONS
| |
| --------------------------------------------NOTE--------------------------------------------
| |
| Separate Condition entry is allowed for each inoperable valve.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve to 4 hours valves OPERABLE inoperable. status.
| |
| B. Required B.1 Be in MODE 3. 6 hours Action and associated AND Completion Time not met. B.2 Be in MODE 4. 12 hours The Note above the ACTIONS Table is a method of modifying how the Completion Time is tracked. If this method of modifying how the Completion Time is tracked was applicable only to a specific Condition, the Note would appear in that Condition rather than at the top of the ACTIONS Table.
| |
| The Note allows Condition A to be entered separately for each inoperable valve, and Completion Times tracked on a per valve basis. When a valve is declared inoperable, Condition A is entered and its Completion Time starts. If subsequent valves are declared inoperable, Condition A is entered for each valve and separate Completion Times start and are tracked for each valve.
| |
| If the Completion Time associated with a valve in Condition A expires, Condition B is entered for that valve. If the Completion Times associated with subsequent valves in Condition A expire, Condition B is entered separately for each valve and separate Completion Times start and are tracked for each valve. If a valve that caused entry into Condition B is restored to OPERABLE status, Condition B is exited for that valve.
| |
| Since the Note in this example allows multiple Condition entry and tracking of separate Completion Times, Completion Time extensions do not apply.
| |
| Palisades Nuclear Plant 1.3-7 Amendment No. XXX
| |
| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-6 (continued)
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One channel A.1 Perform SR 3.x.x.x. Once per 8 hours inoperable.
| |
| OR A.2 Reduce THERMAL 8 hours POWER to 50% RTP.
| |
| B. Required B.1 Be in MODE 3. 6 hours Action and associated Completion Time not met.
| |
| Entry into Condition A offers a choice between Required Action A.1 or A.2. Required Action A.1 has a "once per" Completion Time, which qualifies for the 25% extension, per SR 3.0.2, to each performance after the initial performance. The initial 8 hour interval of Required Action A.1 begins when Condition A is entered and the initial performance of Required Action A.1 must be complete within the first 8 hour interval. If Required Action A.1 is followed and the Required Action is not met within the Completion Time (plus the extension allowed by SR 3.0.2),
| |
| Condition B is entered. If Required Action A.2 is followed and the Completion Time of 8 hours is not met, Condition B is entered.
| |
| If after entry into Condition B, Required Action A.1 or A.2 is met, Condition B is exited and operation may then continue in Condition A.
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| Palisades Nuclear Plant 1.3-8 Amendment No. XXX
| |
| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued)
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Verify affected 1 hour subsystem subsystem isolated.
| |
| inoperable. AND Once per 8 hours thereafter AND A.2 Restore subsystem 72 hours to OPERABLE status.
| |
| B. Required B.1 Be in MODE 3. 6 hours Action and associated AND Completion Time not met. B.2 Be in MODE 5. 36 hours Required Action A.1 has two Completion Times. The 1 hour Completion Time begins at the time the Condition is entered and each "Once per 8 hours thereafter" interval begins upon performance of Required Action A.1.
| |
| If after Condition A is entered, Required Action A.1 is not met within either the initial 1 hour or any subsequent 8 hour interval from the previous performance (plus the extension allowed by SR 3.0.2), Condition B is entered.
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| Palisades Nuclear Plant 1.3-9 Amendment No. XXX
| |
| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued)
| |
| The Completion Time clock for Condition A does not stop after Condition B is entered, but continues from the time Condition A was initially entered. If Required Action A.1 is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not expired.
| |
| IMMEDIATE When "Immediately" is used as a Completion Time, the Required Action COMPLETION TIME should be pursued without delay and in a controlled manner.
| |
| Palisades Nuclear Plant 1.3-10 Amendment No. 270, 272, XXX
| |
| | |
| Frequency 1.4 1.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.
| |
| DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR.
| |
| The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR)
| |
| Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR, as well as certain Notes in the Surveillance column that modify performance requirements.
| |
| Sometimes special situations dictate when the requirements of a Surveillance are to be met. They are otherwise stated conditions allowed by SR 3.0.1. They may be stated as clarifying Notes in the Surveillance, as part of the Surveillance, or both.
| |
| Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated LCO is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only required when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction.
| |
| The use of "met and performed in these instances conveys specific meanings. A Surveillance is "met only when the acceptance criteria are satisfied. Known failure of the requirements of a Surveillance, even without a Surveillance specifically being performed, constitutes a Surveillance not "met. Performance refers only to the requirement to specifically determine the ability to meet the acceptance criteria.
| |
| Palisades Nuclear Plant 1.4-1 Amendment No. 231, 272, XXX
| |
| | |
| Frequency 1.4 1.4 Frequency DESCRIPTION Some Surveillances contain notes that modify the Frequency of (continued) performance or the conditions during which the acceptance criteria must be satisfied. For these Surveillances, the MODE-entry restrictions of SR 3.0.4 may not apply. Such a Surveillance is not required to be performed prior to entering a MODE or other specified condition in the Applicability of the associated LCO if any of the following three conditions are satisfied:
| |
| : a. The Surveillance is not required to be met in the MODE or other specified condition to be entered; or
| |
| : b. The Surveillance is required to be met in the MODE or other specified condition to be entered, but has been performed within the specified Frequency (i.e., it is current) and is known not to be failed; or
| |
| : c. The Surveillance is required to be met, but not performed, in the MODE or other specified condition to be entered, and is known not to be failed.
| |
| Examples 1.4-3, 1.4-4, 1.4-5, and 1.4-6 discuss these special situations.
| |
| EXAMPLES The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LCO (LCO not shown) is MODES 1, 2, and 3.illustrate the type of Frequency statements that appears in the Technical Specifications (TS).
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| Palisades Nuclear Plant 1.4-2 Amendment No. 231, 272, XXX
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| | |
| Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-1 (continued)
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perform CHANNEL CHECKVerify level is within 7 days12 hours limit.
| |
| Example 1.4-1 contains one the type of SR most often encountered in the Technical Specifications (TS). The Frequency specifies an interval of 7 days(12 hours) during which the associated Surveillance must be performed at least one time. Performance of the Surveillance initiates the subsequent interval. Although the Frequency is stated as 7 days12 hours, an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for operational flexibility. The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment is inoperable, a variable is outside specified limits, or the facility plant is outside the Applicability of the LCO). If the interval specified by SR 3.0.2 is exceeded while the facility is in a MODE or other specified condition in the Applicability of the LCO, and the performance of the Surveillance is not otherwise modified (refer to Example 1.4-3), then SR 3.0.3 becomes applicable.
| |
| If the interval as specified by SR 3.0.2 is exceeded while the facility plant is not in a Mode or other specified condition in the Applicability of the LCO for which performance of the SR is required, then SR 3.0.4 becomes applicable. The Surveillance must be performed within the Frequency requirements of SR 3.0.2, as modified by SR 3.0.3, prior to entry into the MODE or other specified condition or the LCO is considered not met (in accordance with SR 3.0.1) and LCO 3.0.4 becomes applicable.
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| Palisades Nuclear Plant 1.4-2 Amendment No. 231, 272, XXX
| |
| | |
| Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-2 (continued)
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits.... Once within 12 hours after 25% RTPPrior to storing a fuel assembly AND 24 hours thereafter Example 1.4-2 has two Frequencies. The first isillustrates a one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency requirements must be met. Each time reactor power is increased from a power level < 25% RTP to 25% RTP, the Surveillance must be performed within 12 hours.
| |
| The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "AND").
| |
| This type of Frequency does not qualify for the extension allowed by SR 3.0.2. "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to
| |
| < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.
| |
| Palisades Nuclear Plant 1.4-4 Amendment No. 231, 272, XXX
| |
| | |
| Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued)
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
| |
| ----------------------------NOTE----------------------------
| |
| Not required to be performed until 12 hours after 25% RTP.
| |
| Perform channel adjustment. 7 days The interval continues, whether or not the plant operation is < 25% RTP between performances.
| |
| As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours after power reaches 25% RTP to perform the Surveillance.
| |
| The Surveillance is still considered to be performed within the "specified Frequency." The interval continues, whether or not the plant operation is
| |
| < 25% RTP between performances. Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours with power 25% RTP.
| |
| Once the plant reaches 25% RTP, 12 hours would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour interval, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.
| |
| Palisades Nuclear Plant 1.4-5 Amendment No. XXX
| |
| | |
| Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-4 (continued)
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
| |
| ----------------------------NOTE----------------------------
| |
| Only required to be met in MODE 1.
| |
| Verify leakage rates are within limits. 24 hours Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the plant is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an otherwise stated exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour interval (plus the extension allowed by SR 3.0.2), but the plant was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR.
| |
| Palisades Nuclear Plant 1.4-6 Amendment No. XXX
| |
| | |
| Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-5 (continued)
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
| |
| ----------------------------NOTE----------------------------
| |
| Only required to be performed in MODE 1.
| |
| Perform complete cycle of the valve. 7 days The interval continues, whether or not the plant operation is in MODE 1, 2, or 3 (the assumed Applicability of the associated LCO) between performances.
| |
| As the Note modifies the required performance of the Surveillance, the Note is construed to be part of the specified Frequency. Should the 7 day interval be exceeded while operation is not in MODE 1, this Note allows entry into and operation in MODES 2 and 3 to perform the Surveillance. The Surveillance is still considered to be performed within the specified Frequency if completed prior to entering MODE 1.
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| Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was not in MODE 1, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not result in entry into MODE 1.
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| Once the plant reaches MODE 1, the requirement for the Surveillance to be performed within its specified Frequency applies and would require that the Surveillance had been performed. If the Surveillance were not performed prior to entering MODE 1, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.
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| Palisades Nuclear Plant 1.4-7 Amendment No. XXX
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| | |
| EXAMPLES EXAMPLE 1.4-6 (continued)
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| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
| |
| ----------------------------NOTE----------------------------
| |
| Not required to be met in MODE 3.
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| Verify parameter is within limits. 24 hours Example 1.4-6 specifies that the requirements of this Surveillance do not have to be met while the plant is in MODE 3 (the assumed Applicability of the associated LCO is MODES 1, 2, and 3). The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an otherwise stated exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour interval (plus the extension allowed by SR 3.0.2), but the plant was in MODE 3, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES to enter MODE 3, even with the 24 hour Frequency exceeded, provided the MODE change does not result in entry into MODE 2. Prior to entering MODE 2 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR.
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| Palisades Nuclear Plant 1.4-8 Amendment No. XXX
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| | |
| SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.01 (Deleted)SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, the Departure from Nucleate Boiling Ratio (DNBR) shall be maintained at or above the following DNB correlation safety limits:
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| Correlation Safety Limit XNB 1.17 ANFP 1.154 HTP 1.141 2.1.1.2 In MODES 1 and 2, the peak Linear Heat Rate (LHR) (adjusted for fuel rod dynamics) shall be maintained at 21.0 kW/ft.
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| 2.1.2 Primary Coolant System (PCS) Pressure SL In MODES 1, 2, 3, 4, 5, and 6, the PCS pressure shall be maintained at 2750 psia.
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| 2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour.
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| 2.2.2 If SL 2.1.2 is violated:
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| 2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour.
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| 2.2.2.2 In MODE 3, 4, 5, or 6, restore compliance within 5 minutes.
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| Palisades Nuclear Plant 2.0-1 Amendment No. XXX
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| | |
| LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2, LCO 3.0.7, LCO 3.0.8, and LCO 3.0.9.
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| LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6.
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| If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required, unless otherwise stated.
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| LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the plant shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the plant, as applicable, in:
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| : a. MODE 3 within 7 hours;
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| : b. MODE 4 within 31 hours; and
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| : c. MODE 5 within 37 hours.
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| Exceptions to this Specification are stated in the individual Specifications.
| |
| Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required.
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| LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4.
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| LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:
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| : a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; Palisades Nuclear Plant 3.0-1 Amendment No. 270, 272, XXX
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| | |
| LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.4 (continued)
| |
| : b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate (exceptions to this Specification are stated in the individual Specifications); or
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| : c. When an allowance is stated in the individual value, parameter, or other Specification.
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| This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the plant.
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| LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.
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| LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.13, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.
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| Palisades Nuclear Plant 3.0-2 Amendment No. XXX
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| | |
| LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.7 Special Test Exception (STE) LCOs in each applicable LCO section allow specified Technical Specifications (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with STE LCOs is optional. When an STE LCO is desired to be met but is not met, the ACTIONS of the STE LCO shall be met. When an STE LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with the other applicable Specifications.
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| LCO 3.0.8 When one or more required snubbers are unable to perform their associated support function(s), any affected supported LCO(s) are not required to be declared not met solely for this reason if risk is assessed and managed, and:
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| : a. the snubbers not able to perform their associated support function(s) are associated with only one train or subsystem of a multiple train or subsystem supported system or are associated with a single train or subsystem supported system and are able to perform their associated support function within 72 hours; or
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| : b. the snubbers not able to perform their associated support function(s) are associated with more than one train or subsystem of a multiple train or subsystem supported system and are able to perform their associated support function within 12 hours.
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| At the end of the specified period the required snubbers must be able to perform their associated support function(s), or the affected supported system LCO(s) shall be declared not met.
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| LCO 3.0.9 When one or more required barriers are unable to perform their related support function(s), any supported system LCO(s) are not required to be declared not met solely for this reason for up to 30 days provided that at least one train or subsystem of the supported system is OPERABLE and supported by barriers capable of providing their related support function(s), and risk is assessed and managed. This specification may be concurrently applied to more than one train or subsystem of a multiple train or subsystem supported system provided at least one train or subsystem of the supported system is OPERABLE and the barriers supporting each of these trains or subsystems provide their related support function(s) for different categories of initiating events.
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| Palisades Nuclear Plant 3.0-3 Amendment No. XXX
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| | |
| LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.9 (continued)
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| If the required OPERABLE train or subsystem becomes inoperable while this specification is in use, it must be restored to OPERABLE status within 24 hours or the provisions of this specification cannot be applied to the trains or subsystems supported by the barriers that cannot perform their related support function(s).
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| At the end of the specified period, the required barriers must be able to perform their related support function(s) or the supported system LCO(s) shall be declared not met.
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| Palisades Nuclear Plant 3.0-4 Amendment No. XXX
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| | |
| SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR.
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| Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.
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| SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.
| |
| For Frequencies specified as "once," the above interval extension does not apply.
| |
| If a Completion Time requires periodic performance on a "once per . . ."
| |
| basis, the above Frequency extension applies to each performance after the initial performance.
| |
| Exceptions to this Specification are stated in the individual Specifications.
| |
| SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. The delay period is only applicable when there is a reasonable expectation the surveillance will be met when performed. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.
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| If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
| |
| When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
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| Palisades Nuclear Plant 3.0-5 Amendment No. 270, 272, XXX
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| | |
| SR Applicability 3.0 3.0 SR APPLICABILITY SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.
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| This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the plant.
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| Palisades Nuclear Plant 3.0-6 Amendment No. 270, 272, XXX
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| | |
| 3.1 REACTIVITY CONTROL SYSTEMS Insert LCO 3.1.1 SHUTDOWN MARGIN (SDM)
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| Insert LCO 3.1.2 Reactivity Balance Insert LCO 3.1.3 Moderator Temperature Coefficient (MTC)
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| Insert LCO 3.1.4 Control Rod Alignment Insert LCO 3.1.5 Shutdown and Part-Length Control Rod Group Insertion Limits Insert LCO 3.1.6 Regulating Rod Group Position Limits Insert LCO 3.1.7 Special Test Exceptions (STE) 3.2 POWER DISTRIBUTION LIMITS Insert LCO 3.2.1 Linear Heat Rate (LHR)
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| Insert LCO 3.2.2 TOTAL RADIAL PEAKING FACTOR (FRT)
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| Insert LCO 3.2.3 QUADRANT POWER TILT (Tq)
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| Insert LCO 3.2.4 AXIAL SHAPE INDEX (ASI) 3.3 INSTRUMENTATION Insert LCO 3.3.1 Reactor Protective System (RPS) Instrumentation Insert LCO 3.3.2 Reactor Protective System (RPS) Logic and Trip Initiation Insert LCO 3.3.3 Engineered Safety Features (ESF) Instrumentation Insert LCO 3.3.4 Engineered Safety Features (ESF) Logic and Manual Initiation Insert LCO 3.3.5 Diesel Generator (DG) - Undervoltage Start (UV Start)
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| Insert LCO 3.3.6 Refueling Containment High Radiation (CHR)
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| Instrumentation Insert LCO 3.3.7 Post Accident Monitoring (PAM) Instrumentation Insert LCO 3.3.8 Alternate Shutdown System Insert LCO 3.3.9 Neutron Flux Monitoring Channels Insert LCO 3.3.10 Engineered Safeguards Room Ventilation (ESRV)
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| Instrumentation
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| | |
| 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| Insert LCO 3.4.1 PCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits Insert LCO 3.4.2 PCS Minimum Temperature for Criticality Insert LCO 3.4.3 PCS Pressure and Temperature (P/T) Limits Insert LCO 3.4.4 PCS Loops - MODES 1 and 2 Insert LCO 3.4.5 PCS Loops - MODE 3 Insert LCO 3.4.6 PCS Loops - MODE 4 Insert LCO 3.4.7 PCS Loops - MODE 5, Loops Filled Insert LCO 3.4.8 PCS Loops - MODE 5, Loops Not Filled Insert LCO 3.4.9 Pressurizer Insert LCO 3.4.10 Pressurizer Safety Valves Insert LCO 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)
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| Insert LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System Insert LCO 3.4.13 PCS Operational LEAKAGE Insert LCO 3.4.14 PCS Pressure Isolation Valve (PIV) Leakage Insert LCO 3.4.15 PCS Leakage Detection Instrumentation Insert LCO 3.4.16 PCS Specific Activity Insert LCO 3.4.17 Steam Generator (SG) Tube Integrity 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
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| Insert LCO 3.5.1 Safety Injection Tanks (SITs)
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| Insert LCO 3.5.2 ECCS - Operating Insert LCO 3.5.3 ECCS - Shutdown Insert LCO 3.5.4 Safety Injection Refueling Water Tank (SIRWT)
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| Insert LCO 3.5.5 Containment Sump Buffering Agent and Weight Requirements
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| | |
| 3.6 CONTAINMENT SYSTEMS Insert LCO 3.6.1 Containment Insert LCO 3.6.2 Containment Air Locks Insert LCO 3.6.3 Containment Isolation Valves Insert LCO 3.6.4 Containment Pressure Insert LCO 3.6.5 Containment Air Temperature Insert LCO 3.6.6 Containment Cooling Systems 3.7 PLANT SYSTEMS Insert LCO 3.7.1 Main Steam Safety Valves (MSSVs)
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| Insert LCO 3.7.2 Main Steam Isolation Valves (MSIVs)
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| Insert LCO 3.7.3 Main Feedwater Regulating Valves (MFRVs) and MFRV Bypass Valves Insert LCO 3.7.4 Atmospheric Dump Valves (ADVs)
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| Insert LCO 3.7.5 Auxiliary Feedwater (AFW) System Insert LCO 3.7.6 Condensate Storage and Supply Insert LCO 3.7.7 Component Cooling Water (CCW) System Insert LCO 3.7.8 Service Water System (SWS)
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| Insert LCO 3.7.9 Ultimate Heat Sink (UHS)
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| Insert LCO 3.7.10 Control Room Ventilation (CRV) Filtration Insert LCO 3.7.11 Control Room Ventilation (CRV) Cooling Insert LCO 3.7.12 Fuel Handling Area Ventilation System Insert LCO 3.7.13 Engineered Safeguards Room Ventilation (ESRV) Dampers
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| SFP Water Level 3.7.14 3.7 FACILITY PLANT SYSTEMS 3.7.14 Spent Fuel Pool (SFP) Water Level LCO 3.7.14 The SFP water level shall be 647 ft elevation.
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| ----------------------------------------------NOTE------------------------------------------
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| SFP level may be below the 647 ft elevation to support fuel cask movement, if the displacement of water by the fuel cask when submerged in the SFP, would raise SFP level to 647 ft elevation.
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| APPLICABILITY: During movement of irradiated fuel assemblies in the SFP, During movement of a fuel cask in or over the SFP.
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| ACTIONS
| |
| ----------------------------------------------------------NOTE-----------------------------------------------------------
| |
| LCO 3.0.3 is not applicable.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. SFP water level not within A.1 Suspend movement of Immediately limit. irradiated fuel assemblies in SFP.
| |
| AND A.2 Suspend movement of Immediately fuel cask in or over the SFP.
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| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the SFP water level is 647 ft elevation. In accordance with the Surveillance Frequency Control Program7 days Palisades Nuclear Plant 3.7.14-1 Amendment No. 271, 272, XXX
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| | |
| SFP Boron Concentration 3.7.15 3.7 FACILITY PLANT SYSTEMS 3.7.15 Spent Fuel Pool (SFP) Boron Concentration LCO 3.7.15 The SFP boron concentration shall be 1720 ppm.
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| APPLICABILITY: When fuel assemblies are stored in the Spent Fuel Pool.
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| ACTIONS
| |
| ----------------------------------------------------------NOTE-----------------------------------------------------------
| |
| LCO 3.0.3 is not applicable.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. SFP boron concentration A.1 Suspend movement of Immediately not within limit. fuel assemblies in the SFP.
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| AND A.2 Initiate action to restore Immediately SFP boron concentration to within limit.
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| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the SFP boron concentration is within limit. In accordance with the Surveillance Frequency Control Program7 days Palisades Nuclear Plant 3.7.15-1 Amendment No. 271, 272
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| | |
| Spent Fuel Pool Storage 3.7.16 3.7 FACILITY PLANT SYSTEMS 3.7.16 Spent Fuel Pool Storage LCO 3.7.16 Storage in the spent fuel pool shall be as follows:
| |
| : a. Each fuel assembly and non-fissile bearing component stored in a Region I Carborundum equipped storage rack shall be within the limitations in Specification 4.3.1.1 and, as applicable, within the requirements of the maximum nominal planar average U-235 enrichment and burnup of Tables 3.7.16-2, 3.7.16-3, 3.7.16-4 or 3.7.16-5,
| |
| : b. Fuel assemblies in a Region I Metamic equipped storage rack shall be within the limitations in Specification 4.3.1.2, and
| |
| : c. The combination of maximum nominal planar average U-235 enrichment, burnup, and decay time of each fuel assembly stored in Region II shall be within the requirements of Table 3.7.16-1.
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| APPLICABILITY: Whenever any fuel assembly or non-fissile bearing component is stored in the spent fuel pool or the north tilt pit.
| |
| ACTIONS
| |
| ----------------------------------------------------------NOTE-----------------------------------------------------------
| |
| LCO 3.0.3 is not applicable.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 Initiate action to restore Immediately not met. the noncomplying fuel assembly or non-fissile bearing component within requirements.
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| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify by administrative means each fuel assembly Prior to storing the or non-fissile bearing component meets fuel fuel assembly or storage requirements. non-fissile bearing component in the spent fuel pool Palisades Nuclear Plant 3.7.16-1 Amendment No. 250, 272, XXX
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| 3.7 PLANT SYSTEMS Insert LCO 3.7.17 Secondary Specific Activity 3.8 ELECTRICAL POWER SYSTEMS Insert LCO 3.8.1 AC Sources - Operating Insert LCO 3.8.2 AC Sources - Shutdown Insert LCO 3.8.3 Diesel Fuel, Lube Oil, and Starting Air Insert LCO 3.8.4 DC Sources - Operating Insert LCO 3.8.5 DC Sources - Shutdown Insert LCO 3.8.6 Battery Cell Parameters Insert LCO 3.8.7 Inverters - Operating Insert LCO 3.8.8 Inverters - Shutdown Insert LCO 3.8.9 Distribution Systems - Operating Insert LCO 3.8.10 Distribution Systems - Shutdown 3.9 REFUELING OPERATIONS Insert LCO 3.9.1 Boron Concentration Insert LCO 3.9.2 Nuclear Instrumentation Insert LCO 3.9.3 Containment Penetrations Insert LCO 3.9.4 Shutdown Cooling (SOC) and Coolant Circulation - High Water Level Insert LCO 3.9.5 Shutdown Cooling (SOC) and Coolant Circulation - Low Water Level Insert LCO 3.9.6 Refueling Cavity Water Level
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| | |
| Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The Palisades Nuclear Plant is located on property owned by Entergy Nuclear Palisades, LLC on the eastern shore of Lake Michigan approximately four and one-half miles south of the southern city limits of South Haven, Michigan. The minimum distance to the boundary of the exclusion area as defined in 10 CFR 100.3 shall be 677 meters.
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| 4.2 (Deleted) Reactor Core 4.2.1 Fuel Assemblies The reactor core shall contain 204 fuel assemblies. Each assembly shall consist of a matrix of zircaloy-4 or M5 clad fuel rods with an initial composition of depleted, natural, or slightly enriched uranium dioxide (UO2) as fuel material.
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| Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. A core plug or plugs may be used to replace one or more fuel assemblies subject to the analysis of the resulting power distribution. Poison may be placed in the fuel bundles for long-term reactivity control.
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| 4.2.2 Control Rod Assemblies The reactor core shall contain 45 control rods. Four of these control rods may consist of part-length absorbers. The control material shall be silver-indium-cadmium, as approved by the NRC.
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| 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The Region I (See Figure B 3.7.16-1) Carborundum equipped fuel storage racks incorporating Regions 1A, 1B, 1C, 1D, and 1E are designed and shall be maintained with:
| |
| : a. Irradiated New or irradiated fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.54 weight percent; Palisades Nuclear Plant 4.0-1 Amendment No. 272, 273, XXX
| |
| | |
| Design Features 4.0 4.3 Fuel Storage 4.3.1 Criticality (continued)
| |
| : 3. Control blades may be stored in both fueled and unfueled locations in Regions 1D and 1E, with no limitation on the number.
| |
| 4.3.1.2 The Region I (See Figure B 3.7.16-1) Metamic equipped fuel storage racks are designed and shall be maintained with:
| |
| : a. Fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.95 weight percent;
| |
| : b. Keff < 1.0 if fully flooded with unborated water, which includes allowances for uncertainties as described in Section 9.11 of the FSAR;
| |
| : c. Keff 0.95 if fully flooded with water borated to 850 ppm, which includes allowances for uncertainties as described in Section 9.11 of the FSAR;
| |
| : d. A nominal 10.25 inch center to center distance between fuel assemblies;
| |
| : e. New or Irradiated irradiated fuel assemblies;
| |
| : f. Two empty rows of storage locations shall exist between the fuel assemblies in a Carborundum equipped rack and the fuel assemblies in an adjacent Metamic equipped rack; and
| |
| : g. A minimum Metamic B10 areal density of 0.02944 g/cm2.
| |
| 4.3.1.3 The Region II fuel storage racks (See Figure B 3.7.16-1) are designed and shall be maintained with:
| |
| : a. Fuel assemblies having maximum nominal planar average U-235 enrichment of 4.60 weight percent;
| |
| : b. Keff < 1.0 if fully flooded with unborated water, which includes allowances for uncertainties as described in Section 9.11 of the FSAR.
| |
| : c. Keff 0.95 if fully flooded with water borated to 850 ppm, which includes allowance for uncertainties as described in Section 9.11 of the FSAR.
| |
| : d. A nominal 9.17 inch center to center distance between fuel assemblies; and
| |
| : e. New or Irradiated irradiated fuel assemblies which meet the maximum nominal planar average U-235 enrichment, burnup, and decay time requirements of Table 3.7.16-1 Palisades Nuclear Plant 4.0-4 Amendment No. 236, 246, XXX
| |
| | |
| Design Features 4.0 4.3 Fuel Storage 4.3.1 Criticality (continued) 4.3.1.4 (Deleted)The new fuel storage racks are designed and shall be maintained with:
| |
| : a. Twenty four unirradiated fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.95 weight percent, and stored in accordance with the pattern shown in Figure 4.3-1, or Thirty six unirradiated fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.05 weight percent, and stored in accordance with the pattern shown in Figure 4.3-1;
| |
| : b. Keff 0.95 when flooded with either full density or low density (optimum moderation) water including allowances for uncertainties as described in Section 9.11 of the FSAR.
| |
| : c. The pitch of the new fuel storage rack lattice being 9.375 inches and every other position in the lattice being permanently occupied by an 8" x 8" structural steel or core plugs, resulting in a nominal 13.26 inch center to center distance between fuel assemblies placed in alternating storage locations.
| |
| 4.3.2 Drainage The spent fuel storage pool cooling system suction and discharge piping is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 644 ft 5 inches.
| |
| 4.3.3 Capacity The spent fuel storage pool and north tilt pit are designed and shall be maintained with a storage capacity limited to no more than 892 fuel assemblies.
| |
| Palisades Nuclear Plant 4.0-5 Amendment No. 207, 236, XXX
| |
| | |
| Design Features 4.0 4.3 Fuel Storage INSERT FIGURE 4.3-1 0 0 0 0 0 0 0 0 0 0 0 0 CENTERLINE - LEGEND PATTERN REPEATS 8 X 8 STEEL BOX BEAM ASSEMBLY STORAGE LOCATION (ENRICHMENT <= 4.95 WT% U-235)
| |
| ASSEMBLY STORAGE LOCATION (ENRICHMENT <= 4.05 WT% U-235)
| |
| Note: If any assemblies containing fuel enrichments greater than 4.05% U-235 are stored in the New Fuel Storage Rack, the center row must remain empty.
| |
| Figure 4.3-1 (page 1 of 1)
| |
| New Fuel Storage Rack Arrangement Palisades Nuclear Plant 4.0-6 Amendment No. 207, 236, XXX
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| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.2 (Deleted)Primary Coolant Sources Outside Containment This program provides controls to minimize leakage to the engineered safeguards rooms, from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident, to as low as practical. The systems include the Containment Spray System, the Safety Injection System, the Shutdown Cooling System, and the containment sump suction piping. This program shall include the following:
| |
| : a. Provisions establishing preventive maintenance and periodic visual inspection requirements, and
| |
| : b. Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.
| |
| : c. The portion of the shutdown cooling system that is outside the containment shall be tested either by use in normal operation or hydrostatically tested at 255 psig.
| |
| : d. Piping from valves CV-3029 and CV-3030 to the discharge of the safety injection pumps and containment spray pumps shall be hydrostatically tested at no less than 100 psig.
| |
| : e. The maximum allowable leakage from the recirculation heat removal systems' components (which include valve stems, flanges and pump seals) shall not exceed 0.2 gallon per minute under the normal hydrostatic head from the SIRW tank.
| |
| 5.5.3 (Deleted) 5.5.4 Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the Offsite Dose Calculation Manual (ODCM), (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
| |
| : a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, Palisades Nuclear Plant 5.0-1 Amendment No. 213, 272
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| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.5 (Deleted) Containment Structural Integrity Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Containment Structural Integrity Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with ASME Boiler and Pressure Vessel Code, Section XI, Subsection IWE and IWL.
| |
| If, as a result of a tendon inspection, corrective retensioning of five percent (8) or more of the total number of dome tendons is necessary to restore their liftoff forces to within the limits, a dome delamination inspection shall be performed within 90 days following such corrective retensioning. The results of this inspection shall be reported to the NRC in accordance with Specification 5.6.7, Containment Structural Integrity Surveillance Report.
| |
| The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Containment Structural Integrity Surveillance Program inspection frequencies.
| |
| 5.5.6 (Deleted) Primary Coolant Pump Flywheel Surveillance Program
| |
| : a. Surveillance of the primary coolant pump flywheels shall consist of a 100% volumetric inspection of the upper flywheels each 10 years.
| |
| : b. The provisions of SR 3.0.2 are not applicable to the Flywheel Testing Program 5.5.7 (Deleted) 5.5.8 (Deleted)Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:
| |
| : a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.
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| Condition monitoring assessments shall be conducted during each outage Palisades Nuclear Plant 5.0-3 Amendment No. 266, 272
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| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.
| |
| : b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
| |
| : 1. Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents.
| |
| This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
| |
| : 2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.
| |
| Leakage is not to exceed 0.3 gpm.
| |
| : 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, PCS Operational LEAKAGE.
| |
| Palisades Nuclear Plant 5.0-4 Amendment No. 266, 272
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program
| |
| : c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged. The following alternative repair criteria shall be applied as an alternate to the 40% depth based criteria:
| |
| : 1. Tubes found by inservice inspection to contain service-induced flaws within 12.5 inches below the bottom of the hot-leg expansion transition or top of the hot-leg tubesheet, whichever is lower, shall be plugged. Flaws located below this elevation may remain in service.
| |
| : 2. Tubes found by inservice inspection to contain service-induced flaws within 13.67 inches below the bottom of the cold-leg expansion transition or top of the cold-leg tubesheet, whichever is lower, shall be plugged. Flaws located below this elevation may remain in service.
| |
| : d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from 12.5 inches below the bottom of the hot-leg expansion transition or top of the hot-leg tubesheet, whichever is lower, to 13.67 inches below the bottom of the cold-leg expansion transition or top of the cold-leg tubesheet, whichever is lower, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet
| |
| --- weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, d.3, and d.4 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
| |
| Palisades Nuclear Plant 5.0-5 Amendment No. 266, 272
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program
| |
| : d. Provisions for SG tube inspections. (continued)
| |
| : 1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.
| |
| : 2. Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.
| |
| : 3. If crack indications are found in any SG tube from 12.5 inches below the bottom of the hot-leg expansion transition or top of the hot-leg tubesheet, whichever is lower, to 13.67 inches below the bottom of the cold-leg expansion transition or top of the cold-leg tubesheet, whichever is lower, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
| |
| : 4. When the alternate repair criteria of TS 5.5.8c.1 are implemented, inspect 100% of the inservice tubes to the hot-leg tubesheet region with the objective of detecting flaws that may satisfy the applicable tube repair criteria of TS 5.5.8c.1 every 24 effective full-power months, or one refueling outage, whichever is less.
| |
| : e. Provisions for monitoring operational primary to secondary LEAKAGE.
| |
| Palisades Nuclear Plant 5.0-6 Amendment No. 266, 272
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 (Deleted) Secondary Water Chemistry Program A program shall be established, implemented and maintained for monitoring of secondary water chemistry to inhibit steam generator tube degradation and shall include:
| |
| : a. Identification of a sampling schedule for the critical variables and control points for these variables,
| |
| : b. Identification of the procedures used to measure the values of the critical variables,
| |
| : c. Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage,
| |
| : d. Procedures for the recording and management of data,
| |
| : e. Procedures defining corrective actions for all off-control point chemistry conditions, and
| |
| : f. A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective actions.
| |
| 5.5.10 Ventilation Filter Testing Program A program shall be established to implement the following required testing of Control Room Ventilation (CRV) and Fuel Handling Area Ventilation (FHAV) systems at the frequencies specified in Regulatory Guide 1.52, Revision 2 (RG 1.52), and in accordance with RG 1.52 and ASME N510-1989, at the system flowrates and tolerances specified below*:
| |
| : a. Demonstrate for each of the ventilation systems that an inplace test of the High Efficiency Particulate Air (HEPA) filters shows a penetration and system bypass < 0.05% for the CRV system and < 1.00% for the FHAV system when tested in accordance with RG 1.52 and ASME N510-1989:
| |
| (Deleted)
| |
| Ventilation System Flowrate (CFM)
| |
| FHAV (single fan operation) 7300 +/- 20%
| |
| FHAV (dual fan operation) 10,000 +/- 20%
| |
| CRV 3,200 +10% -5%
| |
| Palisades Nuclear Plant 5.0-7 Amendment No. 266, 272
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Ventilation Filter Testing Program (Continued
| |
| : b. Demonstrate for each of the ventilation systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 0.05% for the CRV system and < 1.00% for the FHAV system when tested in accordance with RG 1.52 and ASME N510-1989.
| |
| Ventilation System Flowrate (CFM)
| |
| FHAV (dual fan operation) 10,000 +/- 20%
| |
| CRV 3200 +10% -5%
| |
| : c. Demonstrate for each of the ventilation systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in RG 1.52 shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30C and equal to the relative humidity specified as follows:
| |
| Ventilation System Penetration Relative Humidity FHAV 6.00% 95%
| |
| CRV 0.157% 70%
| |
| : d. For each of the ventilation systems, demonstrate the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with RG 1.52 and ASME N510-1989:
| |
| Ventilation System Delta P (In H20) Flowrate (CFM)
| |
| FHAV (dual fan operation) 6.0 10,000 +/- 20%
| |
| CRV 8.0 3200 +10% -5%
| |
| : e. Demonstrate that the heaters for the CRV system dissipates the following specified value +/- 20% when tested in accordance with ASME N510-1989:
| |
| Ventilation System Wattage CRV 15 kW The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Ventilation Filter Testing Program frequencies.
| |
| * Should the 720-hour limitation on charcoal adsorber operation occur during a plant operation requiring the use of the charcoal adsorber - such as refueling - testing may be delayed until the completion of the plant operation or up to 1,500 hours of filter operation; whichever occurs first.
| |
| Palisades Nuclear Plant 5.0-8 Amendment No. 266, 272
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 (Deleted) Fuel Oil Testing Program A fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling requirements, testing requirements, and acceptance criteria, based on the diesel manufacturers specifications and applicable ASTM Standards. The program shall establish the following:
| |
| : a. Acceptability of new fuel oil prior to addition to the Fuel Oil Storage Tank, and acceptability of fuel oil stored in the Fuel Oil Storage Tank, by determining that the fuel oil has the following properties within limits:
| |
| : 1. API gravity or an absolute specific gravity,
| |
| : 2. Kinematic viscosity, and
| |
| : 3. Water and sediment content.
| |
| : b. Other properties of fuel oil stored in the Fuel Oil Storage Tank, specified by the diesel manufacturers or specified for grade 2D fuel oil in ASTM D 975, are within limits.
| |
| The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Fuel Oil Testing Program.
| |
| 5.5.12 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.
| |
| : a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
| |
| : b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
| |
| : 1. A change in the TS incorporated in the license; or
| |
| : 2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
| |
| Palisades Nuclear Plant 5.0-9 Amendment No. 266, 272
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Technical Specifications (TS) Bases Control Program (continued)
| |
| : c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.
| |
| : d. Proposed changes that meet the criteria of Specification 5.5.12.b. above shall be reviewed and approved by the NRC prior to implementation.
| |
| Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e) 5.5.13 (Deleted) Safety Functions Determination Program (SFDP)
| |
| This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:
| |
| : a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;
| |
| : b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;
| |
| : c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and
| |
| : d. Other appropriate limitations and remedial or compensatory actions.5.5.13 Safety Functions Determination Program (SFDP)
| |
| (Continued)
| |
| A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:
| |
| : a. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or
| |
| : b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or Palisades Nuclear Plant 5.0-10 Amendment No. 266, 272
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Safety Functions Determination Program (SFDP) (continued)
| |
| : c. A required system redundant to support system(s) for the supported systems (a) and (b) above is also inoperable.
| |
| The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.
| |
| 5.5.14 (Deleted) Containment Leak Rate Testing Program
| |
| : a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with NEI 94-01, Revision 2-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated October 2008, with the following exceptions:
| |
| : 1. Leakage rate testing is not necessary after opening the Emergency Escape Air Lock doors for post-test restoration or post-test adjustment of the air lock door seals. However, a seal contact check shall be performed instead.
| |
| Emergency Escape Airlock door opening, solely for the purpose of strongback removal and performance of the seal contact check, does not necessitate additional pressure testing.
| |
| : 2. Leakage rate testing at Pa is not necessary after adjustment of the Personnel Air Lock door seals. However, a between-the-seals test shall be performed at 10 psig instead.
| |
| : 3. Leakage rate testing frequency for the Containment 4 inch purge exhaust valves, the 8 inch purge exhaust valves, and the 12 inch air room supply valves may be extended up to 60 months based on component performance.
| |
| : b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 54.2 psig. The containment design pressure is 55 psig.
| |
| : c. The maximum allowable containment leakage rate, La, at Pa, shall be 0.1% of containment air weight per day.
| |
| Palisades Nuclear Plant 5.0-11 Amendment No. 266, 272
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Containment Leak Rate Testing Program (Continued)
| |
| : d. Leakage rate acceptance criteria are:
| |
| : 1. Containment leakage rate acceptance criteria is 1.0 La. During the first plant startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and Type C tests and 0.75 La for Type A tests.
| |
| : 2. Air lock testing acceptance criteria are:
| |
| a) Overall air lock leakage is 1.0 La when tested at Pa and combined with all penetrations and valves subjected to Type B and C tests. However, during the first unit startup following testing performed in accordance with this program, the leakage rate acceptance criteria is < 0.6 La when combined with all penetrations and valves subjected to Type B and C tests.
| |
| b) For each Personnel Air Lock door, leakage is 0.023 La when pressurized to 10 psig.
| |
| c) For each Emergency Escape Air Lock door, a seal contact check , consisting of a verification of continuous contact between the seals and the sealing surfaces, is acceptable.
| |
| : e. Containment OPERABILITY is equivalent to "Containment Integrity" for the purposes of the testing requirements.
| |
| : f. The provisions of SR 3.0.3 are applicable to the Containment Leak Rate Testing Program requirements.
| |
| : g. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.
| |
| Palisades Nuclear Plant 5.0-12 Amendment No. 266, 272
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 (Deleted) Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Ventilation (CRV) Filtration, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident. The program shall include the following elements:
| |
| : a. The definition of the CRE and the CRE boundary.
| |
| : b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.
| |
| : c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.
| |
| : d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CRV Filtration, operating at the flow rate required by the Ventilation Filter Testing Program, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.
| |
| : e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c.
| |
| The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.
| |
| : f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.
| |
| Palisades Nuclear Plant 5.0-14 Amendment No. 266, 272
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.17 (Deleted) Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
| |
| : a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.
| |
| : b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.
| |
| : c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.
| |
| Palisades Nuclear Plant 5.0-15 Amendment No. 266, 272
| |
| | |
| Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.
| |
| 5.6.1 (Deleted) 5.6.2 Radiological Environmental Operating Report The Radiological Environmental Operating Report covering the operation of the facility plant during the previous calendar year shall be submitted before May 15 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.
| |
| 5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering operation of the facility plant in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the facilityplant. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual and Process Control Program, and shall be in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.
| |
| 5.6.4 (Deleted) 5.6.5 (Deleted) CORE OPERATING LIMITS REPORT (COLR)
| |
| : a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
| |
| 3.1.1 Shutdown Margin 3.1.6 Regulating Rod Group Position Limits 3.2.1 Linear Heat Rate Limits 3.2.2 Radial Peaking Factor Limits 3.2.4 ASI Limits 3.4.1 DNB Limits Palisades Nuclear Plant 5.0-16 Amendment No. 261, 272
| |
| | |
| 5.6 Reporting Requirements 5.6.5 COLR (Continued)
| |
| : b. The analytical methods used to determine the core operating limits shall be those approved by the NRC, specifically those described in the latest approved revision of the following documents:
| |
| : 1. EMF-96-029(P)(A) Volumes 1 and 2, Reactor Analysis System for PWRs, Siemens Power Corporation.
| |
| (LCOs 3.1.1, 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 2. ANF-84-73 Appendix B (P)(A), "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," Advanced Nuclear Fuels Corporation. (Bases report not approved) (LCOs 3.1.1, 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 3. XN-NF-82-21(P)(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company.
| |
| (LCOs 3.2.1, 3.2.2, & 3.2.4)
| |
| : 4. EMF-84-093(P)(A), Steam Line Break Methodology for PWRs, Siemens Power Corporation.
| |
| (LCOs 3.1.1, 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 5. XN-75-32(P)(A) Supplements 1 through 4, "Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company. (Bases document not approved)
| |
| (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 6. EMF-2310 (P)(A), Revision 0, Framatome ANP, Inc., May 2001, SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 7. XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company. (LCOs 3.1.6, 3.2.1, & 3.2.2)
| |
| : 8. ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events,"
| |
| Advanced Nuclear Fuels Corporation.
| |
| (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 9. EMF-92-153(P)(A) and Supplement 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel,"
| |
| Siemens Power Corporation. (LCOs 3.2.1, 3.2.2, & 3.2.4)
| |
| Palisades Nuclear Plant 5.0-17 Amendment No. 261, 272
| |
| | |
| 5.6 Reporting Requirements 5.6.5 COLR (Continued)
| |
| : 10. XN-NF-621(P)(A), Exxon Nuclear DNB Correlation for PWR Fuel Designs, Exxon Nuclear Company. (LCOs 3.2.1, 3.2.2, & 3.2.4)
| |
| : 11. XN-NF-82-06(P)(A) and Supplements 2, 4, and 5, Qualification of Exxon Nuclear Fuel for Extended Burnup, Exxon Nuclear Company. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 12. ANF-88-133(P)(A) and Supplement 1, Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWD/MTU, Advanced Nuclear Fuels Corporation.
| |
| (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 13. XN-NF-85-92(P)(A), Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results, Exxon Nuclear Company. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 14. EMF-92-116(P)(A), Generic Mechanical Design Criteria for PWR Fuel Designs, Siemens Power Corporation.
| |
| (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 15. EMF-2087(P)(A), SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications, Siemens Power Corporation.
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| (LCOs 3.1.6, 3.2.1, & 3.2.2)
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| : 16. ANF-87-150 Volume 2, Palisades Modified Reactor Protection System Report: Analysis of Chapter 15 Events, Advanced Nuclear Fuels Corporation. [Approved for use in the Palisades design during the NRC review of license Amendment 118, November 15, 1988] (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.4.1)
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| : 17. EMF-1961(P)(A), Revision 0, Siemens Power Corporation, July 2000, Statistical Setpoint/Transient Methodology for Combustion Engineering Type Reactors. (LCOs 3.1.6, 3.2.1, 3.2.2, 3.2.4, &
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| 3.4.1)
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| : 18. EMF-2328 (P)(A), Revision 0, Framatome ANP, Inc., March 2001, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based.
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| (LCOs 3.1.6, 3.2.1, & 3.2.2)
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| : 19. BAW-2489P, Revised Fuel Assembly Growth Correlation for Palisades. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
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| Palisades Nuclear Plant 5.0-18 Amendment No. 261, 272
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| 5.6 Reporting Requirements 5.6.5 COLR (Continued)
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| : 20. EMF-2103(P)(A), Realistic Large Break LOCA Methodology for Pressurized Water Reactors. (LCOs 3.1.6, 3.2.1, & 3.2.2)
| |
| : 21. BAW-10240(P)-A, Incorporation of M5 Properties in Framatome ANP Approved Methods. (LCOs 3.1.6, 3.2.1, 3.2.2, 3.2.4, & 3.4.1)
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| : c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met.
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| : d. The COLR, including any mid cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC.
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| 5.6.6 (Deleted) Post Accident Monitoring Report When a report is required by LCO 3.3.7, "Post Accident Monitoring Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels to OPERABLE status.
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| 5.6.7 (Deleted) Containment Structural Integrity Surveillance Report Reports shall be submitted to the NRC covering Prestressing, Anchorage, and Dome Delamination tests within 90 days after completion of the tests.
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| 5.6.8 (Deleted) Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.8, Steam Generator (SG) Program. The report shall include:
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| : a. The scope of inspections performed on each SG,
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| : b. Active degradation mechanisms found,
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| : c. Nondestructive examination techniques utilized for each degradation mechanism, Palisades Nuclear Plant 5.0-19 Amendment No. 261, 272
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| | |
| 5.6 Reporting Requirements 5.6.8 Steam Generator Tube Inspection Report (Continued)
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| : d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,
| |
| : e. Number of tubes plugged during the inspection outage for each active degradation mechanism,
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| : f. Total number and percentage of tubes plugged to date,
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| : g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and
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| : h. The effective plugging percentage for all plugging in each SG.
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| i The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.
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| Palisades Nuclear Plant 5.0-15 Amendment No. 196, 272
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| PALISADES PLANT ENVIRONMENTAL PROTECTION PLAN (NON-RADIOLOGICAL)
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| TABLE OF CONTENTS Section Page 1.0 Objectives of the Environmental Protection Plan ..................................... 1-1 2.0 Environmental Protection Issues .............................................................. 2-1 2.1 Aquatic Issues .......................................................................................... 2-1 2.2 Terrestrial Issues ...................................................................................... 2-1 3.0 Consistency Requirements ...................................................................... 3-1 3.1 Facility Plant Design and Operation ......................................................... 3-1 3.2 Reporting Related to the NPDES Permits and State Certification. ........... 3-2 3.3 Changes Required for Compliance with Other Environmental Regulations .............................................................................................. 3-3 4.0 Environmental Conditions ........................................................................ 4-1 4.1 Unusual or Important Environmental Events ............................................ 4-1 4.2 Environmental Monitoring......................................................................... 4-1 5.0 Administrative Procedures ....................................................................... 5-1 5.1 Review and Audit ..................................................................................... 5-1 5.2 Records Retention ................................................................................... 5-1 5.3 Changes in Environmental Protection Plan .............................................. 5-2 5.4 Facility Plant Reporting Requirements ..................................................... 5-2 Amendment No. 63, 272, XXX May 13, 2022
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| | |
| 1.0 Objectives of the Environmental Protection Plan The Environmental Protection Plan (EPP) is to provide for protection of environmental values during handling and storage of spent fuelconstruction and maintenance operation of the nuclear facility. The principal objectives of the EPP are as follows:
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| (1) Verify that the facility is maintainedplant is operated in an environmentally acceptable manner, as established by the FES and other NRC environmental impact assessments.
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| (2) Coordinate NRC requirements and maintain consistency with other Federal, State and local requirements for environmental protection.
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| (3) Keep NRC informed of the environmental effects of handling and storage of spent fuel and maintenance of the facility construction and operation and of actions taken to control those effects.
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| Environmental concerns identified in the FES which relate to water quality matters are regulated by way of the licensee's NPDES permit.
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| 1-1 Amendment No. 272, XXX May 13, 2022
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| | |
| 2.0 Environmental Protection Issues In the final addendum to the FES-OL dated February 1978 the staff considered the environmental impacts associated with the operation of the Palisades Plant.
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| Certain environmental issues were identified which required study or license conditions to resolve environmental concerns and to assure adequate protection of the environment.
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| 2.1 Aquatic Issues Specific aquatic issues raised by the staff in the FES-OL were:
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| The need for aquatic monitoring programs to confirm that thermal mixing occurs as predicted, that chlorine releases are controlled within those discharge concentrations evaluated, and that effects on aquatic biota and water quality due to facility plant operation are no greater than predicted.
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| Aquatic issues are addressed by the effluent limitations, and monitoring requirements are contained in the effective NPDES permit issued by the State of Michigan, Department of Natural Resources. The NRC will rely on this agency for regulation of matters involving water quality and aquatic biota.
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| 2.2 Terrestrial Issues
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| : 1. Potential impacts on the terrestrial environment associated with drift from the mechanical draft cooling towers. (FES-OL addendum Section 6.3) 2-1 Amendment No. 272, XXX May 13, 2022
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| | |
| 3.0 Consistency Requirements 3.1 Facility Plant Design and Operation The licensee may make changes in facility station design or operation or perform tests or experiments affecting the environment provided such changes, tests or experiments do not involve an unreviewed environmental question, and do not involve a change in the Environmental Protection Plan. Changes in facility plant design or operation or performance of tests or experiments which do not affect the environment are not subject to the requirements of this EPP. Activities governed by Section 3.3 are not subject to the requirements of this section.
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| Before engaging in additional construction or operational activities which may affect the environment, the licensee shall prepare and record an environmental evaluation of such activity. When the evaluation indicates that such activity involves an unreviewed environmental question, the licensee shall provide a written evaluation of such activities and obtain prior approval from the Director, Office of Nuclear Reactor Regulation. When such activity involves a change in the Environmental Protection Plan, such activity and change to the Environmental Protection Plan may be implemented only in accordance with an appropriate license amendment as set forth in Section 5.3.
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| A proposed change, test or experiment shall be deemed to involve an unreviewed environmental question if it concerns (1) a matter which may result in a significant increase in any adverse environmental impact previously evaluated in the final environmental statement (FES) as modified by staff's testimony to the Atomic Safety and Licensing Board, supplements to the FES, environmental impact appraisals, or in any decisions of the Atomic Safety and Licensing Board; or (2) a significant change in effluents or power level [in accordance with 10 CFR Part 51.5(b)(2)] or (3) a matter not previously reviewed and evaluated in the documents specified in (1) of this Subsection, which may have a significant adverse environmental impact.
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| 3-1 Amendment No. 272, XXX May 13, 2022
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| | |
| 3.3 Changes Required for Compliance with Other Environmental Regulations Changes in facility plant design or operation and performance of tests or experiments which are required to achieve compliance with other Federal, State, or local environmental regulations are not subject to the requirements of Section 3.1.
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| 3-3 Amendment No. 272, XXX May 13, 2022
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| | |
| 4.0 Environmental Conditions 4.1 Unusual or Important Environmental Events Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to the handling and storage of spent fuel and maintenance of the facilityplant operation shall be recorded and promptly reported to the NRC within 24 hours by telephone, telegraph, or facsimile transmissions followed by a written report per Subsection 5.4.2. The following are examples: excessive bird impaction events, onsite plant or animal disease outbreaks, mortality or unusual occurrence of any species protected by the Endangered Species Act of 1973, fish kills, increase in nuisance organisms or conditions and unanticipated or emergency discharge of waste water or chemical substances.
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| No routine monitoring programs are required to implement this condition.
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| 4.2 Environmental Monitoring 4.2.1 Meteorological Monitoring A meteorological monitoring program shall be conducted in the vicinity of the plant site for at least two years after conversion to cooling towers to document effects of cooling tower operation on meteorological variables. Data on the following meteorological variables shall be obtained from the station network shown in Figure 4.2.1: precipitation, temperature, humidity, solar radiation, downcoming radiation, visibility, wind direction and wind speed. In addition, studies shall be conducted for at least two years to measure affects of cooling tower drift on vegetation by associated salt deposition, icing or other causes.
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| 4-1 Amendment No. 272, XXX May 13, 2022
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| | |
| 5.0 Administrative Procedures 5.1 Review and Audit The licensee shall provide for review and audit of compliance with the Environmental Protection Plan. The audits shall be conducted independently of the individual or groups responsible for performing the specific activity. A description of the organization structure utilized to achieve the independent review and audit function and results of the audit activities shall be maintained and made available for inspection.
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| 5.2 Records Retention Records and logs relative to the environmental aspects of previous plant operation and the handling and storage of spent fuel and maintenance of the facility shall be made and retained in a manner convenient for review and inspection. These records and logs shall be made available to NRC on request.
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| Records of modifications to facility plant structures, systems and components determined to potentially affect the continued protection of the environment shall be retained for the life of the facilityplant. All other records, data and logs relating to this EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.
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| 5-1 Amendment No. 272, XXX May 13, 2022
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| | |
| 5.3 Changes in Environmental Protection Plan Request for change in the Environmental Protection Plan shall include an assessment of the environmental impact of the proposed change and a supporting justification. Implementation of such changes in the EPP shall not commence prior to NRC approval of the proposed changes in the form of a license amendment incorporating the appropriate revision to the Environmental Protection Plan.
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| 5.4 Facility Plant Reporting Requirements 5.4.1 Routine Reports An Annual Environmental Operating Report describing implementation of this EPP for the previous year shall be submitted to the NRC prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following issuance of the operating license. The period of the first report shall begin with the date of issuance of the operating license.
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| The report shall include summaries and analyses of the results of the environmental protection activities required by Subsection 4.2 of this Environmental Protection Plan for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous nonradiological environmental monitoring reports, and an assessment of the observed impacts of the facility plant operation on the environment. If harmful effects or evidence of trends towards irreversible damage to the environment are observed, the licensee shall provide a detailed analysis of the data and a proposed course of action to alleviate the problem.
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| 5-2 Amendment No. 272, XXX May 13, 2022
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| | |
| The Annual Environmental Operating Report shall also include:
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| (a) A list of EPP noncompliances and the corrective actions taken to remedy them.
| |
| (b) A list of all changes in facility station design or operation, tests, and experiments made in accordance with Subsection 3.1 which involved a potentially significant unreviewed environmental issue.
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| (c) A list of nonroutine reports submitted in accordance with Subsection 5.4.2.
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| In the event that some results are not available by the report due date, the report shall be submitted noting and explaining the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
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| 5-3 Amendment No. 272, XXX May 13, 2022
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| | |
| 5.4.2 Nonroutine Reports A written report shall be submitted to the NRC within 30 days of occurrence of nonroutine event. The report shall (a) describe, analyze, and evaluate the event, including extent and magnitude of the impact and facility plant operating characteristics, (b) describe the probable cause of the event, (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e) indicate the agencies notified and their preliminary responses.
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| Events reportable under this subsection which also require reports to other Federal, State or local agencies shall be reported in accordance with those reporting requirements in lieu of the requirements of this subsection. The NRC shall be provided a copy of such report at the time it is submitted to the other agency.
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| 5-4 Amendment No. 272, XXX May 13, 2022
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| | |
| Enclosure Attachment 2 to HDI PNP 2023-030 Page Change Instructions and Retyped Pages for the Palisades Plant Renewed Facility License DPR-20, Appendix A Technical Specifications, and Appendix B Environmental Protection Plan Note, references to "HDI" are replaced by bracketed Palisades Energy, LLC, or Palisades Energy (e.g. [Palisades Energy]) to reflect the change in operating authority per license transfer application conforming amendments.
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| 251 pages follow
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| | |
| Page Change Instructions ATTACHMENT TO LICENSE AMENDMENT NO. XXX RENEWED FACILITY OPERATING LICENSE NO. DPR-20 DOCKET NO. 50-255 Remove the following pages of Palisades Plant Renewed Facility Operating License and replace with the attached revised Palisades Plant Renewed Facility License. The revised pages are identified by amendment number and contains a line in the margin indicating the area of change.
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| REMOVE INSERT Pages 1 through 7 Pages 1 through 8 Remove the following pages of Appendix A, Permanently Defueled Technical Specifications, and replace with the attached revised pages. The revised pages are identified by amendment number and contain a line in the margin indicating the area of change.
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| REMOVE INSERT TS Title Page TS Title Page Page i None Pages 1.1-1 Page 1.1-1 through 1.1-7 Pages 1.2-1 through 1.2-2 Pages 1.2-1 through 1.2-3 Pages 1.3-1 through 1.3-2 Pages 1.3-1 through 1.3-12 Pages 1.4-1 through 1.4-3 Pages 1.4-1 through 1.4-8 Page 2.0-1 Page 2.0-1 Pages 3.0-1 through 3.0-2 Pages 3.0-1 through 3.0-6 New Pages 3.1.1-1 through 3.7.13-1 Page 3.7.14-1 Page 3.7.14-1 Page 3.7.15-1 Page 3.7.15-1 Page 3.7.16-1 Page 3.7.16-1 New Pages 3.7.17-1 through 3.9.6-1 Pages 4.0-1 through 4.0-5 Pages 4.0-1 through 4.0-6 Pages 5.0-7 through 5.0-15 Pages 5.0-7 through 5.0-30 Remove the following pages of Appendix B, Environmental Protection Plan, and replace with the attached revised pages. The revised pages are identified by amendment number and contain a line in the margin indicating the area of change.
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| REMOVE INSERT Cover Sheet Cover Sheet Table of Contents Table of Contents Page 1-1 Page 1-1 Page 2-1 Page 2-1 Pages 3-1 through 3-3 Pages 3-1 through 3-3 Page 4-1 Page 4-1 Pages 5-1 through 5-4 Pages 5-1 through 5-4
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| | |
| HOLTEC PALISADES, LLC
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| [Palisades Energy], LLC DOCKET NO. 50-255 PALISADES NUCLEAR PLANT RENEWED FACILITY OPERATING LICENSE Renewed License No. DPR-20
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| : 1. The Nuclear Regulatory Commission (NRC or the Commission) having previously made the findings set forth in Operating License No. DPR-20, dated February 21, 1991, has now found that:
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| A. The application for Renewed Operating License No. DPR-20 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
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| and the Commissions rules and regulations set forth in 10 CFR Chapter I, and all required notifications to other agencies or bodies have been duly made; B. [deleted];
| |
| C. Actions have been identified and have been or will be taken with respect to:
| |
| (1) managing the effects of aging on the functionality of structures and components that have been identified to require review under 10 CFR 54.21(a)(1) during the period of extended operation, and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21(c), such that there is reasonable assurance that the activities authorized by this renewed operating license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3 for the facility, and that any changes made to the facilitys current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commissions regulations; Renewed License No. DPR-20 Amendment No. XXX
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| | |
| D. The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; E. There is reasonable assurance: (i) that the activities authorized by this renewed operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations set forth in 10 CFR Chapter I; F. Holtec Palisades, LLC (Holtec Palisades) is financially qualified and [Palisades Energy, LLC (Palisades Energy)] is financially and technically qualified to engage in the activities authorized by this renewed operating license in accordance with the Commissions regulations set forth in 10 CFR Chapter I; G. Holtec Palisades and [Palisades Energy] have satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements" of the Commissions regulations; H. The issuance of this renewed operating license will not be inimical to the common defense and security or to the health and safety of the public; I. After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of this renewed Facility Operating License No. DPR-20, subject to the conditions for protection of the environment set forth herein, is in accordance with 10 CFR Part 51 (formerly Appendix D to Part 50), of the Commissions regulations and all applicable requirements have been satisfied; and J. The receipt, possession, and use of source, byproduct, and special nuclear material as authorized by this renewed operating license will be in accordance with 10 CFR Parts 30, 40, and 70.
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| : 2. Renewed Facility Operating License No. DPR-20 is hereby issued to Holtec Palisades and
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| [Palisades Energy] as follows:
| |
| A. This renewed license applies to the Palisades Plant, a pressurized light water moderated and cooled reactor and electrical generating equipment (the facility). The facility is located in Van Buren County, Michigan, and is described in the Palisades Plant Updated Final Safety Analysis Report, as supplemented and amended, and in the Palisades Plant Environmental Report, as supplemented and amended.
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| B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:
| |
| (1) Pursuant to Section 104b of the Act, as amended, and 10 CFR Part 50, Licensing of Production and Utilization Facilities, (a) Holtec Palisades to possess and use, and (b) [Palisades Energy, LLC (Palisades Energy)] to possess, use and operate the facility as a utilization facility at the designated location in Van Buren County, Michigan, in accordance with the procedures and limitation set forth in this license; Renewed License No. DPR-20 Amendment No. 272, 273, XXX
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| | |
| (2) [Palisades Energy], pursuant to the Act and 10 CFR Parts 40 and 70, to receive ,
| |
| possess and use source and special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3) [Palisades Energy], pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use byproduct, source and special nuclear material as sealed sources for reactor startup, reactor instrumentation radiation monitoring equipment calibration, and that were used as fission detectors in amounts as required; (4) [Palisades Energy], pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material for sample analysis or instrument calibration, or associated with radioactive apparatus or components; and (5) [Palisades Energy], pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operations of the facility.
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| C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act; to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
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| (1) [Palisades Energy] is authorized to operate the facility at steady-state reactor core power levels not in excess of 2565.4 Megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.
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| (2) The Technical Specifications contained in Appendix A, as revised through Amendment No. XXX, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. [Palisades Energy] shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
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| (3) Fire Protection
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| [Palisades Energy] shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment requests dated December 12, 2012, November 1, 2017, November 1, 2018, and March 8, 2019, as supplemented by letters dated February 21, 2013, September 30, 2013, October 24, 2013, December 2, 2013, April 2, 2014, May 7, 2014, June 17, 2014, August 14, 2014, November 4, 2014, December 18, 2014, January 24, 2018, and May 28, 2019, as approved in the safety evaluations dated February 27, 2015, February 27, 2018, and August 20, 2019. Except where NRC approval for changes or Renewed License No. DPR-20 Amendment No. 272, 273, XXX
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| | |
| deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
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| (a) Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.
| |
| : 1. Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
| |
| : 2. Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1x10-7/year (yr) for CDF and less than 1x10-8/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
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| (b) Other Changes that May Be Made Without Prior NRC Approval
| |
| : 1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard.
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| The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3, element is functionally equivalent to the corresponding technical Renewed License No. DPR-20 Amendment No. 272, 273, XXX
| |
| | |
| requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.
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| The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is adequate for the hazard. Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:
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| * Fire Alarm and Detection Systems (Section 3.8);
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| * Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);
| |
| * Gaseous Fire Suppression Systems (Section 3.10); and
| |
| * Passive Fire Protection Features (Section 3.11).
| |
| This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.
| |
| : 2. Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation dated February 27, 2015, to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.
| |
| (c) Transition License Conditions
| |
| : 1. Before achieving full compliance with 10 CFR 50.48(c), as Renewed License No. DPR-20 Amendment No. 272, 273, XXX
| |
| | |
| specified by 2, below, risk-informed changes to the licensees fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2. above.
| |
| : 2. The licensee shall implement the modifications to its facility, as described in Table S-2, Plant Modifications Committed, of Entergy Nuclear Operations, Inc. (ENO) letter PNP 2019-028 dated May 28, 2019, to complete the transition to full compliance with 10 CFR 50.48(c) before the end of the refueling outage following the fourth full operating cycle after NRC approval. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.
| |
| : 3. The licensee shall implement the items listed in Table S-3, Implementation Items, of ENO letter PNP 2014-097 dated November 4, 2014, within six months after NRC approval, or six months after a refueling outage if in progress at the time of approval with the exception of Implementation Items 3 and 8 which will be completed once the related modifications are installed and validated in the PRA model.
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| (4) [deleted]
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| (5) [deleted]
| |
| (6) Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:
| |
| : a. Fire fighting response strategy with the following elements:
| |
| : 1. Pre-defined coordinated fire response strategy and guidance
| |
| : 2. Assessment of mutual aid fire fighting assets
| |
| : 3. Designated staging areas for equipment and materials
| |
| : 4. Command and control
| |
| : 5. Training of response personnel
| |
| : b. Operations to mitigate fuel damage considering the following:
| |
| : 1. Protection and use of personnel assets
| |
| : 2. Communications
| |
| : 3. Minimizing fire spread
| |
| : 4. Procedures for implementing integrated fire response strategy
| |
| : 5. Identification of readily-available pre-staged equipment
| |
| : 6. Training on integrated fire response strategy
| |
| : 7. Spent fuel pool mitigation measures
| |
| : c. Actions to minimize release to include consideration of:
| |
| : 1. Water spray scrubbing
| |
| : 2. Dose to onsite responders Renewed License No. DPR-20 Amendment No. 272, 273, XXX
| |
| | |
| (7) [deleted]
| |
| (8) Amendment 257 authorizes the implementation of 10 CFR 50.61a in lieu of 10 CFR 50.61.
| |
| D. The facility has been granted certain exemptions from Appendix J to 10 CFR Part 50, Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors.
| |
| This section contains leakage test requirements, scheduled and acceptance criteria for tests of the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. These exemptions were granted in a {{letter dated|date=December 6, 1989|text=letter dated December 6, 1989}}.
| |
| These exemptions granted pursuant to 10 CFR 50.12, are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. With these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.
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| E. [Palisades Energy] shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Palisades Nuclear Plant Physical Security Plan.
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| [Palisades Energy] shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Palisades CSP was approved by License Amendment No. 243 as supplemented by changes approved by License Amendment Nos. 248, 253, 259, and 264.
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| F. [deleted]
| |
| G. Holtec Palisades and [Palisades Energy] shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.
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| H. [deleted]
| |
| I. [deleted]
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| Renewed License No. DPR-20 Amendment No. 272, 273, XXX
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| | |
| J. All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of American Society for Testing and Materials (ASTM) E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal scheduled, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC, as required by 10 CFR Part 50, Appendix H.
| |
| K. This license is effective as of the date of issuance and shall expire at midnight March 24, 2031.
| |
| FOR THE NUCLEAR REGULATORY COMMISSION
| |
| /RA/
| |
| J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:
| |
| : 1. Appendix A -Technical Specifications
| |
| : 2. Appendix B - Environmental Protection Plan Date of Issuance: January 17, 2007 Renewed License No. DPR-20 Amendment No. 272, 273, XXX
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| | |
| PALISADES PLANT RENEWED FACILITY OPERATING LICENSE DPR-20 APPENDIX A TECHNICAL SPECIFICATIONS As Amended through Amendment No. XXX
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| | |
| Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions
| |
| ----------------------------------------------------------NOTE----------------------------------------------------------
| |
| The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.
| |
| Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.
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| AVERAGE DISINTEGRATION shall be the average (weighted in proportion to the ENERGY - concentration of each radionuclide in the primary coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total noniodine activity in the coolant.
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| AXIAL OFFSET (AO) AO shall be the power generated in the lower half of the core less the power generated in the upper half of the core, divided by the sum of the power generated in the lower and upper halves of the core (determined using the incore monitoring system).
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| AXIAL SHAPE INDEX (ASI) ASI shall be the power generated in the lower half of the core less the power generated in the upper half of the core, divided by the sum of the power generated in the lower and upper halves of the core (determined using the excore monitoring system).
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| CERTIFIED FUEL HANDLER A CERTIFIED FUEL HANDLER is an individual who complies with provisions of the CERTIFIED FUEL HANDLER training and retraining program required by Specification 5.3.2.
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| Palisades Nuclear Plant 1.1-1 Amendment No. 266, 272, XXX
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| | |
| Definitions 1.1 1.1 Definitions CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST.
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| Calibration of instrument channels with Resistance Temperature Detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel.
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| Whenever a RTD or thermocouple sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.
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| The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps.
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| CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
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| CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:
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| : a. Analog and bistable channels - the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, of all devices in the channel required for channel OPERABILITY;
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| : b. Digital channels - the use of diagnostic programs to test digital hardware and the injection of simulated process data into the channel to verify OPERABILITY, of all devices in the channel required for channel OPERABILITY Palisades Nuclear Plant 1.1-2 Amendment No. XXX
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| | |
| Definitions 1.1 1.1 Definitions CHANNEL FUNCTIONAL TEST The CHANNEL FUNCTIONAL TEST may be performed by (continued) means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or control rods within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
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| CORE OPERATING LIMITS The COLR is the plant specific document that provides cycle REPORT (COLR) specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.
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| DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The dose conversion factors used for this calculation shall be those listed in Federal Guidance Report 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion, 1989; (Table 2.1, Exposure-to-Dose Conversion Factors for Inhalation).
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| INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).
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| Palisades Nuclear Plant 1.1-3 Amendment No. XXX
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| | |
| Definitions 1.1 1.1 Definitions LEAKAGE LEAKAGE shall be:
| |
| : a. Identified LEAKAGE
| |
| : 1. LEAKAGE, such as that from pump seals or valve packing (except Primary Coolant Pump seal water leakoff), that is captured and conducted to collection systems or a sump or collecting tank;
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| : 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known not to interfere with the operation of leakage detection systems and not to be pressure boundary LEAKAGE; and
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| : 3. Primary Coolant System (PCS) LEAKAGE through a Steam Generator to the Secondary System (primary to secondary LEAKAGE).
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| : b. Unidentified LEAKAGE All LEAKAGE (except Primary Coolant Pump seal leakoff) that is not identified LEAKAGE;
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| : c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in a PCS component body, pipe wall, or vessel wall.
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| MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average primary coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
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| NON-CERTIFIED OPERATOR A NON-CERTIFIED OPERATOR is a non-licensed operator who complies with the qualification requirements of Specification 5.3.1.
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| Palisades Nuclear Plant 1.1-4 Amendment No. 266, 272, XXX
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| | |
| Definitions 1.1 1.1 Definitions OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
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| PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:
| |
| : a. Described in Chapter 13, Initial Tests and Operation, of the FSAR;
| |
| : b. Authorized under the provisions of 10 CFR 50.59; or
| |
| : c. Otherwise approved by the Nuclear Regulatory Commission.
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| QUADRANT POWER TILT Tq shall be the maximum positive ratio of the power (Tq) generated in any quadrant minus the average quadrant power, to the average quadrant power.
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| RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) primary coolant of 2565.4 MWt.
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| REFUELING BORON REFUELING BORON CONCENTRATION shall be a Primary CONCENTRATION Coolant System boron concentration of 1720 ppm and sufficient to assure the reactor is subcritical by 5% with all control rods withdrawn.
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| Palisades Nuclear Plant 1.1-5 Amendment No. XXX
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| | |
| Definitions 1.1 1.1 Definitions SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
| |
| : a. All full length control rods (shutdown and regulating) are fully inserted except for the single rod of highest reactivity worth, which is assumed to be fully withdrawn.
| |
| However, with all full length control rods verified fully inserted by two independent means, it is not necessary to account for a stuck rod in the SDM calculation. With any full length control rods not capable of being fully inserted, the reactivity worth of these rods must be accounted for in the determination of SDM; and
| |
| : b. There is no change in part length rod position STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
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| THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the primary coolant.
| |
| TOTAL RADIAL FRT shall be the maximum ratio of the individual PEAKING FACTOR fuel pin power to the core average pin power integrated over (FRT) the total core height, including tilt.
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| Palisades Nuclear Plant 1.1-6 Amendment No. XXX
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| | |
| Definitions 1.1 Table 1.1-1 (page 1 of 1)
| |
| MODES
| |
| % RATED AVERAGE PRIMARY REACTIVITY THERMAL MODE TITLE COOLANT CONDITION (a)
| |
| POWER TEMPERATURE (keff)
| |
| (°F) 1 Power Operation 0.99 >5 NA 2 Startup 0.99 5 NA 3 Hot Standby < 0.99 NA 300 4 Hot Shutdown(b) < 0.99 NA 300 > Tave > 200 5 Cold Shutdown(b) < 0.99 NA 200 6 Refueling(c) NA NA NA (a) Excluding decay heat.
| |
| (b) All reactor vessel head closure bolts fully tensioned.
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| (c) One or more reactor vessel head closure bolts less than fully tensioned.
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| Palisades Nuclear Plant 1.1-7 Amendment No. XXX
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| | |
| Logical Connectors 1.2 1.0 USE AND APPLICATION 1.2 Logical Connectors PURPOSE The purpose of this section is to explain the meaning of logical connectors.
| |
| Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies. The only logical connectors that appear in TS are AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings.
| |
| BACKGROUND Several levels of logic may be used to state Required Actions. These levels are identified by the placement (or nesting) of the logical connectors and by the number assigned to each Required Action. The first level of logic is identified by the first digit of the number assigned to a Required Action and the placement of the logical connector in the first level of nesting (i.e., left justified with the number of the Required Action).
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| The successive levels of logic are identified by additional digits of the Required Action number and by successive indentions of the logical connectors.
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| When logical connectors are used to state a Condition, Completion Time, Surveillance, or Frequency, only the first level of logic is used, and the logical connector is left justified with the statement of the Condition, Completion Time, Surveillance, or Frequency.
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| Palisades Nuclear Plant 1.2-1 Amendment No. 189, 272, XXX
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| | |
| Logical Connectors 1.2 1.2 Logical Connectors EXAMPLES The following examples illustrate the use of logical connectors.
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| EXAMPLE 1.2-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Verify . . .
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| AND A.2 Restore . . .
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| In this example the logical connector AND is used to indicate that when in Condition A, both Required Actions A.1 and A.2 must be completed.
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| Palisades Nuclear Plant 1.2-2 Amendment No. 189, 272, XXX
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| | |
| Logical Connectors 1.2 1.2 Logical Connectors EXAMPLES EXAMPLE 1.2-2 (continued)
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| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Trip . . .
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| OR A.2.1 Verify . . .
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| AND A.2.2.1 Reduce . . .
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| OR A.2.2.2 Perform . . .
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| OR A.3 Align . . .
| |
| This example represents a more complicated use of logical connectors.
| |
| Required Actions A.1, A.2, and A.3 are alternative choices, only one of which must be performed as indicated by the use of the logical connector OR and the left justified placement. Any one of these three Actions may be chosen. If A.2 is chosen, then both A.2.1 and A.2.2 must be performed as indicated by the logical connector AND. Required Action A.2.2 is met by performing A.2.2.1 or A.2.2.2. The indented position of the logical connector OR indicates that A.2.2.1 and A.2.2.2 are alternative choices, only one of which must be performed.
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| Palisades Nuclear Plant 1.2-3 Amendment No. XXX
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| | |
| Completion Times 1.3 1.0 USE AND APPLICATION 1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.
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| BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the plant. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s).
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| DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the discovery of a situation (e.g.,
| |
| inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the plant is in a MODE or specified condition stated in the Applicability of the LCO.
| |
| Unless otherwise specified, the Completion Time begins when a senior licensed operator on the operating shift crew with responsibility for plant operations makes the determination that an LCO is not met and an ACTIONS Condition is entered. The "otherwise specified" exceptions are varied, such as a Required Action Note or Surveillance Requirement Note that provides an alternative time to perform specific tasks, such as testing, without starting the Completion Time. While utilizing the Note, should a Condition be applicable for any reason not addressed by the Note, the Completion Time begins. Should the time allowance in the Note be exceeded, the Completion Time begins at that point. The exceptions may also be incorporated into the Completion Time. For example, LCO 3.8.1, "AC Sources - Operating," Required Action B.2, requires declaring required feature(s) supported by an inoperable diesel generator, inoperable when the redundant required feature(s) are inoperable. The Completion Time states, "4 hours from discovery of Condition B concurrent with inoperability of redundant required feature(s)." In this case the Completion Time does not begin until the conditions in the Completion Time are satisfied.
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| Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the plant is not within the LCO Applicability.
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| If situations are discovered that require entry into more than one Condition at a time within a single LCO (multiple Conditions), the Required Actions for each Condition must be performed within the Palisades Nuclear Plant 1.3-1 Amendment No. 270, 272, XXX
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| | |
| Completion Times 1.3 1.3 Completion Times DESCRIPTION associated Completion Time. When in multiple Conditions, separate (continued) Completion Times are tracked for each Condition starting from the discovery of the situation that required entry into the Condition, unless otherwise specified.
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| Once a Condition has been entered, subsequent trains, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition, unless specifically stated. The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition, unless otherwise specified.
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| However, when a subsequent train, subsystem, component, or variable expressed in the Condition is discovered to be inoperable or not within limits, the Completion Time(s) may be extended. To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability:
| |
| : a. Must exist concurrent with the first inoperability; and
| |
| : b. Must remain inoperable or not within limits after the first inoperability is resolved.
| |
| The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either:
| |
| : a. The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours; or
| |
| : b. The stated Completion Time as measured from discovery of the subsequent inoperability.
| |
| The above Completion Time extensions do not apply to those Specifications that have exceptions that allow completely separate re-entry into the Condition (for each train, subsystem, component, or variable expressed in the Condition) and separate tracking of Completion Times based on this re-entry. These exceptions are stated in individual Specifications.
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| The above Completion Time extension does not apply to a Completion Time with a modified "time zero." This modified "time zero" may be expressed as a repetitive time (i.e., "once per 8 hours," where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase "from discovery . . ." Example 1.3-3 illustrates one use of Palisades Nuclear Plant 1.3-2 Amendment No. XXX
| |
| | |
| Completion Times 1.3 1.3 Completion Times DESCRIPTION this type of Completion Time. The 10 day Completion Time specified for (continued) Conditions A and B in Example 1.3-3 may not be extended.
| |
| EXAMPLES The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions.
| |
| EXAMPLE 1.3-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required B.1 Be in MODE 3. 6 hours Action and associated AND Completion Time not met. B.2 Be in MODE 5. 36 hours Condition B has two Required Actions. Each Required Action has its own separate Completion Time. Each Completion Time is referenced to the time that Condition B is entered.
| |
| The Required Actions of Condition B are to be in MODE 3 within 6 hours AND in MODE 5 within 36 hours. A total of 6 hours is allowed for reaching MODE 3 and a total of 36 hours (not 42 hours) is allowed for reaching MODE 5 from the time that Condition B was entered. If MODE 3 is reached within 3 hours, the time allowed for reaching MODE 5 is the next 33 hours because the total time allowed for reaching MODE 5 is 36 hours.
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| If Condition B is entered while in MODE 3, the time allowed for reaching MODE 5 is the next 36 hours.
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| Palisades Nuclear Plant 1.3-3 Amendment No. 270, 272, XXX
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| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-2 (continued)
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| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pump A.1 Restore pump to 7 days inoperable. OPERABLE status.
| |
| B. Required B.1 Be in MODE 3. 6 hours Action and associated AND Completion Time not met. B.2 Be in MODE 5. 36 hours When a pump is declared inoperable, Condition A is entered. If the pump is not restored to OPERABLE status within 7 days, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable pump is restored to OPERABLE status after Condition B is entered, Condition A and B are exited, and therefore, the Required Actions of Condition B may be terminated.
| |
| When a second pump is declared inoperable while the first pump is still inoperable, Condition A is not re-entered for the second pump.
| |
| LCO 3.0.3 is entered, since the ACTIONS do not include a Condition for more than one inoperable pump. The Completion Time clock for Condition A does not stop after LCO 3.0.3 is entered, but continues to be tracked from the time Condition A was initially entered.
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| While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has not expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition A.
| |
| While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition B.
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| Palisades Nuclear Plant 1.3-4 Amendment No. XXX
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| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-2 (continued)
| |
| The Completion Time for Condition B is tracked from the time the Condition A Completion Time expired.
| |
| On restoring one of the pumps to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first pump was declared inoperable. This Completion Time may be extended if the pump restored to OPERABLE status was the first inoperable pump. A 24 hour extension to the stated 7 days is allowed, provided this does not result in the second pump being inoperable for > 7 days.
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| Palisades Nuclear Plant 1.3-5 Amendment No. XXX
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| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued)
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Restore Function X 7 days Function X train to OPERABLE train status. AND inoperable.
| |
| 10 days from discovery of failure to meet the LCO B. One B.1 Restore Function Y 72 hours Function Y train to OPERABLE train status. AND inoperable.
| |
| 10 days from discovery of failure to meet the LCO C. One C.1 Restore Function X 12 hours Function X train to OPERABLE train status.
| |
| inoperable.
| |
| OR AND C.2 Restore Function Y One train to OPERABLE 12 hours Function Y status.
| |
| train inoperable.
| |
| Palisades Nuclear Plant 1.3-6 Amendment No. XXX
| |
| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued)
| |
| When one Function X train and one Function Y train are inoperable, Condition A and Condition B are concurrently applicable. The Completion Times for Condition A and Condition B are tracked separately for each train starting from the time each train was declared inoperable and the Condition was entered. A separate Completion Time is established for Condition C and tracked from the time the second train was declared inoperable (i.e., the time the situation described in Condition C was discovered).
| |
| If Required Action C.2 is completed within the specified Completion Time, Conditions B and C are exited. If the Completion Time for Required Action A.1 has not expired, operation may continue in accordance with Condition A. The remaining Completion Time in Condition A is measured from the time the affected train was declared inoperable (i.e., initial entry into Condition A).
| |
| The Completion Times of Conditions A and B are modified by a logical connector, with a separate 10 day Completion Time measured from the time it was discovered the LCO was not met. In this example, without the separate Completion Time, it would be possible to alternate between Conditions A, B, and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO. The separate Completion Time modified by the phrase "from discovery of failure to meet the LCO" is designed to prevent indefinite continued operation while not meeting the LCO. This Completion Time allows for an exception to the normal "time zero" for beginning the Completion Time "clock." In this instance, the Completion Time "time zero" is specified as commencing at the time the LCO was initially not met, instead of at the time the associated Condition was entered.
| |
| Palisades Nuclear Plant 1.3-7 Amendment No. XXX
| |
| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-4 (continued)
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve(s) to 4 hours valves OPERABLE status.
| |
| inoperable.
| |
| B. Required B.1 Be in MODE 3. 6 hours Action and associated AND Completion Time not met. B.2 Be in MODE 4. 30 hours A single Completion Time is used for any number of valves inoperable at the same time. The Completion Time associated with Condition A is based on the initial entry into Condition A and is not tracked on a per valve basis. Declaring subsequent valves inoperable, while Condition A is still in effect, does not trigger the tracking of separate Completion Times.
| |
| Once one of the valves has been restored to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first valve was declared inoperable. The Completion Time may be extended if the valve restored to OPERABLE status was the first inoperable valve. The Condition A Completion Time may be extended for up to 4 hours provided this does not result in any subsequent valve being inoperable for > 4 hours.
| |
| If the Completion Time of 4 hours (including the extension) expires while one or more valves are still inoperable, Condition B is entered.
| |
| Palisades Nuclear Plant 1.3-8 Amendment No. XXX
| |
| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-5 (continued)
| |
| ACTIONS
| |
| --------------------------------------------NOTE--------------------------------------------
| |
| Separate Condition entry is allowed for each inoperable valve.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve to 4 hours valves OPERABLE inoperable. status.
| |
| B. Required B.1 Be in MODE 3. 6 hours Action and associated AND Completion Time not met. B.2 Be in MODE 4. 12 hours The Note above the ACTIONS Table is a method of modifying how the Completion Time is tracked. If this method of modifying how the Completion Time is tracked was applicable only to a specific Condition, the Note would appear in that Condition rather than at the top of the ACTIONS Table.
| |
| The Note allows Condition A to be entered separately for each inoperable valve, and Completion Times tracked on a per valve basis. When a valve is declared inoperable, Condition A is entered and its Completion Time starts. If subsequent valves are declared inoperable, Condition A is entered for each valve and separate Completion Times start and are tracked for each valve.
| |
| If the Completion Time associated with a valve in Condition A expires, Condition B is entered for that valve. If the Completion Times associated with subsequent valves in Condition A expire, Condition B is entered separately for each valve and separate Completion Times start and are tracked for each valve. If a valve that caused entry into Condition B is restored to OPERABLE status, Condition B is exited for that valve.
| |
| Since the Note in this example allows multiple Condition entry and tracking of separate Completion Times, Completion Time extensions do not apply.
| |
| Palisades Nuclear Plant 1.3-9 Amendment No. XXX
| |
| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLE EXAMPLE 1.3-6 (continued)
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One channel A.1 Perform SR 3.x.x.x. Once per 8 hours inoperable.
| |
| OR A.2 Reduce THERMAL 8 hours POWER to 50% RTP.
| |
| B. Required B.1 Be in MODE 3. 6 hours Action and associated Completion Time not met.
| |
| Entry into Condition A offers a choice between Required Action A.1 or A.2. Required Action A.1 has a "once per" Completion Time, which qualifies for the 25% extension, per SR 3.0.2, to each performance after the initial performance. The initial 8 hour interval of Required Action A.1 begins when Condition A is entered and the initial performance of Required Action A.1 must be complete within the first 8 hour interval. If Required Action A.1 is followed and the Required Action is not met within the Completion Time (plus the extension allowed by SR 3.0.2),
| |
| Condition B is entered. If Required Action A.2 is followed and the Completion Time of 8 hours is not met, Condition B is entered.
| |
| If after entry into Condition B, Required Action A.1 or A.2 is met, Condition B is exited and operation may then continue in Condition A.
| |
| Palisades Nuclear Plant 1.3-10 Amendment No. XXX
| |
| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued)
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Verify affected 1 hour subsystem subsystem isolated.
| |
| inoperable. AND Once per 8 hours thereafter AND A.2 Restore subsystem 72 hours to OPERABLE status.
| |
| B. Required B.1 Be in MODE 3. 6 hours Action and associated AND Completion Time not met. B.2 Be in MODE 5. 36 hours Required Action A.1 has two Completion Times. The 1 hour Completion Time begins at the time the Condition is entered and each "Once per 8 hours thereafter" interval begins upon performance of Required Action A.1.
| |
| If after Condition A is entered, Required Action A.1 is not met within either the initial 1 hour or any subsequent 8 hour interval from the previous performance (plus the extension allowed by SR 3.0.2),
| |
| Condition B is entered.
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| Palisades Nuclear Plant 1.3-11 Amendment No. XXX
| |
| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued)
| |
| The Completion Time clock for Condition A does not stop after Condition B is entered, but continues from the time Condition A was initially entered. If Required Action A.1 is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not expired.
| |
| IMMEDIATE When "Immediately" is used as a Completion Time, the Required Action COMPLETION TIME should be pursued without delay and in a controlled manner.
| |
| Palisades Nuclear Plant 1.3-12 Amendment No. 270, 272, XXX
| |
| | |
| Frequency 1.4 1.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.
| |
| DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR.
| |
| The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR)
| |
| Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR, as well as certain Notes in the Surveillance column that modify performance requirements.
| |
| Sometimes special situations dictate when the requirements of a Surveillance are to be met. They are otherwise stated conditions allowed by SR 3.0.1. They may be stated as clarifying Notes in the Surveillance, as part of the Surveillance, or both.
| |
| Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated LCO is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only required when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction.
| |
| The use of "met and performed in these instances conveys specific meanings. A Surveillance is "met only when the acceptance criteria are satisfied. Known failure of the requirements of a Surveillance, even without a Surveillance specifically being performed, constitutes a Surveillance not "met. Performance refers only to the requirement to specifically determine the ability to meet the acceptance criteria.
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| Palisades Nuclear Plant 1.4-1 Amendment No. 231, 272, XXX
| |
| | |
| Frequency 1.4 1.4 Frequency DESCRIPTION Some Surveillances contain notes that modify the Frequency of (continued) performance or the conditions during which the acceptance criteria must be satisfied. For these Surveillances, the MODE-entry restrictions of SR 3.0.4 may not apply. Such a Surveillance is not required to be performed prior to entering a MODE or other specified condition in the Applicability of the associated LCO if any of the following three conditions are satisfied:
| |
| : a. The Surveillance is not required to be met in the MODE or other specified condition to be entered; or
| |
| : b. The Surveillance is required to be met in the MODE or other specified condition to be entered, but has been performed within the specified Frequency (i.e., it is current) and is known not to be failed; or
| |
| : c. The Surveillance is required to be met, but not performed, in the MODE or other specified condition to be entered, and is known not to be failed.
| |
| Examples 1.4-3, 1.4-4, 1.4-5, and 1.4-6 discuss these special situations.
| |
| EXAMPLES The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LCO (LCO not shown) is MODES 1, 2, and 3.
| |
| Palisades Nuclear Plant 1.4-2 Amendment No. XXX
| |
| | |
| Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-1 (continued)
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perform CHANNEL CHECK. 12 hours Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours) during which the associated Surveillance must be performed at least one time. Performance of the Surveillance initiates the subsequent interval. Although the Frequency is stated as 12 hours, an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for operational flexibility. The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment is inoperable, a variable is outside specified limits, or the plant is outside the Applicability of the LCO). If the interval specified by SR 3.0.2 is exceeded while the facility is in a MODE or other specified condition in the Applicability of the LCO, and the performance of the Surveillance is not otherwise modified (refer to Example 1.4-3), then SR 3.0.3 becomes applicable.
| |
| If the interval as specified by SR 3.0.2 is exceeded while the plant is not in a Mode or other specified condition in the Applicability of the LCO for which performance of the SR is required, then SR 3.0.4 becomes applicable. The Surveillance must be performed within the Frequency requirements of SR 3.0.2, as modified by SR 3.0.3, prior to entry into the MODE or other specified condition or the LCO is considered not met (in accordance with SR 3.0.1) and LCO 3.0.4 becomes applicable.
| |
| Palisades Nuclear Plant 1.4-3 Amendment No. 231, 272, XXX
| |
| | |
| Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-2 (continued)
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits. Once within 12 hours after 25% RTP AND 24 hours thereafter Example 1.4-2 has two Frequencies. The first is a one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency requirements must be met. Each time reactor power is increased from a power level
| |
| < 25% RTP to 25% RTP, the Surveillance must be performed within 12 hours.
| |
| The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "AND").
| |
| This type of Frequency does not qualify for the extension allowed by SR 3.0.2. "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to
| |
| < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.
| |
| Palisades Nuclear Plant 1.4-4 Amendment No. 231, 272, XXX
| |
| | |
| Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued)
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
| |
| ----------------------------NOTE----------------------------
| |
| Not required to be performed until 12 hours after 25% RTP.
| |
| Perform channel adjustment. 7 days The interval continues, whether or not the plant operation is < 25% RTP between performances.
| |
| As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours after power reaches 25% RTP to perform the Surveillance.
| |
| The Surveillance is still considered to be performed within the "specified Frequency." The interval continues, whether or not the plant operation is
| |
| < 25% RTP between performances. Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours with power 25% RTP.
| |
| Once the plant reaches 25% RTP, 12 hours would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour interval, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.
| |
| Palisades Nuclear Plant 1.4-5 Amendment No. XXX
| |
| | |
| Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-4 (continued)
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
| |
| ----------------------------NOTE----------------------------
| |
| Only required to be met in MODE 1.
| |
| Verify leakage rates are within limits. 24 hours Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the plant is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an otherwise stated exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour interval (plus the extension allowed by SR 3.0.2), but the plant was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR.
| |
| Palisades Nuclear Plant 1.4-6 Amendment No. XXX
| |
| | |
| Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-5 (continued)
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
| |
| ----------------------------NOTE----------------------------
| |
| Only required to be performed in MODE 1.
| |
| Perform complete cycle of the valve. 7 days The interval continues, whether or not the plant operation is in MODE 1, 2, or 3 (the assumed Applicability of the associated LCO) between performances.
| |
| As the Note modifies the required performance of the Surveillance, the Note is construed to be part of the specified Frequency. Should the 7 day interval be exceeded while operation is not in MODE 1, this Note allows entry into and operation in MODES 2 and 3 to perform the Surveillance. The Surveillance is still considered to be performed within the specified Frequency if completed prior to entering MODE 1.
| |
| Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was not in MODE 1, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not result in entry into MODE 1.
| |
| Once the plant reaches MODE 1, the requirement for the Surveillance to be performed within its specified Frequency applies and would require that the Surveillance had been performed. If the Surveillance were not performed prior to entering MODE 1, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.
| |
| Palisades Nuclear Plant 1.4-7 Amendment No. XXX
| |
| | |
| Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-6 (continued)
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
| |
| ----------------------------NOTE----------------------------
| |
| Not required to be met in MODE 3.
| |
| Verify parameter is within limits. 24 hours Example 1.4-6 specifies that the requirements of this Surveillance do not have to be met while the plant is in MODE 3 (the assumed Applicability of the associated LCO is MODES 1, 2, and 3). The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an otherwise stated exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour interval (plus the extension allowed by SR 3.0.2), but the plant was in MODE 3, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES to enter MODE 3, even with the 24 hour Frequency exceeded, provided the MODE change does not result in entry into MODE 2. Prior to entering MODE 2 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR.
| |
| Palisades Nuclear Plant 1.4-8 Amendment No. XXX
| |
| | |
| SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, the Departure from Nucleate Boiling Ratio (DNBR) shall be maintained at or above the following DNB correlation safety limits:
| |
| Correlation Safety Limit XNB 1.17 ANFP 1.154 HTP 1.141 2.1.1.2 In MODES 1 and 2, the peak Linear Heat Rate (LHR) (adjusted for fuel rod dynamics) shall be maintained at 21.0 kW/ft.
| |
| 2.1.2 Primary Coolant System (PCS) Pressure SL In MODES 1, 2, 3, 4, 5, and 6, the PCS pressure shall be maintained at 2750 psia.
| |
| 2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour.
| |
| 2.2.2 If SL 2.1.2 is violated:
| |
| 2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour.
| |
| 2.2.2.2 In MODE 3, 4, 5, or 6, restore compliance within 5 minutes.
| |
| Palisades Nuclear Plant 2.0-1 Amendment No. XXX
| |
| | |
| LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2, LCO 3.0.7, LCO 3.0.8, and LCO 3.0.9.
| |
| LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6.
| |
| If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required, unless otherwise stated.
| |
| LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the plant shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the plant, as applicable, in:
| |
| : a. MODE 3 within 7 hours;
| |
| : b. MODE 4 within 31 hours; and
| |
| : c. MODE 5 within 37 hours.
| |
| Exceptions to this Specification are stated in the individual Specifications.
| |
| Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required.
| |
| LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4.
| |
| LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:
| |
| : a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; Palisades Nuclear Plant 3.0-1 Amendment No. 270, 272, XXX
| |
| | |
| LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.4 (continued)
| |
| : b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate (exceptions to this Specification are stated in the individual Specifications); or
| |
| : c. When an allowance is stated in the individual value, parameter, or other Specification.
| |
| This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the plant.
| |
| LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.
| |
| LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.13, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.
| |
| Palisades Nuclear Plant 3.0-2 Amendment No. XXX
| |
| | |
| LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.7 Special Test Exception (STE) LCOs in each applicable LCO section allow specified Technical Specifications (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with STE LCOs is optional. When an STE LCO is desired to be met but is not met, the ACTIONS of the STE LCO shall be met. When an STE LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with the other applicable Specifications.
| |
| LCO 3.0.8 When one or more required snubbers are unable to perform their associated support function(s), any affected supported LCO(s) are not required to be declared not met solely for this reason if risk is assessed and managed, and:
| |
| : a. the snubbers not able to perform their associated support function(s) are associated with only one train or subsystem of a multiple train or subsystem supported system or are associated with a single train or subsystem supported system and are able to perform their associated support function within 72 hours; or
| |
| : b. the snubbers not able to perform their associated support function(s) are associated with more than one train or subsystem of a multiple train or subsystem supported system and are able to perform their associated support function within 12 hours.
| |
| At the end of the specified period the required snubbers must be able to perform their associated support function(s), or the affected supported system LCO(s) shall be declared not met.
| |
| LCO 3.0.9 When one or more required barriers are unable to perform their related support function(s), any supported system LCO(s) are not required to be declared not met solely for this reason for up to 30 days provided that at least one train or subsystem of the supported system is OPERABLE and supported by barriers capable of providing their related support function(s), and risk is assessed and managed. This specification may be concurrently applied to more than one train or subsystem of a multiple train or subsystem supported system provided at least one train or subsystem of the supported system is OPERABLE and the barriers supporting each of these trains or subsystems provide their related support function(s) for different categories of initiating events.
| |
| Palisades Nuclear Plant 3.0-3 Amendment No. XXX
| |
| | |
| LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.9 (continued)
| |
| If the required OPERABLE train or subsystem becomes inoperable while this specification is in use, it must be restored to OPERABLE status within 24 hours or the provisions of this specification cannot be applied to the trains or subsystems supported by the barriers that cannot perform their related support function(s).
| |
| At the end of the specified period, the required barriers must be able to perform their related support function(s) or the supported system LCO(s) shall be declared not met.
| |
| Palisades Nuclear Plant 3.0-4 Amendment No. XXX
| |
| | |
| SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR.
| |
| Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.
| |
| SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.
| |
| For Frequencies specified as "once," the above interval extension does not apply.
| |
| If a Completion Time requires periodic performance on a "once per . . ."
| |
| basis, the above Frequency extension applies to each performance after the initial performance.
| |
| Exceptions to this Specification are stated in the individual Specifications.
| |
| SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. The delay period is only applicable when there is a reasonable expectation the surveillance will be met when performed. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.
| |
| If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
| |
| When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
| |
| Palisades Nuclear Plant 3.0-5 Amendment No. 270, 272, XXX
| |
| | |
| SR Applicability 3.0 3.0 SR APPLICABILITY SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.
| |
| This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the plant.
| |
| Palisades Nuclear Plant 3.0-6 Amendment No. 270, 272, XXX
| |
| | |
| SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)
| |
| LCO 3.1.1 SDM shall be within the limits specified in the COLR.
| |
| APPLICABILITY: MODE 3, 4, and 5.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit. A.1 Initiate boration to 15 minutes restore SDM to within limit.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM to be within limits. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.1.1-1 Amendment No. XXX
| |
| | |
| Reactivity Balance 3.1.2 3.1 REACTIVITY CONTROL SYSTEMS 3.1.2 Reactivity Balance LCO 3.1.2 The core reactivity balance shall be within +/- 1% of predicted values.
| |
| APPLICABILITY: MODE 1.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Core reactivity balance not A.1 Re-evaluate core design 7 days within limit. and safety analysis and determine that the reactor core is acceptable for continued operation.
| |
| AND A.2 Establish appropriate 7 days operating restrictions and SRs.
| |
| B. Required Action and B.1 Be in MODE 2. 6 hours associated Completion Time not met.
| |
| Palisades Nuclear Plant 3.1.2-1 Amendment No. XXX
| |
| | |
| Reactivity Balance 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 ------------------------------NOTE----------------------------
| |
| The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 Effective Full Power Days (EFPD) after each fuel loading.
| |
| Verify overall core reactivity balance is within Prior to entering
| |
| +/- 1% of predicted values. MODE 1 after each fuel loading AND
| |
| ----------NOTE---------
| |
| Only required after initial 60 EFPD In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.1.2-2 Amendment No. XXX
| |
| | |
| MTC 3.1.3 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Moderator Temperature Coefficient (MTC)
| |
| LCO 3.1.3 The MTC shall be maintained less positive than 0.5 E-4 /°F at 2%
| |
| RATED THERMAL POWER (RTP).
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. MTC not within limits. A.1 Be in MODE 3. 6 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Verify MTC is less positive than 0.5 E-4 /°F at Prior to exceeding 2% RTP. 2% RTP after each fuel loading Palisades Nuclear Plant 3.1.3-1 Amendment No. XXX
| |
| | |
| Control Rod Alignment 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Alignment LCO 3.1.4 All control rods, including their position indication channels, shall be OPERABLE and aligned to within 8 inches of all other rods in their respective group, and the control rod position deviation alarm shall be OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One channel of rod position A.1 Perform SR 3.1.4.1 Once within indication inoperable for (rod position 15 minutes following one or more control rods. verification). any rod motion in that group B. Rod position deviation B.1 Perform SR 3.1.4.1 Once within alarm inoperable. (rod position 15 minutes of verification). movement of any control rod C. One control rod misaligned C.1 Perform SR 3.2.2.1 2 hours by > 8 inches. (peaking factor verification).
| |
| OR C.2 Reduce THERMAL 2 hours POWER to 75% RTP.
| |
| D. One full-length control rod D.1 Restore control rod to Prior to entering immovable, but trippable. OPERABLE status. MODE 2 following next MODE 3 entry Palisades Nuclear Plant 3.1.4-1 Amendment No. XXX
| |
| | |
| Control Rod Alignment 3.1.4 ACTIONS E. Required Action and E.1 Be in MODE 3. 6 hours associated Completion Time not met.
| |
| OR One or more control rods inoperable for reasons other than Condition D.
| |
| OR Two or more control rods misaligned by > 8 inches.
| |
| OR Both rod position indication channels inoperable for one or more control rods.
| |
| Palisades Nuclear Plant 3.1.4-2 Amendment No. XXX
| |
| | |
| Control Rod Alignment 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify the position of each control rod to be within In accordance with 8 inches of all other control rods in its group. the Surveillance Frequency Control Program SR 3.1.4.2 Perform a CHANNEL CHECK of the control rod In accordance with position indication channels. the Surveillance Frequency Control Program SR 3.1.4.3 Verify control rod freedom of movement by moving In accordance with each individual full-length control rod that is not fully the Surveillance inserted into the reactor core 6 inches in either Frequency Control direction. Program SR 3.1.4.4 Verify the rod position deviation alarm is In accordance with OPERABLE. the Surveillance Frequency Control Program SR 3.1.4.5 Perform a CHANNEL CALIBRATION of the control In accordance with rod position indication channels. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.1.4-3 Amendment No. XXX
| |
| | |
| Control Rod Alignment 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.6 Verify each full-length control rod drop time is Prior to reactor 2.5 seconds. criticality, after each reinstallation of the reactor head Palisades Nuclear Plant 3.1.4-4 Amendment No. XXX
| |
| | |
| Shutdown and Part-Length Rod Group Insertion Limits 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Shutdown and Part-Length Control Rod Group Insertion Limits LCO 3.1.5 All shutdown and part-length rod groups shall be withdrawn to 128 inches.
| |
| APPLICABILITY: MODE 1, MODE 2 with any regulating rod withdrawn above 5 inches.
| |
| --------------------------------------------NOTE--------------------------------------------
| |
| This LCO is not applicable while performing SR 3.1.4.3 (rod exercise test).
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more shutdown or A.1 Declare affected control Immediately part-length rods not within rod(s) inoperable and limit. enter the applicable Conditions and Required Actions of LCO 3.1.4.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown and part-length rod group is In accordance with withdrawn 128 inches. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.1.5-1 Amendment No. XXX
| |
| | |
| Regulating Rod Group Position Limits 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Regulating Rod Group Position Limits LCO 3.1.6 The Power Dependent Insertion Limit (PDIL) alarm circuit and the Control Rod Out Of Sequence (CROOS) alarm circuit shall be OPERABLE, and the regulating rod groups shall be limited to the withdrawal sequence, overlap, and insertion limits specified in the COLR.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ----------------------------------------------NOTE--------------------------------------------
| |
| This LCO is not applicable while performing SR 3.1.4.3 (rod exercise test).
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Regulating rod groups A.1 Restore regulating rod 2 hours inserted beyond the groups to within limits.
| |
| insertion limit.
| |
| OR A.2 Reduce THERMAL 2 hours POWER to less than or equal to the fraction of RTP allowed by the regulating rod group position and insertion limits specified in the COLR.
| |
| Palisades Nuclear Plant 3.1.6-1 Amendment No. XXX
| |
| | |
| Regulating Rod Group Position Limits 3.1.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Regulating rod groups not B.1 Restore regulating rod 2 hours within sequence or overlap groups to within limits. appropriate sequence and overlap limits.
| |
| C. PDIL or CROOS alarm C.1 Perform SR 3.1.6.1 Once within circuit inoperable. (group position 15 minutes following verification). any rod motion D. Required Action and D.1 Be in MODE 3. 6 hours associated Completion Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify each regulating rod group is within its In accordance with withdrawal sequence, overlap, and insertion limits. the Surveillance Frequency Control Program SR 3.1.6.2 Verify PDIL alarm circuit is OPERABLE. In accordance with the Surveillance Frequency Control Program SR 3.1.6.3 Verify CROOS alarm circuit is OPERABLE. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.1.6-2 Amendment No. .XXX
| |
| | |
| Special Test Exceptions (STE) 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Special Test Exceptions (STE)
| |
| LCO 3.1.7 During the performance of PHYSICS TESTS, the requirements of LCO 3.1.4, "Control Rod Alignment";
| |
| LCO 3.1.5, "Shutdown and Part-Length Rod Group Insertion Limits";
| |
| LCO 3.1.6, "Regulating Rod Group Position Limits"; and LCO 3.4.2, "PCS Minimum Temperature for Criticality may be suspended, provided:
| |
| : a. THERMAL POWER is 2% RTP;
| |
| : b. 1% shutdown reactivity, based on predicted control rod worth, is available for trip insertion; and
| |
| : c. Tave is 500°F.
| |
| APPLICABILITY: MODE 2 during PHYSICS TESTS.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. THERMAL POWER not A.1 Reduce THERMAL 15 minutes within limit. POWER to within limit.
| |
| B. Shutdown reactivity not B.1 Initiate boration to 15 minutes within limit. restore shutdown reactivity to within limit.
| |
| C. Tave not within limit. C.1 Restore Tave to within 15 minutes limit.
| |
| Palisades Nuclear Plant 3.1.7-1 Amendment No. XXX
| |
| | |
| Special Test Exceptions (STE) 3.1.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Suspend PHYSICS 1 hour associated Completion TESTS.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify THERMAL POWER is 2% RTP. In accordance with the Surveillance Frequency Control Program SR 3.1.7.2 Verify Tave is 500°F. In accordance with the Surveillance Frequency Control Program SR 3.1.7.3 Verify 1% shutdown reactivity is available for trip In accordance with insertion. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.1.7-2 Amendment No. XXX
| |
| | |
| LHR 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Linear Heat Rate (LHR)
| |
| LCO 3.2.1 LHR shall be within the limits specified in the COLR, and the Incore Alarm System or Excore Monitoring System shall be OPERABLE to monitor LHR.
| |
| APPLICABILITY: MODE 1 with THERMAL POWER > 25% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LHR, as determined by the A.1 Restore LHR to within 1 hour automatic Incore Alarm limits.
| |
| System, not within limits specified in the COLR, as indicated by four or more coincident incore channels.
| |
| OR LHR, as determined by the Excore Monitoring System, not within limits specified in the COLR.
| |
| OR LHR, as determined by manual incore detector readings, not within limits specified in the COLR.
| |
| Palisades Nuclear Plant 3.2.1-1 Amendment No. XXX
| |
| | |
| LHR 3.2.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Incore Alarm and Excore B.1 Reduce THERMAL 2 hours Monitoring Systems POWER to 85% RTP.
| |
| inoperable for monitoring LHR. AND B.2 Verify LHR is within 4 hours limits using manual incore readings. AND Once per 2 hours thereafter C. Required Action and C.1 Reduce THERMAL 4 hours associated Completion POWER to 25% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 -------------------------------NOTE---------------------------
| |
| Only required to be met when the Incore Alarm System is being used to monitor LHR.
| |
| Verify LHR is within the limits specified in the In accordance with COLR. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.2.1-2 Amendment No. XXX
| |
| | |
| LHR 3.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.2 ------------------------------NOTE----------------------------
| |
| Only required to be met when the Incore Alarm System is being used to monitor LHR.
| |
| Adjust incore alarm setpoints based on a Prior to operation measured power distribution. > 50% RTP after each fuel loading AND In accordance with the Surveillance Frequency Control Program SR 3.2.1.3 -------------------------------NOTE---------------------------
| |
| Only required to be met when the Excore Monitoring System is being used to monitor LHR.
| |
| Verify measured ASI has been within 0.05 of Prior to each initial target ASI for last 24 hours. use of Excore Monitoring System to monitor LHR SR 3.2.1.4 -------------------------------NOTE---------------------------
| |
| Only required to be met when the Excore Monitoring System is being used to monitor LHR.
| |
| Verify THERMAL POWER is less than the APL. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.2.1-3 Amendment No. XXX
| |
| | |
| LHR 3.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.5 -------------------------------NOTE---------------------------
| |
| Only required to be met when the Excore Monitoring System is being used to monitor LHR.
| |
| Verify measured ASI is within 0.05 of target ASI. In accordance with the Surveillance Frequency Control Program SR 3.2.1.6 -------------------------------NOTE---------------------------
| |
| Only required to be met when the Excore Monitoring System is being used to monitor LHR.
| |
| Verify Tq 0.03. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.2.1-4 Amendment No. XXX
| |
| | |
| Radial Peaking 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 TOTAL RADIAL PEAKING FACTOR (FRT)
| |
| LCO 3.2.2 FRT shall be within the limits specified in the COLR.
| |
| APPLICABILITY: MODE 1 with THERMAL POWER > 25% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. FRT not within limits A.1 Restore FRT to within 6 hours specified in the COLR. limits.
| |
| B. Required Action and B.1 Reduce THERMAL 4 hours associated Completion POWER to 25% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify FRT is within limits specified in the COLR. Prior to operation
| |
| > 50% RTP after each fuel loading AND In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.2.2-1 Amendment No. XXX
| |
| | |
| Tq 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 QUADRANT POWER TILT (Tq)
| |
| LCO 3.2.3 Tq shall be 0.05.
| |
| APPLICABILITY: MODE 1 with THERMAL POWER > 25% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Tq > 0.05. A.1 Verify FRT is within the 2 hours limits of LCO 3.2.2, TOTAL RADIAL AND PEAKING FACTOR .
| |
| Once per 8 hours thereafter B. Tq > 0.10. B.1 Reduce THERMAL 4 hours POWER to < 50% RTP.
| |
| C. Required Action and C.1 Reduce THERMAL 4 hours associated Completion POWER to 25% RTP.
| |
| Time not met.
| |
| OR Tq > 0.15.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify Tq is < 0.05. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.2.3-1 Amendment No. XXX
| |
| | |
| ASI 3.2.4 3.2 POWER DISTRIBUTION LIMITS 3.2.4 AXIAL SHAPE INDEX (ASI)
| |
| LCO 3.2.4 The ASI shall be within the limits specified in the COLR.
| |
| APPLICABILITY: MODE 1 with THERMAL POWER > 25% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ASI not within limits A.1 Restore ASI to within 2 hours specified in COLR. limits.
| |
| B. Required Action and B.1 Reduce THERMAL 4 hours associated Completion POWER to 25% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 Verify ASI is within limits specified in the COLR. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.2.4-1 Amendment No. XXX
| |
| | |
| RPS Instrumentation 3.3.1 3.3 INSTRUMENTATION 3.3.1 Reactor Protective System (RPS) Instrumentation LCO 3.3.1 Four RPS trip units, associated instrument channels, and associated Zero Power Mode (ZPM) Bypass removal channels for each Function in Table 3.3.1-1 shall be OPERABLE.
| |
| APPLICABILITY: According to Table 3.3.1-1.
| |
| ACTIONS
| |
| -----------------------------------------------------------NOTE----------------------------------------------------------
| |
| Separate Condition entry is allowed for each Function.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. -------------NOTE-------------- A.1 Place affected trip unit in 7 days Not applicable to High trip.
| |
| Startup Rate, Loss of Load, or ZPM Bypass Removal Functions.
| |
| One or more Functions with one RPS trip unit or associated instrument channel inoperable.
| |
| B. One High Startup Rate trip B.1 Restore trip unit and Prior to entering unit or associated associated instrument MODE 2 from instrument channel channel to OPERABLE MODE 3 inoperable. status.
| |
| Palisades Nuclear Plant 3.3.1-1 Amendment No. XXX
| |
| | |
| RPS Instrumentation 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One Loss of Load trip unit C.1 Restore trip unit and Prior to increasing or associated instrument associated instrument THERMAL POWER channel inoperable. channel to OPERABLE to 17% RTP status. following entry into MODE 3 D. One or more ZPM Bypass D.1 Remove the affected Immediately Removal channels ZPM Bypasses.
| |
| inoperable.
| |
| OR D.2 Declare affected trip Immediately units inoperable.
| |
| E. -------------NOTE-------------- E.1 Place one trip unit in trip. 1 hour Not applicable to ZPM Bypass Removal AND Function.
| |
| ----------------------------------- -----------------NOTE-------------------
| |
| Not applicable to High Startup One or more Functions Rate or Loss of Load Functions.
| |
| with two RPS trip units or ---------------------------------------------
| |
| associated instrument channels inoperable. E.2 Restore one trip unit and 7 days associated instrument channel to OPERABLE status.
| |
| F. Two power range F.1 Restrict THERMAL 2 hours channels inoperable. POWER to 70% RTP.
| |
| Palisades Nuclear Plant 3.3.1-2 Amendment No. XXX
| |
| | |
| RPS Instrumentation 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and G.1 Be in MODE 3. 6 hours associated Completion Time not met. AND OR G.2.1 Verify no more than one 6 hours full-length control rod is Control room ambient air capable of being temperature > 90oF. withdrawn.
| |
| OR G.2.2 Verify PCS boron 6 hours concentration is at REFUELING BORON CONCENTRATION.
| |
| SURVEILLANCE REQUIREMENTS
| |
| -----------------------------------------------------------NOTE----------------------------------------------------------
| |
| Refer to Table 3.3.1-1 to determine which SR shall be performed for each Function.
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform a CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.1.2 Verify control room temperature is 90°F. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.3.1-3 Amendment No. XXX
| |
| | |
| RPS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.3 -----------------------------NOTE-----------------------------
| |
| Not required to be performed until 12 hours after THERMAL POWER is 15% RTP.
| |
| Perform calibration (heat balance only) and adjust In accordance with the power range excore and T power channels to the Surveillance agree with calorimetric calculation if the absolute Frequency Control difference is 1.5%. Program SR 3.3.1.4 -----------------------------NOTE-----------------------------
| |
| Not required to be performed until 12 hours after THERMAL POWER is 25% RTP.
| |
| Calibrate the power range excore channels using In accordance with the incore detectors. the Surveillance Frequency Control Program SR 3.3.1.5 Perform a CHANNEL FUNCTIONAL TEST and In accordance with verify the Thermal Margin Monitor Constants. the Surveillance Frequency Control Program SR 3.3.1.6 Perform a calibration check of the power range In accordance with excore channels with a test signal. the Surveillance Frequency Control Program SR 3.3.1.7 Perform a CHANNEL FUNCTIONAL TEST of High Once within 7 days Startup Rate and Loss of Load Functions. prior to each reactor startup Palisades Nuclear Plant 3.3.1-4 Amendment No. XXX
| |
| | |
| RPS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.8 -----------------------------NOTE-----------------------------
| |
| Neutron detectors are excluded from the CHANNEL CALIBRATION.
| |
| Perform a CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.3.1-5 Amendment No. XXX
| |
| | |
| RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 2)
| |
| Reactor Protective System Instrumentation APPLICABLE SURVEILLANCE ALLOWABLE FUNCTION MODES REQUIREMENTS VALUE
| |
| : 1. Variable High Power Trip 1,2,3(a),4(a),5(a) SR 3.3.1.1 15% RTP above SR 3.3.1.2 current THERMAL SR 3.3.1.3 POWER with a SR 3.3.1.4 minimum of 30%
| |
| SR 3.3.1.5 RTP and a SR 3.3.1.6 maximum of SR 3.3.1.8 109.4% RTP
| |
| : 2. High Startup Rate Trip(b) 1,2 SR 3.3.1.1 NA SR 3.3.1.7 SR 3.3.1.8
| |
| : 3. Low Primary Coolant System Flow Trip(c) 1,2,3(a),4(a),5(a) SR 3.3.1.1 95%
| |
| SR 3.3.1.5 SR 3.3.1.8
| |
| : 4. Low Steam Generator A Level Trip 1,2,3(a),4(a),5(a) SR 3.3.1.1 25.9% narrow SR 3.3.1.5 range SR 3.3.1.8
| |
| : 5. Low Steam Generator B Level Trip 1,2,3(a),4(a),5(a) SR 3.3.1.1 25.9% narrow SR 3.3.1.5 range SR 3.3.1.8
| |
| : 6. Low Steam Generator A Pressure Trip(c) 1,2,3(a),4(a),5(a) SR 3.3.1.1 500 psia SR 3.3.1.5 SR 3.3.1.8
| |
| : 7. Low Steam Generator B Pressure Trip(c) 1,2,3(a),4(a),5(a) SR 3.3.1.1 500 psia SR 3.3.1.5 SR 3.3.1.8
| |
| : 8. High Pressurizer Pressure Trip 1,2,3(a),4(a),5(a) SR 3.3.1.1 2255 psia SR 3.3.1.5 SR 3.3.1.8 (a) With more than one full-length control rod capable of being withdrawn and PCS boron concentration less than REFUELING BORON CONCENTRATION.
| |
| (b) Trip may be bypassed when Wide Range Power is < 1E-4% RTP or when THERMAL POWER is > 13% RTP.
| |
| (c) Trips may be bypassed when Wide Range Power is < 1E-4% RTP. Bypass shall be automatically removed when Wide Range Power is 1E-4% RTP.
| |
| Palisades Nuclear Plant 3.3.1-6 Amendment No. XXX
| |
| | |
| RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 2)
| |
| Reactor Protective System Instrumentation APPLICABLE SURVEILLANCE ALLOWABLE FUNCTION MODES REQUIREMENTS VALUE
| |
| : 9. Thermal Margin/
| |
| Low Pressure Trip(c) 1,2,3(a),4(a),5(a) SR 3.3.1.1 Table 3.3.1-2 SR 3.3.1.2 SR 3.3.1.3 SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.8
| |
| : 10. Loss of Load Trip 1(d) SR 3.3.1.7 NA SR 3.3.1.8
| |
| : 11. Containment High Pressure Trip 1,2,3(a),4(a),5(a) SR 3.3.1.5 3.70 psig SR 3.3.1.8
| |
| : 12. Zero Power Mode Bypass Automatic Removal 1,2,3(a),4(a),5(a) SR 3.3.1.8 NA (a) With more than one full-length control rod capable of being withdrawn and PCS boron concentration less than REFUELING BORON CONCENTRATION.
| |
| (c) Trips may be bypassed when Wide Range Power is < 1E-4% RTP. Bypass shall be automatically removed when Wide Range Power is 1E-4% RTP.
| |
| (d) When THERMAL POWER is 17% RTP.
| |
| Palisades Nuclear Plant 3.3.1-7 Amendment No. XXX
| |
| | |
| RPS Instrumentation 3.3.1 Table 3.3.1-2 (page 1 of 1)
| |
| Thermal Margin/Low Pressure Trip Function Allowable Value The Allowable Value for the Thermal Margin/Low Pressure Trip, Ptrip, is the higher of two values, Pmin and Pvar, both in psia:
| |
| Pmin = 1750 Pvar = 2012(QA)(QR1) + 17.0(Tin) - 9559 Where:
| |
| QA = - 0.720(ASI) + 1.028; when - 0.628 ASI < - 0.100 QA = - 0.333(ASI) + 1.067; when - 0.100 ASI < + 0.200 QA = + 0.375(ASI) + 0.925; when + 0.200 ASI + 0.565 ASI = Measured ASI when Q 0.0625 ASI = 0.0 when Q < 0.0625 QR1 = 0.412(Q) + 0.588; when Q 1.0 QR1 = Q; when Q > 1.0 Q = THERMAL POWER/RATED THERMAL POWER Tin = Maximum primary coolant inlet temperature, in °F ASI, Tin, and Q are the existing values as measured by the associated instrument channel.
| |
| Palisades Nuclear Plant 3.3.1-8 Amendment No. XXX
| |
| | |
| RPS Logic and Trip Initiation 3.3.2 3.3 INSTRUMENTATION 3.3.2 Reactor Protective System (RPS) Logic and Trip Initiation LCO 3.3.2 Six channels of RPS Matrix Logic, four channels of RPS Trip Initiation Logic, and two channels of RPS Manual Trip shall be OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2, MODES 3, 4, and 5, with more than one full-length control rod capable of being withdrawn and Primary Coolant System (PCS) boron concentration less than REFUELING BORON CONCENTRATION.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One Matrix Logic channel A.1 Restore channel to 48 hours inoperable. OPERABLE status.
| |
| B. One channel of Trip B.1 De-energize the affected 1 hour Initiation Logic inoperable. clutch power supplies.
| |
| C. One channel of Manual C.1 Restore channel to Prior to entering Trip inoperable. OPERABLE status. MODE 2 from MODE 3 D. Two channels of Trip D.1 De-energize the affected Immediately Initiation Logic affecting clutch power supplies.
| |
| the same trip leg inoperable.
| |
| Palisades Nuclear Plant 3.3.2-1 Amendment No. XXX
| |
| | |
| RPS Logic and Trip Initiation 3.3.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Be in MODE 3. 6 hours associated Completion Time not met. AND OR E.2.1 Verify no more than one 6 hours full-length control rod is One or more Functions capable of being with two or more Manual withdrawn.
| |
| Trip, Matrix Logic or Trip Initiation Logic channels OR inoperable for reasons other than Condition D.
| |
| E.2.2 Verify PCS boron 6 hours concentration is at REFUELING BORON CONCENTRATION.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.2.1 Perform a CHANNEL FUNCTIONAL TEST on In accordance with each RPS Matrix Logic channel and each RPS the Surveillance Trip Initiation Logic channel. Frequency Control Program SR 3.3.2.2 Perform a CHANNEL FUNCTIONAL TEST on Once within 7 days each RPS Manual Trip channel. prior to each reactor startup Palisades Nuclear Plant 3.3.2-2 Amendment No. XXX
| |
| | |
| ESF Instrumentation 3.3.3 3.3 INSTRUMENTATION 3.3.3 Engineered Safety Features (ESF) Instrumentation LCO 3.3.3 Four ESF bistables and associated instrument channels for each Function in Table 3.3.3-1 shall be OPERABLE.
| |
| APPLICABILITY: As specified in Table 3.3.3-1.
| |
| ACTIONS
| |
| -----------------------------------------------------------NOTE----------------------------------------------------------
| |
| Separate Condition entry is allowed for each Function.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. ---------------NOTE--------------- A.1 Place affected bistable 7 days Not applicable to RAS. in trip.
| |
| One or more Functions with one ESF bistable or associated instrument channel inoperable.
| |
| B. ---------------NOTE--------------- 8 hours Not applicable to RAS. B.1 Place one bistable in
| |
| -------------------------------------- trip.
| |
| One or more Functions with AND two ESF bistables or associated instrument B.2 Restore one bistable 7 days channels inoperable. and associated instrument channel to OPERABLE status.
| |
| Palisades Nuclear Plant 3.3.3-1 Amendment No. XXX
| |
| | |
| ESF Instrumentation 3.3.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One RAS bistable or C.1 Bypass affected 8 hours associated instrument bistable.
| |
| channel inoperable.
| |
| AND C.2 Restore bistable and 7 days associated instrument channel to OPERABLE status.
| |
| D. Required Action and D.1 Be in MODE 3. 6 hours associated Completion Time not met for Functions 1, 2, 3, AND 4, or 7.
| |
| D.2 Be in MODE 4. 30 hours E. Required Action and E.1 Be in MODE 3. 6 hours associated Completion Time not met for Functions 5 or 6. AND E.2 Be in MODE 5. 36 hours Palisades Nuclear Plant 3.3.3-2 Amendment No. XXX
| |
| | |
| ESF Instrumentation 3.3.3 SURVEILLANCE REQUIREMENTS
| |
| -----------------------------------------------------------NOTE----------------------------------------------------------
| |
| Refer to Table 3.3.3-1 to determine which SR shall be performed for each Function.
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform a CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.3.2 Perform a CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.3.3 Perform a CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.3.3-3 Amendment No. XXX
| |
| | |
| ESF Instrumentation 3.3.3 Table 3.3.3-1 (page 1 of 2)
| |
| Engineered Safety Features Instrumentation APPLICABLE SURVEILLANCE ALLOWABLE FUNCTION MODES REQUIREMENTS VALUE
| |
| : 1. Safety Injection Signal (SIS)
| |
| : a. Pressurizer Low Pressure 1,2,3 SR 3.3.3.1 1593 psia SR 3.3.3.2 SR 3.3.3.3
| |
| : 2. Steam Generator Low Pressure Signal (SGLP)
| |
| : a. Steam Generator A Low 1,2(a),3(a) SR 3.3.3.1 500 psia Pressure SR 3.3.3.2 SR 3.3.3.3
| |
| : b. Steam Generator B Low 1,2(a),3(a) SR 3.3.3.1 500 psia Pressure SR 3.3.3.2 SR 3.3.3.3
| |
| : 3. Recirculation Actuation Signal (RAS)
| |
| : a. SIRWT Low Level 1,2,3 SR 3.3.3.3 21 inches and 27 inches above tank bottom
| |
| : 4. Auxiliary Feedwater Actuation Signal (AFAS)
| |
| : a. Steam Generator A Low 1,2,3 SR 3.3.3.1 25.9%
| |
| Level SR 3.3.3.2 narrow range SR 3.3.3.3
| |
| : b. Steam Generator B Low 1,2,3 SR 3.3.3.1 25.9%
| |
| Level SR 3.3.3.2 narrow range SR 3.3.3.3 (a) Not required to be OPERABLE when all Main Steam Isolation Valves (MSIVs) are closed and deactivated, and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated, or isolated by closed manual valves.
| |
| Palisades Nuclear Plant 3.3.3-4 Amendment No. XXX
| |
| | |
| ESF Instrumentation 3.3.3 Table 3.3.3-1 (page 2 of 2)
| |
| Engineered Safety Features Instrumentation APPLICABLE SURVEILLANCE ALLOWABLE FUNCTION MODES REQUIREMENTS VALUE
| |
| : 5. Containment High Pressure (CHP)
| |
| : a. Containment High Pressure Left Train 1,2,3,4 SR 3.3.3.2 3.7 psig SR 3.3.3.3 and 4.3 psig
| |
| : b. Containment High Pressure Right Train 1,2,3,4 SR 3.3.3.2 3.7 psig SR 3.3.3.3 and 4.3 psig
| |
| : 6. Containment High Radiation Signal (CHR)
| |
| : a. Containment High Radiation 1,2,3,4 SR 3.3.3.1 20 R/hour SR 3.3.3.2 SR 3.3.3.3
| |
| : 7. Automatic Bypass Removals
| |
| : a. Pressurizer Low Pressure 1,2,3 SR 3.3.3.3 1700 psia Bypass
| |
| : b. Steam Generator A Low 1,2(a),3(a) SR 3.3.3.3 565 psia Pressure Bypass
| |
| : c. Steam Generator B Low SR 3.3.3.3 565 psia 1,2(a),3(a)
| |
| Pressure Bypass (a) Not required to be OPERABLE when all Main Steam Isolation Valves (MSIVs) are closed and deactivated, and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated, or isolated by closed manual valves.
| |
| Palisades Nuclear Plant 3.3.3-5 Amendment No. XXX
| |
| | |
| ESF Logic and Manual Initiation 3.3.4 3.3 INSTRUMENTATION 3.3.4 Engineered Safety Features (ESF) Logic and Manual Initiation LCO 3.3.4 Two ESF Manual Initiation and two ESF Actuation Logic channels and associated bypass removal channels shall be OPERABLE for each ESF Function specified in Table 3.3.4-1.
| |
| APPLICABILITY: According to Table 3.3.4-1.
| |
| ACTIONS
| |
| ----------------------------------------------------------NOTE-----------------------------------------------------------
| |
| Separate Condition entry is allowed for each Function.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions with A.1 Restore channel to 48 hours one Manual Initiation, Bypass OPERABLE status.
| |
| Removal, or Actuation Logic channel inoperable.
| |
| B. One or more Functions with B.1 Be in MODE 3. 6 hours two Manual Initiation, Bypass Removal, or Actuation Logic AND channels inoperable for Functions 1, 2, 3, or 4. B.2 Be in MODE 4. 30 hours OR Required Action and associated Completion Time of Condition A not met for Functions 1, 2, 3, or 4.
| |
| Palisades Nuclear Plant 3.3.4-1 Amendment No. XXX
| |
| | |
| ESF Logic and Manual Initiation 3.3.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One or more Functions with C.1 Be in MODE 3. 6 hours two Manual Initiation, or Actuation Logic channels AND inoperable for Functions 5 or 6. C.2 Be in MODE 5. 36 hours OR Required Action and associated Completion Time of Condition A not met for Functions 5 or 6.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1 Perform functional test of each SIS actuation In accordance with channel normal and standby power functions. the Surveillance Frequency Control Program SR 3.3.4.2 Perform a CHANNEL FUNCTIONAL TEST of each In accordance with AFAS actuation logic channel. the Surveillance Frequency Control Program SR 3.3.4.3 Perform a CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.3.4-2 Amendment No. XXX
| |
| | |
| ESF Logic and Manual Initiation 3.3.4 Table 3.3.4-1 (page 1 of 1)
| |
| Engineered Safety Features Actuation Logic and Manual Initiation APPLICABLE FUNCTION MODES
| |
| : 1. Safety Injection Signal (SIS)(a) 1,2,3
| |
| : 2. Steam Generator Low Pressure Signal 1,2(d),3(d)
| |
| (SGLP)(b)(c)
| |
| : 3. Recirculation Actuation Signal (RAS) 1,2,3
| |
| : 4. Auxiliary Feedwater Actuation Signal (AFAS) 1,2,3
| |
| : 5. Containment High Pressure Signal (CHP)(c) 1,2,3,4
| |
| : 6. Containment High Radiation Signal (CHR) 1,2,3,4 (a) SIS actuation by Pressurizer Low Pressure may be manually bypassed when pressurizer pressure is 1700 psia. The bypass shall be automatically removed whenever pressurizer pressure is > 1700 psia.
| |
| (b) SGLP actuation may be manually bypassed when SG pressure is 565 psia. The bypass shall be automatically removed whenever steam generator pressure is
| |
| > 565 psia.
| |
| (c) Manual Initiation may be achieved by individual component controls.
| |
| (d) Not required to be OPERABLE when all Main Steam Isolation Valves (MSIVs) are closed and deactivated, and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated, or isolated by closed manual valves.
| |
| Palisades Nuclear Plant 3.3.4-3 Amendment No. XXX
| |
| | |
| DG - UV Start 3.3.5 3.3 INSTRUMENTATION 3.3.5 Diesel Generator (DG) - Undervoltage Start (UV Start)
| |
| LCO 3.3.5 Three channels of Loss of Voltage Function and three channels of Degraded Voltage Function auto-initiation instrumentation and associated logic channels for each DG shall be OPERABLE.
| |
| APPLICABILITY: When associated DG is required to be OPERABLE.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Enter applicable Immediately with one channel per DG Conditions and inoperable. Required Actions for the associated DG made inoperable by DG - UV Start instrumentation.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.5.1 Perform a CHANNEL FUNCTIONAL TEST on In accordance with each DG-UV start logic channel. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.3.5-1 Amendment No. XXX
| |
| | |
| DG - UV Start 3.3.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.5.2 Perform CHANNEL CALIBRATION on each Loss In accordance with of Voltage and Degraded Voltage channel with the Surveillance setpoints as follows: Frequency Control Program
| |
| : a. Degraded Voltage Function 2187 V and 2264 V
| |
| : 1. Time delay (degraded voltage sensing relay): 0.5 seconds and 0.8 seconds; and
| |
| : 2. Time delay (degraded voltage sensing relay plus time delay relay): 6.2 seconds and 7.1 seconds.
| |
| : b. Loss of Voltage Function 1780 V and 1940 V Time delay: 5.45 seconds and 8.15 seconds at 1400 V.
| |
| Palisades Nuclear Plant 3.3.5-2 Amendment No. XXX
| |
| | |
| Refueling CHR Instrumentation 3.3.6 3.3 INSTRUMENTATION 3.3.6 Refueling Containment High Radiation (CHR) Instrumentation LCO 3.3.6 Two Refueling CHR Automatic Actuation Function channels and two CHR Manual Actuation Function channels shall be OPERABLE.
| |
| APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Place the affected 4 hours with one channel channel in trip.
| |
| inoperable.
| |
| OR A.2.1 Suspend CORE 4 hours ALTERATIONS.
| |
| AND A.2.2 Suspend movement of 4 hours irradiated fuel assemblies within containment.
| |
| B. One or more Functions B.1 Suspend CORE Immediately with two channels ALTERATIONS.
| |
| inoperable.
| |
| AND B.2 Suspend movement of Immediately irradiated fuel assemblies within containment.
| |
| Palisades Nuclear Plant 3.3.6-1 Amendment No. XXX
| |
| | |
| Refueling CHR Instrumentation 3.3.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform a CHANNEL CHECK of each refueling In accordance with CHR monitor channel. the Surveillance Frequency Control Program SR 3.3.6.2 Perform a CHANNEL FUNCTIONAL TEST of each In accordance with refueling CHR monitor channel. the Surveillance Frequency Control Program SR 3.3.6.3 Perform a CHANNEL FUNCTIONAL TEST of each In accordance with CHR Manual Initiation channel. the Surveillance Frequency Control Program SR 3.3.6.4 Perform a CHANNEL CALIBRATION of each In accordance with refueling CHR monitor channel. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.3.6-2 Amendment No. XXX
| |
| | |
| PAM Instrumentation 3.3.7 3.3 INSTRUMENTATION 3.3.7 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.7 The PAM instrumentation for each Function in Table 3.3.7-1 shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS
| |
| -----------------------------------------------------------NOTE--------------------------------------------------------
| |
| Separate Condition entry is allowed for each Function.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Restore required 30 days with one required channel channel to OPERABLE inoperable. status.
| |
| B. Required Action and B.1 Initiate action in Immediately associated Completion accordance with Time of Condition A not Specification 5.6.6.
| |
| met.
| |
| C. One or more Functions C.1 Restore one channel to 7 days with two required channels OPERABLE status.
| |
| inoperable.
| |
| Palisades Nuclear Plant 3.3.7-1 Amendment No. XXX
| |
| | |
| PAM Instrumentation 3.3.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. (Not Used)
| |
| E. Required Action and E.1 Enter the Condition Immediately associated Completion referenced in Time of Condition C not Table 3.3.7-1 for the met. channel.
| |
| F. As required by Required F.1 Be in MODE 3. 6 hours Action E.1 and referenced in Table 3.3.7-1. AND F.2 Be in MODE 4. 30 hours G. As required by Required G.1 Initiate action in Immediately Action E.1 and referenced accordance with in Table 3.3.7-1. Specification 5.6.6.
| |
| Palisades Nuclear Plant 3.3.7-2 Amendment No. XXX
| |
| | |
| PAM Instrumentation 3.3.7 SURVEILLANCE REQUIREMENTS
| |
| ---------------------------------------------------------NOTE------------------------------------------------------------
| |
| These SRs apply to each PAM instrumentation Function in Table 3.3.7-1.
| |
| SURVEILLANCE FREQUENCY SR 3.3.7.1 Perform CHANNEL CHECK for each required In accordance with instrumentation channel that is normally energized. the Surveillance Frequency Control Program SR 3.3.7.2 -----------------------------NOTE------------------------------
| |
| Neutron detectors are excluded from the CHANNEL CALIBRATION.
| |
| Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.3.7-3 Amendment No. XXX
| |
| | |
| PAM Instrumentation 3.3.7 Table 3.3.7-1 (page 1 of 1)
| |
| Post Accident Monitoring Instrumentation CONDITIONS REFERENCED REQUIRED FROM REQUIRED FUNCTION CHANNELS ACTION E.1
| |
| : 1. Primary Coolant System Hot Leg Temperature (wide range) 2 F
| |
| : 2. Primary Coolant System Cold Leg 2 F Temperature (wide range)
| |
| : 3. Wide Range Neutron Flux 2 F
| |
| : 4. Containment Floor Water Level 2 F (wide range)
| |
| : 5. Subcooled Margin Monitor 2 F
| |
| : 6. Pressurizer Level (wide range) 2 F
| |
| : 7. (Deleted)
| |
| : 8. Condensate Storage Tank Level 2 F
| |
| : 9. Primary Coolant System Pressure 2 F (wide range)
| |
| : 10. Containment Pressure (wide range) 2 F
| |
| : 11. Steam Generator A Water Level 2 F (wide range)
| |
| : 12. Steam Generator B Water Level 2 F (wide range)
| |
| : 13. Steam Generator A Pressure 2 F
| |
| : 14. Steam Generator B Pressure 2 F
| |
| : 15. Containment Isolation Valve Position 1 per valve(a) F
| |
| : 16. Core Exit Temperature - Quadrant 1 4 F
| |
| : 17. Core Exit Temperature - Quadrant 2 4 F
| |
| : 18. Core Exit Temperature - Quadrant 3 4 F
| |
| : 19. Core Exit Temperature - Quadrant 4 4 F
| |
| : 20. Reactor Vessel Water Level 2 G
| |
| : 21. Containment Area Radiation 2 G (high range)
| |
| (a) Not required for isolation valves whose associated penetration is isolated by at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.
| |
| Palisades Nuclear Plant 3.3.7-4 Amendment No. XXX
| |
| | |
| Alternate Shutdown System 3.3.8 3.3 INSTRUMENTATION 3.3.8 Alternate Shutdown System LCO 3.3.8 The Alternate Shutdown System Functions in Table 3.3.8-1 shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS
| |
| ---------------------------------------------------------NOTE----------------------------------------------------------
| |
| Separate Condition entry is allowed for each Function.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required 30 days Functions inoperable. Functions to OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 4. 30 hours Palisades Nuclear Plant 3.3.8-1 Amendment No. XXX
| |
| | |
| Alternate Shutdown System 3.3.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.8.1 Perform CHANNEL FUNCTIONAL TEST of the Once within 7 days Source Range Neutron Flux Function. prior to each reactor startup SR 3.3.8.2 Verify each required control circuit and transfer In accordance with switch is capable of performing the intended the Surveillance function. Frequency Control Program SR 3.3.8.3 -----------------------------NOTES---------------------------
| |
| : 1. Not required for Functions 16, 17, and 18.
| |
| : 2. Neutron detectors are excluded from the CHANNEL CALIBRATION.
| |
| Perform CHANNEL CALIBRATION for each In accordance with required instrumentation channel. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.3.8-2 Amendment No. XXX
| |
| | |
| Alternate Shutdown System 3.3.8 Table 3.3.8-1 (page 1 of 1)
| |
| Alternate Shutdown System Instrumentation and Controls REQUIRED FUNCTION, INSTRUMENT OR CONTROL PARAMATER CHANNELS
| |
| : 1. Source Range Neutron Flux 1
| |
| : 2. Pressurizer Pressure 1
| |
| : 3. Pressurizer Level 1
| |
| : 4. Primary Coolant System (PCS) #1 Hot Leg Temperature 1
| |
| : 5. PCS #2 Hot Leg Temperature 1
| |
| : 6. PCS #1 Cold Leg Temperature 1
| |
| : 7. PCS #2 Cold Leg Temperature 1
| |
| : 8. Steam Generator (SG) A Pressure 1
| |
| : 9. SG B Pressure 1
| |
| : 10. SG A Wide Range Level 1
| |
| : 11. SG B Wide Range Level 1
| |
| : 12. Safety Injection Refueling Water (SIRW) Tank Level 1
| |
| : 13. Auxiliary Feedwater (AFW) Flow Indication to SG A 1
| |
| : 14. AFW Flow Indication to SG B 1
| |
| : 15. AFW Low Suction Pressure Alarm (P-8B) 1
| |
| : 16. AFW Pump P-8B Steam Supply Valve Control 1
| |
| : 17. AFW Flow Control to SG A 1
| |
| : 18. AFW Flow Control to SG B 1 Palisades Nuclear Plant 3.3.8-3 Amendment No. XXX
| |
| | |
| Neutron Flux Monitoring Channels 3.3.9 3.3 INSTRUMENTATION 3.3.9 Neutron Flux Monitoring Channels LCO 3.3.9 Two channels of neutron flux monitoring instrumentation shall be OPERABLE.
| |
| APPLICABILITY: MODES 3, 4, and 5.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Suspend all operations Immediately channel(s) inoperable. involving positive reactivity additions.
| |
| AND A.2 Perform SDM 4 hours verification in accordance with AND SR 3.1.1.1.
| |
| Once per 12 hours thereafter Palisades Nuclear Plant 3.3.9-1 Amendment No. XXX
| |
| | |
| Neutron Flux Monitoring Channels 3.3.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.9.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.9.2 -----------------------------NOTE------------------------------
| |
| Neutron detectors are excluded from the CHANNEL CALIBRATION.
| |
| Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.3.9-2 Amendment No. XXX
| |
| | |
| ESRV Instrumentation 3.3.10 3.3 INSTRUMENTATION 3.3.10 Engineered Safeguards Room Ventilation (ESRV) Instrumentation LCO 3.3.10 Two channels of ESRV Instrumentation shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS
| |
| -----------------------------------------------------------NOTE----------------------------------------------------------
| |
| Separate Condition entry is allowed for each channel.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Initiate action to isolate Immediately inoperable. the associated ESRV System.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.10.1 Perform a CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.10.2 Perform a CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.10.3 Perform a CHANNEL CALIBRATION. In accordance with the Surveillance Verify high radiation setpoint on each ESRV Frequency Control Instrumentation radiation monitoring channel is Program 2.2E+5 cpm.
| |
| Palisades Nuclear Plant 3.3.10-1 Amendment No. XXX
| |
| | |
| PCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.1 PCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 PCS DNB parameters for pressurizer pressure, cold leg temperature, and PCS total flow rate shall be within the limits specified in the COLR.
| |
| APPLICABILITY: MODE 1.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer pressure, PCS A.1 Restore parameter(s) to 2 hours cold leg temperature, or within limit.
| |
| PCS total flow rate not within limits.
| |
| B. Required Action and B.1 Be in MODE 2. 6 hours associated Completion Time not met.
| |
| Palisades Nuclear Plant 3.4.1-1 Amendment No. XXX
| |
| | |
| PCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure within the limits In accordance with specified in the COLR. the Surveillance Frequency Control Program SR 3.4.1.2 Verify PCS cold leg temperature within the limit In accordance with specified in the COLR. the Surveillance Frequency Control Program SR 3.4.1.3 ------------------------------NOTE----------------------------
| |
| Not required to be performed until 31 EFPD after THERMAL POWER is 90% RTP.
| |
| Verify PCS total flow rate within the limit specified In accordance with in the COLR. the Surveillance Frequency Control Program AND After each plugging of 10 or more steam generator tubes Palisades Nuclear Plant 3.4.1-2 Amendment No. XXX
| |
| | |
| PCS Minimum Temperature for Criticality 3.4.2 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.2 PCS Minimum Temperature for Criticality LCO 3.4.2 Each PCS loop average temperature (Tave) shall be 525°F.
| |
| APPLICABILITY: MODE 1 MODE 2 with Keff > 1.0.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 30 minutes A. Tave in one or more PCS A.1 Be in MODE 2 with Keff loops not within limit. < 1.0.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 Verify PCS Tave in each loop > 525oF. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.2-1 Amendment No. XXX
| |
| | |
| PCS P/T Limits 3.4.3 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.3 PCS Pressure and Temperature (P/T) Limits LCO 3.4.3 PCS pressure, PCS temperature, and PCS heatup and cooldown rates shall be maintained within the limits of Figure 3.4.3-1 and Figure 3.4.3-2.
| |
| APPLICABILITY: At all times.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. -------------NOTE-------------- A.1 Restore parameter(s) to 30 minutes Required Action A.2 shall within limits.
| |
| be completed whenever this Condition is entered. AND A.2 Determine PCS is 72 hours Requirements of LCO not acceptable for continued met in MODE 1, 2, 3, or 4. operation.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time of Condition A not AND met.
| |
| B.2 Be in MODE 5 with 36 hours PCS pressure
| |
| < 270 psia.
| |
| Palisades Nuclear Plant 3.4.3-1 Amendment No. XXX
| |
| | |
| PCS P/T Limits 3.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. --------------NOTE------------- C.1 Initiate action to restore Immediately Required Action C.2 shall parameter(s) to within be completed whenever limits.
| |
| this Condition is entered.
| |
| ----------------------------------- AND Requirements of LCO not C.2 Determine PCS is Prior to entering met any time in other than acceptable for continued MODE 4 MODE 1, 2, 3, or 4. operation.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 -------------------------------NOTE---------------------------
| |
| Only required to be performed during PCS heatup and cooldown operations.
| |
| Verify PCS pressure, PCS temperature, and PCS In accordance with heatup and cooldown rates are within the limits of the Surveillance Figure 3.4.3-1 and Figure 3.4.3-2. Frequency Control Program Palisades Nuclear Plant 3.4.3-2 Amendment No. XXX
| |
| | |
| PCS P/T Limits 3.4.3 2500--------------------
| |
| 2250------------------
| |
| 2000
| |
| -+- 1/J F/Hr
| |
| - 21/J F/Hr
| |
| <C 1750 -+- 41/J F/Hr
| |
| ( /)
| |
| Q_
| |
| --&- 61/J F/Hr CD
| |
| : 1. -1/4- 81/J F/Hr
| |
| : 0) 1500 0)
| |
| CD
| |
| -+- ll/JI/J F/Hr 1.
| |
| Q_
| |
| 1.
| |
| CD 1250
| |
| ....N 1.
| |
| 0) 0)
| |
| CD 1.
| |
| Q_
| |
| 1000 750 RV Inlet Av_g. Hrly.
| |
| Tempereture H/0 L1m1 t T 170*F 20.F/Hr.
| |
| 250 T } 170*F 40.F/Hr.
| |
| 350 > T >2509F sa*F/Hr.*
| |
| T 3509F 100*F/Hr.
| |
| *When shutdown ooohi:,g 1soletion velves M0-3015 end M0-3016 ere open, PCS heetu~ rete shell be me1nte1ned 41D'"F /Hr.
| |
| 0 __,_.............................--....-.......,......,i-,-,.......,,-+-,...,....--,-+-.-,,......,.-+-,-......-r-1 50 100 150 200 250 300 350 400 450 RV Inlet Temperoture, F Figure 3.4.3-1 (Page 1 of 1)
| |
| Pressure - Temperature Limits for Heatups Applicable up to 42.1 EFPY Palisades Nuclear Plant 3.4.3-3 Amendment No. XXX
| |
| | |
| PCS P/T Limits 3.4.3 2500-------------------,--------....----,
| |
| 2250-------------+--+----+------i------l 2000
| |
| --+- IIJ F/Hr
| |
| - 2IIJ F/Hr
| |
| <C
| |
| ...... 1750 --A- 4IIJ F/Hr (I) 0.. --e-- 6IIJ F/Hr G)
| |
| L 1500 - 8IIJ F/Hr G)
| |
| --+- 1IIJIIJ F/Hr L
| |
| 0..
| |
| L G)
| |
| N 1250 L
| |
| G)
| |
| L 1000 0..
| |
| 750 RV Inlet Av8. Hrly.
| |
| 500 Tempereture C/ L1m1t T 170*F 40.F/Hr.
| |
| 250 T > 170*F 40.F/Hr.
| |
| 350 > T > 250*F GrF/Hr.
| |
| 250 T 350*F 100.F/Hr.
| |
| 0 ..........---'-----......-1,...,....,~...-------..................- .................
| |
| 50 100 150 200 250 300 350 400 450 RV Inlet Tempereture, F Additional restrictions when head is on reactor vessel:
| |
| : 1. Maintain average core exit temperature:
| |
| 135°F > T > 110°F for > 3 hours.
| |
| : 2. Following completion of item 1, maintain average hourly cooldown (C/D) limit of 20°F/hour based on core exit temperature indication.
| |
| Figure 3.4.3-2 (Page 1 of 1)
| |
| Pressure - Temperature Limits for Cooldown Applicable up to 42.1 EFPY Palisades Nuclear Plant 3.4.3-4 Amendment No. XXX
| |
| | |
| PCS Loops - MODES 1 and 2 3.4.4 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.4 PCS Loops - MODES 1 and 2 LCO 3.4.4 Two PCS loops shall be OPERABLE and in operation.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of LCO not A.1 Be in MODE 3. 6 hours met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify each PCS loop is in operation. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.4-1 Amendment No. XXX
| |
| | |
| PCS Loops - MODE 3 3.4.5 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.5 PCS Loops - MODE 3 LCO 3.4.5 Two PCS loops shall be OPERABLE and one PCS loop shall be in operation.
| |
| --------------------------------------------NOTES------------------------------------------
| |
| : 1. All primary coolant pumps may not be in operation for 1 hour per 8 hour period, provided:
| |
| : a. No operations are permitted that would cause reduction of the PCS boron concentration; and
| |
| : b. Core outlet temperature is maintained at least 10°F below saturation temperature.
| |
| : 2. Forced circulation (starting the first primary coolant pump) shall not be initiated unless one of the following conditions is met:
| |
| : a. PCS cold leg temperature (Tc) is > 430°F;
| |
| : b. Steam Generator (SG) secondary temperature is equal to or less than the reactor inlet temperature (Tc);
| |
| : c. SG secondary temperature is < 100°F above Tc, and shutdown cooling is isolated from the PCS, and PCS heatup/cooldown rate is 10°F/hour; or
| |
| : d. SG secondary temperature is < 100°F above Tc, and shutdown cooling is isolated from the PCS, and pressurizer level is 57%.
| |
| APPLICABILITY: MODE 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required PCS loop A.1 Restore required PCS 72 hours inoperable. loop to OPERABLE status.
| |
| Palisades Nuclear Plant 3.4.5-1 Amendment No. XXX
| |
| | |
| PCS Loops - MODE 3 3.4.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 4. 24 hours associated Completion Time of Condition A not met.
| |
| C. No PCS loop OPERABLE. C.1 Suspend all operations Immediately involving a reduction of OR PCS boron concentration.
| |
| No PCS loop in operation.
| |
| AND C.2 Initiate action to restore Immediately one PCS loop to OPERABLE status and operation.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required PCS loop is in operation. In accordance with the Surveillance Frequency Control Program SR 3.4.5.2 Verify secondary side water level in each steam In accordance with generator -84%. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.5-2 Amendment No. XXX
| |
| | |
| PCS Loops - MODE 3 3.4.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.3 Verify correct breaker alignment and indicated In accordance with power available to the required primary coolant the Surveillance pump that is not in operation. Frequency Control Program Palisades Nuclear Plant 3.4.5-3 Amendment No. XXX
| |
| | |
| PCS Loops - MODE 4 3.4.6 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.6 PCS Loops - MODE 4 LCO 3.4.6 Two loops or trains consisting of any combination of PCS loops and Shutdown Cooling (SDC) trains shall be OPERABLE, and either:
| |
| : a. One PCS loop shall be in operation; or
| |
| : b. One SDC train shall be in operation with 2810 gpm flow through the reactor core.
| |
| -------------------------------------------NOTES-------------------------------------------
| |
| : 1. All Primary Coolant Pumps (PCPs) and SDC pumps may not be in operation for 1 hour per 8 hour period, provided:
| |
| : a. No operations are permitted that would cause reduction of the PCS boron concentration; and
| |
| : b. Core outlet temperature is maintained at least 10°F below saturation temperature.
| |
| : 2. Forced circulation (starting the first PCP) shall not be initiated unless one of the following conditions is met:
| |
| : a. Steam Generator (SG) secondary temperature is equal to or less than the reactor inlet temperature (Tc),
| |
| : b. SG secondary temperature is < 100°F above Tc, and shutdown cooling is isolated from the PCS, and PCS heatup/cooldown rate is 10°F/hour,
| |
| : c. SG secondary temperature is < 100°F above Tc, and shutdown cooling is isolated from the PCS, and pressurizer level is 57%.
| |
| : 3. Primary coolant pumps P-50A and P-50B shall not be operated simultaneously.
| |
| APPLICABILITY: MODE 4.
| |
| Palisades Nuclear Plant 3.4.6-1 Amendment No. XXX
| |
| | |
| PCS Loops - MODE 4 3.4.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One PCS loop inoperable. A.1 Initiate action to restore Immediately a second PCS loop or AND one SDC train to OPERABLE status.
| |
| Two SDC trains inoperable.
| |
| B. One SDC train inoperable. B.1 Be in MODE 5. 24 hours AND Two PCS loops inoperable.
| |
| C. No PCS loops or SDC C.1 Suspend all operations Immediately trains OPERABLE. involving reduction of PCS boron OR concentration.
| |
| No PCS loop in operation AND with SDC flow through the reactor core not within C.2.1 Initiate action to restore Immediately limits. one PCS loop to OPERABLE status and operation.
| |
| OR C.2.2 Initiate action to restore Immediately one SDC train to OPERABLE status and operation with 2810 gpm flow through the reactor core.
| |
| Palisades Nuclear Plant 3.4.6-2 Amendment No. XXX
| |
| | |
| PCS Loops - MODE 4 3.4.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify one SDC train is in operation with In accordance with 2810 gpm flow through the reactor core, or one the Surveillance PCS loop is in operation. Frequency Control Program SR 3.4.6.2 Verify secondary side water level in required SG(s) In accordance with is -84%. the Surveillance Frequency Control Program SR 3.4.6.3 Verify correct breaker alignment and indicated In accordance with power available to the required pump that is not in the Surveillance operation. Frequency Control Program Palisades Nuclear Plant 3.4.6-3 Amendment No. XXX
| |
| | |
| PCS Loops - MODE 5, Loops Filled 3.4.7 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.7 PCS Loops - MODE 5, Loops Filled LCO 3.4.7 One Shutdown Cooling (SDC) train shall be OPERABLE and in operation with 2810 gpm flow through the reactor core, and either:
| |
| : a. One additional SDC train shall be OPERABLE; or
| |
| : b. The secondary side water level of each Steam Generator (SG) shall be -84%.
| |
| -------------------------------------------NOTES-------------------------------------------
| |
| : 1. The SDC pump of the train in operation may not be in operation for 1 hour per 8 hour period provided:
| |
| : a. No operations are permitted that would cause reduction of the PCS boron concentration; and
| |
| : b. Core outlet temperature is maintained at least 10°F below saturation temperature.
| |
| : 2. Both SDC trains may be inoperable for up to 2 hours for surveillance testing or maintenance provided:
| |
| : a. One SDC train is providing the required flow through the reactor core;
| |
| : b. Core outlet temperature is maintained at least 10°F below saturation temperature; and
| |
| : c. Each SG secondary side water level is -84%.
| |
| : 3. Forced circulation (starting the first primary coolant pump) shall not be initiated unless one of the following conditions is met:
| |
| : a. SG secondary temperature is equal to or less than the reactor inlet temperature (Tc);
| |
| : b. SG secondary temperature is < 100°F above Tc, and shutdown cooling is isolated from the PCS, and PCS heatup/cooldown rate is 10°F/hour; or
| |
| : c. SG secondary temperature is < 100°F above Tc, and shutdown cooling is isolated from the PCS, and pressurizer level is 57%.
| |
| : 4. Primary coolant pumps P-50A and P-50B shall not be operated simultaneously.
| |
| : 5. All SDC trains may not be in operation during planned heatup to MODE 4 when at least one PCS loop is in operation.
| |
| APPLICABILITY: MODE 5 with PCS loops filled.
| |
| Palisades Nuclear Plant 3.4.7-1 Amendment No. XXX
| |
| | |
| PCS Loops - MODE 5, Loops Filled 3.4.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SDC train inoperable. A.1 Initiate action to restore Immediately a second SDC train to AND OPERABLE status.
| |
| Any SG with secondary OR side water level not within limit. A.2 Initiate action to restore Immediately SG secondary side water levels to within limits.
| |
| B. Two SDC trains B.1 Suspend all operations Immediately inoperable. involving reduction in PCS boron OR concentration.
| |
| SDC flow through the AND reactor core not within limits. B.2 Initiate action to restore Immediately one SDC train to OPERABLE status and operation with 2810 gpm flow through the reactor core.
| |
| Palisades Nuclear Plant 3.4.7-2 Amendment No. XXX
| |
| | |
| PCS Loops - MODE 5, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 Verify one SDC train is in operation with In accordance with 2810 gpm flow through the reactor core. the Surveillance Frequency Control Program SR 3.4.7.2 Verify required SG secondary side water level is In accordance with
| |
| - 84%. the Surveillance Frequency Control Program SR 3.4.7.3 Verify correct breaker alignment and indicated In accordance with power available to the required SDC pump that is the Surveillance not in operation. Frequency Control Program Palisades Nuclear Plant 3.4.7-3 Amendment No. XXX
| |
| | |
| PCS Loops - MODE 5, Loops Not Filled 3.4.8 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.8 PCS Loops - MODE 5, Loops Not Filled LCO 3.4.8 Two Shutdown Cooling (SDC) trains shall be OPERABLE, and either:
| |
| : a. One SDC train in operation with 2810 gpm flow through the reactor core; or
| |
| : b. One SDC train in operation with 650 gpm flow through the reactor core with two of the three charging pumps incapable of reducing the boron concentration in the PCS below the minimum value necessary to maintain the required SHUTDOWN MARGIN.
| |
| ---------------------------------------------NOTES-----------------------------------------
| |
| : 1. All SDC pumps may not be in operation for 1 hour provided:
| |
| : a. No operations are permitted that would cause a reduction of the PCS boron concentration;
| |
| : b. Core outlet temperature is maintained > 10°F below saturation temperature; and
| |
| : c. No draining operations to further reduce the PCS water volume are permitted.
| |
| : 2. One SDC train may be inoperable for 2 hours for surveillance testing provided the other SDC train is OPERABLE and in operation.
| |
| APPLICABILITY: MODE 5 with PCS loops not filled.
| |
| Palisades Nuclear Plant 3.4.8-1 Amendment No. XXX
| |
| | |
| PCS Loops - MODE 5, Loops Not Filled 3.4.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SDC train inoperable. A.1 Initiate action to restore Immediately SDC train to OPERABLE status.
| |
| B. Two SDC trains B.1 Suspend all operations Immediately inoperable. involving reduction of PCS boron OR concentration.
| |
| SDC flow through the AND reactor core not within limits. B.2 Initiate action to restore Immediately one SDC train to OPERABLE status and operation with SDC flow through the reactor core within limit.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 -------------------------------NOTE---------------------------
| |
| Only required to be met when complying with LCO 3.4.8.a.
| |
| Verify one SDC train is in operation with In accordance with 2810 gpm flow through the reactor core. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.8-2 Amendment No. XXX
| |
| | |
| PCS Loops - MODE 5, Loops Not Filled 3.4.8 SURVEILLANCE FREQUENCY SR 3.4.8.2 -------------------------------NOTE---------------------------
| |
| Only required to be met when complying with LCO 3.4.8.b.
| |
| Verify one SDC train is in operation with In accordance with 650 gpm flow through the reactor core. the Surveillance Frequency Control Program SR 3.4.8.3 --------------------------------NOTE--------------------------
| |
| Only required to be met when complying with LCO 3.4.8.b.
| |
| Verify two of three charging pumps are incapable In accordance with of reducing the boron concentration in the PCS the Surveillance below the minimum value necessary to maintain Frequency Control the required SHUTDOWN MARGIN. Program SR 3.4.8.4 Verify correct breaker alignment and indicated In accordance with power available to the SDC pump that is not in the Surveillance operation. Frequency Control Program Palisades Nuclear Plant 3.4.8-3 Amendment No. XXX
| |
| | |
| Pressurizer 3.4.9 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.9 Pressurizer LCO 3.4.9 The pressurizer shall be OPERABLE with:
| |
| : a. Pressurizer water level < 62.8%;
| |
| ---------------------------------------------NOTE-------------------------------------------
| |
| The pressurizer water level limit does not apply in MODE 3 until after a bubble has been established in the pressurizer and the pressurizer water level has been lowered to within its normal operating band.
| |
| : b. 375 kW of pressurizer heater capacity available from electrical bus 1D, and
| |
| : c. 375 kW of pressurizer heater capacity available from electrical bus 1E with the capability of being powered from an emergency power supply.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer water level not A.1 Be in MODE 3 with 6 hours within limit. reactor tripped.
| |
| AND A.2 Be in MODE 4. 30 hours Palisades Nuclear Plant 3.4.9-1 Amendment No. XXX
| |
| | |
| Pressurizer 3.4.9 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. < 375 kW pressurizer B.1 Restore required 72 hours heater capacity available pressurizer heaters to from electrical bus 1D, or OPERABLE status.
| |
| electrical bus 1E, OR Required pressurizer heater capacity from electrical bus 1E not capable of being powered from an emergency power supply.
| |
| C. ------------NOTE--------------- C.1 Restore at least 24 hours Not applicable when the electrical bus 1D or remaining electrical bus 1D electrical bus 1E or electrical bus 1E required pressurizer required pressurizer heaters to OPERABLE heaters intentionally made status.
| |
| inoperable.
| |
| < 375 kW pressurizer heater capacity available from electrical bus 1D, and electrical bus 1E, OR
| |
| < 375 kW pressurizer heater capacity available from electrical bus 1D, and required pressurizer heater capacity from electrical bus 1E not capable of being powered from an emergency power supply.
| |
| Palisades Nuclear Plant 3.4.9-2 Amendment No. XXX
| |
| | |
| Pressurizer 3.4.9 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 6 hours associated Completion Time of Condition B or C AND not met.
| |
| D.2 Be in MODE 4. 30 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 -------------------------------NOTE---------------------------
| |
| Not required to be met until 1 hour after establishing a bubble in the pressurizer and the pressurizer water level has been lowered to within its normal operating band.
| |
| Verify pressurizer water level is < 62.8%. In accordance with the Surveillance Frequency Control Program SR 3.4.9.2 Verify the capacity of pressurizer heaters from In accordance with electrical bus 1D, and electrical bus 1E is the Surveillance 375 kW. Frequency Control Program SR 3.4.9.3 Verify the required pressurizer heater capacity In accordance with from electrical bus 1E is capable of being powered the Surveillance from an emergency power supply. Frequency Control Program Palisades Nuclear Plant 3.4.9-3 Amendment No. XXX
| |
| | |
| Pressurizer Safety Valves 3.4.10 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.10 Pressurizer Safety Valves LCO 3.4.10 Three pressurizer safety valves shall be OPERABLE with lift settings as specified in Table 3.4.10-1.
| |
| APPLICABILITY: MODES 1 and 2, MODE 3 with all PCS cold leg temperatures 430°F.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pressurizer safety A.1 Restore valve to 15 minutes valve inoperable. OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND OR B.2 Reduce any PCS cold 12 hours leg temperature Two or more pressurizer < 430°F.
| |
| safety valves inoperable.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each pressurizer safety valve is OPERABLE In accordance with in accordance with the INSERVICE TESTING the INSERVICE PROGRAM. Following testing, lift settings shall be TESTING within +/- 1% of required setpoint. PROGRAM Palisades Nuclear Plant 3.4.10-1 Amendment No. XXX
| |
| | |
| Pressurizer Safety Valves 3.4.10 Table 3.4.10-1 (page 1 of 1)
| |
| Pressurizer Safety Valve Lift Settings VALVE NUMBER LIFT SETTING (psia +/- 3%)
| |
| RV-1039 2580 RV-1040 2540 RV-1041 2500 Palisades Nuclear Plant 3.4.10-2 Amendment No. XXX
| |
| | |
| Pressurizer PORVs 3.4.11 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)
| |
| LCO 3.4.11 Each PORV and associated block valve shall be OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2, MODE 3 with all PCS cold leg temperatures 430°F.
| |
| ACTIONS
| |
| -----------------------------------------------------------NOTE--------------------------------------------------------
| |
| Separate Condition entry is allowed for each PORV.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One PORV inoperable. A.1 Close associated block 1 hour valve.
| |
| AND A.2 Restore PORV to 72 hours OPERABLE status.
| |
| B. One block valve B.1 Place associated PORV 1 hour inoperable. in manual control.
| |
| AND B.2 Restore block valve to 72 hours OPERABLE status.
| |
| Palisades Nuclear Plant 3.4.11-1 Amendment No. XXX
| |
| | |
| Pressurizer PORVs 3.4.11 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Two PORVs inoperable. C.1 Close associated block 1 hour valves.
| |
| AND C.2 Restore at least one 2 hours PORV to OPERABLE status.
| |
| D. Two block valves D.1 Place associated 1 hour inoperable. PORVs in manual control.
| |
| AND D.2 Restore at least one 2 hours block valve to OPERABLE status.
| |
| E. Required Action and E.1 Be in MODE 3. 6 hours associated Completion Time not met.
| |
| Palisades Nuclear Plant 3.4.11-2 Amendment No. XXX
| |
| | |
| Pressurizer PORVs 3.4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 Perform a complete cycle of each block valve. Once prior to entering MODE 4 from MODE 5 if not performed within previous 92 days SR 3.4.11.2 Perform a complete cycle of each PORV with PCS In accordance with average temperature > 200°F. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.11-3 Amendment No. XXX
| |
| | |
| LTOP System 3.4.12 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.12 Low Temperature Overpressure Protection (LTOP) System LCO 3.4.12 An LTOP System shall be OPERABLE with:
| |
| : a. Both High Pressure Safety Injection (HPSI) pumps incapable of injecting into the PCS, and
| |
| ------------------------------------------NOTES-----------------------------------------
| |
| : 1. LCO 3.4.12.a is only required when any PCS cold leg temperature is < 300°F.
| |
| : 2. LCO 3.4.12.a does not prohibit the use of the HPSI pumps for emergency addition of makeup to the PCS.
| |
| : b. One of the following pressure relief capabilities:
| |
| : 1. Two Power Operated Relief Valves (PORVs) with lift settings as specified in Figure 3.4.12-1; or
| |
| : 2. The PCS depressurized and a PCS vent capable of relieving 167 gpm at a pressure of 315 psia.
| |
| APPLICABILITY: MODE 3 when any PCS cold leg temperature is < 430°F, MODES 4 and 5, MODE 6 when the reactor vessel head is on.
| |
| ACTIONS
| |
| -----------------------------------------------------------NOTE------------------------------------------------------
| |
| LCO 3.0.4.b is not applicable to PORVs when entering MODE 4.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or two HPSI pumps A.1 Initiate action to verify Immediately capable of injecting into no HPSI pump is the PCS. capable of injecting into the PCS.
| |
| Palisades Nuclear Plant 3.4.12-1 Amendment No. XXX
| |
| | |
| LTOP System 3.4.12 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One required PORV B.1 Restore required PORV 7 days inoperable and pressurizer to OPERABLE status.
| |
| water level 57%.
| |
| C. One required PORV C.1 Restore required PORV 24 hours inoperable and pressurizer to OPERABLE status.
| |
| water level > 57%.
| |
| D. Two required PORVs D.1 Depressurize PCS and 8 hours inoperable. establish PCS vent capable of relieving OR 167 gpm at a PCS pressure of 315 psia.
| |
| Required Action and associated Completion Time not met.
| |
| OR LTOP System inoperable for any reason other than Condition A, B, or C.
| |
| Palisades Nuclear Plant 3.4.12-2 Amendment No. XXX
| |
| | |
| LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 -------------------------------NOTE-------------------------
| |
| Only required to be met when complying with LCO 3.4.12.a.
| |
| Verify both HPSI pumps are incapable of injecting In accordance with into the PCS. the Surveillance Frequency Control Program SR 3.4.12.2 Verify required PCS vent, capable of relieving In accordance with 167 gpm at a PCS pressure of 315 psia, is open. the Surveillance Frequency Control Program SR 3.4.12.3 Verify PORV block valve is open for each required In accordance with PORV. the Surveillance Frequency Control Program SR 3.4.12.4 -------------------------------NOTE-------------------------
| |
| Not required to be performed until 12 hours after decreasing any PCS cold leg temperature to
| |
| < 430°F.
| |
| Perform CHANNEL FUNCTIONAL TEST on each In accordance with required PORV, excluding actuation. the Surveillance Frequency Control Program SR 3.4.12.5 Perform CHANNEL CALIBRATION on each In accordance with required PORV actuation channel. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.12-3 Amendment No. XXX
| |
| | |
| LTOP System 3.4.12 2500 2250 I
| |
| 2000 I
| |
| 1750
| |
| <C C l) 0..
| |
| G) * --
| |
| I 1.
| |
| .,ii 1500 G)
| |
| : 1. --
| |
| 0..
| |
| II 1.
| |
| 1250 1.
| |
| ii I 1000 -
| |
| I/
| |
| 1.
| |
| 0.. -
| |
| /'
| |
| 750 500
| |
| - L,....,-"""
| |
| -- l,...-----"'"'
| |
| 250 0 I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I 50 100 150 200 250 300 350 400 450 RV Inlet Temperoture, F Figure 3.4.12-1 (Page 1 of 1)
| |
| LTOP Setpoint Limit Applicable up to 42.1 EFPY Palisades Nuclear Plant 3.4.12-4 Amendment No. XXX
| |
| | |
| PCS Operational LEAKAGE 3.4.13 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.13 PCS Operational LEAKAGE LCO 3.4.13 PCS operational LEAKAGE shall be limited to:
| |
| : a. No pressure boundary LEAKAGE;
| |
| : b. 1 gpm unidentified LEAKAGE;
| |
| : c. 10 gpm identified LEAKAGE; and
| |
| : d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. PCS operational A.1 Reduce LEAKAGE to 4 hours LEAKAGE not within limits within limits.
| |
| for reasons other than pressure boundary LEAKAGE or primary to secondary leakage.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND OR B.2 Be in MODE 5. 36 hours Pressure boundary LEAKAGE exists.
| |
| OR Primary to secondary LEAKAGE not within limit.
| |
| Palisades Nuclear Plant 3.4.13-1 Amendment No. XXX
| |
| | |
| PCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 -------------------------------NOTES------------------------- ----------NOTE--------
| |
| : 1. Not required to be performed in MODE 3 or 4 Only required to be until 12 hours of steady state operation. performed during steady state
| |
| : 2. Not applicable to primary to secondary operation LEAKAGE. ---------------------------
| |
| Verify PCS operational LEAKAGE is within limits In accordance with by performance of PCS water inventory balance. the Surveillance Frequency Control Program SR 3.4.13.2 ------------------------------- NOTE ---------------------------
| |
| Not required to be performed until 12 hours after establishment of steady state operation.
| |
| Verify primary to secondary LEAKAGE is < 150 In accordance with gallons per day through any one SG. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.13-2 Amendment No. XXX
| |
| | |
| PCS PIV Leakage 3.4.14 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.14 PCS Pressure Isolation Valve (PIV) Leakage LCO 3.4.14 Leakage from each PCS PIV shall be within limits and both Shutdown Cooling (SDC) suction valve interlocks shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3, MODE 4, except during the SDC mode of operation, or transition to or from, the SDC mode of operation.
| |
| ACTIONS
| |
| ----------------------------------------------------------NOTES---------------------------------------------------------
| |
| : 1. Separate Condition entry is allowed for each flow path.
| |
| : 2. Enter applicable Conditions and Required Actions for systems made inoperable by an inoperable PIV.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more flow paths ------------------NOTE------------------
| |
| with leakage from one or Each valve used to satisfy more PCS PIVs not within Required Action A.1 must have limit. been verified to meet SR 3.4.14.1 and be on the PCS pressure boundary or the high pressure portion of the system.
| |
| (continued)
| |
| Palisades Nuclear Plant 3.4.14-1 Amendment No. XXX
| |
| | |
| PCS PIV Leakage 3.4.14 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.1 Isolate the high pressure 4 hours portion of the affected system from the low pressure portion by use of one closed manual, deactivated automatic, or check valve.
| |
| AND A.2 Restore PCS PIV 72 hours to within limits.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time for Condition A not AND met.
| |
| B.2 Be in MODE 5. 36 hours C. One or both SDC suction C.1 Isolate the affected 4 hours valve interlocks penetration by use of inoperable. one closed deactivated valve.
| |
| Palisades Nuclear Plant 3.4.14-2 Amendment No. XXX
| |
| | |
| PCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 -------------------------------NOTES--------------------------
| |
| : 1. Only required to be performed in MODES 1 and 2.
| |
| : 2. Leakage rates 5.0 gpm are unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible leakage rate of 5.0 gpm by 50%
| |
| or greater.
| |
| : 3. Minimum test differential pressure shall not be less than 150 psid.
| |
| Verify leakage from each PCS PIV is equivalent to In accordance with 5 gpm at a PCS pressure of 2060 psia. the Surveillance Frequency Control Program AND Once prior to entering MODE 2 whenever the plant has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months SR 3.4.14.2 Verify each SDC suction valve interlock prevents its In accordance with associated valve from being opened with a the Surveillance simulated or actual PCS pressure signal 280 psia. Frequency Control Program Palisades Nuclear Plant 3.4.14-3 Amendment No. XXX
| |
| | |
| PCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.3 -------------------------------NOTE----------------------------
| |
| Only required to be performed in MODES 1 and 2.
| |
| Prior to entering Verify each of the four Low Pressure Safety MODE 2 after each Injection (LPSI) check valves are closed. use of the LPSI check valves for SDC Palisades Nuclear Plant 3.4.14-4 Amendment No. XXX
| |
| | |
| PCS Leakage Detection Instrumentation 3.4.15 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.15 PCS Leakage Detection Instrumentation LCO 3.4.15 Three of the following PCS leakage detection instrumentation channels shall be OPERABLE:
| |
| : a. One containment sump level indicating channel;
| |
| : b. One containment atmosphere gaseous activity monitoring channel;
| |
| : c. One containment air cooler condensate level switch channel;
| |
| : d. One containment atmosphere humidity monitoring channel.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or two required leak A.1 Perform SR 3.4.13.1 Once per 24 hours detection instrument (PCS water inventory channels inoperable. balance).
| |
| AND A.2 Restore inoperable 30 days channel(s) to OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours Palisades Nuclear Plant 3.4.15-1 Amendment No. XXX
| |
| | |
| PCS Leakage Detection Instrumentation 3.4.15 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. All required channels C.1 Enter LCO 3.0.3. Immediately inoperable.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.15.1 Perform CHANNEL CHECK of the required In accordance with containment sump level indicator. the Surveillance Frequency Control Program SR 3.4.15.2 Perform CHANNEL CHECK of the required In accordance with containment atmosphere gaseous activity monitor. the Surveillance Frequency Control Program SR 3.4.15.3 Perform CHANNEL CHECK of the required In accordance with containment atmosphere humidity monitor. the Surveillance Frequency Control Program SR 3.4.15.4 Perform CHANNEL FUNCTIONAL TEST of the In accordance with required containment air cooler condensate level the Surveillance switch. Frequency Control Program SR 3.4.15.5 Perform CHANNEL CALIBRATION of the required In accordance with containment sump level indicator. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.15-2 Amendment No. XXX
| |
| | |
| PCS Leakage Detection Instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.15.6 Perform CHANNEL CALIBRATION of the required In accordance with containment atmosphere gaseous activity monitor. the Surveillance Frequency Control Program SR 3.4.15.7 Perform CHANNEL CALIBRATION of the required In accordance with containment atmosphere humidity monitor. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.15-3 Amendment No. XXX
| |
| | |
| PCS Specific Activity 3.4.16 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.16 PCS Specific Activity LCO 3.4.16 The specific activity of the primary coolant shall be within limits.
| |
| APPLICABILITY: MODES 1 and 2, MODE 3 with PCS average temperature (Tave) 500°F.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT I-131 -----------------NOTE-------------------
| |
| > 1.0 µCi/gm. LCO 3.0.4.c is applicable.
| |
| A.1 Verify DOSE Once per 4 hours EQUIVALENT I-131
| |
| < 40 µCi/gm.
| |
| AND A.2 Restore DOSE 48 hours EQUIVALENT I-131 to within limit.
| |
| Palisades Nuclear Plant 3.4.16-1 Amendment No. XXX
| |
| | |
| PCS Specific Activity 3.4.16 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3 with 6 hours associated Completion Tave < 500°F.
| |
| Time of Condition A not met.
| |
| OR DOSE EQUIVALENT I-131 40 µCi/gm.
| |
| OR Gross specific activity of the primary coolant not within limit.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify primary coolant gross specific activity In accordance with 100/ µCi/gm. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.4.16-2 Amendment No. XXX
| |
| | |
| PCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.2 -------------------------------NOTE---------------------------
| |
| Only required to be performed in MODE 1.
| |
| In accordance with Verify primary coolant DOSE EQUIVALENT I-131 the Surveillance specific activity 1.0 µCi/gm. Frequency Control Program AND Once between 2 and 6 hours after THERMAL POWER change of 15% RTP within a 1 hour period SR 3.4.16.3 -------------------------------NOTE---------------------------
| |
| Not required to be performed until 31 days after a minimum of 2 EFPD and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 48 hours.
| |
| Determine from a sample taken in MODE 1 after In accordance with a minimum of 2 EFPD and 20 days of MODE 1 the Surveillance operation have elapsed since the reactor was last Frequency Control subcritical for 48 hours. Program Palisades Nuclear Plant 3.4.16-3 Amendment No. XXX
| |
| | |
| SG Tube Integrity 3.4.17 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.
| |
| AND All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS
| |
| ------------------------------------------------------------NOTE-----------------------------------------------------------
| |
| Separate Condition entry is allowed for each SG tube.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube repair affected tube(s) is criteria and not plugged maintained until the next in accordance with the refueling outage or SG Steam Generator tube inspection.
| |
| Program.
| |
| AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program. next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time of Condition A not AND met.
| |
| B.2 Be in MODE 5. 36 hours OR SG tube integrity not maintained.
| |
| Palisades Nuclear Plant 3.4.17-1 Amendment No. XXX
| |
| | |
| SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the In accordance Steam Generator Program. with the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube repair criteria is plugged in accordance with the MODE 4 following Steam Generator Program. a SG tube inspection Palisades Nuclear Plant 3.4.17-2 Amendment No. XXX
| |
| | |
| SITs 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Safety Injection Tanks (SITs)
| |
| LCO 3.5.1 Four SITs shall be OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SIT inoperable due to A.1 Restore SIT to 72 hours boron concentration not OPERABLE status.
| |
| within limits.
| |
| OR One SIT inoperable due to the inability to verify level or pressure.
| |
| B. One SIT inoperable for B.1 Restore SIT to 24 hour reasons other than OPERABLE status.
| |
| Condition A.
| |
| C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time of Condition A or B not met.
| |
| D. Two or more SITs D.1 Enter LCO 3.0.3. Immediately inoperable.
| |
| Palisades Nuclear Plant 3.5.1-1 Amendment No. XXX
| |
| | |
| SITs 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each SIT isolation valve is fully open. In accordance with the Surveillance Frequency Control Program SR 3.5.1.2 Verify borated water volume in each SIT is In accordance with 1040 ft3 and 1176 ft3. the Surveillance Frequency Control Program SR 3.5.1.3 Verify nitrogen cover pressure in each SIT is In accordance with 200 psig. the Surveillance Frequency Control Program SR 3.5.1.4 Verify boron concentration in each SIT is In accordance with 1720 ppm and 2500 ppm. the Surveillance Frequency Control Program SR 3.5.1.5 Verify power is removed from each SIT isolation In accordance with valve operator. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.5.1-2 Amendment No. XXX
| |
| | |
| ECCS - Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2, MODE 3 with Primary Coolant System (PCS) temperature 325°F.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One LPSI subsystem A.1 Restore LPSI 7 days inoperable. subsystem to OPERABLE status.
| |
| B. One or more ECCS trains B.1 Restore train(s) to 72 hours inoperable for reasons OPERABLE status.
| |
| other than Condition A.
| |
| C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time of Condition A or B AND not met.
| |
| C.2 Reduce PCS 24 hours temperature to < 325°F.
| |
| D. Less than 100% of the D.1 Enter LCO 3.0.3. Immediately required ECCS flow available.
| |
| Palisades Nuclear Plant 3.5.2-1 Amendment No. XXX
| |
| | |
| ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify the following valves and hand switches are In accordance with in the open position. the Surveillance Frequency Control Valve/Hand Program Switch Number Function CV-3027 SIRWT Recirc Valve HS-3027A Hand Switch For CV-3027 HS-3027B Hand Switch For CV-3027 CV-3056 SIRWT Recirc Valve HS-3056A Hand Switch For CV-3056 HS-3056B Hand Switch For CV-3056 SR 3.5.2.2 Verify each ECCS manual, power operated, and In accordance with automatic valve in the flow path, that is not locked, the Surveillance sealed, or otherwise secured in position, is in the Frequency Control correct position. Program SR 3.5.2.3 Verify CV-3006, "SDC Flow Control Valve," is open In accordance with and its air supply is isolated. the Surveillance Frequency Control Program SR 3.5.2.4 Verify each ECCS pump's developed head at the In accordance with test flow point is greater than or equal to the the INSERVICE required developed head. TESTING PROGRAM SR 3.5.2.5 Verify each ECCS automatic valve that is not In accordance with locked, sealed, or otherwise secured in position, in the Surveillance the flow path actuates to the correct position on an Frequency Control actual or simulated actuation signal. Program Palisades Nuclear Plant 3.5.2-2 Amendment No. XXX
| |
| | |
| ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.6 Verify each ECCS pump starts automatically on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program SR 3.5.2.7 Verify each LPSI pump stops on an actual or In accordance with simulated actuation signal. the Surveillance Frequency Control Program SR 3.5.2.8 Verify, for each ECCS throttle valve listed below, In accordance with each position stop is in the correct position. the Surveillance Frequency Control Valve Number Function Program MO-3008 LPSI to Cold leg 1A MO-3010 LPSI to Cold leg 1B MO-3012 LPSI to Cold leg 2A MO-3014 LPSI to Cold leg 2B MO-3082 HPSI to Hot leg 1 MO-3083 HPSI to Hot leg 1 SR 3.5.2.9 Verify, by visual inspection, the containment sump In accordance with passive strainer assemblies are not restricted by the Surveillance debris, and the containment sump passive strainer Frequency Control assemblies and other containment sump entrance Program pathways show no evidence of structural distress or abnormal corrosion.
| |
| Palisades Nuclear Plant 3.5.2-3 Amendment No. XXX
| |
| | |
| ECCS - Shutdown 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.3 ECCS - Shutdown LCO 3.5.3 One Low Pressure Safety Injection (LPSI) train shall be OPERABLE.
| |
| ------------------------------------------NOTE----------------------------------------------
| |
| A LPSI train may be considered OPERABLE during alignment and operation for shutdown cooling if capable of being manually realigned to the ECCS mode of operation.
| |
| APPLICABILITY: MODE 3 with Primary Coolant System (PCS) temperature < 325°F, MODE 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required LPSI train A.1 Initiate action to restore Immediately inoperable. one LPSI train to OPERABLE status.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 The following SRs of Specification 3.5.2, "ECCS - In accordance with Operating," are applicable: applicable SRs SR 3.5.2.2 SR 3.5.2.9 SR 3.5.2.4 Palisades Nuclear Plant 3.5.3-1 Amendment No. XXX
| |
| | |
| SIRWT 3.5.4 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.4 Safety Injection Refueling Water Tank (SIRWT)
| |
| LCO 3.5.4 The SIRWT shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SIRWT boron A.1 Restore SIRWT to 8 hours concentration not within OPERABLE status.
| |
| limits.
| |
| OR SIRWT borated water temperature not within limits.
| |
| B. SIRWT inoperable for B.1 Restore SIRWT to 1 hour reasons other than OPERABLE status.
| |
| Condition A.
| |
| C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time not met. AND C.2 Be in MODE 5. 36 hours Palisades Nuclear Plant 3.5.4-1 Amendment No. XXX
| |
| | |
| SIRWT 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 Verify SIRWT borated water temperature is 40°F In accordance with and 100°F. the Surveillance Frequency Control Program SR 3.5.4.2 --------------------------------NOTE--------------------------
| |
| Only required to be met in MODES 1, 2, and 3.
| |
| Verify SIRWT borated water volume is In accordance with 250,000 gallons. the Surveillance Frequency Control Program SR 3.5.4.3 --------------------------------NOTE--------------------------
| |
| Only required to be met in MODE 4.
| |
| Verify SIRWT borated water volume is In accordance with 200,000 gallons. the Surveillance Frequency Control Program SR 3.5.4.4 Verify SIRWT boron concentration is 1720 ppm In accordance with and 2500 ppm. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.5.4-2 Amendment No. XXX
| |
| | |
| STB 3.5.5 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.5 Containment Sump Buffering Agent and Weight Requirements LCO 3.5.5 Buffer baskets shall contain 8,186 lbs and 10,553 lbs of Sodium Tetraborate Decahydrate (STB) Na2B4O7 10H2O.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. STB not within limits. A.1 Restore STB to within limits. 72 hours B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 4. 30 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.5.1 Verify the STB baskets contain 8,186 lbs and In accordance with 10,553 lbs of equivalent weight sodium tetraborate the Surveillance decahydrate. Frequency Control Program SR 3.5.5.2 Verify that a sample from the STB baskets provides In accordance with adequate pH adjustment of borated water. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.5.5-1 Amendment No. XXX
| |
| | |
| Containment 3.6.1 3.6 CONTAINMENT SYSTEMS 3.6.1 Containment LCO 3.6.1 Containment shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment inoperable. A.1 Restore containment to 1 hour OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.1 Perform required visual examinations and leakage In accordance with rate testing, except for containment air lock testing, the Containment in accordance with the Containment Leak Leak Rate Testing Rate Testing Program. Program SR 3.6.1.2 Verify containment structural integrity in In accordance with accordance with the Containment Structural the Containment Integrity Surveillance Program. Structural Integrity Surveillance Program Palisades Nuclear Plant 3.6.1-1 Amendment No. XXX
| |
| | |
| Containment Air Locks 3.6.2 3.6 CONTAINMENT SYSTEMS 3.6.2 Containment Air Locks LCO 3.6.2 Two containment air locks shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS
| |
| -----------------------------------------------------------NOTES--------------------------------------------------------
| |
| : 1. Entry and exit is permissible through a locked air lock door to perform repairs on the affected air lock components.
| |
| : 2. Separate Condition entry is allowed for each air lock.
| |
| : 3. Enter applicable Conditions and Required Actions of LCO 3.6.1, "Containment," when leakage results in exceeding the overall containment leakage rate acceptance criteria.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more containment -----------------NOTES----------------
| |
| air locks with one 1. Required Actions A.1, A.2, containment air lock door and A.3 are not applicable if inoperable. both doors in the same air lock are inoperable and Condition C is entered.
| |
| : 2. Entry and exit is permissible for 7 days under administrative controls if both air locks are inoperable.
| |
| A.1 Verify the OPERABLE 1 hour door is closed in the affected air lock.
| |
| AND (continued)
| |
| Palisades Nuclear Plant 3.6.2-1 Amendment No. XXX
| |
| | |
| Containment Air Locks 3.6.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 Lock the OPERABLE 24 hours door closed in the affected air lock.
| |
| AND
| |
| ------------------NOTE-----------------
| |
| Air lock doors in high radiation areas may be verified locked closed by administrative means.
| |
| A.3 Verify the OPERABLE Once per 31 days door is locked closed in the affected air lock.
| |
| B. One or more containment -----------------NOTES----------------
| |
| air locks with containment 1. Required Actions B.1, B.2, air lock interlock and B.3 are not applicable if mechanism inoperable. both doors in the same air lock are inoperable and Condition C is entered.
| |
| : 2. Entry and exit of containment is permissible under the control of a dedicated individual.
| |
| B.1 Verify an OPERABLE 1 hour door is closed in the affected air lock.
| |
| AND (continued)
| |
| Palisades Nuclear Plant 3.6.2-2 Amendment No. XXX
| |
| | |
| Containment Air Locks 3.6.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Lock an OPERABLE 24 hours door closed in the affected air lock.
| |
| AND
| |
| ------------------NOTE-----------------
| |
| Air lock doors in high radiation areas may be verified locked closed by administrative means.
| |
| B.3 Verify an OPERABLE Once per 31 days door is locked closed in the affected air lock.
| |
| C. One or more containment C.1 Initiate action to Immediately air locks inoperable for evaluate overall reasons other than containment leakage Condition A or B. rate per LCO 3.6.1.
| |
| AND C.2 Verify a door is closed 1 hour in the affected air lock.
| |
| AND C.3 Restore air lock to 24 hours OPERABLE status.
| |
| D. Required Action and D.1 Be in MODE 3. 6 hours associated Completion Time not met. AND D.2 Be in MODE 5. 36 hours Palisades Nuclear Plant 3.6.2-3 Amendment No. XXX
| |
| | |
| Containment Air Locks 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1 ----------------------------NOTES---------------------------
| |
| : 1. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.
| |
| : 2. Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.
| |
| In accordance with Perform required air lock leakage rate testing in the Containment Leak accordance with the Containment Leak Rate Rate Testing Program Testing Program.
| |
| SR 3.6.2.2 Verify only one door in the air lock can be opened In accordance with at a time. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.6.2-4 Amendment No. XXX
| |
| | |
| Containment Isolation Valves 3.6.3 3.6 CONTAINMENT SYSTEMS 3.6.3 Containment Isolation Valves LCO 3.6.3 Each containment isolation valve shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS
| |
| -----------------------------------------------------------NOTES--------------------------------------------------------
| |
| : 1. Penetration flow paths, except for 8 inch purge exhaust valves and 12 inch air room supply valves penetration flow paths, may be unisolated intermittently under administrative controls.
| |
| : 2. Separate Condition entry is allowed for each penetration flow path.
| |
| : 3. Enter applicable Conditions and Required Actions for system(s) made inoperable by containment isolation valves.
| |
| : 4. Enter applicable Conditions and Required Actions of LCO 3.6.1, "Containment," when leakage results in exceeding the overall containment leakage rate acceptance criteria.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. --------------NOTE------------- A.1 Isolate the affected 4 hours Only applicable to penetration flow path by penetration flow paths with use of at least one two containment isolation closed and de-activated valves. automatic valve, closed
| |
| ----------------------------------- manual valve, blind flange, or check valve One or more penetration with flow through the flow paths with one valve secured.
| |
| containment isolation valve inoperable (except for AND purge exhaust valve or air room supply valve not locked closed).
| |
| (continued)
| |
| Palisades Nuclear Plant 3.6.3-1 Amendment No. XXX
| |
| | |
| Containment Isolation Valves 3.6.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) --------------------NOTE----------------
| |
| Isolation devices in high radiation areas may be verified by use of administrative means.
| |
| A.2 Verify the affected Once per 31 days for penetration flow path is isolation devices isolated. outside containment AND Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days for isolation devices inside containment B. --------------NOTE------------- B.1 Isolate the affected 1 hour Only applicable to penetration flow path by penetration flow paths with use of at least one two containment isolation closed and de-activated valves. automatic valve, closed
| |
| ----------------------------------- manual valve, or blind flange.
| |
| One or more penetration flow paths with two containment isolation valves inoperable (except for purge exhaust valve or air room supply valve not locked closed).
| |
| Palisades Nuclear Plant 3.6.3-2 Amendment No. XXX
| |
| | |
| Containment Isolation Valves 3.6.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.1 Isolate the affected 72 hours C. --------------NOTE------------- penetration flow path by Only applicable to use of at least one penetration flow paths with closed and de-activated only one containment automatic valve, closed isolation valve and a manual valve, or blind closed system. flange.
| |
| AND One or more penetration flow paths with one -------------------NOTE-----------------
| |
| containment isolation valve Isolation devices in high radiation inoperable. areas may be verified by use of administrative means.
| |
| C.2 Verify the affected Once per 31 days penetration flow path is isolated.
| |
| D. One or more purge D.1 Lock closed the affected 1 hour exhaust or air room supply valves.
| |
| valves not locked closed.
| |
| E. Required Action and E.1 Be in MODE 3. 6 hours associated Completion Time not met. AND E.2 Be in MODE 5. 36 hours Palisades Nuclear Plant 3.6.3-3 Amendment No. XXX
| |
| | |
| Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each 8 inch purge valve and 12 inch air In accordance with room supply valve is locked closed. the Surveillance Frequency Control Program SR 3.6.3.2 -----------------------------NOTE----------------------------
| |
| Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| Verify each manual containment isolation valve In accordance with and blind flange that is located outside the Surveillance containment and not locked, sealed, or otherwise Frequency Control secured in position, and is required to be closed Program during accident conditions, is closed, except for containment isolation valves that are open under administrative controls.
| |
| SR 3.6.3.3 ----------------------------NOTE-----------------------------
| |
| Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| Verify each manual containment isolation valve Prior to entering and blind flange that is located inside containment MODE 4 from and not locked, sealed, or otherwise secured in MODE 5 if not position, and required to be closed during performed within the accident conditions, is closed, except for previous 92 days containment isolation valves that are open under administrative controls.
| |
| Palisades Nuclear Plant 3.6.3-4 Amendment No. XXX
| |
| | |
| Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.4 Verify the isolation time of each automatic power In accordance with operated containment isolation valve is within the INSERVICE limits. TESTING PROGRAM SR 3.6.3.5 Verify each containment 8 inch purge exhaust and In accordance with 12 inch air room supply valve is closed by the Surveillance performance of a leakage rate test. Frequency Control Program SR 3.6.3.6 Verify each automatic containment isolation valve In accordance with that is not locked, sealed, or otherwise secured in the Surveillance position, actuates to the isolation position on an Frequency Control actual or simulated actuation signal. Program Palisades Nuclear Plant 3.6.3-5 Amendment No. XXX
| |
| | |
| Containment Pressure 3.6.4 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure LCO 3.6.4 Containment pressure shall be 1.0 psig in MODES 1 and 2 and 1.5 psig in MODES 3 and 4.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment pressure not A.1 Restore containment 1 hour within limit. pressure to within limit.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1 Verify containment pressure is within limit. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.6.4-1 Amendment No. XXX
| |
| | |
| Containment Air Temperature 3.6.5 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature LCO 3.6.5 Containment average air temperature shall be 140°F.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment average air A.1 Restore containment 8 hours temperature not within average air temperature limit. to within limit.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment average air temperature is In accordance with within limit. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.6.5-1 Amendment No. XXX
| |
| | |
| Containment Cooling Systems 3.6.6 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Cooling Systems LCO 3.6.6 Two containment cooling trains shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more containment A.1 Restore train(s) to 72 hours cooling trains inoperable. OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time of Condition A not AND met.
| |
| B.2 Be in MODE 4. 30 hours C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately required post accident containment cooling capability available.
| |
| Palisades Nuclear Plant 3.6.6-1 Amendment No. XXX
| |
| | |
| Containment Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Verify each containment spray manual, power In accordance with operated, and automatic valve in the flow path the Surveillance that is not locked, sealed, or otherwise secured in Frequency Control position is in the correct position. Program SR 3.6.6.2 Operate each Containment Air Cooler Fan Unit for In accordance with 15 minutes. the Surveillance Frequency Control Program SR 3.6.6.3 Verify the containment spray piping is full of water In accordance with to the 735 ft elevation in the containment spray the Surveillance header. Frequency Control Program SR 3.6.6.4 Verify total service water flow rate, when aligned In accordance with for accident conditions, is 4800 gpm to the Surveillance Containment Air Coolers VHX-1, VHX-2, and Frequency Control VHX-3. Program SR 3.6.6.5 Verify each containment spray pumps developed In accordance with head at the flow test point is greater than or equal the INSERVICE to the required developed head. TESTING PROGRAM SR 3.6.6.6 Verify each automatic containment spray valve in In accordance with the flow path that is not locked, sealed, or the Surveillance otherwise secured in position, actuates to its Frequency Control correct position on an actual or simulated Program actuation signal.
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| Palisades Nuclear Plant 3.6.6-2 Amendment No. XXX
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| | |
| Containment Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.7 Verify each containment spray pump starts In accordance with automatically on an actual or simulated actuation the Surveillance signal. Frequency Control Program SR 3.6.6.8 Verify each containment cooling fan starts In accordance with automatically on an actual or simulated actuation the Surveillance signal. Frequency Control Program SR 3.6.6.9 Verify each spray nozzle is unobstructed. Following maintenance which could result in nozzle blockage Palisades Nuclear Plant 3.6.6-3 Amendment No. XXX
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| | |
| MSSVs 3.7.1 3.7 PLANT SYSTEMS 3.7.1 Main Steam Safety Valves (MSSVs)
| |
| LCO 3.7.1 Twenty-three MSSVs shall be OPERABLE as specified in Table 3.7.1-1.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required 4 hours MSSVs inoperable. MSSVs to OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 4. 30 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 Verify each required MSSV lift setting is within the In accordance with limits of Table 3.7.1-1 in accordance with the the INSERVICE INSERVICE TESTING PROGRAM. Following TESTING PROGRAM testing, lift settings shall be within +/- 1%.
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| Palisades Nuclear Plant 3.7.1-1 Amendment No. XXX
| |
| | |
| MSSVs 3.7.1 Table 3.7.1-1 (page 1 of 1)
| |
| Main Steam Safety Valve Lift Settings VALVE NUMBER LIFT SETTING Steam Generator A Steam Generator B (psig +/- 3%)
| |
| RV-0703 RV-0701 RV-0704 RV-0702 1025 RV-0705 RV-0707 RV-0706 RV-0708 RV-0713 RV-0709 RV-0714 RV-0710 1005 RV-0715 RV-0711 RV-0716 RV-0712 RV-0717 RV-0719 RV-0718 RV-0720 985 RV-0723 RV-0721 RV-0724 RV-0722 Palisades Nuclear Plant 3.7.1-2 Amendment No. XXX
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| | |
| MSIVs 3.7.2 3.7 PLANT SYSTEMS 3.7.2 Main Steam Isolation Valves (MSIVs)
| |
| LCO 3.7.2 Two MSIVs shall be OPERABLE.
| |
| APPLICABILITY: MODE 1, MODES 2 and 3 except when both MSIVs are closed and de-activated.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One MSIV inoperable in A.1 Restore MSIV to 8 hours MODE 1. OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 2. 6 hours Associated Completion Time of Condition A not met.
| |
| C. --------------NOTE------------- C.1 Close MSIV. 8 hours Separate Condition entry is allowed for each MSIV. AND C.2 Verify MSIV is closed. Once per 7 days One or more MSIVs inoperable in MODE 2 or 3.
| |
| D. Required Action and D.1 Be in MODE 3. 6 hours associated Completion Time of Condition C not AND met.
| |
| D.2 Be in MODE 4. 30 hours Palisades Nuclear Plant 3.7.2-1 Amendment No. XXX
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| | |
| MSIVs 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify closure time of each MSIV is 5 seconds on In accordance with an actual or simulated actuation signal from each the Surveillance train under no flow conditions. Frequency Control Program Palisades Nuclear Plant 3.7.2-2 Amendment No. XXX
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| | |
| MFRVs and MFRV Bypass Valves 3.7.3 3.7 PLANT SYSTEMS 3.7.3 Main Feedwater Regulating Valves (MFRVs) and MFRV Bypass Valves LCO 3.7.3 Two MFRVs and two MFRV bypass valves shall be OPERABLE.
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| APPLICABILITY: MODES 1, 2, and 3 except when both MFRVs and both MFRV bypass valves are either closed and de-activated, or isolated by closed manually actuated valves.
| |
| ACTIONS
| |
| ------------------------------------------------------------NOTE---------------------------------------------------------
| |
| Separate Condition entry is allowed for each valve.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more MFRVs or A.1 Close or isolate 8 hours MFRV bypass valves inoperable MFRV(s) or inoperable. MFRV bypass valve(s).
| |
| AND A.2 Verify inoperable Once per 7 days MFRV(s) or MFRV bypass valve(s) is closed or isolated.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 4. 30 hours Palisades Nuclear Plant 3.7.3-1 Amendment No. XXX
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| | |
| MFRVs and MFRV Bypass Valves 3.7.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Verify the closure time of each MFRV and MFRV In accordance with bypass valve is 22 seconds on a actual or the Surveillance simulated actuation signal. Frequency Control Program Palisades Nuclear Plant 3.7.3-2 Amendment No. XXX
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| | |
| ADVs 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Atmospheric Dump Valves (ADVs)
| |
| LCO 3.7.4 One ADV per steam generator shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is being relied upon for heat removal.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required ADV A.1 Restore ADV to 7 days inoperable. OPERABLE status.
| |
| B. Two required ADVs B.1 Restore one ADV to 24 hours inoperable. OPERABLE status.
| |
| C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time not met. AND C.2 Be in MODE 4 without 30 hours reliance upon steam generator for heat removal.
| |
| Palisades Nuclear Plant 3.7.4-1 Amendment No. XXX
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| | |
| ADVs 3.7.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Verify one complete cycle of each ADV. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.7.4-2 Amendment No. XXX
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| | |
| AFW System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Auxiliary Feedwater (AFW) System LCO 3.7.5 Two AFW trains shall be OPERABLE.
| |
| --------------------------------------------NOTES------------------------------------------
| |
| : 1. Only one AFW train, which includes a motor driven pump, is required to be OPERABLE in MODE 4.
| |
| : 2. The steam driven pump is only required to be OPERABLE prior to making the reactor critical.
| |
| : 3. Two AFW pumps may be placed in manual for testing, for a period of up to 4 hours.
| |
| APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.
| |
| ACTIONS
| |
| -----------------------------------------------------------NOTE------------------------------------------------------
| |
| LCO 3.0.4.b is not applicable.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more AFW A.1 Restore train(s) to 72 hours trains inoperable in OPERABLE status.
| |
| MODE 1, 2, or 3.
| |
| Palisades Nuclear Plant 3.7.5-1 Amendment No. XXX
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| | |
| AFW System 3.7.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time of Condition A not AND met.
| |
| B.2 Be in MODE 4. 30 hours OR Less than 100% of the required AFW flow available to either steam generator.
| |
| OR Less than two AFW pumps OPERABLE in MODE 1, 2, OR 3.
| |
| C. Less than 100% of the ------------------NOTE------------------
| |
| required AFW flow LCO 3.0.3 and all other LCO available, to both steam Required Actions requiring MODE generators. changes or power reductions are suspended until at least 100% of the required AFW flow is available.
| |
| C.1 Initiate action to restore Immediately one AFW train to OPERABLE status.
| |
| Palisades Nuclear Plant 3.7.5-2 Amendment No. XXX
| |
| | |
| AFW System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each required AFW manual, power In accordance with operated, and automatic valve in each water flow the Surveillance path and in the steam supply flow path to the Frequency Control steam turbine driven pump, that is not locked, Program sealed, or otherwise secured in position, is in the correct position.
| |
| SR 3.7.5.2 ------------------------NOTE--------------------------------
| |
| Not required to be met for the turbine driven AFW pump in MODE 3 below 800 psig in the steam generators.
| |
| Verify the developed head of each required AFW In accordance with pump at the flow test point is greater than or equal the INSERVICE to the required developed head. TESTING PROGRAM SR 3.7.5.3 ----------------------------NOTE----------------------------
| |
| Only required to be met in MODES 1, 2 or 3 when AFW is not in operation.
| |
| Verify each AFW automatic valve that is not In accordance with locked, sealed, or otherwise secured in position, the Surveillance actuates to the correct position on an actual or Frequency Control simulated actuation signal. Program SR 3.7.5.4 ------------------------------NOTE----------------------------
| |
| Only required to be met in MODES 1, 2, and 3.
| |
| Verify each required AFW pump starts In accordance with automatically on an actual or simulated actuation the Surveillance signal. Frequency Control Program Palisades Nuclear Plant 3.7.5-3 Amendment No. XXX
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| | |
| Condensate Storage and Supply 3.7.6 3.7 PLANT SYSTEMS 3.7.6 Condensate Storage and Supply LCO 3.7.6 The combined useable volume of the Condensate Storage Tank (CST) and Primary Makeup Storage Tank (T-81) shall be 100,000 gallons.
| |
| APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Condensate volume not A.1 Verify OPERABILITY of 4 hours within limit. backup water supplies.
| |
| AND Once per 12 hours thereafter AND A.2 Restore condensate 7 days volume to within limit.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 4 without 30 hours reliance on steam generator for heat removal.
| |
| Palisades Nuclear Plant 3.7.6-1 Amendment No. XXX
| |
| | |
| Condensate Storage and Supply 3.7.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify condensate useable volume is In accordance with 100,000 gallons. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.7.6-2 Amendment No. XXX
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| | |
| CCW System 3.7.7 3.7 PLANT SYSTEMS 3.7.7 Component Cooling Water (CCW) System LCO 3.7.7 Two CCW trains shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more CCW trains A.1 Restore train(s) to 72 hours inoperable. OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time of Condition A not AND met.
| |
| B.2 Be in MODE 5. 36 hours C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately required post accident CCW cooling capability available.
| |
| Palisades Nuclear Plant 3.7.7-1 Amendment No. XXX
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| | |
| CCW System 3.7.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 ------------------------------NOTE----------------------------
| |
| Isolation of CCW flow to individual components does not render the CCW System inoperable.
| |
| Verify each CCW manual, power operated, and In accordance with automatic valve in the flow path servicing safety the Surveillance related equipment, that is not locked, sealed, or Frequency Control otherwise secured in position, is in the correct Program position.
| |
| SR 3.7.7.2 -----------------------------NOTE-----------------------------
| |
| Only required to be met in MODES 1, 2, and 3.
| |
| Verify each CCW automatic valve in the flow path In accordance with that is not locked, sealed, or otherwise secured in the Surveillance position, actuates to the correct position on an Frequency Control actual or simulated actuation signal. Program SR 3.7.7.3 -----------------------------NOTE-----------------------------
| |
| Only required to be met in MODES 1, 2, and 3.
| |
| Verify each CCW pump starts automatically on an In accordance with actual or simulated actuation signal in the with the Surveillance standby power available mode. Frequency Control Program Palisades Nuclear Plant 3.7.7-2 Amendment No. XXX
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| | |
| SWS 3.7.8 3.7 PLANT SYSTEMS 3.7.8 Service Water System (SWS)
| |
| LCO 3.7.8 Two SWS trains shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SWS trains A.1 Restore train(s) to 72 hours inoperable. OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time of Condition A not AND met.
| |
| B.2 Be in MODE 5. 36 hours C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately required post accident SWS cooling capability available.
| |
| Palisades Nuclear Plant 3.7.8-1 Amendment No. XXX
| |
| | |
| SWS 3.7.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8.1 -----------------------------NOTE-----------------------------
| |
| Isolation of SWS flow to individual components does not render SWS inoperable.
| |
| Verify each SWS manual, power operated, and In accordance with automatic valve in the flow path servicing safety the Surveillance related equipment, that is not locked, sealed, or Frequency Control otherwise secured in position, is in the correct Program position.
| |
| SR 3.7.8.2 ----------------------------NOTE------------------------------
| |
| Only required to be met in MODES 1, 2, and 3.
| |
| Verify each SWS automatic valve in the flow path In accordance with that is not locked, sealed, or otherwise secured in the Surveillance position, actuates to the correct position on an Frequency Control actual or simulated actuation signal. Program SR 3.7.8.3 -----------------------------NOTE-----------------------------
| |
| Only required to be met in MODES 1, 2, and 3.
| |
| Verify each SWS pump starts automatically on an In accordance with actual or simulated actuation signal in the with the Surveillance standby power available mode. Frequency Control Program Palisades Nuclear Plant 3.7.8-2 Amendment No. XXX
| |
| | |
| UHS 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Ultimate Heat Sink (UHS)
| |
| LCO 3.7.9 The UHS shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. UHS inoperable. A.1 Be in MODE 3. 6 hours AND A.2 Be in MODE 5. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Verify water level of UHS is 568.25 ft above In accordance with mean sea level. the Surveillance Frequency Control Program SR 3.7.9.2 Verify water temperature of UHS is 85°F. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.7.9-1 Amendment No. XXX
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| | |
| CRV Filtration 3.7.10 3.7 PLANT SYSTEMS 3.7.10 Control Room Ventilation (CRV) Filtration LCO 3.7.10 Two CRV Filtration trains shall be OPERABLE.
| |
| ----------------------------------NOTE--------------------------------------------
| |
| The control room envelope (CRE) boundary may be opened intermittently under administrative control.
| |
| APPLICABILITY: MODES 1, 2, 3, 4, During CORE ALTERATIONS, During movement of irradiated fuel assemblies, During movement of a fuel cask in or over the Spent Fuel Pool (SFP).
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CRV Filtration train A.1 Restore CRV Filtration 7 days inoperable for reasons train to OPERABLE status.
| |
| other than Condition B.
| |
| B. One or more CRV B.1 Initiate action to implement Immediately Filtration trains mitigating actions.
| |
| inoperable due to inoperable CRE AND boundary in MODE 1, 2, 3, or 4. B.2 Verify mitigating actions 24 hours ensure CRE occupant radiological exposures will not exceed limits, and CRE occupants are protected from chemical and smoke hazards.
| |
| AND B.3 Restore CRE boundary to 90 days OPERABLE status.
| |
| Palisades Nuclear Plant 3.7.10-1 Amendment No. XXX
| |
| | |
| CRV Filtration 3.7.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. ------------NOTE------------ C.1 Initiate action to implement Immediately Not applicable when mitigating actions.
| |
| second CRV Filtration train intentionally made AND inoperable.
| |
| --------------------------------- C.2 Verify LCO 3.4.16, PCS 1 hour Specific Activity, is met.
| |
| Two CRV Filtration trains AND inoperable in MODE 1, 2, 3, or 4 for reasons C.3 Restore at least one CRV 24 hour other than Condition B. Filtration train to OPERABLE status.
| |
| D. Required Action and D.1 Place OPERABLE CRV Immediately associated Completion Filtration train in Time of Condition A not emergency mode.
| |
| met during CORE ALTERATIONS, during OR movement of irradiated fuel assemblies, or D.2.1 Suspend CORE Immediately during movement of a ALTERATIONS.
| |
| fuel cask in or over the SFP. AND D.2.2 Suspend movement of Immediately irradiated fuel assemblies.
| |
| AND D.2.3 Suspend movement of a Immediately fuel cask in or over the SFP.
| |
| Palisades Nuclear Plant 3.7.10-2 Amendment No. XXX
| |
| | |
| CRV Filtration 3.7.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. Two CRV Filtration E.1 Suspend CORE Immediately trains inoperable during ALTERATIONS.
| |
| CORE ALTERATIONS, during movement of AND irradiated fuel assemblies, or during E.2 Suspend movement of Immediately movement of a fuel cask irradiated fuel assemblies.
| |
| in or over the SFP.
| |
| AND OR E.3 Suspend movement of a Immediately One or more CRV fuel cask in or over the Filtration trains SFP.
| |
| inoperable due to an inoperable CRE boundary during CORE ALTERATIONS, during movement of irradiated fuel assemblies, or during movement of a fuel cask in or over the SFP.
| |
| F. Required Action and F.1 Be in MODE 3. 6 hours associated Completion Time of Condition A, B, AND or C not met in MODE 1, 2, 3, or 4. F.2 Be in MODE 5. 36 hours Palisades Nuclear Plant 3.7.10-3 Amendment No. XXX
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| | |
| CRV Filtration 3.7.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Operate each CRV Filtration train for In accordance with 10 continuous hours with associated heater the Surveillance (VHX-26A or VHX-26B) operating. Frequency Control Program SR 3.7.10.2 Perform required CRV Filtration filter testing in In accordance with accordance with the Ventilation Filter Testing the Ventilation Filter Program. Testing Program SR 3.7.10.3 -----------------------------NOTE-----------------------------
| |
| Only required to be met in MODES 1, 2, 3, and 4, and during movement of irradiated fuel assemblies in containment.
| |
| Verify each CRV Filtration train actuates on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program SR 3.7.10.4 Perform required CRE unfiltered air inleakage In accordance with testing in accordance with the Control Room the Control Room Envelope Habitability Program. Envelope Habitability Program Palisades Nuclear Plant 3.7.10-4 Amendment No. XXX
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| | |
| CRV Cooling 3.7.11 3.7 PLANT SYSTEMS 3.7.11 Control Room Ventilation (CRV) Cooling LCO 3.7.11 Two CRV Cooling trains shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, 4, During CORE ALTERATIONS, During movement of irradiated fuel assemblies, During movement of a fuel cask in or over the Spent Fuel Pool (SFP).
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CRV Cooling train A.1 Restore CRV Cooling 30 days inoperable. train to OPERABLE status.
| |
| B. -------------NOTE-------------- B.1 Restore at least one 24 hours Not applicable when CRV Cooling train to second CRV Cooling train OPERABLE status.
| |
| intentionally made inoperable.
| |
| Two CRV Cooling trains inoperable in MODE 1, 2, 3, or 4.
| |
| Palisades Nuclear Plant 3.7.11-1 Amendment No. XXX
| |
| | |
| CRV Cooling 3.7.11 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time of Condition A or B AND not met in MODE 1, 2, 3, or 4. C.2 Be in MODE 5. 36 hours D. Required Action and D.1 Place OPERABLE CRV Immediately associated Completion Cooling train in Time of Condition A not operation.
| |
| met during CORE ALTERATIONS, during OR movement of irradiated fuel assemblies, or D.2.1 Suspend CORE Immediately movement of a fuel cask in ALTERATIONS.
| |
| or over the SFP.
| |
| AND D.2.2 Suspend movement of Immediately irradiated fuel assemblies.
| |
| AND D.2.3 Suspend movement of a Immediately fuel cask in or over the SFP.
| |
| Palisades Nuclear Plant 3.7.11-2 Amendment No. XXX
| |
| | |
| CRV Cooling 3.7.11 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. Two CRV Cooling trains E.1 Suspend CORE Immediately inoperable during CORE ALTERATIONS.
| |
| ALTERATIONS, during movement of irradiated AND fuel assemblies, or movement of a fuel cask in E.2 Suspend movement of Immediately or over the SFP. irradiated fuel assemblies.
| |
| AND E.3 Suspend movement of a Immediately fuel cask in or over the SFP.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.11.1 Verify each CRV Cooling train has the capability to In accordance with remove the assumed heat load. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.7.11-3 Amendment No. XXX
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| | |
| Fuel Handling Area Ventilation System 3.7.12 3.7 PLANT SYSTEMS 3.7.12 Fuel Handling Area Ventilation System LCO 3.7.12 The Fuel Handling Area Ventilation System shall be OPERABLE with one fuel handling area exhaust fan aligned to the emergency filter bank and in operation.
| |
| APPLICABILITY: During movement of irradiated fuel assemblies in the fuel handling building when irradiated fuel assemblies with < 30 days decay time are in the fuel handling building, During movement of a fuel cask in or over the SFP when irradiated fuel assemblies with < 90 days decay time are in the fuel handling building, During CORE ALTERATIONS when irradiated fuel assemblies with
| |
| < 30 days decay time are in the containment with the equipment hatch open, During movement of irradiated fuel assemblies in the containment when irradiated fuel assemblies with < 30 days decay time are in the containment with the equipment hatch open.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel Handling Area A.1 Suspend movement of Immediately Ventilation System not fuel assemblies.
| |
| aligned or in operation.
| |
| AND OR A.2 Suspend CORE Immediately Fuel Handling Area ALTERATIONS.
| |
| Ventilation System inoperable. AND A.3 Suspend movement of a Immediately fuel cask in or over the SFP.
| |
| Palisades Nuclear Plant 3.7.12-1 Amendment No. XXX
| |
| | |
| Fuel Handling Area Ventilation System 3.7.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Perform required Fuel Handling Area Ventilation In accordance with System filter testing in accordance with the the Ventilation Filter Ventilation Filter Testing Program. Testing Program SR 3.7.12.2 Verify the flow rate of the Fuel Handling Area In accordance with Ventilation System, when aligned to the the Surveillance emergency filter bank, is 5840 cfm and Frequency Control 8760 cfm. Program Palisades Nuclear Plant 3.7.12-2 Amendment No. XXX
| |
| | |
| ESRV Dampers 3.7.13 3.7 PLANT SYSTEMS 3.7.13 Engineered Safeguards Room Ventilation (ESRV) Dampers LCO 3.7.13 Two ESRV Damper trains shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more ESRV A.1 Initiate action to isolate Immediately Damper trains inoperable. associated ESRV Damper train(s).
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify each ESRV Damper train closes on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.7.13-1 Amendment No. XXX
| |
| | |
| SFP Water Level 3.7.14 3.7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool (SFP) Water Level LCO 3.7.14 The SFP water level shall be 647 ft elevation.
| |
| ----------------------------------------------NOTE------------------------------------------
| |
| SFP level may be below the 647 ft elevation to support fuel cask movement, if the displacement of water by the fuel cask when submerged in the SFP, would raise SFP level to 647 ft elevation.
| |
| APPLICABILITY: During movement of irradiated fuel assemblies in the SFP, During movement of a fuel cask in or over the SFP.
| |
| ACTIONS
| |
| ----------------------------------------------------------NOTE-----------------------------------------------------------
| |
| LCO 3.0.3 is not applicable.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. SFP water level not within A.1 Suspend movement of Immediately limit. irradiated fuel assemblies in SFP.
| |
| AND A.2 Suspend movement of Immediately fuel cask in or over the SFP.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the SFP water level is 647 ft elevation. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.7.14-1 Amendment No. 271, 272, XXX
| |
| | |
| SFP Boron Concentration 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool (SFP) Boron Concentration LCO 3.7.15 The SFP boron concentration shall be 1720 ppm.
| |
| APPLICABILITY: When fuel assemblies are stored in the Spent Fuel Pool.
| |
| ACTIONS
| |
| ----------------------------------------------------------NOTE-----------------------------------------------------------
| |
| LCO 3.0.3 is not applicable.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. SFP boron concentration A.1 Suspend movement of Immediately not within limit. fuel assemblies in the SFP.
| |
| AND A.2 Initiate action to restore Immediately SFP boron concentration to within limit.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the SFP boron concentration is within limit. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.7.15-1 Amendment No. 271, 272, XXX
| |
| | |
| Spent Fuel Pool Storage 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Pool Storage LCO 3.7.16 Storage in the spent fuel pool shall be as follows:
| |
| : a. Each fuel assembly and non-fissile bearing component stored in a Region I Carborundum equipped storage rack shall be within the limitations in Specification 4.3.1.1 and, as applicable, within the requirements of the maximum nominal planar average U-235 enrichment and burnup of Tables 3.7.16-2, 3.7.16-3, 3.7.16-4 or 3.7.16-5,
| |
| : b. Fuel assemblies in a Region I Metamic equipped storage rack shall be within the limitations in Specification 4.3.1.2, and
| |
| : c. The combination of maximum nominal planar average U-235 enrichment, burnup, and decay time of each fuel assembly stored in Region II shall be within the requirements of Table 3.7.16-1.
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| APPLICABILITY: Whenever any fuel assembly or non-fissile bearing component is stored in the spent fuel pool or the north tilt pit.
| |
| ACTIONS
| |
| ----------------------------------------------------------NOTE-----------------------------------------------------------
| |
| LCO 3.0.3 is not applicable.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 Initiate action to restore Immediately not met. the noncomplying fuel assembly or non-fissile bearing component within requirements.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify by administrative means each fuel assembly Prior to storing the or non-fissile bearing component meets fuel fuel assembly or storage requirements. non-fissile bearing component in the spent fuel pool Palisades Nuclear Plant 3.7.16-1 Amendment No. 250, 272, XXX
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| | |
| Secondary Specific Activity 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Secondary Specific Activity LCO 3.7.17 The specific activity of the secondary coolant shall be 0.10 µCi/gm DOSE EQUIVALENT I-131.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Specific activity not within A.1 Be in MODE 3. 6 hours limit.
| |
| AND A.2 Be in MODE 5. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify the specific activity of the secondary coolant is In accordance with within limit. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.7.17-1 Amendment No. XXX
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| | |
| AC Sources - Operating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources - Operating LCO 3.8.1 The following AC electrical sources shall be OPERABLE:
| |
| : a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System; and
| |
| : b. Two Diesel Generators (DGs) each capable of supplying one train of the onsite Class 1E AC Electrical Power Distribution System.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS
| |
| -----------------------------------------------------------NOTE------------------------------------------------------
| |
| LCO 3.0.4.b is not applicable to DGs.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One offsite circuit A.1 Perform SR 3.8.1.1 1 hour inoperable. (offsite source check) for OPERABLE offsite AND circuit.
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| Once per 8 hours AND thereafter A.2 Restore offsite circuit to 72 hours OPERABLE status.
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| AND 10 days from discovery of failure to meet LCO Palisades Nuclear Plant 3.8.1-1 Amendment No. XXX
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| | |
| AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One DG inoperable. B.1 Perform SR 3.8.1.1 1 hour (offsite source check) for the OPERABLE AND offsite circuit(s).
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| Once per 8 hours thereafter AND B.2 Declare required 4 hours from feature(s) supported by discovery of the inoperable DG Condition B inoperable when its concurrent with redundant required inoperability of feature(s) is inoperable. redundant required feature(s)
| |
| AND B.3.1 Determine OPERABLE 24 hours DG is not inoperable due to common cause failure.
| |
| OR B.3.2 Perform SR 3.8.1.2 24 hours (start test) for OPERABLE DG.
| |
| AND B.4 Restore DG to 7 days OPERABLE status.
| |
| AND 10 days from discovery of failure to meet LCO Palisades Nuclear Plant 3.8.1-2 Amendment No. XXX
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| | |
| AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Two offsite circuits C.1 Declare required 12 hours from inoperable. feature(s) inoperable discovery of when its redundant Condition C required feature(s) is concurrent with inoperable. inoperability of redundant required feature(s)
| |
| AND C.2 Restore one offsite 24 hours circuit to OPERABLE status.
| |
| D. One offsite circuit ------------------NOTE------------------
| |
| inoperable. Enter applicable Conditions and Required Actions of LCO 3.8.9, AND "Distribution Systems -
| |
| Operating," when Condition D is One DG inoperable. entered with no AC power source to any train.
| |
| D.1 Restore offsite circuit to 12 hours OPERABLE status.
| |
| OR D.2 Restore DG to 12 hours OPERABLE status.
| |
| E. Two DGs inoperable. E.1 Restore one DG to 2 hours OPERABLE status.
| |
| Palisades Nuclear Plant 3.8.1-3 Amendment No. XXX
| |
| | |
| AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and F.1 Be in MODE 3. 6 hours Associated Completion Time of Condition A, B, C, AND D, or E not met.
| |
| F.2 Be in MODE 5. 36 hours G. Three or more AC sources G.1 Enter LCO 3.0.3. Immediately inoperable.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and voltage for In accordance with each offsite circuit. the Surveillance Frequency Control Program SR 3.8.1.2 Verify each DG starts from standby conditions and In accordance with achieves: the Surveillance Frequency Control
| |
| : a. In 10 seconds, ready-to-load status; and Program
| |
| : b. Steady state voltage 2280 V and 2520 V, and frequency 59.5 Hz and 61.2 Hz.
| |
| Palisades Nuclear Plant 3.8.1-4 Amendment No. XXX
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.3 -----------------------------NOTES---------------------------
| |
| : 1. Momentary transients outside the load range do not invalidate this test.
| |
| : 2. This Surveillance shall be conducted on only one DG at a time.
| |
| : 3. This Surveillance shall be preceded by and immediately follow without shutdown a successful performance of SR 3.8.1.2.
| |
| Verify each DG is synchronized and loaded, and In accordance with operates for 60 minutes: the Surveillance Frequency Control
| |
| : a. For 15 minutes loaded to greater than or Program equal to peak accident load; and
| |
| : b. For the remainder of the test at a load 2300 kW and 2500 kW.
| |
| SR 3.8.1.4 Verify each day tank contains 2500 gallons of In accordance with fuel oil. the Surveillance Frequency Control Program SR 3.8.1.5 Verify each DG rejects a load greater than or equal In accordance with to its associated single largest post-accident load, the Surveillance and: Frequency Control Program
| |
| : a. Following load rejection, the frequency is 68 Hz;
| |
| : b. Within 3 seconds following load rejection, the voltage is 2280 V and 2640 V; and
| |
| : c. Within 3 seconds following load rejection, the frequency is 59.5 Hz and 61.5 Hz.
| |
| Palisades Nuclear Plant 3.8.1-5 Amendment No. XXX
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.6 Verify each DG, operating at a power factor 0.9, In accordance with does not trip, and voltage is maintained 4000 V the Surveillance during and following a load rejection of 2300 kW Frequency Control and 2500 kW. Program SR 3.8.1.7 -----------------------------NOTE-----------------------------
| |
| This Surveillance shall not be performed in MODE 1, 2, 3, or 4.
| |
| Verify on an actual or simulated loss of offsite In accordance with power signal: the Surveillance Frequency Control
| |
| : a. De-energization of emergency buses; Program
| |
| : b. Load shedding from emergency buses;
| |
| : c. DG auto-starts from standby condition and:
| |
| : 1. energizes permanently connected loads in 10 seconds,
| |
| : 2. energizes auto-connected shutdown loads through automatic load sequencer,
| |
| : 3. maintains steady state voltage 2280 V and 2520 V,
| |
| : 4. maintains steady state frequency 59.5 Hz and 61.2 Hz, and
| |
| : 5. supplies permanently connected loads for 5 minutes.
| |
| Palisades Nuclear Plant 3.8.1-6 Amendment No. XXX
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.8 -----------------------------NOTE-----------------------------
| |
| Momentary transients outside the load and power factor ranges do not invalidate this test.
| |
| In accordance with Verify each DG, operating at a power factor 0.9, the Surveillance operates for 24 hours: Frequency Control Program
| |
| : a. For 100 minutes loaded its peak accident loading; and
| |
| : b. For the remaining hours of the test loaded 2300 kW and 2500 kW.
| |
| SR 3.8.1.9 -----------------------------NOTE----------------------------
| |
| This Surveillance shall not be performed in MODE 1, 2, 3, or 4.
| |
| Verify each DG:
| |
| In accordance with
| |
| : a. Synchronizes with offsite power source while the Surveillance supplying its associated 2400 V bus upon a Frequency Control simulated restoration of offsite power; Program
| |
| : b. Transfers loads to offsite power source; and
| |
| : c. Returns to ready-to-load operation.
| |
| Palisades Nuclear Plant 3.8.1-7 Amendment No. XXX
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.10 -----------------------------NOTE-----------------------------
| |
| This Surveillance shall not be performed in MODE 1, 2, 3, or 4.
| |
| Verify the time of each sequenced load is within In accordance with
| |
| +/- 0.3 seconds of design timing for each automatic the Surveillance load sequencer. Frequency Control Program SR 3.8.1.11 -----------------------------NOTE-----------------------------
| |
| This Surveillance shall not be performed in MODE 1, 2, 3, or 4.
| |
| Verify on an actual or simulated loss of offsite In accordance with power signal in conjunction with an actual or the Surveillance simulated safety injection signal: Frequency Control Program
| |
| : a. De-energization of emergency buses;
| |
| : b. Load shedding from emergency buses;
| |
| : c. DG auto-starts from standby condition and:
| |
| : 1. energizes permanently connected loads in 10 seconds,
| |
| : 2. energizes auto-connected emergency loads through its automatic load sequencer,
| |
| : 3. achieves steady state voltage 2280 V and 2520 V,
| |
| : 4. achieves steady state frequency 59.5 Hz and 61.2 Hz, and
| |
| : 5. supplies permanently connected loads for 5 minutes.
| |
| Palisades Nuclear Plant 3.8.1-8 Amendment No. XXX
| |
| | |
| AC Sources - Shutdown 3.8.2 3.8 ELECTRICAL POWER SYSTEMS 3.8.2 AC Sources - Shutdown LCO 3.8.2 The following AC electrical power sources shall be OPERABLE:
| |
| : a. One qualified circuit between the offsite transmission network and the onsite Class 1E AC electrical power distribution subsystem(s) required by LCO 3.8.10, "Distribution Systems - Shutdown"; and
| |
| : b. One Diesel Generator (DG) capable of supplying one train of the onsite Class 1E AC electrical power distribution subsystem(s) required by LCO 3.8.10.
| |
| APPLICABILITY: MODES 5 and 6, During movement of irradiated fuel assemblies.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. The required offsite circuit -----------------NOTE-------------------
| |
| inoperable. Enter applicable Conditions and Required Actions of LCO 3.8.10, with one required train de-energized as a result of Condition A.
| |
| A.1 Declare affected Immediately required feature(s) with no offsite power available inoperable.
| |
| OR A.2.1 Suspend CORE Immediately ALTERATIONS.
| |
| AND (continued)
| |
| Palisades Nuclear Plant 3.8.2-1 Amendment No. XXX
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| | |
| AC Sources - Shutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend movement of Immediately irradiated fuel assemblies.
| |
| AND A.2.3 Initiate action to Immediately suspend operations involving positive reactivity additions.
| |
| AND A.2.4 Initiate action to restore Immediately required offsite power circuit to OPERABLE status.
| |
| B. The required DG B.1 Suspend CORE Immediately inoperable. ALTERATIONS.
| |
| AND B.2 Suspend movement of Immediately irradiated fuel assemblies.
| |
| AND B.3 Initiate action to Immediately suspend operations involving positive reactivity additions.
| |
| AND B.4 Initiate action to restore Immediately required DG to OPERABLE status.
| |
| Palisades Nuclear Plant 3.8.2-2 Amendment No. XXX
| |
| | |
| AC Sources - Shutdown 3.8.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.2.1 For AC sources required to be OPERABLE, the In accordance with following SRs of Specification 3.8.1, "AC Sources - applicable SRs Operating" are applicable:
| |
| SR 3.8.1.1 SR 3.8.1.2 SR 3.8.1.4.
| |
| Palisades Nuclear Plant 3.8.2-3 Amendment No. XXX
| |
| | |
| Diesel Fuel, Lube Oil, and Starting Air 3.8.3 3.8 ELECTRICAL POWER SYSTEMS 3.8.3 Diesel Fuel, Lube Oil, and Starting Air LCO 3.8.3 For each Diesel Generator (DG):
| |
| : a. The stored diesel fuel oil, lube oil, and starting air subsystem shall be within limits, and
| |
| : b. Both diesel fuel oil transfer systems shall be OPERABLE.
| |
| APPLICABILITY: When associated DG is required to be OPERABLE.
| |
| ACTIONS
| |
| ----------------------------------------------------------NOTE-----------------------------------------------------------
| |
| Separate Condition entry is allowed for each DG.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel oil inventory less than A.1 Restore fuel oil 48 hours a 7 day supply and greater inventory to within limits.
| |
| than a 6 day supply.
| |
| B. Stored lube oil inventory B.1 Restore stored lube oil 48 hours less than a 7 day supply inventory to within limits.
| |
| and greater than a 6 day supply.
| |
| C. Fuel transfer system C.1 Restore fuel transfer 12 hours (P-18A) inoperable. system to OPERABLE status.
| |
| D. Fuel transfer system D.1 Restore fuel transfer 7 days (P-18B) inoperable. system to OPERABLE status.
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| Palisades Nuclear Plant 3.8.3-1 Amendment No. XXX
| |
| | |
| Diesel Fuel, Lube Oil, and Starting Air 3.8.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. Both fuel transfer systems E.1 Restore one fuel 8 hours inoperable. transfer system to OPERABLE status.
| |
| F. Fuel oil properties other F.1 Restore stored fuel oil 30 days than viscosity, and water properties to within and sediment, not within limits.
| |
| limits.
| |
| G. Required Action and G.1 Declare associated Immediately associated Completion DG(s) inoperable.
| |
| Time not met.
| |
| OR Stored diesel fuel oil, lube oil, or starting air subsystem not within limits for reasons other than Condition A, B, or F.
| |
| Palisades Nuclear Plant 3.8.3-2 Amendment No. XXX
| |
| | |
| Diesel Fuel, Lube Oil, and Starting Air 3.8.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify the fuel oil storage subsystem contains a In accordance with 7 day supply of fuel. the Surveillance Frequency Control Program SR 3.8.3.2 Verify stored lube oil inventory is a 7 day supply. In accordance with the Surveillance Frequency Control Program SR 3.8.3.3 Verify fuel oil properties of new and stored fuel oil are In accordance with tested in accordance with, and maintained within the the Fuel Oil limits of, the Fuel Oil Testing Program. Testing Program SR 3.8.3.4 Verify each DG air start receiver pressure is In accordance with 200 psig. the Surveillance Frequency Control Program SR 3.8.3.5 Check for and remove excess accumulated water from In accordance with the fuel oil storage tank. the Surveillance Frequency Control Program SR 3.8.3.6 Verify the fuel oil transfer system operates to transfer In accordance with fuel oil from the fuel oil storage tank to each DG day the Surveillance tank and engine mounted tank. Frequency Control Program Palisades Nuclear Plant 3.8.3-3 Amendment No. XXX
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| | |
| DC Sources - Operating 3.8.4 3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources - Operating LCO 3.8.4 The Left Train and Right Train DC electrical power sources shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required DC electrical A.1 Verify functional 2 hours power source battery cross-connected battery charger inoperable. charger is connected supplying power to the affected DC train.
| |
| AND A.2 Restore required DC 7 days electrical power source battery charger to OPERABLE status.
| |
| B. One required DC electrical B.1 Verify OPERABLE 2 hours power source battery directly connected and inoperable. functional cross-connected battery chargers are connected supplying power to the affected DC train.
| |
| AND B.2 Restore required DC 24 hours electrical power source battery to OPERABLE status.
| |
| Palisades Nuclear Plant 3.8.4-1 Amendment No. XXX
| |
| | |
| DC Sources - Operating 3.8.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time not met. AND C.2 Be in MODE 5. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is 125 V on float In accordance with charge. the Surveillance Frequency Control Program SR 3.8.4.2 Verify no visible corrosion at battery terminals and In accordance with connectors. the Surveillance Frequency Control OR Program Verify battery connection resistance is 50 µohm for inter-cell connections, 360 µohm for inter-rack connections, and 360 µohm for inter-tier connections.
| |
| SR 3.8.4.3 Inspect battery cells, cell plates, and racks for In accordance with visual indication of physical damage or abnormal the Surveillance deterioration that could degrade battery Frequency Control performance. Program Palisades Nuclear Plant 3.8.4-2 Amendment No. XXX
| |
| | |
| DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.4 Remove visible terminal corrosion and verify In accordance with battery cell to cell and terminal connections are the Surveillance coated with anti-corrosion material. Frequency Control Program SR 3.8.4.5 Verify battery connection resistance is In accordance with 50 µohm for inter-cell connections, 360 µohm the Surveillance for inter-rack connections, and 360 µohm for Frequency Control inter-tier connections. Program SR 3.8.4.6 Verify each required battery charger supplies In accordance with 180 amps at 125 V for 8 hours. the Surveillance Frequency Control Program SR 3.8.4.7 -----------------------------NOTES---------------------------
| |
| : 1. The modified performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test in SR 3.8.4.7.
| |
| : 2. This Surveillance shall not be performed in MODE 1, 2, 3, or 4.
| |
| Verify battery capacity is adequate to supply, and maintain in OPERABLE status, the required In accordance with emergency loads for the design duty cycle when the Surveillance subjected to a battery service test. Frequency Control Program Palisades Nuclear Plant 3.8.4-3 Amendment No. XXX
| |
| | |
| DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.8 -------------------------------NOTE---------------------------
| |
| This Surveillance shall not be performed in MODE 1, 2, 3, or 4.
| |
| Verify battery capacity is 80% of the In accordance with manufacturer's rating when subjected to a the Surveillance performance discharge test or a modified Frequency Control performance discharge test. Program AND 12 months when battery shows degradation or has reached 85% of the expected life with capacity < 100% of manufacturer's rating AND 24 months when battery has reached 85% of the expected life with capacity 100% of manufacturer's rating Palisades Nuclear Plant 3.8.4-4 Amendment No. XXX
| |
| | |
| DC Sources - Shutdown 3.8.5 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources - Shutdown LCO 3.8.5 DC electrical power source(s) shall be OPERABLE to support the DC electrical power distribution subsystem(s) required by LCO 3.8.10, "Distribution Systems Shutdown."
| |
| APPLICABILITY: MODES 5 and 6, During movement of irradiated fuel assemblies.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required DC A.1 Declare affected Immediately electrical power sources required feature(s) inoperable. inoperable.
| |
| OR A.2.1 Suspend CORE Immediately ALTERATIONS.
| |
| AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies.
| |
| AND A.2.3 Initiate action to Immediately suspend operations involving positive reactivity additions.
| |
| AND (continued)
| |
| Palisades Nuclear Plant 3.8.5-1 Amendment No. XXX
| |
| | |
| DC Sources - Shutdown 3.8.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.4 Initiate action to restore Immediately required DC electrical power source(s) to OPERABLE status.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.5.1 For DC sources required to be OPERABLE, the In accordance with following SRs are applicable: applicable SRs SR 3.8.4.1 SR 3.8.4.3 SR 3.8.4.5 SR 3.8.4.2 SR 3.8.4.4 SR 3.8.4.6.
| |
| Palisades Nuclear Plant 3.8.5-2 Amendment No. XXX
| |
| | |
| Battery Cell Parameters 3.8.6 3.8 ELECTRICAL POWER SYSTEMS 3.8.6 Battery Cell Parameters LCO 3.8.6 Battery cell parameters for the Left Train and Right Train batteries shall be within limits.
| |
| APPLICABILITY: When associated DC electrical power source(s) are required to be OPERABLE.
| |
| ACTIONS
| |
| -----------------------------------------------------------NOTE----------------------------------------------------------
| |
| Separate Condition entry is allowed for each battery.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more batteries with A.1 Verify pilot cells 1 hour one or more battery cell electrolyte level and float parameters not within voltage meet Category A or B limits. Table 3.8.6-1 Category C limits.
| |
| AND A.2 Verify battery cell 24 hours parameters meet Table 3.8.6-1 AND Category C limits.
| |
| Once per 7 days thereafter AND A.3 Restore battery cell 31 days parameters to Category A and B limits of Table 3.8.6-1.
| |
| Palisades Nuclear Plant 3.8.6-1 Amendment No. XXX
| |
| | |
| Battery Cell Parameters 3.8.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Declare associated Immediately associated Completion battery inoperable.
| |
| Time of Condition A not met.
| |
| OR One or more batteries with average electrolyte temperature of the representative cells
| |
| < 70°F.
| |
| OR One or more batteries with one or more battery cell parameters not within Category C limits.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters meet Table 3.8.6-1 In accordance with Category A limits. the Surveillance Frequency Control Program SR 3.8.6.2 Verify average electrolyte temperature of In accordance with representative cells is 70°F. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.8.6-2 Amendment No. XXX
| |
| | |
| Battery Cell Parameters 3.8.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.3 Verify battery cell parameters meet Table 3.8.6-1 In accordance with Category B limits. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.8.6-3 Amendment No. XXX
| |
| | |
| Battery Cell Parameters 3.8.6 Table 3.8.6-1 (page 1 of 1)
| |
| Battery Surveillance Requirements CATEGORY A: CATEGORY B: CATEGORY C:
| |
| NORMAL LIMITS NORMAL LIMITS ALLOWABLE FOR EACH FOR EACH LIMITS FOR EACH DESIGNATED CONNECTED CONNECTED PARAMETER PILOT CELL CELL CELL Electrolyte Level > Minimum level > Minimum level Above top of plates, indication mark, and indication mark, and and not overflowing 1/4 inch above 1/4 inch above maximum level maximum level indication mark(a) indication mark(a)
| |
| Float Voltage 2.13 V 2.13 V > 2.07 V Specific Gravity(b)(c) 1.205 1.200 Not more than 0.020 below AND average connected cells Average of connected cells AND 1.205 Average of all connected cells 1.195 (a) It is acceptable for the electrolyte level to temporarily increase above the specified maximum during equalizing charges provided it is not overflowing.
| |
| (b) Corrected for electrolyte temperature and level. Level correction is not required, however, when battery charging is < 2 amps when on float charge.
| |
| (c) A battery charging current of < 2 amps when on float charge is acceptable for meeting specific gravity limits.
| |
| Palisades Nuclear Plant 3.8.6-4 Amendment No. XXX
| |
| | |
| Inverters - Operating 3.8.7 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 Inverters - Operating LCO 3.8.7 Four inverters shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One inverter inoperable. -----------------NOTE------------------
| |
| Enter applicable Conditions and Required Actions of LCO 3.8.9, "Distribution Systems -
| |
| Operating" with any Preferred AC bus de-energized.
| |
| A.1 Restore inverter to 24 hours OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct inverter voltage, frequency, and In accordance with alignment to Preferred AC buses. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.8.7-1 Amendment No. XXX
| |
| | |
| Inverters - Shutdown 3.8.8 3.8 ELECTRICAL POWER SYSTEMS 3.8.8 Inverters - Shutdown LCO 3.8.8 Inverter(s) shall be OPERABLE to support the onsite Class 1E Preferred AC bus electrical power distribution subsystem(s) required by LCO 3.8.10, "Distribution Systems Shutdown."
| |
| APPLICABILITY: MODES 5 and 6, During movement of irradiated fuel assemblies.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare affected Immediately inverters inoperable. required feature(s) inoperable.
| |
| OR A.2.1 Suspend CORE Immediately ALTERATIONS.
| |
| AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies.
| |
| AND A.2.3 Initiate action to Immediately suspend operations involving positive reactivity additions.
| |
| AND A.2.4 Initiate action to restore Immediately required inverters to OPERABLE status.
| |
| Palisades Nuclear Plant 3.8.8-1 Amendment No. XXX
| |
| | |
| Inverters - Shutdown 3.8.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct inverter voltage, frequency, and In accordance with alignment to required Preferred AC buses. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.8.8-2 Amendment No. XXX
| |
| | |
| Distribution Systems - Operating 3.8.9 3.8 ELECTRICAL POWER SYSTEMS 3.8.9 Distribution Systems - Operating LCO 3.8.9 Left Train and Right Train AC, DC, and Preferred AC bus electrical power distribution subsystems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more AC electrical A.1 Restore AC electrical 8 hours power distribution power distribution subsystems in one train subsystem(s) to AND inoperable. OPERABLE status.
| |
| 16 hours from discovery of failure to meet LCO B. One Preferred AC bus B.1 Restore Preferred AC 8 hours inoperable. bus to OPERABLE status. AND 16 hours from discovery of failure to meet LCO C. One or more DC electrical C.1 Restore DC electrical 8 hours power distribution power distribution subsystems in one train subsystem(s) to AND inoperable. OPERABLE status.
| |
| 16 hours from discovery of failure to meet LCO Palisades Nuclear Plant 3.8.9-1 Amendment No. XXX
| |
| | |
| Distribution Systems - Operating 3.8.9 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 6 hours associated Completion Time not met. AND D.2 Be in MODE 5. 36 hours E. Two or more inoperable E.1 Enter LCO 3.0.3. Immediately distribution subsystems that result in a loss of function.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to In accordance with required AC, DC, and Preferred AC bus electrical the Surveillance power distribution subsystems. Frequency Control Program Palisades Nuclear Plant 3.8.9-2 Amendment No. XXX
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| | |
| Distribution Systems - Shutdown 3.8.10 3.8 ELECTRICAL POWER SYSTEMS 3.8.10 Distribution Systems - Shutdown LCO 3.8.10 The necessary portion of AC, DC, and Preferred AC bus electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.
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| APPLICABILITY: MODES 5 and 6, During movement of irradiated fuel assemblies.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required AC, A.1 Declare associated Immediately DC, or Preferred AC bus supported required electrical power distribution feature(s) inoperable.
| |
| subsystems inoperable.
| |
| OR A.2.1 Suspend CORE Immediately ALTERATIONS.
| |
| AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies.
| |
| AND A.2.3 Initiate action to Immediately suspend operations involving positive reactivity additions.
| |
| AND (continued)
| |
| Palisades Nuclear Plant 3.8.10-1 Amendment No. XXX
| |
| | |
| Distribution Systems - Shutdown 3.8.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.4 Initiate actions to restore Immediately required AC, DC, and Preferred AC bus electrical power distribution subsystems to OPERABLE status.
| |
| AND A.2.5 Declare associated Immediately required shutdown cooling train inoperable and not in operation.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and voltage to In accordance with required AC, DC, and Preferred AC bus electrical the Surveillance power distribution subsystems. Frequency Control Program Palisades Nuclear Plant 3.8.10-2 Amendment No. XXX
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| | |
| Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LCO 3.9.1 Boron concentrations of the Primary Coolant System and the refueling cavity shall be maintained at the REFUELING BORON CONCENTRATION.
| |
| APPLICABILITY: MODE 6.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration not A.1 Suspend CORE Immediately within limit. ALTERATIONS.
| |
| AND A.2 Suspend positive Immediately reactivity additions.
| |
| AND A.3 Initiate action to restore Immediately boron concentration to within limit.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Verify boron concentration is at the REFUELING In accordance with BORON CONCENTRATION. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.9.1-1 Amendment No. XXX
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| | |
| Nuclear Instrumentation 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Nuclear Instrumentation LCO 3.9.2 Two source range channels shall be OPERABLE.
| |
| APPLICABILITY: MODE 6.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One source range channel A.1 Suspend CORE Immediately inoperable. ALTERATIONS.
| |
| AND A.2 Suspend positive Immediately reactivity additions.
| |
| B. Two source range B.1 Initiate action to restore Immediately channels inoperable. one source range channel to OPERABLE status.
| |
| AND B.2 Perform SR 3.9.1.1 Once per (PCS boron 12 hours concentration verification).
| |
| Palisades Nuclear Plant 3.9.2-1 Amendment No. XXX
| |
| | |
| Nuclear Instrumentation 3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.9.2.2 -----------------------------NOTE------------------------------
| |
| Neutron detectors are excluded from the CHANNEL CALIBRATION.
| |
| Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program Palisades Nuclear Plant 3.9.2-2 Amendment No. XXX
| |
| | |
| Containment Penetrations 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Containment Penetrations LCO 3.9.3 The containment penetrations shall be in the following status:
| |
| : a. The equipment hatch closed and held in place by four bolts;
| |
| ----------------------------------------------NOTE------------------------------------------
| |
| The equipment hatch is only required to be closed when the Fuel Handling Area Ventilation System is not in compliance with LCO 3.7.12, "Fuel Handling Area Ventilation System."
| |
| : b. One door in the personnel air lock closed;
| |
| --------------------------------------------NOTE--------------------------------------------
| |
| One door in the personnel air lock is only required to be closed when the equipment hatch is closed.
| |
| : c. One door in the emergency air lock closed; and
| |
| : d. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:
| |
| : 1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or
| |
| : 2. capable of being closed by an OPERABLE Refueling Containment High Radiation Initiation signal.
| |
| APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment.
| |
| Palisades Nuclear Plant 3.9.3-1 Amendment No. XXX
| |
| | |
| Containment Penetrations 3.9.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more containment A.1 Suspend CORE Immediately penetrations not in ALTERATIONS.
| |
| required status.
| |
| AND A.2 Suspend movement of Immediately irradiated fuel assemblies within containment.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify each required to be met containment In accordance with penetration is in the required status. the Surveillance Frequency Control Program SR 3.9.3.2 -----------------------------NOTE-----------------------------
| |
| Only required to be met for unisolated containment penetrations.
| |
| Verify each required automatic isolation valve In accordance with closes on an actual or simulated Refueling the Surveillance Containment High Radiation signal. Frequency Control Program Palisades Nuclear Plant 3.9.3-2 Amendment No. XXX
| |
| | |
| SDC and Coolant Circulation - High Water Level 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Shutdown Cooling (SDC) and Coolant Circulation - High Water Level LCO 3.9.4 One SDC train shall be OPERABLE and in operation.
| |
| --------------------------------------------NOTES------------------------------------------
| |
| : 1. The required SDC train may not be in operation for 1 hour per 8 hour period, provided no operations are permitted that would cause reduction of the Primary Coolant System boron concentration.
| |
| : 2. The required SDC train may be made inoperable for 2 hours per 8 hour period for testing or maintenance, provided one SDC train is in operation providing flow through the reactor core, and core outlet temperature is 200°F.
| |
| APPLICABILITY: MODE 6 with the refueling cavity water level 647 ft elevation.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required SDC train A.1 Initiate action to restore Immediately inoperable or not in SDC train to operation. OPERABLE status and operation.
| |
| AND A.2 Suspend operations Immediately involving a reduction in primary coolant boron concentration.
| |
| AND (continued)
| |
| Palisades Nuclear Plant 3.9.4-1 Amendment No. XXX
| |
| | |
| SDC and Coolant Circulation - High Water Level 3.9.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Suspend loading Immediately irradiated fuel assemblies in the core.
| |
| AND A.4 Close all containment 4 hours penetrations providing direct access from containment atmosphere to outside atmosphere.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.4.1 Verify one SDC train is in operation and circulating In accordance with primary coolant at a flow rate of 1000 gpm. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.9.4-2 Amendment No. XXX
| |
| | |
| SDC and Coolant Circulation - Low Water Level 3.9.5 3.9 REFUELING OPERATIONS 3.9.5 Shutdown Cooling (SDC) and Coolant Circulation - Low Water Level LCO 3.9.5 Two SDC trains shall be OPERABLE, and one SDC train shall be in operation.
| |
| APPLICABILITY: MODE 6 with the refueling cavity water level < 647 ft elevation.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SDC train inoperable. A.1 Initiate action to restore Immediately SDC train to OPERABLE status.
| |
| OR A.2 Initiate action to Immediately establish the refueling cavity water level 647 ft elevation.
| |
| Palisades Nuclear Plant 3.9.5-1 Amendment No. XXX
| |
| | |
| SDC and Coolant Circulation - Low Water Level 3.9.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. No SDC train OPERABLE B.1 Suspend operations Immediately or in operation. involving a reduction in primary coolant boron concentration.
| |
| AND B.2 Initiate action to restore Immediately one SDC train to OPERABLE status and to operation.
| |
| AND B.3 Initiate action to close Immediately all containment penetrations providing direct access from containment atmosphere to outside atmosphere.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1 Verify one SDC train is in operation and In accordance with circulating primary coolant at a flow rate of the Surveillance 1000 gpm. Frequency Control Program SR 3.9.5.2 Verify correct breaker alignment and indicated In accordance with power available to the required SDC pump that is the Surveillance not in operation. Frequency Control Program Palisades Nuclear Plant 3.9.5-2 Amendment No. XXX
| |
| | |
| Refueling Cavity Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Refueling Cavity Water Level LCO 3.9.6 The refueling cavity water level shall be maintained 647 ft elevation.
| |
| APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refueling cavity water A.1 Suspend CORE Immediately level not within limit. ALTERATIONS.
| |
| AND A.2 Suspend movement of Immediately irradiated fuel assemblies within containment.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify refueling cavity water level is 647 ft In accordance with elevation. the Surveillance Frequency Control Program Palisades Nuclear Plant 3.9.6-1 Amendment No. XXX
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| | |
| Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The Palisades Nuclear Plant is located on property owned by Holtec Palisades, LLC on the eastern shore of Lake Michigan approximately four and one-half miles south of the southern city limits of South Haven, Michigan. The minimum distance to the boundary of the exclusion area as defined in 10 CFR 100.3 shall be 677 meters.
| |
| 4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor core shall contain 204 fuel assemblies. Each assembly shall consist of a matrix of zircaloy-4 or M5 clad fuel rods with an initial composition of depleted, natural, or slightly enriched uranium dioxide (UO2) as fuel material.
| |
| Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. A core plug or plugs may be used to replace one or more fuel assemblies subject to the analysis of the resulting power distribution. Poison may be placed in the fuel bundles for long-term reactivity control.
| |
| 4.2.2 Control Rod Assemblies The reactor core shall contain 45 control rods. Four of these control rods may consist of part-length absorbers. The control material shall be silver-indium-cadmium, as approved by the NRC.
| |
| 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The Region I (See Figure B 3.7.16-1) Carborundum equipped fuel storage racks incorporating Regions 1A, 1B, 1C, 1D, and 1E are designed and shall be maintained with:
| |
| : a. New or irradiated fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.54 weight percent; Palisades Nuclear Plant 4.0-1 Amendment No. 272, 273, XXX
| |
| | |
| Design Features 4.0 4.3 Fuel Storage 4.3.1 Criticality (continued)
| |
| : b. Keff < 1.0 if fully flooded with unborated water, which includes allowances for uncertainties as described in Section 9.11 of the FSAR;
| |
| : c. Keff 0.95 if fully flooded with water borated to 850 ppm, which includes allowances for uncertainties as described in Section 9.11 of the FSAR;
| |
| : d. Regions 1A, 1B, and 1C have a nominal 10.25 inch center to center distance between fuel assemblies;
| |
| : e. Regions 1D and 1E have a nominal 11.25 inch by 10.69 inch center to center distance between fuel assemblies;
| |
| : f. Region 1A is defined as a subregion of the Region I storage racks located in the main spent fuel pool and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1A shall be in a maximum of two-of-four checkerboard loading pattern of two fuel assemblies (or fissile bearing components) and two empty cells.
| |
| Designated empty cells may contain non-fuel bearing components in accordance with Section 4.3.1.1m.2. below;
| |
| : g. Region 1B is defined as a subregion of the Region I storage racks located in the main spent fuel pool and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1B shall be in a maximum of three-of-four loading pattern consisting of three fuel assemblies (or fissile bearing components) and one empty cell. Fuel assemblies in Region 1B shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-2. Designated empty cells may contain non-fuel bearing components in accordance with Section 4.3.1.1m.2. below;
| |
| : h. Region 1C is defined as a subregion of the Region I storage racks located in the main spent fuel pool and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1C may be in a maximum of four-of-four loading pattern with no required empty cells. Fuel assemblies in Region 1C shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-3;
| |
| : i. Interface requirements for the main spent fuel pool between Region 1A, 1B, and 1C are as follows. Region 1A, 1B, and 1C can be distributed in Region I, in the main spent fuel pool, in any manner provided that any two-by-two grouping of storage cells and the assemblies in them correspond to the requirements of 4.3.1.1f., 4.3.1.1g., or 4.3.1.1h. above; Palisades Nuclear Plant 4.0-2 Amendment No. 246, 272, XXX
| |
| | |
| Design Features 4.0 4.3 Fuel Storage 4.3.1 Criticality (continued)
| |
| : j. Region 1D is defined as a subregion of the Region I storage rack located in the north tilt pit and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1D may be in a maximum of three-of-four loading pattern consisting of three fuel assemblies (or fissile bearing components) and one empty cell. Fuel assemblies in Region 1D shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-4;
| |
| : k. Region 1E is defined as a subregion of the Region I storage rack located in the north tilt pit and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1E may be in a maximum of four-of-four loading pattern with no required empty cells. Fuel assemblies in Region 1E shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-5;
| |
| : l. Interface requirements for the north tilt pit between Region 1D and 1E are as follows. Region 1D and 1E can be distributed in Region I in the north tilt pit in any manner provided that any two-by-two grouping of storage cells and the assemblies in them correspond to the requirements of 4.3.1.1j. or 4.3.1.1k.
| |
| above;
| |
| : m. Non-fissile bearing component restrictions are as follows:
| |
| : 1. Non-fissile material components may be stored in any designated fuel location in Region 1A, 1B, 1C, 1D, or 1E without restriction.
| |
| : 2. The following non-fuel bearing components (NFBC) may be stored face adjacent to fuel in any designated empty cell in Region 1A or 1B.
| |
| (i) The gauge dummy assembly and the lead dummy assembly may be stored face adjacent to fuel in any designated empty cells with no minimum required separation distance.
| |
| (ii) A component comprised primarily of stainless steel that displaces less than 30 square inches of water in any plane within the active fuel region may be stored in any designated empty cell as long as the NFBC is at least ten locations away from any other NFBC that is in a designated empty cell, with the exception of 4.3.1.1m.2.(i) above.
| |
| Palisades Nuclear Plant 4.0-3 Amendment No. 250, 272, XXX
| |
| | |
| Design Features 4.0 4.3 Fuel Storage 4.3.1 Criticality (continued)
| |
| : 3. Control blades may be stored in both fueled and unfueled locations in Regions 1D and 1E, with no limitation on the number.
| |
| 4.3.1.2 The Region I (See Figure B 3.7.16-1) Metamic equipped fuel storage racks are designed and shall be maintained with:
| |
| : a. Fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.95 weight percent;
| |
| : b. Keff < 1.0 if fully flooded with unborated water, which includes allowances for uncertainties as described in Section 9.11 of the FSAR;
| |
| : c. Keff 0.95 if fully flooded with water borated to 850 ppm, which includes allowances for uncertainties as described in Section 9.11 of the FSAR;
| |
| : d. A nominal 10.25 inch center to center distance between fuel assemblies;
| |
| : e. New or irradiated fuel assemblies;
| |
| : f. Two empty rows of storage locations shall exist between the fuel assemblies in a Carborundum equipped rack and the fuel assemblies in an adjacent Metamic equipped rack; and
| |
| : g. A minimum Metamic B10 areal density of 0.02944 g/cm2.
| |
| 4.3.1.3 The Region II fuel storage racks (See Figure B 3.7.16-1) are designed and shall be maintained with:
| |
| : a. Fuel assemblies having maximum nominal planar average U-235 enrichment of 4.60 weight percent;
| |
| : b. Keff < 1.0 if fully flooded with unborated water, which includes allowances for uncertainties as described in Section 9.11 of the FSAR.
| |
| : c. Keff 0.95 if fully flooded with water borated to 850 ppm, which includes allowance for uncertainties as described in Section 9.11 of the FSAR.
| |
| : d. A nominal 9.17 inch center to center distance between fuel assemblies; and Palisades Nuclear Plant 4.0-4 Amendment No. 250, 272, XXX
| |
| | |
| Design Features 4.0 4.3 Fuel Storage 4.3.1 Criticality (continued)
| |
| : e. New or irradiated fuel assemblies which meet the maximum nominal planar average U-235 enrichment, burnup, and decay time requirements of Table 3.7.16-1 4.3.1.4 The new fuel storage racks are designed and shall be maintained with:
| |
| : a. Twenty four unirradiated fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.95 weight percent, and stored in accordance with the pattern shown in Figure 4.3-1, or Thirty six unirradiated fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.05 weight percent, and stored in accordance with the pattern shown in Figure 4.3-1;
| |
| : b. Keff 0.95 when flooded with either full density or low density (optimum moderation) water including allowances for uncertainties as described in Section 9.11 of the FSAR.
| |
| : c. The pitch of the new fuel storage rack lattice being 9.375 inches and every other position in the lattice being permanently occupied by an 8" x 8" structural steel or core plugs, resulting in a nominal 13.26 inch center to center distance between fuel assemblies placed in alternating storage locations.
| |
| 4.3.2 Drainage The spent fuel storage pool cooling system suction and discharge piping is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 644 ft 5 inches.
| |
| 4.3.3 Capacity The spent fuel storage pool and north tilt pit are designed and shall be maintained with a storage capacity limited to no more than 892 fuel assemblies.
| |
| Palisades Nuclear Plant 4.0-5 Amendment No. 250, 272, XXX
| |
| | |
| Design Features 4.0 0 0 0 0 0 0 0 0 0 0 0 0 CENTERLINE - LEGEND PATTERN REPEATS 8 X 8 STEEL BOX BEAM ASSEMBLY STORAGE LOCATION (ENRICHMENT <= 4.95 WT% U-235)
| |
| ASSEMBLY STORAGE LOCATION (ENRICHMENT <= 4.05 WT% U-235)
| |
| Note: If any assemblies containing fuel enrichments greater than 4.05% U-235 are stored in the New Fuel Storage Rack, the center row must remain empty.
| |
| Figure 4.3-1 (page 1 of 1)
| |
| New Fuel Storage Rack Arrangement Palisades Nuclear Plant 4.0-6 Amendment No. XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage to the engineered safeguards rooms, from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident, to as low as practical. The systems include the Containment Spray System, the Safety Injection System, the Shutdown Cooling System, and the containment sump suction piping. This program shall include the following:
| |
| : a. Provisions establishing preventive maintenance and periodic visual inspection requirements, and
| |
| : b. Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.
| |
| : c. The portion of the shutdown cooling system that is outside the containment shall be tested either by use in normal operation or hydrostatically tested at 255 psig.
| |
| : d. Piping from valves CV-3029 and CV-3030 to the discharge of the safety injection pumps and containment spray pumps shall be hydrostatically tested at no less than 100 psig.
| |
| : e. The maximum allowable leakage from the recirculation heat removal systems' components (which include valve stems, flanges and pump seals) shall not exceed 0.2 gallon per minute under the normal hydrostatic head from the SIRW tank.
| |
| 5.5.3 (Deleted)
| |
| Palisades Nuclear Plant 5.0-7 Amendment No. 272, XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the Offsite Dose Calculation Manual (ODCM), (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
| |
| : a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,
| |
| : b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402.
| |
| : c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM,
| |
| : d. Limitation on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each plant to unrestricted areas conforming to 10 CFR 50, Appendix I,
| |
| : e. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be in accordance with the following:
| |
| : 1. For noble gases: a dose rate 500 mrem/yr to the whole body and a dose rate 3000 mrem/yr to the skin, and
| |
| : 2. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rate 1500 mrem/yr to any organ;
| |
| : f. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary conforming to 10 CFR 50, Appendix I, Palisades Nuclear Plant 5.0-8 Amendment No. 213, 272, XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)
| |
| : g. Limitations on the annual and quarterly doses to a member of the public from Iodine-131, Iodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each plant to areas beyond the site boundary conforming to 10 CFR 50, Appendix I,
| |
| : h. Limitations on the annual doses or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR 190.
| |
| The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency.
| |
| 5.5.5 Containment Structural Integrity Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Containment Structural Integrity Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with ASME Boiler and Pressure Vessel Code, Section XI, Subsection IWE and IWL.
| |
| If, as a result of a tendon inspection, corrective retensioning of five percent (8) or more of the total number of dome tendons is necessary to restore their liftoff forces to within the limits, a dome delamination inspection shall be performed within 90 days following such corrective retensioning. The results of this inspection shall be reported to the NRC in accordance with Specification 5.6.7, Containment Structural Integrity Surveillance Report.
| |
| The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Containment Structural Integrity Surveillance Program inspection frequencies.
| |
| 5.5.6 Primary Coolant Pump Flywheel Surveillance Program
| |
| : a. Surveillance of the primary coolant pump flywheels shall consist of a 100% volumetric inspection of the upper flywheels each 10 years.
| |
| : b. The provisions of SR 3.0.2 are not applicable to the Flywheel Testing Program Palisades Nuclear Plant 5.0-9 Amendment No. 271, 272, XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 (Deleted) 5.5.8 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:
| |
| : a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.
| |
| Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.
| |
| : b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
| |
| : 1. Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents.
| |
| This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
| |
| Palisades Nuclear Plant 5.0-10 Amendment No. 272, XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program
| |
| : b. Performance criteria for SG tube integrity. (continued)
| |
| : 2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.
| |
| Leakage is not to exceed 0.3 gpm.
| |
| : 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, PCS Operational LEAKAGE.
| |
| : c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged. The following alternative repair criteria shall be applied as an alternate to the 40% depth based criteria:
| |
| : 1. Tubes found by inservice inspection to contain service induced flaws within 12.5 inches below the bottom of the hot-leg expansion transition or top of the hot-leg tubesheet, whichever is lower, shall be plugged. Flaws located below this elevation may remain in service.
| |
| : 2. Tubes found by inservice inspection to contain service induced flaws within 13.67 inches below the bottom of the cold-leg expansion transition or top of the cold-leg tubesheet, whichever is lower, shall be plugged. Flaws located below this elevation may remain in service.
| |
| : d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from 12.5 inches below the bottom of the hot-leg expansion transition or top of the hot-leg tubesheet, whichever is lower, to 13.67 inches below the bottom of the cold-leg expansion transition or top of the cold-leg tubesheet, whichever is lower, and that may satisfy the applicable tube repair criteria. The tube to tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, d.3, and d.4 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and Palisades Nuclear Plant 5.0-11 Amendment No. XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program
| |
| : d. Provisions for SG tube inspections. (continued) location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
| |
| : 1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.
| |
| : 2. Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.
| |
| : 3. If crack indications are found in any SG tube from 12.5 inches below the bottom of the hot-leg expansion transition or top of the hot-leg tubesheet, whichever is lower, to 13.67 inches below the bottom of the cold-leg expansion transition or top of the cold-leg tubesheet, whichever is lower, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
| |
| : 4. When the alternate repair criteria of TS 5.5.8c.1 are implemented, inspect 100% of the inservice tubes to the hot-leg tubesheet region with the objective of detecting flaws that may satisfy the applicable tube repair criteria of TS 5.5.8c.1 every 24 effective full power months, or one refueling outage, whichever is less.
| |
| : e. Provisions for monitoring operational primary to secondary LEAKAGE.
| |
| Palisades Nuclear Plant 5.0-12 Amendment No. XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Secondary Water Chemistry Program A program shall be established, implemented and maintained for monitoring of secondary water chemistry to inhibit steam generator tube degradation and shall include:
| |
| : a. Identification of a sampling schedule for the critical variables and control points for these variables,
| |
| : b. Identification of the procedures used to measure the values of the critical variables,
| |
| : c. Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage,
| |
| : d. Procedures for the recording and management of data,
| |
| : e. Procedures defining corrective actions for all off-control point chemistry conditions, and
| |
| : f. A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective actions.
| |
| 5.5.10 Ventilation Filter Testing Program A program shall be established to implement the following required testing of Control Room Ventilation (CRV) and Fuel Handling Area Ventilation (FHAV) systems at the frequencies specified in Regulatory Guide 1.52, Revision 2 (RG 1.52), and in accordance with RG 1.52 and ASME N510-1989, at the system flowrates and tolerances specified below*:
| |
| : a. Demonstrate for each of the ventilation systems that an inplace test of the High Efficiency Particulate Air (HEPA) filters shows a penetration and system bypass < 0.05% for the CRV system and < 1.00% for the FHAV system when tested in accordance with RG 1.52 and ASME N510-1989:
| |
| Ventilation System Flowrate (CFM)
| |
| FHAV (single fan operation) 7300 +/- 20%
| |
| FHAV (dual fan operation) 10,000 +/- 20%
| |
| CRV 3,200 +10% -5%
| |
| Palisades Nuclear Plant 5.0-13 Amendment No. XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Ventilation Filter Testing Program (Continued)
| |
| : b. Demonstrate for each of the ventilation systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 0.05% for the CRV system and < 1.00% for the FHAV system when tested in accordance with RG 1.52 and ASME N510-1989.
| |
| Ventilation System Flowrate (CFM)
| |
| FHAV (dual fan operation) 10,000 +/- 20%
| |
| CRV 3200 +10% -5%
| |
| : c. Demonstrate for each of the ventilation systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in RG 1.52 shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and equal to the relative humidity specified as follows:
| |
| Ventilation System Penetration Relative Humidity FHAV 6.00% 95%
| |
| CRV 0.157% 70%
| |
| : d. For each of the ventilation systems, demonstrate the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with RG 1.52 and ASME N510-1989:
| |
| Ventilation System Delta P (In H20) Flowrate (CFM)
| |
| FHAV (dual fan operation) 6.0 10,000 +/- 20%
| |
| CRV 8.0 3200 +10% -5%
| |
| : e. Demonstrate that the heaters for the CRV system dissipates the following specified value +/- 20% when tested in accordance with ASME N510-1989:
| |
| Ventilation System Wattage CRV 15 kW The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Ventilation Filter Testing Program frequencies.
| |
| * Should the 720-hour limitation on charcoal adsorber operation occur during a plant operation requiring the use of the charcoal adsorber - such as refueling - testing may be delayed until the completion of the plant operation or up to 1,500 hours of filter operation; whichever occurs first.
| |
| Palisades Nuclear Plant 5.0-14 Amendment No. XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Fuel Oil Testing Program A fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling requirements, testing requirements, and acceptance criteria, based on the diesel manufacturers specifications and applicable ASTM Standards. The program shall establish the following:
| |
| : a. Acceptability of new fuel oil prior to addition to the Fuel Oil Storage Tank, and acceptability of fuel oil stored in the Fuel Oil Storage Tank, by determining that the fuel oil has the following properties within limits:
| |
| : 1. API gravity or an absolute specific gravity,
| |
| : 2. Kinematic viscosity, and
| |
| : 3. Water and sediment content.
| |
| : b. Other properties of fuel oil stored in the Fuel Oil Storage Tank, specified by the diesel manufacturers or specified for grade 2D fuel oil in ASTM D 975, are within limits.
| |
| The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Fuel Oil Testing Program.
| |
| 5.5.12 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.
| |
| : a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
| |
| : b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
| |
| : 1. A change in the TS incorporated in the license; or
| |
| : 2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
| |
| Palisades Nuclear Plant 5.0-15 Amendment No. 271, 272, XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Technical Specifications (TS) Bases Control Program (continued)
| |
| : c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.
| |
| : d. Proposed changes that meet the criteria of Specification 5.5.12.b. above shall be reviewed and approved by the NRC prior to implementation.
| |
| Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
| |
| 5.5.13 Safety Functions Determination Program (SFDP)
| |
| This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:
| |
| : a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;
| |
| : b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;
| |
| : c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and
| |
| : d. Other appropriate limitations and remedial or compensatory actions.
| |
| A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:
| |
| : a. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or
| |
| : b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or
| |
| : c. A required system redundant to support system(s) for the supported systems (a) and (b) above is also inoperable.
| |
| Palisades Nuclear Plant 5.0-16 Amendment No. 271, 272, XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Safety Functions Determination Program (SFDP) (Continued)
| |
| The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.
| |
| 5.5.14 Containment Leak Rate Testing Program
| |
| : a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with NEI 94-01, Revision 2-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated October 2008, with the following exceptions:
| |
| : 1. Leakage rate testing is not necessary after opening the Emergency Escape Air Lock doors for post-test restoration or post-test adjustment of the air lock door seals. However, a seal contact check shall be performed instead.
| |
| Emergency Escape Airlock door opening, solely for the purpose of strongback removal and performance of the seal contact check, does not necessitate additional pressure testing.
| |
| : 2. Leakage rate testing at Pa is not necessary after adjustment of the Personnel Air Lock door seals. However, a between-the-seals test shall be performed at 10 psig instead.
| |
| : 3. Leakage rate testing frequency for the Containment 4 inch purge exhaust valves, the 8 inch purge exhaust valves, and the 12 inch air room supply valves may be extended up to 60 months based on component performance.
| |
| : b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 54.2 psig. The containment design pressure is 55 psig.
| |
| : c. The maximum allowable containment leakage rate, La, at Pa, shall be 0.1% of containment air weight per day.
| |
| Palisades Nuclear Plant 5.0-17 Amendment No. XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Containment Leak Rate Testing Program (Continued)
| |
| : d. Leakage rate acceptance criteria are:
| |
| : 1. Containment leakage rate acceptance criteria is 1.0 La. During the first plant startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and Type C tests and 0.75 La for Type A tests.
| |
| : 2. Air lock testing acceptance criteria are:
| |
| a) Overall air lock leakage is 1.0 La when tested at Pa and combined with all penetrations and valves subjected to Type B and C tests. However, during the first unit startup following testing performed in accordance with this program, the leakage rate acceptance criteria is < 0.6 La when combined with all penetrations and valves subjected to Type B and C tests.
| |
| b) For each Personnel Air Lock door, leakage is 0.023 La when pressurized to 10 psig.
| |
| c) For each Emergency Escape Air Lock door, a seal contact check , consisting of a verification of continuous contact between the seals and the sealing surfaces, is acceptable.
| |
| di. Containment OPERABILITY is equivalent to "Containment Integrity" for the purposes of the testing requirements.
| |
| dii. The provisions of SR 3.0.3 are applicable to the Containment Leak Rate Testing Program requirements.
| |
| diii. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.
| |
| Palisades Nuclear Plant 5.0-18 Amendment No. XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Process Control Program
| |
| : a. The Process Control Program shall contain the current formula, sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR 20, 10 CFR 71, Federal and State regulations, and other requirements governing the disposal of the radioactive waste.
| |
| : b. Changes to the Process Control Program:
| |
| : 1. Shall be documented and records of reviews performed shall be retained as required by the Quality Program. This documentation shall contain:
| |
| a) Sufficient information to support the change together with the appropriate analyses or evaluation justifying the change(s) and b) A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.
| |
| : 2. Shall become effective after approval by the plant manager.
| |
| Palisades Nuclear Plant 5.0-19 Amendment No. 271, 272, XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Ventilation (CRV) Filtration, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident. The program shall include the following elements:
| |
| : a. The definition of the CRE and the CRE boundary.
| |
| : b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.
| |
| : c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.
| |
| : d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CRV Filtration, operating at the flow rate required by the Ventilation Filter Testing Program, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.
| |
| : e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c.
| |
| The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.
| |
| : f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.
| |
| Palisades Nuclear Plant 5.0-20 Amendment No. XXX
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.17 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
| |
| : a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.
| |
| : b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.
| |
| : c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.
| |
| Palisades Nuclear Plant 5.0-21 Amendment No. XXX
| |
| | |
| Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.
| |
| 5.6.1 (Deleted) 5.6.2 Radiological Environmental Operating Report The Radiological Environmental Operating Report covering the operation of the plant during the previous calendar year shall be submitted before May 15 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.
| |
| 5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering operation of the plant in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant.
| |
| The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual and Process Control Program, and shall be in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.
| |
| 5.6.4 (Deleted) 5.6.5 CORE OPERATING LIMITS REPORT (COLR)
| |
| : a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
| |
| 3.1.1 Shutdown Margin 3.1.6 Regulating Rod Group Position Limits 3.2.1 Linear Heat Rate Limits 3.2.2 Radial Peaking Factor Limits 3.2.4 ASI Limits 3.4.1 DNB Limits Palisades Nuclear Plant 5.0-22 Amendment No. 261, 272, XXX
| |
| | |
| Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (Continued)
| |
| : b. The analytical methods used to determine the core operating limits shall be those approved by the NRC, specifically those described in the latest approved revision of the following documents:
| |
| : 1. EMF-96-029(P)(A) Volumes 1 and 2, Reactor Analysis System for PWRs, Siemens Power Corporation.
| |
| (LCOs 3.1.1, 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 2. ANF-84-73 Appendix B (P)(A), "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," Advanced Nuclear Fuels Corporation. (Bases report not approved) (LCOs 3.1.1, 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 3. XN-NF-82-21(P)(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations,"
| |
| Exxon Nuclear Company.
| |
| (LCOs 3.2.1, 3.2.2, & 3.2.4)
| |
| : 4. EMF-84-093(P)(A), Steam Line Break Methodology for PWRs, Siemens Power Corporation.
| |
| (LCOs 3.1.1, 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 5. XN-75-32(P)(A) Supplements 1 through 4, "Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company. (Bases document not approved)
| |
| (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 6. EMF-2310 (P)(A), Revision 0, Framatome ANP, Inc., May 2001, SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 7. XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company. (LCOs 3.1.6, 3.2.1, & 3.2.2)
| |
| : 8. ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 9. EMF-92-153(P)(A) and Supplement 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation. (LCOs 3.2.1, 3.2.2, & 3.2.4)
| |
| Palisades Nuclear Plant 5.0-23 Amendment No. XXX
| |
| | |
| Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (Continued)
| |
| : 10. XN-NF-621(P)(A), Exxon Nuclear DNB Correlation for PWR Fuel Designs, Exxon Nuclear Company. (LCOs 3.2.1, 3.2.2, & 3.2.4)
| |
| : 11. XN-NF-82-06(P)(A) and Supplements 2, 4, and 5, Qualification of Exxon Nuclear Fuel for Extended Burnup, Exxon Nuclear Company. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 12. ANF-88-133(P)(A) and Supplement 1, Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWD/MTU, Advanced Nuclear Fuels Corporation. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 13. XN-NF-85-92(P)(A), Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results, Exxon Nuclear Company. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 14. EMF-92-116(P)(A), Generic Mechanical Design Criteria for PWR Fuel Designs, Siemens Power Corporation.
| |
| (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 15. EMF-2087(P)(A), SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications, Siemens Power Corporation.
| |
| (LCOs 3.1.6, 3.2.1, & 3.2.2)
| |
| : 16. ANF-87-150 Volume 2, Palisades Modified Reactor Protection System Report: Analysis of Chapter 15 Events, Advanced Nuclear Fuels Corporation. [Approved for use in the Palisades design during the NRC review of license Amendment 118, November 15, 1988] (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.4.1)
| |
| : 17. EMF-1961(P)(A), Revision 0, Siemens Power Corporation, July 2000, Statistical Setpoint/Transient Methodology for Combustion Engineering Type Reactors. (LCOs 3.1.6, 3.2.1, 3.2.2, 3.2.4, &
| |
| 3.4.1)
| |
| : 18. EMF-2328 (P)(A), Revision 0, Framatome ANP, Inc., March 2001, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based.
| |
| (LCOs 3.1.6, 3.2.1, & 3.2.2)
| |
| : 19. BAW-2489P, Revised Fuel Assembly Growth Correlation for Palisades. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
| |
| : 20. EMF-2103(P)(A), Realistic Large Break LOCA Methodology for Pressurized Water Reactors. (LCOs 3.1.6, 3.2.1, & 3.2.2)
| |
| Palisades Nuclear Plant 5.0-24 Amendment No. XXX
| |
| | |
| Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (Continued)
| |
| : 21. BAW-10240(P)-A, Incorporation of M5 Properties in Framatome ANP Approved Methods. (LCOs 3.1.6, 3.2.1, 3.2.2, 3.2.4, & 3.4.1)
| |
| : c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met.
| |
| : d. The COLR, including any mid cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC.
| |
| 5.6.6 Post Accident Monitoring Report When a report is required by LCO 3.3.7, "Post Accident Monitoring Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels to OPERABLE status.
| |
| 5.6.7 Containment Structural Integrity Surveillance Report Reports shall be submitted to the NRC covering Prestressing, Anchorage, and Dome Delamination tests within 90 days after completion of the tests.
| |
| 5.6.8 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.8, Steam Generator (SG) Program. The report shall include:
| |
| : a. The scope of inspections performed on each SG,
| |
| : b. Active degradation mechanisms found,
| |
| : c. Nondestructive examination techniques utilized for each degradation mechanism,
| |
| : d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,
| |
| : e. Number of tubes plugged during the inspection outage for each active degradation mechanism, Palisades Nuclear Plant 5.0-25 Amendment No. XXX
| |
| | |
| Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.8 Steam Generator Tube Inspection Report (continued)
| |
| : f. Total number and percentage of tubes plugged to date,
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| : g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and
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| : h. The effective plugging percentage for all plugging in each SG.
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| i The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.
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| Palisades Nuclear Plant 5.0-26 Amendment No. XXX
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| High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:
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| 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation
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| : a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.
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| : b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP), or equivalent, that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
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| : c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP, or equivalent, while performing their assigned duties, provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.
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| : d. Each individual or group entering such an area shall possess:
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| : 1. A radiation monitoring device that continuously displays radiation dose rates in the area; or
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| : 2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the devices dose alarm setpoint is reached, with an appropriate alarm setpoint, or
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| : 3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or Palisades Nuclear Plant 5.0-27 Amendment No. 196, 272, XXX
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| High Radiation Area 5.7 5.7 High Radiation Area 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation (continued)
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| : 4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area, and who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance, as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.
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| : e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.
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| Palisades Nuclear Plant 5.0-28 Amendment No. 196, 272, XXX
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| High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation
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| : a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:
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| : 1. All such door and gate keys shall be maintained under the administrative control of the shift manager, radiation protection manager, or his or her designee.
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| : 2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.
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| : b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
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| : c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP, or equivalent, while performing radiation surveys in such areas, provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.
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| : d. Each individual or group entering such an area shall possess:
| |
| : 1. A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the devices dose alarm setpoint is reached, with an appropriate alarm setpoint, or
| |
| : 2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, and with the means to communicate with and control every individual in the area, or Palisades Nuclear Plant 5.0-29 Amendment No. 266, 272, XXX
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| High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued)
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| : 3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; and who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance, as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.
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| : 4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the As Low As is Reasonably Achievable principle, a radiation monitoring device that continuously displays radiation dose rates in the area.
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| : e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.
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| : f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.
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| Palisades Nuclear Plant 5.0-30 Amendment No. 196, 272, XXX
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| PALISADES PLANT RENEWED FACILITY OPERATING LICENSE DPR-20 APPENDIX B ENVIRONMENTAL PROTECTION PLAN (NON-RADIOLOGICAL)
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| Amendment No. 176, 272, XXX
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| PALISADES PLANT ENVIRONMENTAL PROTECTION PLAN (NON-RADIOLOGICAL)
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| TABLE OF CONTENTS Section Page 1.0 Objectives of the Environmental Protection Plan ..................................... 1-1 2.0 Environmental Protection Issues .............................................................. 2-1 2.1 Aquatic Issues .......................................................................................... 2-1 2.2 Terrestrial Issues...................................................................................... 2-1 3.0 Consistency Requirements ...................................................................... 3-1 3.1 Plant Design and Operation ..................................................................... 3-1 3.2 Reporting Related to the NPDES Permits and State Certification. ........... 3-2 3.3 Changes Required for Compliance with Other Environmental Regulations .............................................................................................. 3-3 4.0 Environmental Conditions ........................................................................ 4-1 4.1 Unusual or Important Environmental Events ............................................ 4-1 4.2 Environmental Monitoring......................................................................... 4-1 5.0 Administrative Procedures ....................................................................... 5-1 5.1 Review and Audit ..................................................................................... 5-1 5.2 Records Retention ................................................................................... 5-1 5.3 Changes in Environmental Protection Plan .............................................. 5-2 5.4 Plant Reporting Requirements ................................................................. 5-2 Amendment No. 63, 272,XXX May 13, 2022
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| | |
| 1.0 Objectives of the Environmental Protection Plan The Environmental Protection Plan (EPP) is to provide for protection of environmental values during construction and operation of the nuclear facility.
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| The principal objectives of the EPP are as follows:
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| (1) Verify that the plant is operated in an environmentally acceptable manner, as established by the FES and other NRC environmental impact assessments.
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| (2) Coordinate NRC requirements and maintain consistency with other Federal, State and local requirements for environmental protection.
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| (3) Keep NRC informed of the environmental effects of facility construction and operation and of actions taken to control those effects.
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| Environmental concerns identified in the FES which relate to water quality matters are regulated by way of the licensee's NPDES permit.
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| 1-1 Amendment No. 272, XXX May 13, 2022
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| 2.0 Environmental Protection Issues In the final addendum to the FES-OL dated February 1978 the staff considered the environmental impacts associated with the operation of the Palisades Plant.
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| Certain environmental issues were identified which required study or license conditions to resolve environmental concerns and to assure adequate protection of the environment.
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| 2.1 Aquatic Issues Specific aquatic issues raised by the staff in the FES-OL were:
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| The need for aquatic monitoring programs to confirm that thermal mixing occurs as predicted, that chlorine releases are controlled within those discharge concentrations evaluated, and that effects on aquatic biota and water quality due to plant operation are no greater than predicted.
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| Aquatic issues are addressed by the effluent limitations, and monitoring requirements are contained in the effective NPDES permit issued by the State of Michigan, Department of Natural Resources. The NRC will rely on this agency for regulation of matters involving water quality and aquatic biota.
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| 2.2 Terrestrial Issues
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| : 1. Potential impacts on the terrestrial environment associated with drift from the mechanical draft cooling towers. (FES-OL addendum Section 6.3) 2-1 Amendment No. 272, XXX May 13, 2022
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| 3.0 Consistency Requirements 3.1 Plant Design and Operation The licensee may make changes in station design or operation or perform tests or experiments affecting the environment provided such changes, tests or experiments do not involve an unreviewed environmental question, and do not involve a change in the Environmental Protection Plan. Changes in plant design or operation or performance of tests or experiments which do not affect the environment are not subject to the requirements of this EPP. Activities governed by Section 3.3 are not subject to the requirements of this section.
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| Before engaging in additional construction or operational activities which may affect the environment, the licensee shall prepare and record an environmental evaluation of such activity. When the evaluation indicates that such activity involves an unreviewed environmental question, the licensee shall provide a written evaluation of such activities and obtain prior approval from the Director, Office of Nuclear Reactor Regulation. When such activity involves a change in the Environmental Protection Plan, such activity and change to the Environmental Protection Plan may be implemented only in accordance with an appropriate license amendment as set forth in Section 5.3.
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| A proposed change, test or experiment shall be deemed to involve an unreviewed environmental question if it concerns (1) a matter which may result in a significant increase in any adverse environmental impact previously evaluated in the final environmental statement (FES) as modified by staff's testimony to the Atomic Safety and Licensing Board, supplements to the FES, environmental impact appraisals, or in any decisions of the Atomic Safety and Licensing Board; or (2) a significant change in effluents or power level [in accordance with 10 CFR Part 51.5(b)(2)] or (3) a matter not previously reviewed and evaluated in the documents specified in (1) of this Subsection, which may have a significant adverse environmental impact.
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| 3-1 Amendment No. 272, XXX May 13, 2022
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| The licensee shall maintain records of changes in facility design or operation and of tests and experiments carried out pursuant to this Subsection. These records shall include a written evaluation which provide bases for the determination that the change, test, or experiment does not involve an unreviewed environmental question nor constitute a decrease in the effectiveness of this EPP to meet the objectives specified in Section 1.0. The licensee shall include as part of his Annual Environmental Operating Report (per Subsection 5.4.1) brief descriptions, analyses, interpretations, and evaluations of such changes, tests and experiments.
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| 3.2 Reporting Related to the NPDES Permits and State Certifications Violations of the NPDES Permit or the State certification (pursuant to Section 401 of the Clean Mater Act) shall be reported to the NRC by submittal of copies of the reports required by the NPDES Permit or certification.
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| Changes and additions to the NPDES Permit or the State certification shall be reported to the NRC within 30 days following the date the change is approved. If a permit or certification, in part or in its entirety, is appealed and stayed, the NRC shall be notified within 30 days following the date the stay is granted.
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| The NRC shall be notified of changes to the effective NPDES Permit proposed by the licensee by providing NRC with a copy of the proposed change at the same time it is submitted to the permitting agency. The notification of a licensee-initiated change shall include a copy of the requested revision submitted to the permitting agency. The licensee shall provide the NRC a copy of the application for renewal of the NPDES permit at the same time the application is submitted to the permitting agency.
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| 3-2 Amendment No. 272 May 13, 2022
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| 3.3 Changes Required for Compliance with Other Environmental Regulations Changes in plant design or operation and performance of tests or experiments which are required to achieve compliance with other Federal, State, or local environmental regulations are not subject to the requirements of Section 3.1.
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| 3-3 Amendment No. 272, XXX May 13, 2022
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| 4.0 Environmental Conditions 4.1 Unusual or Important Environmental Events Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to plant operation shall be recorded and promptly reported to the NRC within 24 hours by telephone, telegraph, or facsimile transmissions followed by a written report per Subsection 5.4.2. The following are examples: excessive bird impaction events, onsite plant or animal disease outbreaks, mortality or unusual occurrence of any species protected by the Endangered Species Act of 1973, fish kills, increase in nuisance organisms or conditions and unanticipated or emergency discharge of waste water or chemical substances.
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| No routine monitoring programs are required to implement this condition.
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| 4.2 Environmental Monitoring 4.2.1 Meteorological Monitoring A meteorological monitoring program shall be conducted in the vicinity of the plant site for at least two years after conversion to cooling towers to document effects of cooling tower operation on meteorological variables. Data on the following meteorological variables shall be obtained from the station network shown in Figure 4.2.1: precipitation, temperature, humidity, solar radiation, downcoming radiation, visibility, wind direction and wind speed. In addition, studies shall be conducted for at least two years to measure affects of cooling tower drift on vegetation by associated salt deposition, icing or other causes.
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| 4-1 Amendment No. 272, XXX May 13, 2022
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| 5.0 Administrative Procedures 5.1 Review and Audit The licensee shall provide for review and audit of compliance with the Environmental Protection Plan. The audits shall be conducted independently of the individual or groups responsible for performing the specific activity. A description of the organization structure utilized to achieve the independent review and audit function and results of the audit activities shall be maintained and made available for inspection.
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| 5.2 Records Retention Records and logs relative to the environmental aspects of plant operation shall be made and retained in a manner convenient for review and inspection.
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| These records and logs shall be made available to NRC on request.
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| Records of modifications to plant structures, systems and components determined to potentially affect the continued protection of the environment shall be retained for the life of the plant. All other records, data and logs relating to this EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.
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| 5-1 Amendment No. 272, XXX May 13, 2022
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| 5.3 Changes in Environmental Protection Plan Request for change in the Environmental Protection Plan shall include an assessment of the environmental impact of the proposed change and a supporting justification. Implementation of such changes in the EPP shall not commence prior to NRC approval of the proposed changes in the form of a license amendment incorporating the appropriate revision to the Environmental Protection Plan.
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| 5.4 Plant Reporting Requirements 5.4.1 Routine Reports An Annual Environmental Operating Report describing implementation of this EPP for the previous year shall be submitted to the NRC prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following issuance of the operating license. The period of the first report shall begin with the date of issuance of the operating license.
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| The report shall include summaries and analyses of the results of the environmental protection activities required by Subsection 4.2 of this Environmental Protection Plan for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous nonradiological environmental monitoring reports, and an assessment of the observed impacts of the plant operation on the environment. If harmful effects or evidence of trends towards irreversible damage to the environment are observed, the licensee shall provide a detailed analysis of the data and a proposed course of action to alleviate the problem.
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| 5-2 Amendment No. 272, XXX May 13, 2022
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| The Annual Environmental Operating Report shall also include:
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| (a) A list of EPP noncompliances and the corrective actions taken to remedy them.
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| (b) A list of all changes in station design or operation, tests, and experiments made in accordance with Subsection 3.1 which involved a potentially significant unreviewed environmental issue.
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| (c) A list of nonroutine reports submitted in accordance with Subsection 5.4.2.
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| In the event that some results are not available by the report due date, the report shall be submitted noting and explaining the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
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| 5-3 Amendment No. 272, XXX May 13, 2022
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| 5.4.2 Nonroutine Reports A written report shall be submitted to the NRC within 30 days of occurrence of nonroutine event. The report shall (a) describe, analyze, and evaluate the event, including extent and magnitude of the impact and plant operating characteristics, (b) describe the probable cause of the event, (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e) indicate the agencies notified and their preliminary responses.
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| Events reportable under this subsection which also require reports to other Federal, State or local agencies shall be reported in accordance with those reporting requirements in lieu of the requirements of this subsection. The NRC shall be provided a copy of such report at the time it is submitted to the other agency.
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| 5-4 Amendment No. 272, XXX May 13, 2022
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| Enclosure Attachment 3 to HDI PNP 2023-030 Proposed Technical Specifications Bases Changes (for information only) 586 pages follow
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| INSERT Bases 2.1.1 Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)
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| B 2.1.1 Reactor Core SLs BASES BACKGROUND The Palisades Nuclear Plant design criteria (Ref. 1) requires, and these SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and Anticipated Operational Occurrences (AOOs). This is accomplished by having a Departure from Nucleate Boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence level (95/95 DNB criterion) that DNB will not occur and by requiring that fuel centerline temperature stays below the melting temperature.
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| The restrictions of this SL prevent overheating of the fuel and cladding and possible cladding perforation that would result in the release of fission products to the primary coolant. Overheating of the fuel is prevented by maintaining the steady state, peak Linear Heat Rate (LHR) below the level at which fuel centerline melting occurs.
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| Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
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| Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the primary coolant.
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| Operation above the boundary of the nucleate boiling regime beyond onset of DNB could result in excessive cladding temperature because of the resultant sharp reduction in the heat transfer coefficient in the transition and film boiling regimes. If a steam film is allowed to form, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form.
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| This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the primary coolant.
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| Palisades Nuclear Plant B 2.1.1-1 Amendment No. XXX Revised XX/XX/20XX
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| INSERT Bases 2.1.1 Reactor Core SLs B 2.1.1 BASES BACKGROUND The Reactor Protective System (RPS), in combination with the LCOs, is (continued) designed to prevent any anticipated combination of transient conditions for Primary Coolant System (PCS) temperature, pressure, and THERMAL POWER level that would result in a violation of the reactor core SLs.
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| APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation and AOOs. The reactor core SLs are established to preclude violation of the following fuel design criteria:
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| : a. There must be at least a 95% probability at a 95% confidence level (95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and
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| : b. The hot fuel pellet in the core must not experience centerline fuel melting.
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| Palisades uses three DNB correlations; the XNB, ANFP, and HTP detailed in References 3 through 8. The DNB correlations are used solely as analytical tools to ensure that plant conditions will not degrade to the point where DNB could be challenged. The XNB correlation is used for non-High Thermal Performance (HTP) assemblies (assemblies loaded prior to cycle 9), when the non-HTP assemblies could have been limiting. The XNB correlation provides administrative justification for using non-HTP assemblies in Palisades low leakage core design. The ANFP and HTP correlations are used for Palisades High Thermal Performance (HTP) fuel assemblies (assemblies loaded in cycle 9 and later).
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| The HTP correlation can be used when the calculated reactor coolant conditions fall within the correlation's applicable coolant condition ranges. Outside of the applicable range of the HTP correlation, the ANFP correlation can be used. The ANFP correlation may be used over a broader range of coolant conditions than the HTP correlation.
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| The HTP correlation is an extension of the ANFP correlation and incorporates the results of test sections designed to represent HTP fuel design for CE plants.
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| Palisades Nuclear Plant B 2.1.1-2 Amendment No. XXX Revised XX/XX/20XX
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| INSERT Bases 2.1.1 Reactor Core SLs B 2.1.1 BASES APPLICABLE The prediction of DNB is a function of several measured parameters.
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| SAFETY ANALYSES The following trip functions and LCOs, limit these measured parameters (continued) to protect the Palisades reactor from approaching conditions that could lead to DNB:
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| Parameter Protection Core Flow Rate Low PCS Flow Trip Core Power Variable High Power Trip PCS Pressure/Core Power TM/LP Trip Core Inlet Temperature Tinlet LCO Axial Shape Index (ASI) ASI LCO Assembly Power Incore Power Monitoring (LHR and FRT LCOs)
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| The RPS setpoints, LCO 3.3.1, "Reactor Protective System (RPS)
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| Instrumentation," in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for PCS temperature, pressure, and THERMAL POWER level that would result in a Departure from Nucleate Boiling Ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.
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| The SL represents a design requirement for establishing the protection system trip setpoints identified previously. LCO 3.2.1, "Linear Heat Rate (LHR)," and LCO 3.2.2, TOTAL RADIAL PEAKING FACTOR (FRT )," or the assumed initial conditions of the safety analyses (as indicated in the FSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.
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| SAFETY LIMITS SL 2.1.1.1 and SL 2.1.1.2 ensure that the minimum DNBR is not less than the safety analyses limit and that fuel centerline temperature remains below melting.
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| The minimum value of the DNBR during normal operation and design basis AOOs is limited to the following DNB correlation safety limit:
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| Correlation Safety Limit XNB 1.17 ANFP 1.154 HTP 1.141 The fuel centerline melt LHR value assumed in the safety analysis is 21 kw/ft. Operation 21 kw/ft maintains the dynamically adjusted peak LHR and ensures that fuel centerline melt will not occur during normal operating conditions or design AOOs.
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| Palisades Nuclear Plant B 2.1.1-3 Amendment No. XXX Revised XX/XX/20XX
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| INSERT Bases 2.1.1 Reactor Core SLs B 2.1.1 BASES APPLICABILITY SL 2.1.1.1 and SL 2.1.1.2 only apply in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions are available to prevent PCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the plant into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1.
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| In MODES 3, 4, 5, and 6, a reactor core SL is not required, since the reactor is not generating significant THERMAL POWER.
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| SAFETY LIMIT The following violation responses are applicable to the reactor VIOLATIONS core SLs.
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| 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, the requirement to go to MODE 3 places the plant in a MODE in which this SL is not applicable.
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| The allowed Completion Time of 1 hour recognizes the importance of bringing the plant to a MODE where this SL is not applicable and reduces the probability of fuel damage.
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| REFERENCES 1. FSAR, Section 5.1
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| : 2. FSAR, Chapter 14
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| : 3. XN-NF-621(A), Rev 1
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| : 4. XN-NF-709
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| : 5. ANF-1224(A), May 1989
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| : 6. ANF-89-192, January 1990
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| : 7. XN-NF-82-21, Rev 1
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| : 8. EMF-92-153(A) and Supplement 1, March 1994 Palisades Nuclear Plant B 2.1.1-4 Amendment No. XXX Revised XX/XX/20XX
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| INSERT Bases 2.1.2 PCS Pressure SLs B 2.1.2 B 2.0 SAFETY LIMITS (SLs)
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| B 2.1.2 Primary Coolant System (PCS) Pressure SL BASES BACKGROUND The SL on PCS pressure protects the integrity of the PCS against overpressurization. In the event of fuel cladding failure, fission products are released into the primary coolant. The PCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on PCS pressure, continued PCS integrity is ensured. According to Palisades Nuclear Plant design criteria (Ref. 1), the Primary Coolant Pressure Boundary (PCPB) design conditions are not to be exceeded during normal operation and Anticipated Operational Occurrences (AOOs). Also, according to Palisades Nuclear Plant design criteria (Ref. 1), reactivity accidents, including rod ejection, do not result in damage to the PCPB greater than limited local yielding.
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| The design pressure of the PCS is 2500 psia. During normal operation and AOOs, the PCS pressure is kept from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2) and by the piping, valve, and fitting limit of 120% of design pressure (Ref. 6). The initial hydrostatic test was conducted at 125% of design pressure (3125 psia) to verify the integrity of the primary coolant system (Ref. 2). Following inception of plant operation PCS components shall be pressure tested, in accordance with the requirements of ASME Code, Section XI (Ref. 3).
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| Overpressurization of the PCS could result in a breach of the PCPB. If this occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to applicable limits on radioactive releases specified in 10 CFR 100 and 10 CFR 50.67 (Refs. 4 and 7).
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| Palisades Nuclear Plant B 2.1.2-1 Amendment No. XXX Revised XX/XX/20XX
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| INSERT Bases 2.1.2 PCS Pressure SLs B 2.1.2 BASES APPLICABLE The PCS primary safety valves, the Main Steam Safety Valves SAFETY ANALYSES (MSSVs), and the High Pressurizer Pressure trip have settings established to ensure that the PCS pressure SL will not be exceeded.
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| The PCS primary safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code for Nuclear Power Plant Components (Ref. 2). The transient that establishes the required relief capacity, and hence the valve size requirements and lift settings, is a complete loss of external load without a direct reactor trip. During the transient, no control actions are assumed except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the secondary plant safety valve settings, and nominal feedwater supply is maintained.
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| The Reactor Protective System (RPS) trip setpoints (LCO 3.3.1, Reactor Protective System (RPS) Instrumentation"), together with the settings of the MSSVs (LCO 3.7.1, "Main Steam Safety Valves (MSSVs)") and the primary safety valves, provide pressure protection for normal operation and AOOs. In particular, the High Pressurizer Pressure Trip setpoint is specifically set to provide protection against overpressurization (Ref. 5). Conservative values for all system parameters, delay times and core moderator coefficient are assumed.
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| More specifically, for the limiting case, no credit is taken for operation of any other pressure relieving system including the following:
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| : a. Pressurizer Power Operated Relief Valves (PORVs);
| |
| : b. Turbine Bypass Control System;
| |
| : c. Atmospheric Steam Dump Valves;
| |
| : d. Pressurizer Level Control System; or
| |
| : e. Pressurizer Pressure Control System.
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| SAFETY LIMITS The maximum transient pressure allowable in the PCS pressure vessel under the ASME Code, Section III, is 110% of design pressure. The maximum transient pressure allowable in the PCS piping, valves, and fittings under 120% of design pressure (Ref. 6). The most limiting of these two allowances is the 110% of design pressure; therefore, the SL on maximum allowable PCS pressure is established at 2750 psia.
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| Palisades Nuclear Plant B 2.1.2-2 Amendment No. XXX Revised XX/XX/20XX
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| INSERT Bases 2.1.2 PCS Pressure SLs B 2.1.2 BASES APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, 5, and 6 because this SL could be approached or exceeded in these MODES due to overpressurization events. In MODE 6 with the reactor vessel head installed and the reactor vessel head closure bolts less than fully tensioned the potential for an over pressurization event still exists. Although overpressurization of the PCS is impossible once the reactor vessel head is removed, the requirements of this SL apply as long as fuel is in the reactor. Once all the fuel has been removed from the reactor, the requirements of SL 2.1.2 no longer apply.
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| SAFETY LIMIT The following SL violation responses are applicable to the PCS VIOLATIONS pressure SLs.
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| 2.2.2.1 If the PCS pressure SL is violated when the reactor is in MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour.
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| With PCS pressure greater than the value specified in SL 2.1.2 in MODE 1 or 2, the pressure must be reduced to below this value. A pressure greater than the value specified in SL 2.1.2 exceeds 110% of the PCS design pressure and may challenge system integrity.
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| The allowed Completion Time of 1 hour provides the operator time to complete the necessary actions to reduce PCS pressure by terminating the cause of the pressure increase, removing mass or energy from the PCS, or a combination of these actions, and to establish MODE 3 conditions.
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| 2.2.2.2 If the PCS pressure SL is exceeded in MODE 3, 4, 5 or 6, PCS pressure must be restored to within the SL value within 5 minutes.
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| Exceeding the PCS pressure SL in MODE 3, 4, 5 or 6 is potentially more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. This action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.
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| Palisades Nuclear Plant B 2.1.2-3 Amendment No. XXX Revised XX/XX/20XX
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| INSERT Bases 2.1.2 PCS Pressure SLs B 2.1.2 BASES REFERENCES 1. FSAR, Section 5.1
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| : 2. ASME, Boiler and Pressure Vessel Code, Section III, Article NB-7000
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| : 3. ASME, Boiler and Pressure Vessel Code, Section XI, Article IWX-5000
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| : 4. 10 CFR 100
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| : 5. FSAR, Section 4.3
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| : 6. ASA B31.1-1955, Code for Pressure Piping, 1967
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| : 7. 10 CFR 50.67 Palisades Nuclear Plant B 2.1.2-4 Amendment No. XXX Revised XX/XX/20XX
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| LCO Applicability B 3.0 BASES B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY through BASES 9 LCO LCO 3.0.1 and LCO 3.0.2 establish the general requirements applicable to all Specifications and apply at all times unless otherwise stated.
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| LCO 3.0.1 LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e., when the facility is in the specified conditions of the Applicability
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| _ _ _ _ _\ _____, 9 ~.___ _ _____,
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| statement of each Specification).
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| MODES or other plant LCO 3.0.2 LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCO are not met. This Specification establishes that:
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| , unless otherwise
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| : a. Completion of the Required Actions within thespecified specified Completion Times constitutes compliance with a Specification; and LCO 3.0.2 INSERT
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| : b. Completion of the Required Actions is not required when an LCO is met within the specified Completion Time, unless otherwise specified.
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| Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.
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| Palisades Nuclear Plant B 3.0-1 Amendment No. 272 Revised 06/15/2022
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| LCO 3.0.2 INSERT A There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits.
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| If this type of Required Action is not completed within the specified Completion Time, a shutdown may be required to place the plant in a MODE or condition in which the Specification is not applicable. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.)
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| The second type of Required Action specifies the remedial measures that permit continued operation of the plant that is not further restricted by the Completion Time. In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.
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| Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.
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| The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Conditions no longer exist. The individual LCO's ACTIONS specify the Required Actions where this is the case. An example of this is in LCO 3.4.3, "PCS Pressure and Temperature (P/T)
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| Limits."
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| The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally. The reasons for intentionally relying on the ACTIONS include, but are not limited to, performance of Surveillances, preventive maintenance, corrective maintenance, or investigation of operational problems. Entering ACTIONS for these reasons must be done in a manner that does not compromise safety. Intentional entry into ACTIONS should not be made for operational convenience. Additionally, if intentional entry into ACTIONS would result in redundant equipment being inoperable, alternatives should be used instead.
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| Doing so limits the time both subsystems/trains of a safety function are inoperable and limits the time conditions exist which may result in LCO 3.0.3 being entered. Individual Specifications may specify a time limit for performing an SR when equipment is removed from service or bypassed for testing. In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed.
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| When a change in MODE or other specified condition is required to comply with Required Actions, the plant may enter a MODE or other specified condition in which another Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable and the ACTIONS Condition(s) are entered.
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| INSERT NEW LCO 3.0.3.BASES LCO 3.0.3 LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and:
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| : a. An associated Required Action and Completion Time is not met and no other Condition applies; or
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| : b. The condition of the plant is not specifically addressed by the associated ACTIONS. This means that no combination of Conditions stated in the ACTIONS can be made that exactly corresponds to the actual condition of the plant. Sometimes, possible combinations of Conditions are such that entering LCO 3.0.3 is warranted; in such cases, the ACTIONS specifically state a Condition corresponding to such combinations and also that LCO 3.0.3 be entered immediately.
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| This Specification delineates the time limits for placing the plant in a safe MODE or other specified condition when operation cannot be maintained within the limits for safe operation as defined by the LCO and its ACTIONS. It is not intended to be used as an operational convenience that permits routine voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable.
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| Upon entering LCO 3.0.3, 1 hour is allowed to prepare for an orderly shutdown before initiating a change in plant operation. This includes time to permit the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The time limits specified to enter lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the capabilities of the plant, assuming that only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the Primary Coolant System and the potential for a plant upset that could challenge safety systems under conditions to which this Specification applies. The use and interpretation of specified times to complete the actions of LCO 3.0.3 are consistent with the discussion of Section 1.3, Completion Times.
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| A plant shutdown required in accordance with LCO 3.0.3 may be terminated and LCO 3.0.3 exited if any of the following occurs:
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| : a. The LCO is now met.
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| : b. The LCO is no longer applicable.
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| : c. A Condition exists for which the Required Actions have now been performed.
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| : d. ACTIONS exist that do not have expired Completion Times. These Completion Times are applicable from the point in time that the
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| INSERT NEW LCO 3.0.3.BASES Condition is initially entered and not from the time LCO 3.0.3 is exited.
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| The time limits of Specification 3.0.3 allow 37 hours for the plant to be in MODE 5 when a shutdown is required during MODE 1 operation. If the plant is in a lower MODE of operation when a shutdown is required, the time limit for entering the next lower MODE applies. If a lower MODE is entered in less time than allowed, however, the total allowable time to entering MODE 5, or other applicable MODE, is not reduced. For example, if MODE 3 is entered in 2 hours, then the time allowed for entering MODE 4 is the next 29 hours, because the total time for entering MODE 4 is not reduced from the allowable limit of 31 hours. Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to enter a lower MODE of operation in less than the total time allowed.
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| In MODES 1, 2, 3, and 4, LCO 3.0.3 provides actions for Conditions not covered in other Specifications. The requirements of LCO 3.0.3 do not apply in MODES 5 and 6 because the plant is already in the most restrictive Condition required by LCO 3.0.3.
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| The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, 3, or 4) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken. Exceptions to LCO 3.0.3 are provided in instances where requiring a plant shutdown, in accordance with LCO 3.0.3, would not provide appropriate remedial measures for the associated condition of the plant.
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| An example of this is in LCO 3.7.14, "Spent Fuel Pool Water Level."
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| LCO 3.7.14 has an Applicability of "During movement of irradiated fuel assemblies in the spent fuel pool." Therefore, this LCO can be applicable in any or all MODES. If the LCO and the Required Actions of LCO 3.7.14 are not met while in MODE 1, 2, or 3, there is no safety benefit to be gained by placing the plant in a shutdown condition. The Required Action of LCO 3.7.14 of "Suspend movement of irradiated fuel assemblies in spent fuel pool" is the appropriate Required Action to complete in lieu of the actions of LCO 3.0.3. These exceptions are addressed in the individual Specifications.
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| INSERT NEW LCO 3.0.4.BASES LCO 3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It allows placing the plant in a MODE or other specified condition stated in that Applicability (e.g., the Applicability desired to be entered) when plant conditions are such that the requirements of the LCO would not be met, in accordance with either LCO 3.0.4.a, LCO 3.0.4.b, or LCO 3.0.4.c.
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| LCO 3.0.4.a allows entry into a MODE or other specified condition in the Applicability with the LCO not met when the associated ACTIONS to be entered following entry into the MODE or other specified condition in the Applicability will permit continued operation within the MODE or other specified condition for an unlimited period of time. Compliance with ACTIONS that permit continued operation of the plant for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the plant before or after the MODE change. Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made and the Required Actions followed after entry into the Applicability.
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| For example, LCO 3.0.4.a may be used when the Required Action to be entered states that an inoperable instrument channel must be placed in the trip condition within the Completion Time. Transition into a MODE or other specified in condition in the Applicability may be made in accordance with LCO 3.0.4 and the channel is subsequently placed in the tripped condition within the Completion Time, which begins when the Applicability is entered. If the instrument channel cannot be placed in the tripped condition and the subsequent default ACTION ("Required Action and associated Completion Time not met") allows the OPERABLE train to be placed in operation, use of LCO 3.0.4.a is acceptable because the subsequent ACTIONS to be entered following entry into the MODE include ACTIONS (place the OPERABLE train in operation) that permit safe plant operation for an unlimited period of time in the MODE or other specified condition to be entered.
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| LCO 3.0.4.b allows entry into a MODE or other specified condition in the (continued) Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate.
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| The risk assessment may use quantitative, qualitative, or blended approaches, and the risk assessment will be conducted using the plant program, procedures, and criteria in place to implement 10 CFR50.65(a)(4), which requires that risk impacts of maintenance activities to be assessed and managed. The risk assessment, for purposes of LCO 3.0.4.b, must take into account all inoperable Technical Specification equipment regardless of whether the equipment is included in the normal 10 CFR 50.65(a)(4) risk assessment scope.
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| INSERT NEW LCO 3.0.4.BASES The risk assessments will be conducted using the procedures and guidance endorsed by Regulatory Guide 1.182, Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants.
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| Regulatory Guide 1.182 endorses the guidance in Section 11 of NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants. These documents address general guidance for conduct of the risk assessment, quantitative and qualitative guidelines for establishing risk management actions, and example risk management actions. These include actions to plan and conduct other activities in a manner that controls overall risk, increased risk awareness by shift and management personnel, actions to reduce the duration of the condition, actions to minimize the magnitude of risk increases (establishment of backup success paths or compensatory measures), and determination that the proposed MODE change is acceptable. Consideration should also be given to the probability of completing restoration such that the requirements of the LCO would be met prior to the expiration of ACTIONS Completion Times that would require exiting the Applicability.
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| LCO 3.0.4.b may be used with single, or multiple systems and components unavailable. NUMARC 93-01 provides guidance relative to consideration of simultaneous unavailability of multiple systems and components.
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| The results of the risk assessment shall be considered in determining the acceptability of entering the MODE or other specified condition in the Applicability, and any corresponding risk management actions. The LCO 3.0.4.b risk assessments do not have to be documented.
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| The Technical Specifications allow continued operation with equipment unavailable in MODE 1 for the duration of the Completion Time. Since this is allowable, and since in general the risk impact in that particular MODE bounds the risk of transitioning into and through the applicable MODES or other specified conditions in the Applicability of the LCO, the use of the LCO 3.0.4.b allowance should be generally acceptable, as long as the risk is assessed and managed as stated above. However, there is a small subset of systems and components that have been determined to be more important to risk and use of the LCO 3.0.4.b allowance is prohibited. The LCOs governing these system and components contain Notes prohibiting the use of LCO 3.0.4.b by stating that LCO 3.0.4.b is not applicable.
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| LCO 3.0.4.c allows entry into a MODE or other specified condition in the Applicability with the LCO not met based on a Note in the Specification which states LCO 3.0.4.c is applicable. These specific allowances permit entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time and a risk assessment has not been performed. This allowance may apply to all the ACTIONS or to a
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| INSERT NEW LCO 3.0.4.BASES specific Required Action of a Specification. The risk assessments performed to justify the use of LCO 3.0.4.b usually only consider systems and components. For this reason, LCO 3.0.4.c is typically applied to Specifications which describe values and parameters (e.g., primary coolant system specific activity), and may be applied to other Specifications based on NRC plant-specific approval.
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| The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.
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| The provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any plant shutdown. In this context, a plant shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, MODE 3 to MODE 4, and MODE 4 to MODE 5.
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| Upon entry into a MODE or other specified condition in the Applicability with the LCO not met, LCO 3.0.1 and 3.0.2 require entry into the applicable Conditions and Required Actions until the Condition is resolved, until the LCO is met, or until the plant is not within the Applicability of the Technical Specification.
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| Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, utilizing LCO 3.0.4 is not a violation of SR 3.0.1 or SR 3.0.4 for any Surveillances that have not been performed on inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO.
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| INSERT NEW LCO 3.0.5.BASES LCO 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with the applicable Required Action(s)) to allow the performance of required testing to demonstrate:
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| : a. The OPERABILITY of the equipment being returned to service; or
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| : b. The OPERABILITY of other equipment.
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| The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to perform the required testing to demonstrate OPERABILITY. This Specification does not provide time to perform any other preventive or corrective maintenance. LCO 3.0.5 should not be used in lieu of other practicable alternatives that comply with Required Actions and that do not require changing the MODE or other specified conditions in the Applicability in order to demonstrate equipment is OPERABLE. LCO 3.0.5 is not intended to be used repeatedly.
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| An example of demonstrating equipment is OPERABLE with the Required Actions not met is opening a manual valve that was closed to comply with Required Actions to isolate a flowpath with excessive Primary Coolant System (PCS) Pressure Isolation Valve (PIV) leakage in order to perform testing to demonstrate that PCS PIV leakage is now within limit.
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| Examples of demonstrating equipment OPERABILITY include instances in which it is necessary to take an inoperable channel or trip system out of a tripped condition that was directed by a Required Action, if there is no Required Action Note for this purpose. An example of verifying OPERABILITY of equipment removed from service is taking a tripped channel out of the tripped condition to permit the logic to function and indicate the appropriate response during performance of required testing on the inoperable channel. Examples of demonstrating the OPERABILITY of other equipment are taking an inoperable channel or trip system out of the tripped condition;
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| : 1) to prevent the trip function from occurring during the performance of required testing on another channel in the other trip system, or
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| : 2) to permit the logic to function and indicate the appropriate response during the performance of required testing on another channel in the same trip system.
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| The administrative controls in LCO 3.0.5 apply in all cases to systems or components in Chapter 3 of the Technical Specifications, as long as the testing could not be conducted while complying with the Required Actions. This includes the realignment or repositioning of redundant or alternate equipment or trains previously manipulated to comply with
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| INSERT NEW LCO 3.0.5.BASES ACTIONS, as well as equipment removed from service or declared inoperable to comply with ACTIONS.
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| INSERT NEW LCO 3.0.6.BASES LCO 3.0.6 LCO 3.0.6 establishes an exception to LCO 3.0.2 for supported systems that have a support system LCO specified in the Technical Specifications (TS). This exception is provided because LCO 3.0.2 would require that the Conditions and Required Actions of the associated inoperable supported system LCO be entered solely due to the inoperability of the support system. This exception is justified because the actions that are required to ensure the plant is maintained in a safe condition are specified in the support system LCO's Required Actions. These Required Actions may include entering the supported system's Conditions and Required Actions or may specify other Required Actions.
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| When a support system is inoperable and there is an LCO specified for it in the TS, the supported system(s) are required to be declared inoperable if determined to be inoperable as a result of the support system inoperability. However, it is not necessary to enter into the supported systems' Conditions and Required Actions unless directed to do so by the support system's Required Actions. The potential confusion and inconsistency of requirements related to the entry into multiple support and supported systems' LCOs Conditions and Required Actions are eliminated by providing all the actions that are necessary to ensure the plant is maintained in a safe condition in the support system's Required Actions.
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| However, there are instances where a support system's Required Action may either direct a supported system to be declared inoperable or direct entry into Conditions and Required Actions for the supported system.
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| This may occur immediately or after some specified delay to perform some other Required Action. Regardless of whether it is immediate or after some delay, when a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.
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| Specification 5.5.13, "Safety Functions Determination Program (SFDP),"
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| ensures loss of safety function is detected and appropriate actions are taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other limitations, remedial actions, or compensatory actions may be identified as a result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions. The SFDP implements the requirements of LCO 3.0.6.
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| Cross train checks to identify a loss of safety function for those support systems that support multiple and redundant safety systems are required.
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| The cross train check verifies that the supported systems of the redundant OPERABLE support system are OPERABLE, thereby ensuring safety function is retained.
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| INSERT NEW LCO 3.0.6.BASES If this evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
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| INSERT NEW LCO 3.0.7.BASES LCO 3.0.7 Special tests and operations are required at various times over the plants life to demonstrate performance characteristics, to perform maintenance activities, and to perform special evaluations. Because TS normally preclude these tests and operations, Special Test Exceptions (STEs) allow specified requirements to be changed or suspended under controlled conditions. STEs are included in applicable sections of the Specifications. Unless otherwise specified, all other TS requirements remain unchanged and in effect as applicable. This will ensure that all appropriate requirements of the MODE or other specified condition not directly associated with or required to be changed or suspended to perform the special test or operation will remain in effect.
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| The Applicability of an STE LCO represents a condition not necessarily in compliance with the normal requirements of the TS. Compliance with STE LCO is optional.
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| A special test may be performed under either the provisions of the appropriate STE LCO or the other applicable TS requirements. If it is desired to perform the special test under the provisions of the STE LCO, the requirements of the STE LCO shall be followed. This includes the SRs specified in the STE LCO.
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| Some of the STE LCO require that one or more of the LCO for normal operation be met (i.e., meeting the STE LCO requires meeting the specified normal LCO). The Applicability, ACTIONS, and SRs of the specified normal LCO, however, are not required to be met in order to meet the STE LCO when it is in effect. This means that, upon failure to meet a specified normal LCO, the associated ACTIONS of the STE LCO apply, in lieu of the ACTIONS of the normal LCO. Exceptions to the above do exist. There are instances when the Applicability of the specified normal LCO must be met, where its ACTIONS must be taken, where certain of its Surveillances must be performed, or where all of these requirements must be met concurrently with the requirements of the STE LCO.
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| Unless the SRs of the specified normal LCO are suspended or changed by the special test, those SRs that are necessary to meet the specified normal LCO must be met prior to performing the special test. During the conduct of the special test, those Surveillances need not be performed unless specified by the ACTIONS or SRs of the STE LCO.
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| ACTIONS for STE LCO provide appropriate remedial measures upon failure to meet the STE LCO. Upon failure to meet these ACTIONS, suspend the performance of the special test and enter the ACTIONS for all LCOs that are then not met. Entry into LCO 3.0.3 may possibly be required, but this determination should not be made by considering only the failure to meet the ACTIONS of the STE LCO.
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| INSERT NEW LCO 3.0.8.BASES LCO 3.0.8 LCO 3.0.8 establishes conditions under which systems are considered to remain (continued) capable of performing their intended safety function when associated snubbers are not capable of providing their associated support function(s). This LCO states that the supported system is not considered to be inoperable solely due to one or more snubbers not capable of performing their associated support function(s). This is appropriate because a limited length of time is allowed for maintenance, testing, or repair of one or more snubbers not capable of performing their associated support function(s) and appropriate compensatory measures are specified in the snubber requirements, which are located outside of the Technical Specifications (TS). The snubber requirements do not meet the criteria in 10 CFR 50.36(c)(2)(ii), and, as such, are appropriate for administrative control.
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| If the allowed time expires and the snubber(s) are unable to perform their associated support function(s), the affected supported systems LCO(s) must be declared not met and the Conditions and Required Actions entered in accordance with LCO 3.0.2.
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| Every time that the provisions of LCO 3.0.8 are applied it is required to confirm that at least one train (or subsystem) of systems supported by the inoperable snubbers would remain capable of performing their required safety or support functions for postulated design loads other than seismic loads. LCO 3.0.8 does not apply to non-seismic snubbers (i.e., seismic vs non-seismic), implementation of this restriction, and the associated plant configuration shall be available on a recoverable basis for NRC staff inspection. SEP-SNB-PLP-001, Snubber Examination and Testing Program, may be used as a reference for application of LCO 3.0.8 to site specific snubbers.
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| LCO 3.0.8.a applies when one or more snubbers are not capable of providing their associated support function(s) to a single train or subsystem of a multiple train or subsystem supported system or to a single train or subsystem supported system. LCO 3.0.8.a allows 72 hours to restore the snubber(s) before declaring the supported system inoperable. The 72 hour Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function and due to the availability of the redundant train of the supported system.
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| When applying LCO 3.0.8.a at least one AFW train (including a minimum set of supporting equipment required for its successful operation), or some alternative means of core cooling, not associated with the inoperable snubber(s), must be available. Implementation of this restriction and the associated plant configuration shall be available on a recoverable basis for NRC staff inspection.
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| INSERT NEW LCO 3.0.8.BASES LCO 3.0.8.b applies when one or more snubbers are not capable of providing their associated support function(s) to more than one train or subsystem of a multiple train or subsystem supported system. LCO 3.0.8.b allows 12 hours to restore the snubber(s) before declaring the supported system inoperable. The 12 hour Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function.
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| When applying LCO 3.0.8.b at least one AFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the inoperable snubber(s), or some alternative means of core cooling (e.g., F&B, fire water system or aggressive secondary cooldown using the steam generators) must be available.
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| Implementation of this restriction and the associated plant configuration shall be available on a recoverable basis for NRC staff inspection.
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| LCO 3.0.8 requires that risk be assessed and managed. Industry and NRC guidance on the implementation of 10 CFR 50.65(a)(4) (the Maintenance Rule) does not address seismic risk. However, use of LCO 3.0.8 should be considered with respect to other plant maintenance activities, and integrated into the existing Maintenance Rule process to the extent possible so that maintenance on any unaffected train or subsystem is properly controlled, and emergent issues are properly addressed. The risk assessment need not be quantified, but may be a qualitative awareness of the vulnerability of systems and components when one or more snubbers are not able to perform their associated support function.
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| INSERT NEW LCO 3.0.9.BASES LCO 3.0.9 LCO 3.0.9 establishes conditions under which systems described in the Technical Specifications are considered to remain OPERABLE when required barriers are not capable of providing their related support function(s).
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| Barriers are doors, walls, floor plugs, curbs, hatches, installed structures or components, or other devices, not explicitly described in Technical Specifications, that support the performance of the safety function of systems described in the Technical Specifications. This LCO states that the supported system is not considered to be inoperable solely due to required barriers not capable of performing their related support function(s) under the described conditions. LCO 3.0.9 allows 30 days before declaring the supported system(s) inoperable and the LCO(s) associated with the supported system(s) not met. A maximum time is placed on each use of this allowance to ensure that as required barriers are found or are otherwise made unavailable, they are restored. However, the allowable duration may be less than the specified maximum time based on the risk assessment.
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| If the allowed time expires and the barriers are unable to perform their related support function(s), the supported systems LCO(s) must be declared not met and the Conditions and Required Actions entered in accordance with LCO 3.0.2.
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| This provision does not apply to barriers which support ventilation systems or to fire barriers. The Technical Specifications for ventilation systems provide specific Conditions for inoperable barriers. Fire barriers are addressed by other regulatory requirements and associated plant programs. This provision does not apply to barriers which are not required to support system OPERABILITY (see NRC Regulatory Issue Summary 2001-09, "Control of Hazard Barriers," dated April 2, 2001).
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| The provisions of LCO 3.0.9 are justified because of the low risk associated with required barriers not being capable of performing their related support function. This provision is based on consideration of the following initiating event categories:
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| x Loss of coolant accidents; x High energy line breaks; x Feedwater line breaks; x Internal flooding; x External flooding; x Turbine missile ejection; and x Tornado or high wind.
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| The risk impact of the barriers which cannot perform their related support function(s) must be addressed pursuant to the risk assessment and management provision of the Maintenance Rule, 10 CFR 50.65 (a)(4),
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| and the associated implementation guidance, Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants." Regulatory Guide 1.182 endorses the guidance in Section 11 of NUMARC 93-01, "Industry Guideline for Monitoring the
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| | |
| INSERT NEW LCO 3.0.9.BASES Effectiveness of Maintenance at Nuclear Power Plants." This guidance provides for the consideration of dynamic plant configuration issues, emergent conditions, and other aspects pertinent to plant operation with the barriers unable to perform their related support function(s). These considerations may result in risk management and other compensatory actions being required during the period that barriers are unable to perform their related support function(s).
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| LCO 3.0.9 may be applied to one or more trains or subsystems of a system supported by barriers that cannot provide their related support function(s), provided that risk is assessed and managed (including consideration of the effects on Large Early Release and from external events). If applied concurrently to more than one train or subsystem of a multiple train or subsystem supported system, the barriers supporting each of these trains or subsystems must provide their related support function(s) for different categories of initiating events. For example, LCO 3.0.9 may be applied for up to 30 days for more than one train of a multiple train supported system if the affected barrier for one train protects against internal flooding and the affected barrier for the other train protects against tornado missiles. In this example, the affected barrier may be the same physical barrier but serve different protection functions for each train.
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| If during the time that LCO 3.0.9 is being used, the required OPERABLE train or subsystem becomes inoperable, it must be restored to OPERABLE status within 24 hours. Otherwise, the train(s) or subsystem(s) supported by barriers that cannot perform their related support function(s) must be declared inoperable and the associated LCOs declared not met. This 24 hour period provides time to respond to emergent conditions that would otherwise likely lead to entry into LCO 3.0.3 and a rapid plant shutdown, which is not justified given the low probability of an initiating event which would require the barrier(s) not capable of performing their related support function(s). During this 24 hour period, the plant risk associated with the existing conditions is assessed and managed in accordance with 10 CFR 50.65(a)(4).
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| | |
| LCO Applicability B 3.0 BASES B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY BASES SRs SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated. SR MODES or other 3.0.2 and SR 3.0.3 apply in Chapter 5 only when invoked by a Chapter 5 specification.
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| the OPERABILITY of systems and components, and SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes a failure to meet an LCO. Surveillances may be performed by means of any series of sequential, overlapping, or total steps provided the entire Surveillance is performed within the specified Frequency.
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| INSERT SR 3.0.1.A The LCO is assumed to be met when the SRs have been met. Nothing in this Specification, however, is to be construed as implying that the LCO is met when the Surveillance(s) are known to be not met between MODE or other Surveillance performances. plant
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| , unless otherwise specified Surveillances do not have to be performed when the facility is in a specified condition for which the requirements of the associated LCO are not applicable. ~ , - - - - - - - - - - ,
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| INSERT SR 3.0.1.B INSERT SR 3.0.1.C Surveillances do not have to be performed on variables that are outside
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| - ~ ~ -
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| their specified -
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| limits because- -
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| the ACTIONS -
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| define -measures the remedial -
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| that apply. Surveillances have to be met and performed in accordance with SR 3.0.2, to restore variables within their specified limits.
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| SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances.
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| plant operating SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers facility conditions that may not be suitable for conducting the Surveillance (e.g., ongoing Surveillance or maintenance activities).
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| transient conditions or other and any Required Action with a Completion Time that requires the periodic performance of Palisades Nuclear Plant B 3.0-2 the Required Action on a "Once Amendment per . . ."
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| No. 272 interval Revised 06/15/2022
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| | |
| INSERT SR 3.0.1.A Additionally, the definitions related to instrument testing (e.g., CHANNEL CALIBRATION) specify that these tests are performed by means of any series of sequential, overlapping, or total steps.
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| Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when:
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| : a. The systems or components are known to be inoperable, although still meeting the SRs; or
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| : b. The requirements of the Surveillance(s) are known to be not met between required Surveillance performances.
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| INSERT SR 3.0.1.B The SRs associated with a Special Test Exception (STE) are only applicable when the STE is used as an allowable exception to the requirements of a Specification.
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| Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this case, the unplanned event may be credited as fulfilling the performance of the SR. This allowance includes those SRs whose performance is normally precluded in a given MODE or other specified condition.
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| INSERT SR 3.0.1.C Surveillances, including Surveillances invoked by Required Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply. Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status.
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| Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary plant parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post
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| | |
| maintenance tests can be completed.
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| An example of this process is:
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| : a. High Pressure Safety Injection (HPSI) maintenance during shutdown that requires system functional tests at a specified pressure.
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| Provided other appropriate testing is satisfactorily completed, startup can proceed with HPSI considered OPERABLE. This allows operation to reach the specified pressure to complete the necessary post maintenance testing.
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| | |
| LCO Applicability B 3.0 BASES SR 3.0.2 When a Section 5.5, Programs and Manuals, specification states that (continued) the provisions of SR 3.0.2 are applicable, a 25% extension of the testing interval, whether stated in the specification or incorporated by reference, is permitted.
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| The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the INSERT SR 3.0.2.A SRs. (other than those consistent with
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| ._____~
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| refueling intervals)
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| The provisions of SR 3.0.2 are not intended to be used repeatedly to extend Surveillance intervals or periodic Completion Time intervals beyond those specified.
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| SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring an affected variable affectedoutside the specified equipment limitsorwhen a Surveillance has not been performed inoperable within the specified Frequency. A delay period of up to 24 hours or up to the limit of the specified Frequency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.
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| When a Section 5.5, Programs and Manuals, specification states that the provisions of SR 3.0.3 are applicable, it permits the flexibility to defer declaring the testing requirement not met in accordance with SR 3.0.3 when the testing has not been completed within the testing interval (including the allowance of SR 3.0.2 if invoked by the Section 5.5 specification).
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| This delay period provides an adequate time to perform Surveillances that have been missed. This delay period permits the performance of a Surveillance before complying with Required Actions or other remedial measures that might preclude performance of the Surveillance.
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| Palisades Nuclear Plant B 3.0-3 Amendment No. 272 Revised 06/15/2022
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| | |
| INSERT SR 3.0.2.A The exceptions to SR 3.0.2 are those Surveillances for which the 25%
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| extension of the interval specified in the Frequency does not apply. These exceptions are stated in the individual Specifications. Examples of where SR 3.0.2 does not apply are the Containment Leak Rate Testing Program required by 10 CFR 50, Appendix J, and the American Society of Mechanical Engineers (ASME) Code inservice testing required by 10 CFR 50.55a. These programs establish testing requirements and frequencies in accordance with the requirements of regulations. The TS cannot, in and of themselves, extend a test interval specified in the regulations directly or by reference.
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| As stated in SR 3.0.2, the 25% extension allowed by SR 3.0.2 may be applied to Required Actions whose Completion Time is stated as once per . . . however, the 25% extension does not apply to the initial performance of a Required Action with a periodic Completion Time that requires performance on a "once per . . ." basis. The 25% extension applies to each performance of the Required Action after the initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25% extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.
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| | |
| LCO Applicability B 3.0 BASES or
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| , operating situations, plant requirements of regulations SR(e.g.,3.0.3 prior to The basis for this delay period includes consideration of facility entering MODE (continued) 1 after conditions, adequate planning, availability of personnel, the time required each fuel loading, or in to perform the Surveillance, the safety significance of the delay in accordance with 10 CFR completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the unit 50, Appendix J, as modified by approved verification of conformance with the requirements. When a Surveillance exemptions, etc.) with a Frequency based not on time intervals, but upon specified facility conditions, is discovered to not have been performed when specified, SR 3.0.3 allows for the full delay period of up to the specified Frequency to perform the Surveillance. However, since there is not a time interval specified, the missed Surveillance should be performed at the first SR 3.0.3 provides a time reasonable opportunity.
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| limit for, and allowances equipment is OPERABLE or that for the performance of, SR 3.0.3 is only applicable if there is a reasonable expectation the Surveillances that associated variables are within limits, and it is expected that the become applicable as a Surveillance will be met when performed. Many factors should be INSERT consequence of MODE considered, such as the period of time since the Surveillance was last SR 3.0.3.A changes imposed by performed, or whether the Surveillance, or a portion thereof, has ever Required Actions. been performed, and any other indications, tests, or activities that might support the expectation that the Surveillance will be met when performed.
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| Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence. Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used repeatedly to extend Surveillance intervals.
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| plant plant While up to 24 hours or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity. The determination of the first reasonable opportunity should include consideration of the impact on facility risk (from delaying the Surveillance unit as well as any facility configuration changes required to perform the Surveillance) and impact on any analysis assumptions, in addition to facility conditions, planning, availability of personnel, and the time required to perform the Surveillance. All missed Surveillances will be placed in the licensee's Corrective Action Program. INSERT SR 3.0.3.B equipment is considered inoperable or the If a Surveillance is not completed within the allowed delay period, then
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| .__________.I\
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| the variable is considered outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin
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| .--------J immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon the failure of the Surveillance.
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| equipment is inoperable, or the Palisades Nuclear Plant B 3.0-4 Amendment No. 272 Revised 06/15/2022
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| | |
| INSERT SR 3.0.3.A An example of the use of SR 3.0.3 would be a relay contact that was not tested as required in accordance with a particular SR, but previous successful performances of the SR included the relay contact; the adjacent, physically connected relay contacts were tested during the SR performance; the subject relay contact has been tested by another SR; or historical operation of the subject relay contact has been successful. It is not sufficient to infer the behavior of the associated equipment from the performance of similar equipment. The rigor of determining whether there is a reasonable expectation a Surveillance will be met when performed should increase based on the length of time since the last performance of the Surveillance. If the Surveillance has been performed recently, a review of the Surveillance history and equipment performance may be sufficient to support a reasonable expectation that the Surveillance will be met when performed. For Surveillances that have not been performed for a long period or that have never been performed, a rigorous evaluation based on objective evidence should provide a high degree of confidence that the equipment is OPERABLE. The evaluation should be documented in sufficient detail to allow a knowledgeable individual to understand the basis for the determination.
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| INSERT SR 3.0.3.B This risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance, NRC Regulatory Guide 1.182, Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants. This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management action up to and including plant shutdown. The missed Surveillance should be treated as an emergent condition as discussed in the Regulatory Guide. The risk evaluation may use quantitative, qualitative, or blended methods. The degree of depth and rigor of the evaluation should be commensurate with the importance of the component. Missed Surveillances for important components should be analyzed quantitatively. If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the safest course of action.
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| | |
| system and component OPERABILITY requirements and LCO Applicability B 3.0 BASES SR 3.0.3 Completion of the Surveillance within the delay period allowed by this (continued) Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1.
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| systems and components MODE or other ensure safe operation of SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable the SRsplant must be met before entry into a specified Condition in the Applicability.
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| MODES or other This Specification ensures that variable limits are met before entry into specified conditions in the Applicability for which these variables ensure
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| ..------ safe- handling
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| -- and
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| , storage
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| ~ of - spent
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| - fuel.
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| INSERT SR 3.0.4.A The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring variables within specified limits before entering an associated specified condition in the Applicability.
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| system, subsystem, inoperable or division, component, However, in certain circumstances, failing to meet an SR will not result in device, or SR 3.0.4 restricting a MODE change or other specified condition
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| . . . . . _ _ _ change._ When
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| ~ a variable is outside its specified limits, the associated SR(s) are not required to be performed, per SR 3.0.1, which states that surveillances do not have to be performed on variables that are outside their specified limits. When a variable is outside its specified limit, SR 3.0.4 does not apply to the associated SR(s) since the requirement for the SR(s) to be performed is removed. Therefore, failing to perform the Surveillance(s) within the specified Frequency does not result in an SR 3.0.4 restriction to changing specified conditions of the Applicability.
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| However, since the LCO is not met in this instance, LCO 3.0.4 will govern any restrictions that may (or may not) apply to specified condition changes.
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| equipment is inoperable SR 3.0.4 does not restrict changing specified conditions of the Applicability when a Surveillance has not been performed within the specified Frequency, providing the requirement to declare the LCO not met has been delayed in accordance with SR 3.0.3.
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| inoperable equipment MODES or other MODES or other MODE or other INSERT SR 3.0.4.B Palisades Nuclear Plant B 3.0-5 Amendment No. 272 Revised 06/15/2022
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| | |
| INSERT SR 3.0.4.A The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability. A provision is included to allow entry into a MODE or other specified condition in the Applicability when an LCO is not met due to Surveillance not being met in accordance with LCO 3.0.4.
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| INSERT SR 3.0.4.B The provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any plant shutdown. In this context, a plant shutdown is defined as a change in MODE or specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, MODE 3 to MODE 4, and MODE 4 to MODE 5.
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| | |
| LCO Applicability MODE or other B 3.0 BASES SR 3.0.4 The precise requirements for performance of SRs are specified such that (continued) exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition(s) specified in a Surveillance procedure require entry into the specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance.
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| A Surveillance that could not be performed until after entering the LCOs Applicability, would have its Frequency specified such that it is not "due" until the specific conditions needed are met. Alternately, the Surveillance may be stated in the form of a Note as not required (to be met or performed) until a particular event, condition, or time has been reached.
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| Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency.
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| Palisades Nuclear Plant B 3.0-6 Amendment No. 272 Revised 06/15/2022
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| | |
| INSERT Bases 3.1.1 SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)
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| BASES BACKGROUND The reactivity control systems must be redundant and capable of maintaining the reactor core subcritical when shut down under cold conditions, in accordance with the Palisades Nuclear Plant design criteria (Ref. 1). Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown events and Anticipated Operational Occurrences (AOOs). As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion of all full-length control rods, assuming that the single control rod of highest reactivity worth remains fully withdrawn. Once all full-length control rods have been verified to be at or below the lower electrical limit, the penalty for the control rod of highest reactivity worth fully withdrawn no longer must be applied.
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| The Palisades Nuclear Plant design criteria requires that two separate reactivity control systems be provided, and that one of these systems be capable of maintaining the core subcritical under cold conditions.
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| These requirements are provided by the use of movable control rods and soluble boric acid in the Primary Coolant System (PCS). The Rod Control System provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel design limits, assuming that the control rod of highest reactivity worth remains fully withdrawn.
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| The soluble boron system can compensate for fuel depletion during operation and all xenon burnout reactivity changes, and maintain the reactor subcritical under cold conditions.
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| During MODES 1 and 2, SDM control is ensured by operating with the shutdown rods within the limits of LCO 3.1.5, Shutdown and Part-Length Rod Group Insertion Limits, and the regulating rods within the limits of LCO 3.1.6, Regulating Rod Group Position Limits. When the plant is in MODES 3, 4, 5, and 6 the SDM requirements are met by means of adjustments to the PCS boron concentration.
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| Palisades Nuclear Plant B 3.1.1-1 Revised 01/29/2020
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| | |
| INSERT Bases 3.1.1 SDM B 3.1.1 BASES APPLICABLE The minimum required SDM is assumed as an initial condition in safety SAFETY ANALYSES analysis. The safety analysis (Ref. 2) establishes an SDM that ensures specified acceptable fuel design limits are not exceeded for normal operation and AOOs, with the assumption that the control rod of highest reactivity worth is fully withdrawn following a reactor trip. For MODE 5, the primary safety analysis that relies on the SDM limits is the boron dilution analysis.
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| The acceptance criteria for the SDM requirements are that specified acceptable fuel design limits are maintained. This is done by ensuring that:
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| : a. The reactor can be made subcritical from all operating conditions, transients, and Design Basis Events;
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| : b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits (Departure from Nucleate Boiling Ratio (DNBR), fuel centerline temperature limit AOOs, and 280 cal/gm energy deposition for the control rod ejection accident); and
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| : c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
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| The most limiting accident for the SDM requirements are based on a Main Steam Line Break (MSLB), as described in the accident analysis (Ref. 2). The increased steam flow resulting from a pipe break in the main steam system causes an increased energy removal from the affected Steam Generator (SG), and consequently the PCS. This results in a reduction of the primary coolant temperature. The resultant coolant shrinkage causes a reduction in pressure. In the presence of a negative moderator temperature coefficient, this cooldown causes an increase in core reactivity. The most limiting MSLB with respect to potential fuel damage is a guillotine break of a main steam line initiated at the end of core life. The positive reactivity addition from the moderator temperature decrease will terminate when the affected SG boils dry, thus terminating PCS heat removal and cooldown. Following the MSLB, a post trip return to power may occur; however, THERMAL POWER does not violate the Safety Limit (SL) requirement of SL 2.1.1.
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| In addition to the limiting MSLB transient, the SDM requirement for MODES 3 and 4 must also protect against an inadvertent boron dilution; (Ref. 3) and an uncontrolled control rod bank withdrawal from subcritical conditions (Ref. 5).
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| Palisades Nuclear Plant B 3.1.1-2 Revised 01/29/2020
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| | |
| INSERT Bases 3.1.1 SDM B 3.1.1 BASES APPLICABLE Each of these events is discussed below.
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| SAFETY ANALYSES (continued) In the boron dilution analysis, the required SDM defines the reactivity difference between an initial subcritical boron concentration and the corresponding critical boron concentration. These values, in conjunction with the configuration of the PCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life when critical boron concentrations are highest.
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| The withdrawal of a control rod bank from subcritical conditions adds reactivity to the reactor core, causing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures and pressure. The withdrawal of control rod banks also produce a time dependent redistribution of core power.
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| Depending on the system initial conditions and reactivity insertion rate, the uncontrolled control rod banks withdrawal transient is terminated by either a high power trip or a high pressurizer pressure trip. In all cases, power level, PCS pressure, linear heat rate, and the DNBR do not exceed allowable limits.
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| SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| LCO The MSLB (Ref. 2) and the boron dilution (Ref. 3) accidents are the most limiting analyses that establish the value for SDM. For MSLB accidents, if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed applicable 10 CFR 50.67 limits (Ref. 4). For the boron dilution accident, if the LCO is violated, then the minimum required time assumed for operator action to terminate dilution may no longer be applicable.
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| SDM is a core physics design condition that can be ensured through full-length control rod positioning (regulating and shutdown rods) and through the soluble boron concentration.
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| APPLICABILITY In MODE 3, 4 and 5, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. In MODES 1 and 2, SDM is ensured by complying with LCO 3.1.5, and LCO 3.1.6. In MODE 6, the shutdown reactivity requirements are given in LCO 3.9.1, "Boron Concentration."
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| Palisades Nuclear Plant B 3.1.1-3 Revised 01/29/2020
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| | |
| INSERT Bases 3.1.1 SDM B 3.1.1 BASES ACTIONS A.1 If the SDM requirements are not met, boration must be initiated promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. It is assumed that boration will be continued until the SDM requirements are met.
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| In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative to raise the boron concentration of the PCS as soon as possible, the boron injection flow should be a highly concentrated solution, such as that normally found in the concentrated boric acid storage tank. The operator should borate with the best source available for the plant conditions.
| |
| In determining the boration flow rate, the time in core life must be considered. For instance, the most difficult time in core life to increase the PCS boron concentration is at the beginning of cycle, when the boron concentration may approach or exceed 2000 ppm. Assuming that a value of 1% must be recovered and a boration flow rate of 35 gpm, it is possible to increase the boron concentration of the PCS by 100 ppm in approximately 25 minutes. If a boron worth of 1.0 E-4
| |
| /ppm is assumed, this combination of parameters will increase the SDM by 1% . These boration parameters of 35 gpm and 100 ppm represent typical values and are provided for the purpose of offering a specific example.
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| SURVEILLANCE SR 3.1.1.1 REQUIREMENTS SDM is verified by a reactivity balance calculation, considering the listed reactivity effects:
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| : a. PCS boron concentration;
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| : b. Control rod positions;
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| : c. PCS average temperature;
| |
| : d. Fuel burnup based on gross thermal energy generation;
| |
| : e. Xenon concentration; and Palisades Nuclear Plant B 3.1.1-4 Revised 01/29/2020
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| | |
| INSERT Bases 3.1.1 SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 (continued)
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| REQUIREMENTS (continued) f. Isothermal Temperature Coefficient (ITC).
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| Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical and the fuel temperature will be changing at the same rate as the PCS.
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| Samarium is not considered in the reactivity analysis since the analysis assumes that the negative reactivity due to Samarium is offset by the positive reactivity of Plutonium built in.
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| SR 3.1.1.1 requires SDM to be within the limits specified in the COLR.
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| This SDM value ensures the consequences of an MSLB, will be acceptable as a result of a cooldown of the PCS which adds positive reactivity in the presence of a negative moderator temperature coefficient as well as the other events described in the Applicable Safety Analysis. As such, the requirements of this SR must be met whenever the plant is in MODES 3, 4, and 5.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Section 5.1
| |
| : 2. FSAR, Section 14.14
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| : 3. FSAR, Section 14.3
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| : 4. 10 CFR 50.67
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| : 5. FSAR, Section 14.2 Palisades Nuclear Plant B 3.1.1-5 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.1.2 Reactivity Balance B 3.1.2 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.2 Reactivity Balance BASES BACKGROUND According to the Palisades Nuclear Plant design criteria (Ref. 1),
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| reactivity shall be controllable, such that, subcriticality is maintained under cold conditions, and acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Therefore, reactivity balance is used as a measure of the predicted versus measured core reactivity during power operation. The periodic confirmation of core reactivity is necessary to ensure that Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity difference could be the result of unanticipated changes in fuel, control rod worth, or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SHUTDOWN MARGIN (SDM) or violation of acceptable fuel design limits. Comparing predicted versus measured core reactivity validates the nuclear methods used in the safety analysis and supports the SDM demonstrations (LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") in ensuring the reactor can be brought safely to cold, subcritical conditions.
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| When the reactor core is critical, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and measured reactivity is convenient under such a balance, since parameters are being maintained relatively stable under steady state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons such as burnable absorbers. Excess reactivity can be inferred from the critical boron curve, which provides an indication of the soluble boron concentration in the Primary Coolant System (PCS) versus cycle burnup. Periodic measurement of the PCS boron concentration for comparison with the predicted value with other variables fixed (such as control rod height, temperature, pressure, and power) provides a convenient method of ensuring that core reactivity is within design expectations, and that the calculational models used to generate the safety analysis are adequate.
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| Palisades Nuclear Plant B 3.1.2-1 Revised 01/29/2020
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| INSERT Bases 3.1.2 Reactivity Balance B 3.1.2 BASES BACKGROUND In order to achieve the required fuel cycle energy output, the uranium (continued) enrichment in the new fuel loading and in the fuel remaining from the previous cycle, provides excess positive reactivity beyond that required to sustain steady state operation throughout the cycle. When the reactor is critical at RTP and moderator temperature, the excess positive reactivity is compensated by burnable poisons, full-length control rods, neutron poisons (mainly xenon and samarium) in the fuel, and the PCS boron concentration.
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| When the core is producing THERMAL POWER, the fuel is being depleted and excess reactivity is decreasing. As the fuel depletes, the PCS boron concentration is reduced to decrease negative reactivity and maintain constant THERMAL POWER. The critical boron curve is based on steady state operation at RTP. Therefore, deviations from the predicted critical boron curve may indicate deficiencies in the design analysis, deficiencies in the calculational models, or abnormal core conditions, and must be evaluated.
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| APPLICABLE Accurate prediction of core reactivity is either an explicit or implicit SAFETY ANALYSES assumption in the accident analysis evaluations. Every accident evaluation (Ref. 2) is, therefore, dependent upon accurate evaluation of core reactivity. In particular, SDM and reactivity transients, such as control rod withdrawal accidents or control rod ejection accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks. Monitoring reactivity balance additionally ensures that the nuclear methods provide an accurate representation of the core reactivity.
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| Design calculations and safety analyses are performed for each fuel cycle for the purpose of predetermining reactivity behavior and the PCS boron concentration requirements for reactivity control during fuel depletion.
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| Palisades Nuclear Plant B 3.1.2-2 Revised 01/29/2020
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| INSERT Bases 3.1.2 Reactivity Balance B 3.1.2 BASES APPLICABLE The comparison between measured and predicted initial core reactivity SAFETY ANALYSES provides a normalization for calculational models used to predict core (continued) reactivity. If the measured and predicted PCS boron concentrations for identical core conditions at Beginning Of Cycle (BOC) are not within design tolerances, then the assumptions used in the reload cycle design analysis or the calculational models used to predict soluble boron requirements may not be accurate. If reasonable agreement between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to the measured boron concentration.
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| Thereafter, any significant deviations in the measured boron concentration from the predicted critical boron curve that develop during fuel depletion may be an indication that the calculational model is not adequate for core burnups beyond BOC, or that an unexpected change in core conditions has occurred.
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| The normalization of predicted PCS boron concentration to the measured value is typically performed after reaching RTP following startup from a refueling outage, with the control rods in their normal positions for power operation. The normalization is performed at BOC conditions, so that core reactivity relative to predicted values can be continually monitored and evaluated as core conditions change during the cycle.
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| The reactivity balance satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| LCO The reactivity balance limit is established to ensure plant operation is maintained within the assumptions of the safety analyses. Large differences between actual and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the nuclear design methodology are larger than expected. A limit on the reactivity balance of +/- 1% has been established, based on engineering judgment. A 1% deviation in reactivity from that predicted is larger than expected for normal operation and should therefore be evaluated.
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| Palisades Nuclear Plant B 3.1.2-3 Revised 01/29/2020
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| INSERT Bases 3.1.2 Reactivity Balance B 3.1.2 BASES LCO When measured core reactivity is within +/- 1% of the predicted value (continued) at steady state thermal conditions, the core is considered to be operating within acceptable design limits. Since deviations from the limits are normally detected by comparing predicted and measured steady state PCS critical boron concentrations, the difference between measured and predicted values would be approximately 100 ppm (depending on the boron worth) before the limit is reached. These values are well within the uncertainty limits for analysis of boron concentration samples, so that spurious violations of the limit due to uncertainty in measuring the PCS boron concentration is unlikely.
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| APPLICABILITY The limits on core reactivity must be maintained during MODES 1 and 2 because a reactivity balance must exist when the reactor is critical or producing THERMAL POWER. As the fuel depletes, core conditions are changing, and confirmation of the reactivity balance ensures the core is operating as designed. This specification does not apply in MODE 2 because enough operating margin exists to limit the effects of a reactivity anomaly, and THERMAL POWER is low enough
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| ( 5% RTP) such that reactivity anomalies are unlikely to occur. This Specification does not apply in MODES 3, 4, and 5 because the reactor is shut down and the reactivity balance is not changing.
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| In MODE 6, fuel loading results in a continually changing core reactivity.
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| Boron concentration requirements (LCO 3.9.1, "Boron Concentration")
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| ensure that fuel movements are performed within the bounds of the safety analysis.
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| ACTIONS A.1 and A.2 Should an imbalance develop between measured and predicted core reactivity, an evaluation of the core design and safety analysis must be performed. Core conditions are evaluated to determine their consistency with input to design calculations. Measured core and process parameters are evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models are reviewed to verify that they are adequate for representation of the core conditions.
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| Palisades Nuclear Plant B 3.1.2-4 Revised 01/29/2020
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| INSERT Bases 3.1.2 Reactivity Balance B 3.1.2 BASES ACTIONS A.1 and A.2 (continued)
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| The required Completion Time of 7 days is based on the low probability of a DBA occurring during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis.
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| Following evaluations of the core design and safety analysis, the cause of the reactivity imbalance may be resolved. If the cause of the reactivity imbalance is a mismatch in core conditions at the time of PCS boron concentration sampling, then a recalculation of the PCS boron concentration requirements may be performed to demonstrate that core reactivity is behaving as expected. If an unexpected physical change in the condition of the core has occurred, it must be evaluated and corrected, if possible. If the cause of the reactivity imbalance is in the calculation technique, then the calculational models must be revised to provide more accurate predictions. If any of these results are demonstrated, and it is concluded that the reactor core is acceptable for continued operation, then the critical boron curve may be renormalized, and power operation may continue. If operational restrictions or additional SRs are necessary to ensure the reactor core is acceptable for continued operation, then they must be defined.
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| The required Completion Time of 7 days is adequate for preparing whatever operating restrictions or Surveillances that may be required to allow continued reactor operation.
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| B.1 If the Required Actions for Condition A are not met within 7 days, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours. The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 2 from full power conditions in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.1.2-5 Revised 01/29/2020
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| Reactivity Balance INSERT Bases 3.1.2 B 3.1.2 BASES SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Core reactivity is verified by periodic comparisons of measured and predicted PCS boron concentrations. The comparison is made considering that other core conditions are fixed or stable including control rod position, moderator temperature, fuel temperature, fuel depletion, and xenon concentration. The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note in the Surveillance column which indicates that if the normalization of predicted core reactivity to the measured value is to occur, it must take place within the first 60 Effective Full Power Days (EFPD) after each refueling. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. A second Note, "only required after initial 60 EFPD," is added to the Frequency column to allow this.
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| REFERENCES 1. FSAR, Section 5.1
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| : 2. FSAR, Chapter 14 Palisades Nuclear Plant B 3.1.2-6 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.1.3 MTC B 3.1.3 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Moderator Temperature Coefficient (MTC)
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| BASES BACKGROUND According to Palisades Nuclear Plant design criteria (Ref. 1), the reactor core and its interaction with the Primary Coolant System (PCS) must be designed for inherently stable power operation, even in the possible event of an accident. In particular, the net reactivity feedback in the system must compensate for any unintended or rapid reactivity increases.
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| The MTC relates a change in core reactivity to a change in primary coolant temperature. A positive MTC means that reactivity increases with increasing moderator temperature; conversely, a negative MTC means that reactivity decreases with increasing moderator temperature. The reactor is designed to operate with a negative MTC over the largest possible range of fuel cycle operation. Therefore, a coolant temperature increase will cause a reactivity decrease. Reactivity increases that cause a coolant temperature increase will thus be self limiting, and stable power operation will result.
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| MTC values are predicted at selected burnups during the safety evaluation analysis and are confirmed to be acceptable by the measurement performed as part of startup testing following a refueling.
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| Both initial and reload cores are designed so that the Beginning Of Cycle (BOC) MTC is less positive than that allowed by the LCO. The actual value of the MTC is dependent on core characteristics, such as fuel loading and primary coolant soluble boron concentration. The core design may require additional fixed distributed poisons (lumped burnable poison assemblies) to yield an MTC at BOC within the range analyzed in the plant accident analysis. The End Of Cycle (EOC) MTC is also limited by the requirements of the accident analysis. However, the safety analysis assumptions for the MTC at EOC are assumed by confirming the BOC MTC measurement is within limits which indicates the core is behaving as predicted.
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| Palisades Nuclear Plant B 3.1.3-1 Amendment No. 189
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| INSERT Bases 3.1.3 MTC B 3.1.3 BASES APPLICABLE The acceptance criteria for the specified MTC are:
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| SAFETY ANALYSES
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| : a. The MTC values must remain within the bounds of those used in the accident analysis (Ref. 2); and
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| : b. The MTC must be such that inherently stable power operations result during normal operation and during accidents, such as overheating and overcooling events.
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| Reference 2 contains analyses of accidents that result in both overheating and overcooling of the reactor core. MTC is one of the controlling parameters for core reactivity in these accidents. Both the most positive value and most negative value of the MTC are important to safety, and both values must be bounded. Values used in the analyses consider worst case conditions, such as very large soluble boron concentrations, to ensure the accident results are bounding (Ref. 3).
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| Accidents that cause core overheating, either by decreased heat removal or increased power production, must be evaluated for results when the MTC is positive. Examples of reactivity accidents that cause increased power production include the control rod bank withdrawal transient from either partial or RATED THERMAL POWER. The limiting overheating event relative to plant response is based on the maximum difference between core power and steam generator heat removal during a transient. Several events discussed in Reference 2 are analyzed with a positive MTC.
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| Accidents that cause core overcooling must be evaluated for results when the MTC is most negative. The event that produces the most rapid cooldown of the PCS, and is therefore the most limiting event with respect to the negative MTC, is a Main Steam Line Break (MSLB) event.
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| Following the reactor trip for the postulated EOC MSLB event, the large moderator temperature reduction combined with the large negative MTC may produce reactivity increases that are as much as the shutdown reactivity. When this occurs, a substantial fraction of core power (approximately 12% RTP) is produced.
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| The MTC satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| Palisades Nuclear Plant B 3.1.3-2 Amendment No. 189
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| INSERT Bases 3.1.3 MTC B 3.1.3 BASES LCO LCO 3.1.3 requires the MTC to be < 0.5 E-4 /°F at 2% RTP to ensure the core operates within the assumptions of the accident analysis. During the reload core safety evaluation, the MTC is analyzed to determine that its values remain within the bounds of the original accident analysis during operation. The limit on a positive MTC ensures that core overheating accidents will not violate the accident analysis assumptions.
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| MTC is a core physics parameter determined by the fuel and fuel cycle design and cannot be easily controlled once the core design is fixed.
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| During operation, therefore, the LCO can only be ensured through measurement. The surveillance check at BOC on the MTC provide confirmation that the MTC is behaving as anticipated, so that the acceptance criteria are met.
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| APPLICABILITY In MODE 1, the MTC must be maintained to ensure that any accident initiated from THERMAL POWER operation will not violate the design assumptions of the accident analysis. In MODE 2, the limits must also be maintained to ensure startup and subcritical accidents, such as the uncontrolled full-length control rod or group withdrawal, will not violate the assumptions of the accident analysis. The measurement of MTC in MODE 2 prior to exceeding 2% RTP is used to confirm that the core is behaving as analyzed. This ensures that the MTC will remain within the analyzed range while operating in MODES 1 and 2. In MODES 3, 4, 5, and 6, this LCO is not applicable, since no Design Basis Accidents (DBAs) using the MTC as an analysis assumption are initiated from these MODES. However, the variation of the MTC, with temperature in MODES 3, 4, and 5, for DBAs initiated in MODES 1 and 2, is accounted for in the subject accident analysis. The variation of the MTC, with temperature assumed in the safety analysis, is accepted as valid once the BOC measurement is used for normalization.
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| Palisades Nuclear Plant B 3.1.3-3 Amendment No. 189
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| INSERT Bases 3.1.3 MTC B 3.1.3 BASES ACTIONS A.1 MTC is a function of the fuel and fuel cycle designs, and cannot be controlled directly once the designs have been implemented in the core.
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| If MTC exceeds its limits, the reactor must be placed in MODE 3. This eliminates the potential for violation of the accident analysis bounds. The associated Completion Time of 6 hours is reasonable, considering the probability of an accident occurring during the time period that would require an MTC value within the LCO limits, and the time for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
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| SURVEILLANCE SR 3.1.3.1 REQUIREMENTS The SR for measurement of the MTC at the beginning of each fuel cycle provides for confirmation of the limiting MTC values. The MTC changes smoothly from most positive (or least negative) to most negative value during fuel cycle operation as the PCS boron concentration is reduced to compensate for fuel depletion. The requirement for measurement prior to operation > 2% RTP satisfies the confirmatory check on the most positive (or least negative) MTC value. It also confirms that the core is behaving as analyzed which ensures that the MTC will remain within the analysis limits for the remainder of the fuel cycle.
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| REFERENCES 1. FSAR, Section 5.1
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| : 2. FSAR, Chapter 14
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| : 3. FSAR, Section 3.3 Palisades Nuclear Plant B 3.1.3-4 Amendment No. 189
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| INSERT Bases 3.1.4 Control Rod Alignment B 3.1.4 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 Control Rod Alignment BASES BACKGROUND The OPERABILITY (e.g., trippability) of the shutdown and regulating rods is an initial assumption in all safety analyses that assume full-length control rod insertion upon reactor trip. Maximum control rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM.
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| The Palisades Nuclear Plant design criteria contain the applicable criteria for these reactivity and power distribution design requirements (Ref. 1).
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| Mechanical or electrical failures may cause a control rod to become inoperable or to become misaligned from its group. Control rod misalignment may cause increased power peaking, due to the asymmetric reactivity distribution, and a reduction in the total available control rod worth for reactor shutdown. Therefore, control rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.
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| Limits on control rod alignment and OPERABILITY have been established, and all control rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.
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| Control rods are moved by their Control Rod Drive Mechanisms (CRDMs). Each CRDM moves its rod at a fixed rate of approximately 46 inches per minute. Although the ability to move a full-length control rod by its drive mechanism is not an initial assumption used in the safety analyses, it is required to support OPERABILITY. As such, the inability to move a full-length control rod results in that full-length control rod being inoperable.
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| The control rods are arranged into groups that are radially symmetric.
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| Therefore, movement of the control rod groups does not introduce radial asymmetries in the core power distribution. The shutdown and regulating rods provide the required reactivity worth for immediate reactor shutdown upon a reactor trip. The regulating rods also provide reactivity (power level) control during normal operation and transients.
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| Palisades Nuclear Plant B 3.1.4-1 Revised 01/29/2020
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| INSERT Bases 3.1.4 Control Rod Alignment B 3.1.4 BASES BACKGROUND The axial position of shutdown and regulating rods is indicated by two (continued) separate and independent systems, which are 1) synchro based position indication system, and 2) the reed switch based position indication system.
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| The synchro based position indication system measures the phase angle of a synchro geared to the CRDM rack. Full control rod travel corresponds to less than 1 turn of the synchro. Each control rod has its own synchro. The Primary Information Processor (PIP) node scans and converts synchro outputs into inches of control rod withdrawal. The resolution of this system is approximately 0.5 inches. Each synchro also has cam operated limit switches that provide input to the matrix indication lights of control rod status indication for various key positions.
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| The reed switch based position indication system is referred to as the Secondary Position Indication (SPI) system. This system provides a highly accurate indication of actual control rod position, but at a lower precision than the synchros. The reed switches are wired so that the voltage read across the reed switch stack is proportional to rod position.
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| The reed switches are spaced along a tube with a center-to-center spacing distance of 1.5 inches. The resolution of the SPI reed switch stacks is 1.5 inches. The reed switches also provide input to the matrix indication lights that provide control rod status indication for various key positions. To increase the reliability of the system, there are redundant reed switches that prevent false indication in the event an individual reed switch fails.
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| A control rod position deviation alarm is provided to alert the operator when any two control rods in the same group are more than 8 inches apart. This helps to ensure any control rod misalignments are minimized. The alarm can be generated by either the SPI system or PIP node since the SPI system, in conjunction with the host computer, is redundant to the PIP node in the task of control rod measurements, control rod monitoring, and limit processing.
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| Palisades Nuclear Plant B 3.1.4-2 Revised 01/29/2020
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| INSERT Bases 3.1.4 Control Rod Alignment B 3.1.4 BASES APPLICABLE Control rod misalignment accidents are analyzed in the safety analysis SAFETY ANALYSES (Refs. 3 and 4). The accident analysis defines control rod misoperation as any event, with the exception of sequential group withdrawals, which could result from a single malfunction in the reactivity control systems.
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| For example, control rod misalignment may be caused by a malfunction of the Rod Control System, or by operator error. A stuck rod may be caused by mechanical jamming. Inadvertent withdrawal of a single control rod may be caused by an electrical or mechanical failure in the Rod Control System. A dropped control rod could be caused by an electrical or mechanical failure in the CRDM.
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| The acceptance criteria for addressing control rod inoperability/misalignment are that:
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| : a. There shall be no violations of:
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| : 1. Specified Acceptable Fuel Design Limits (SAFDL), or
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| : 2. Primary Coolant System (PCS) pressure boundary integrity; and
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| : b. The core must remain subcritical after accident transients.
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| Three types of misoperations are discussed in the safety analysis (Ref. 4). During movement of a group, one control rod may stop moving while the other control rods in the group continue. This condition may cause excessive power peaking. The second type of misoperations occurs if one control rod fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is held in the remaining control rods to meet the SDM requirement with the maximum worth rod stuck fully withdrawn.
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| If a control rod is stuck in the fully withdrawn position, its worth is added to the SDM requirement, since the safety analysis does not take two stuck rods into account. The third type of misoperations occurs when one rod drops partially or fully into the reactor core. This event causes an initial power reduction followed by a return towards the original power, due to positive reactivity feedback from the negative moderator temperature coefficient. Increased peaking during the power increase may result in excessive local Linear Heat Rates (LHRs).
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| Palisades Nuclear Plant B 3.1.4-3 Revised 01/29/2020
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| INSERT Bases 3.1.4 Control Rod Alignment B 3.1.4 BASES APPLICABLE The most limiting static misalignment occurs when Bank 4 is fully SAFETY ANALYSES inserted with one rod fully withdrawn ([Bank 4 is 99 inches out of (continued) alignment with the rated Power Dependent Insertion Limit [PDIL].) This event was bounded by the dropped full-length control rod event (Ref. 4).
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| Since the control rod drop incidents result in the most rapid approach to SAFDLs caused by a control rod misoperation, the accident analysis analyzed a single full-length control rod drop.
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| The above control rod misoperations may or may not result in an automatic reactor trip. In the case of the full-length rod drop, a prompt decrease in core average power and a distortion in radial power are initially produced, which, when conservatively coupled, result in a local power and heat flux increase, and a decrease in DNBR parameters.
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| The results of the control rod misoperation analysis show that during the most limiting misoperation events, no violations of the SAFDLs, fuel centerline temperature, or PCS pressure occur.
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| Control rod alignment satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2).
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| LCO The limits on shutdown, regulating, and part-length rod alignments ensure that the assumptions in the safety analysis will remain valid. The requirements on OPERABILITY ensure that upon reactor trip, the full-length control rods will be available and will be inserted to provide enough negative reactivity to shut down the reactor. The OPERABILITY requirements also ensure that the control rod banks maintain the correct alignment and that each full-length control rod is capable of being moved by its CRDM. The OPERABILITY requirement for the part-length rods is that they are fully withdrawn.
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| The requirement is to maintain the control rod alignment to within 8 inches between any control rod and all other rods in its group. To help ensure this requirement is met, the control rod position deviation alarm generated by either the PIP node or the SPI system, must be OPERABLE and provide an alarm when any control rod becomes misaligned
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| > 8 inches from any other rod in its group. The safety analysis assumes a total misalignment from fully withdrawn to fully inserted. This case bounds the safety analysis for a single rod in any intermediate position.
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| Palisades Nuclear Plant B 3.1.4-4 Revised 01/29/2020
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| INSERT Bases 3.1.4 Control Rod Alignment B 3.1.4 BASES LCO The primary rod position indication system is considered OPERABLE, (continued) for purposes of this specification, if the digital position readout or the PPC display provides valid rod position indication, or if the cam operated red matrix light (regulating and part-length rods only) gives positive (ON) indication of rod position. The secondary rod position indication system is considered OPERABLE, for purposes of this specification, if the magnetically operated reed switches are providing valid indication of rod position either via the plant process computer or by taking direct readings of the output from the magnetic reed switches or if the reed switch operated red matrix light (shutdown rods only) gives positive (ON) indication of rod position.
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| Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDM, any of which may constitute initial conditions inconsistent with the safety analysis.
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| APPLICABILITY The requirements on control rod OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which neutron (or fission) power is generated, and the OPERABILITY (e.g., trippability) and alignment of control rods have the potential to affect the safety of the plant. In MODES 3, 4, 5, and 6, the alignment limits do not apply because the reactor is shut down and not producing fission power. In the shutdown MODES, the OPERABILITY of the shutdown and regulating rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the PCS. See LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," for SDM in MODES 3, 4, and 5, and LCO 3.9.1, "Boron Concentration," for boron concentration requirements during refueling.
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| ACTIONS LCOs 3.1.4, 3.1.5, and 3.1.6, and their ACTIONS were written to support each other. The combined intent is to assure the following:
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| : 1. There is adequate SDM available in withdrawn control rods to assure the reactor is shutdown by, and remains shutdown following, a reactor trip,
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| : 2. The control rod positioning does not cause unacceptable axial or radial flux peaking, and
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| : 3. The programmed rod withdrawal sequence and group overlap result in reactivity insertion rates within the assumptions of the Inadvertent Control Rod Bank Withdrawal Analyses.
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| Palisades Nuclear Plant B 3.1.4-5 Revised 01/29/2020
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| INSERT Bases 3.1.4 Control Rod Alignment B 3.1.4 BASES ACTIONS The ACTIONS for rods that are mispositioned (misaligned or inserted (continued) beyond the limit) were written assuming that an OPERABLE rod discovered to be mispositioned would simply be re-positioned correctly.
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| While the associated Conditions would have to be entered, the rod could be re-positioned (thus exiting the LCO) without taking any other Required Action. A rod that remains mispositioned was assumed to be inoperable. The analyses account for operation with one (and only one) mispositioned rod (a dropped rod being the limiting case). With more than one mispositioned rod, the plant would be outside the bounds of the analyses and must be shutdown.
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| If a rod is discovered to be misaligned (ie, there is more than 8 inches between it and any other rod in its group, but all remaining rods in that group are within 8 inches of each other) Condition 3.1.4 C allows 2 hours to restore the rod alignment (thus exiting the LCO), perform SR 3.2.2.1 (verification that radial peaking is within limits), or reduce power to 75%
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| RTP.
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| If one or more shutdown rods are inserted beyond the insertion limit, Condition 3.1.5 A is entered; the rods are declared inoperable and Condition 3.1.4 D (when one rod is immovable but trippable) or Condition 3.1.4 E (when a movable rod is inserted beyond its insertion limit, or when more than one rod is inoperable for any reason) must be entered.
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| If the rods can be moved, they should be withdrawn and all Conditions exited.
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| If one rod cannot be moved (but is still considered trippable),
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| operation may continue in accordance with Condition 3.1.4 D (and 3.1.4 C if it is misaligned).
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| If more than one rod cannot be moved, Condition 3.1.4 E must also be entered. The plant must be in MODE 3 in 6 hours in accordance with ACTION 3.1.4 E.1.
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| If one or more part length rods are inserted beyond the limit, Condition 3.1.5 A is entered; the rods are declared inoperable and Condition 3.1.4 E is entered (and 3.1.4 C if it is misaligned). Condition 3.1.4 D is not applicable to part-length rods since it only addresses full-length rods.
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| If the rods can be moved, they should be withdrawn and all Conditions exited.
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| Palisades Nuclear Plant B 3.1.4-6 Revised 01/29/2020
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| INSERT Bases 3.1.4 Control Rod Alignment B 3.1.4 BASES ACTIONS If any part-length rods are inserted beyond the limit and cannot (continued) be moved, the plant must be placed in MODE 3 in 6 hours in accordance with ACTION 3.1.4 E.1.
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| If one or more OPERABLE regulating rods are inserted beyond the limit, Condition 3.1.6 A is entered.
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| The rods must be restored to within limits (by rod withdrawal or power reduction) within two hours.
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| If a rod cannot be moved, it must be considered inoperable and Condition 3.1.4 D must be entered (and 3.1.4 C if it is misaligned).
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| Condition 3.1.4 D allows continued operation with one inoperable, but trippable, rod until the next reactor shutdown (MODE 3 entry). If more than one rod cannot be moved, Condition 3.1.4 E must be entered. The plant must be in MODE 3 in 6 hours in accordance with ACTION 3.1.4 E.1.
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| The analyses do not account for the possibility of more then one rod failing to insert on a trip. While boron concentration might be adjusted to restore SHUTDOWN MARGIN, if two adjacent rods fail to insert that portion of the core could remain excessively reactive. Since the analyses must assume that one rod fails to insert, operation may not continue with a known untrippable rod. A shutdown would be required by Condition 3.1.4 E.
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| A.1 Rod position indication is required to allow verification that the rods are positioned and aligned as assumed in the safety analysis. If one rod position indication channel is inoperable for one or more control rods then SR 3.1.4.1 (rod position verification) is required to be performed once within 15 minutes following any rod motion in that group. This ensures that the rods are positioned as required.
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| Palisades Nuclear Plant B 3.1.4-7 Revised 01/29/2020
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| INSERT Bases 3.1.4 Control Rod Alignment B 3.1.4 BASES ACTIONS B.1 (continued)
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| When the control rod deviation alarm is inoperable, performing SR 3.1.4.1, once within 15 minutes of movement of any control rod, ensures improper control rod alignments are identified before unacceptable flux distributions occur. The specified Completion Times take into account other information continuously available to the operator in the control room, so that during control rod movement, deviations can be detected, and the protection provided by the control rod and deviation circuit is not required.
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| C.1 and C.2 Condition C addresses the situation where one rod in a group is misaligned, ie. there is more than 8 inches between that rod and any other rod in its group, but all remaining rods in that group are within 8 inches of each other.
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| A full-length control rod may become misaligned yet remain trippable. In this condition, the control rod can still perform its required function of adding negative reactivity should a reactor trip be necessary.
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| Regulating rod alignment can be restored by either aligning the misaligned rod(s) to within 8 inches of all other rods in its group or, aligning the misaligned rods group to within 8 inches of the misaligned rod if allowed by the rod group insertion limits. Shutdown rod alignment can be restored by aligning the misaligned rod to within 8 inches of all other rods in its group.
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| If one control rod is misaligned by > 8 inches continued operation in MODES 1 and 2 may continue, provided, within 2 hours, the TOTAL RADIAL PEAKING FACTOR has been verified acceptable in accordance with SR 3.2.2.1, or the power is reduced to 75% RTP.
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| Xenon redistribution in the core starts to occur as soon as a rod becomes misaligned. Reducing THERMAL POWER to 75% RTP ensures acceptable power distributions are maintained.
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| Palisades Nuclear Plant B 3.1.4-8 Revised 01/29/2020
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| INSERT Bases 3.1.4 Control Rod Alignment B 3.1.4 BASES ACTIONS C.1 and C.2 (continued)
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| For small misalignments of the control rods, there is:
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| : a. A small effect on the time dependent long-term power distributions relative to those used in generating LCOs and Limiting Safety System Settings (LSSS) setpoints;
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| : b. A negligible effect on the available SDM; and
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| : c. A small effect on the ejected rod worth used in the accident analysis.
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| With a large control rod misalignment, however, this misalignment would cause distortion of the core power distribution. This distortion may, in turn, have a significant effect on the time dependent, long-term power distributions relative to those used in generating LCOs and LSSS setpoints.
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| The effect on the available SDM and the ejected rod worth used in the accident analysis remains small.
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| In both cases, a 2-hour time period is sufficient to:
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| : a. Identify cause of a misaligned rod;
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| : b. Take appropriate corrective action to realign the rods; and
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| : c. Minimize the effects of xenon redistribution.
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| The Palisades analysis for rod misalignment is bounded by a single dropped rod. Therefore, rod misalignments are limited to one rod being misaligned from its group. If a full-length control rod is untrippable, it is not available for reactivity insertion during a reactor trip. With an untrippable full-length control rod, meeting the insertion limits of LCO 3.1.5, Shutdown and Part-Length Rod Group Insertion Limits, and LCO 3.1.6, Regulating Rod Group Position Limits, does not ensure that adequate SDM exists and therefore, the Actions of Condition E must be met.
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| Palisades Nuclear Plant B 3.1.4-9 Revised 01/29/2020
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| INSERT Bases 3.1.4 Control Rod Alignment B 3.1.4 BASES ACTIONS D.1 (continued)
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| Condition D is entered whenever it is discovered that a single full-length control rod cannot be moved by its operator, yet the control rod is still capable of being tripped (or is fully inserted). Although the ability to move a full-length control rod is not an initial assumption used in the safety analyses, it does relate to full-length control rod OPERABILITY. The inability to move a full-length control rod by its operator may be indicative of a systemic failure (other than trippability) that could potentially affect other rods. Thus, declaring a full-length control rod inoperable in this instance is conservative since it limits the number of full-length control rods that cannot be moved by their operators to only one. The Completion Time to restore an inoperable control rod to OPERABLE status is stated as prior to entering MODE 2 following next MODE 3 entry.
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| This Completion Time allows unrestricted operation in MODES 1 and 2 while conservatively preventing a reactor startup with an immovable full-length control rod.
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| E.1 If the Required Action or associated Completion Time of Condition A, Condition B, Condition C, or Condition D is not met; one or more control rods are inoperable for reasons other than Condition D (ie, one full length control rod is inoperable for reasons other than being immovable but trippable, or more than one control rod, whether full length or part length, are inoperable for any reasons); or two or more control rods are misaligned by > 8 inches, or two channels of control rod position indication are inoperable for one or more control rods, the plant is required to be brought to MODE 3. By being brought to MODE 3, the plant is brought outside its MODE of applicability. Continued operation is not allowed in the case of more than one control rod misaligned from any other rod in its group by > 8 inches, or two or more rods inoperable. This is because these cases may be indicative of a loss of SDM and power re-distribution, and a loss of safety function, respectively.
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| Also, if no rod position indication exists for one or more control rods, continued operation is not allowed because the safety analysis assumptions of rod position cannot be ensured.
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| When a Required Action cannot be completed within the required Completion Time, a controlled shutdown should be commenced. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.1.4-10 Revised 01/29/2020
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| INSERT Bases 3.1.4 Control Rod Alignment B 3.1.4 BASES SURVEILLANCE SR 3.1.4.1 REQUIREMENTS Verification that individual control rod positions are within 8 inches of all other control rods in the group allows the operator to detect a control rod that is beginning to deviate from its expected position. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.1.4.2 OPERABILITY of two control rod position indicator channels is required to determine control rod positions, and thereby ensure compliance with the control rod alignment and insertion limits. Performance of a CHANNEL CHECK on the primary and secondary control rod position indication channels provides confidence in the accuracy of the rod position indication systems. The control rod "full in" and "full out" lights, which correspond to the lower electrical limit and the upper electrical limit respectively, provide an additional means for determining the control rod positions when the control rods are at either their fully inserted or fully withdrawn positions.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.1.4-11 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.1.4 Control Rod Alignment B 3.1.4 BASES SURVEILLANCE SR 3.1.4.3 REQUIREMENTS (continued) Verifying each full-length control rod is trippable would require that each full-length control rod be tripped. In MODES 1 and 2, tripping each full-length control rod would result in radial or axial power tilts, or oscillations. Therefore, individual full-length control rods are exercised to provide increased confidence that all full-length control rods continue to be trippable, even if they are not regularly tripped. A movement of 6 inches is adequate to demonstrate motion without exceeding the alignment limit when only one control rod is being moved. At any time, if a control rod(s) is immovable, a determination of the trippability of the control rod(s) must be made, and appropriate action taken. Condition 3.1.4 D would apply whenever it is discovered that a single full-length control rod cannot be moved by its operator, yet the control rod is still capable of being tripped (or is fully inserted.) The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.1.4.4 Demonstrating the rod position deviation alarm is OPERABLE verifies the alarm is functional. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.1.4.5 Performance of a CHANNEL CALIBRATION of each control rod position indication channel ensures the channel is OPERABLE and capable of indicating control rod position over the entire length of the control rod's travel with the exception of the secondary rod position indicating channel dead band near the bottom of travel. This dead band exists because the control rod drive mechanism housing seismic support prevents operation of the reed switches. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.1.4-12 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.1.4 Control Rod Alignment B 3.1.4 BASES SURVEILLANCE SR 3.1.4.6 REQUIREMENTS (continued) Verification of full-length control rod drop times determines that the maximum control rod drop time is consistent with the assumed drop time used in that safety analysis (Ref. 2). The 2.5-second acceptance criteria is measured from the time the CRDM clutch is deenergized by the reactor protection system or test switch to 90% insertion. This time is bounded by that assumed in the safety analysis (Ref.2). Measuring drop times prior to reactor criticality, after reactor vessel head reinstallation, ensures that reactor internals and CRDMs will not interfere with full-length control rod motion or drop time and that no degradation in these systems has occurred that would adversely affect full-length control rod motion or drop time. Individual full-length control rods whose drop times are greater than safety analysis assumptions are not OPERABLE. This SR is performed prior to criticality, based on the need to perform this Surveillance under the conditions that apply during a plant outage and because of the potential for an unplanned plant transient if the Surveillance were performed with the reactor at power.
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| REFERENCES 1. FSAR, Section 5.1
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| : 2. FSAR, Section 14.1
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| : 3. FSAR, Section 14.4
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| : 4. FSAR, Section 14.6 Palisades Nuclear Plant B 3.1.4-13 Revised 01/29/2020
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| INSERT Bases 3.1.5 Shutdown and Part-Length Rod Group Insertion Limits B 3.1.5 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Shutdown and Part-Length Rod Group Insertion Limits BASES BACKGROUND The insertion limits of the shutdown rods are initial assumptions in all safety analyses that assume full-length control rod insertion upon reactor trip.
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| The insertion limits directly affect core power distributions and assumptions of available SDM, ejected rod worth, and initial reactivity insertion rate.
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| The Palisades Nuclear Plant design criteria (Ref. 1) and 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors, contain the applicable criteria for these reactivity and power distribution design requirements. Limits on shutdown rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the reactivity limits, ejected rod worth, and SDM limits are preserved.
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| The shutdown rods are arranged into groups that are radially symmetric.
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| Therefore, movement of the shutdown rod groups does not introduce radial asymmetries in the core power distribution. The shutdown and regulating rod groups provide the required reactivity worth for immediate reactor shutdown upon a reactor trip.
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| The Palisades Nuclear Plant has four part-length control rods installed.
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| The part-length rods are required to remain completely withdrawn during power operation. The part-length rods do not insert on a reactor trip.
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| The design calculations are performed with the assumption that the shutdown rod groups are withdrawn prior to the regulating rod groups. The shutdown rods can be fully withdrawn without the core going critical. This provides available negative reactivity for SDM in the event of boration errors. All control rod groups are controlled manually by the control room operator. During normal plant operation, the shutdown rod groups are fully withdrawn. The shutdown rod groups must be completely withdrawn from the core prior to withdrawing any regulating rods during an approach to criticality. The shutdown rod groups are then left in this position until the reactor is shut down.
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| They affect core power, burnup distribution, and add negative reactivity to shut down the reactor upon receipt of a reactor trip signal.
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| Palisades Nuclear Plant B 3.1.5-1 Revised 01/29/2020
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| INSERT Bases 3.1.5 Shutdown and Part-Length Rod Group Insertion Limits B 3.1.5 BASES APPLICABLE Accident analysis assumes that the shutdown rod groups are fully SAFETY ANALYSES withdrawn any time the reactor is critical. This ensures that:
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| : a. The minimum SDM is maintained; and
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| : b. The potential effects of a control rod ejection accident are limited to acceptable limits.
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| Control rods are considered fully withdrawn at 128 inches, since this position places them in an insignificant reactivity worth region of the integral worth curve for each bank.
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| On a reactor trip, all full-length control rods (shutdown and regulating),
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| except the most reactive rod, are assumed to insert into the core. The shutdown and regulating rod groups shall be at or above their insertion limits and available to insert the required amount of negative reactivity on a reactor trip signal. The regulating rods may be partially inserted in the core as allowed by LCO 3.1.6, "Regulating Rod Group Position Limits." The shutdown rod group insertion limit is established to ensure that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM)) following a reactor trip from full power.
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| The combination of regulating rod and shutdown rods (less the most reactive rod, which is assumed to remain fully withdrawn) is sufficient to take the reactor from full power conditions at rated temperature to zero power, and to maintain the required SDM at rated no load temperature (Ref. 2). The shutdown rod group insertion limit also limits the reactivity worth of an ejected shutdown rod.
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| The acceptance criteria for addressing shutdown rods as well as regulating rod insertion limits and inoperability or misalignment are that:
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| : a. There be no violation of:
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| : 1. Specified acceptable fuel design limits, or
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| : 2. Primary Coolant System pressure boundary damage; and
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| : b. The core remains subcritical after accident transients.
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| Palisades Nuclear Plant B 3.1.5-2 Revised 01/29/2020
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| INSERT Bases 3.1.5 Shutdown and Part-Length Rod Group Insertion Limits B 3.1.5 BASES APPLICABLE As such, the shutdown and part-length rod group insertion limits affect SAFETY ANALYSES safety analyses involving core reactivity, ejected rod worth, and SDM (continued) (Ref. 2). The part-length control rods have the potential to cause power distribution envelopes to be exceeded if inserted while the reactor is critical. Therefore, they must remain withdrawn in accordance with the limits of the LCO (Ref. 3).
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| The shutdown and part-length rod group insertion limits satisfy Criterion 2 of 10 CFR 50.36(c)(2).
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| LCO The shutdown and part-length rod groups must be within their insertion limits any time the reactor is critical or approaching criticality. For a control rod group to be considered above its insertion limit, all OPERABLE rods in that group, which are not misaligned, must be above the insertion limit (inoperable and misaligned rods are addressed by LCO 3.1.4). Maintaining the shutdown rod groups within their insertion limits ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip. Maintaining the part-length rod group within its insertion limit ensures that the power distribution envelope is maintained.
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| APPLICABILITY The shutdown and part-length rod groups must be within their insertion limits, with the reactor in MODES 1 and 2. In MODE 2 the Applicability begins anytime any regulating rod is withdrawn above 5 inches. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip. In MODE 4, 5, or 6, the shutdown rod groups are inserted in the core to at least the lower electrical limit and contribute to the SDM. In MODE 3 the shutdown rod groups may be withdrawn in preparation of a reactor startup. Refer to LCO 3.1.1, SHUTDOWN MARGIN (SDM), for SDM requirements in MODES 3, 4, and 5.
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| LCO 3.9.1, "Boron Concentration," ensures adequate SDM in MODE 6.
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| The Applicability has been modified by a Note indicating the LCO requirement is suspended during SR 3.1.4.3 (rod exercise test). Control rod exercising verifies the freedom of the rods to move, and requires the individual shutdown rods to move below the LCO limits for their group.
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| Only the full-length rods are required to be tested by SR 3.1.4.3. The part-length rods may also be moved however, if a part-length rod is moved below the limit of the associated LCO, the Required Actions of Condition A must be taken. Positioning of an individual control rod within its group is addressed by LCO 3.1.4, Control Rod Alignment.
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| Palisades Nuclear Plant B 3.1.5-3 Revised 01/29/2020
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| INSERT Bases 3.1.5 Shutdown and Part-Length Rod Group Insertion Limits B 3.1.5 BASES ACTIONS LCOs 3.1.4, 3.1.5, and 3.1.6, and their ACTIONS were written to support each other. The combined intent is to assure the following:
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| : 1. There is adequate SDM available in withdrawn control rods to assure the reactor is shutdown by, and remains shutdown following, a reactor trip,
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| : 2. The control rod positioning does not cause unacceptable axial or radial flux peaking, and
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| : 3. The programmed rod withdrawal sequence and group overlap result in reactivity insertion rates within the assumptions of the Inadvertent Control Rod Bank Withdrawal Analyses.
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| The ACTIONS for rods that are mispositioned (misaligned or inserted beyond the limit) were written assuming that an OPERABLE rod discovered to be mispositioned would simply be re-positioned correctly.
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| While the associated Conditions would have to be entered, the rod could be re-positioned (thus exiting the LCO) without taking any other Required Action. A rod that remains mispositioned was assumed to be inoperable. The analyses account for operation with one (and only one) mispositioned rod (a dropped rod being the limiting case). With more than one mispositioned rod, the plant would be outside the bounds of the analyses and must be shutdown.
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| If a rod is discovered to be misaligned (ie, there is more than 8 inches between it and any other rod in its group, but all remaining rods in that group are within 8 inches of each other) Condition 3.1.4 C allows 2 hours to restore the rod alignment (thus exiting the LCO), perform SR 3.2.2.1 (verification that radial peaking is within limits), or reduce power to 75% RTP.
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| If one or more shutdown rods are inserted beyond the insertion limit, Condition 3.1.5 A is entered; the rods are declared inoperable and Condition 3.1.4 D (when one rod is immovable but trippable) or Condition 3.1.4 E (when a movable rod is inserted beyond its insertion limit, or when more than one rod is inoperable for any reason) must be entered.
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| If the rods can be moved, they should be withdrawn and all Conditions exited.
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| If one rod cannot be moved (but is still considered trippable),
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| operation may continue in accordance with Condition 3.1.4 D (and 3.1.4 C if it is misaligned).
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| Palisades Nuclear Plant B 3.1.5-4 Revised 01/29/2020
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| INSERT Bases 3.1.5 Shutdown and Part-Length Rod Group Insertion Limits B 3.1.5 BASES ACTIONS If more than one rod cannot be moved, Condition 3.1.4 E must also (continued) be entered. The plant must be in MODE 3 in 6 hours in accordance with ACTION 3.1.4 E.1.
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| If one or more part-length rods are inserted beyond the limit, Condition 3.1.5 A is entered; the rods are declared inoperable and Condition 3.1.4 E is entered (and 3.1.4 C if it is misaligned). Condition 3.1.4 D is not applicable to part-length rods since it only addresses full-length rods.
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| If the rods can be moved, they should be withdrawn and all Conditions exited.
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| If any part-length rods are inserted beyond the limit and cannot be moved, the plant must be placed in MODE 3 in 6 hours in accordance with ACTION 3.1.4 E.1.
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| If one or more OPERABLE regulating rods are inserted beyond the limit, Condition 3.1.6 A is entered; The rods must be restored to within limits (by rod withdrawal or power reduction) within two hours.
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| If a rod cannot be moved, it must be considered inoperable and Condition 3.1.4 D must be entered (and 3.1.4 C if it is misaligned).
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| Condition 3.1.4 D allows continued operation with one inoperable, but trippable, rod until the next reactor shutdown (MODE 3 entry). If more than one rod cannot be moved, Condition 3.1.4 E must be entered.
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| The plant must be in MODE 3 in 6 hours in accordance with ACTION 3.1.4 E.1.
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| The analyses do not account for the possibility of more then one rod failing to insert on a trip. While boron concentration might be adjusted to restore SHUTDOWN MARGIN, if two adjacent rods fail to insert that portion of the core could remain excessively reactive. Since the analyses must assume that one rod fails to insert, operation may not continue with a known untrippable rod. A shutdown would be required by Condition 3.1.4 E.
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| Palisades Nuclear Plant B 3.1.5-5 Revised 01/29/2020
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| INSERT Bases 3.1.5 Shutdown and Part-Length Rod Group Insertion Limits B 3.1.5 BASES ACTIONS A.1 (continued)
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| Prior to entering this condition, the shutdown and part-length rod groups were fully withdrawn. If a shutdown rod group is then inserted into the core, its potential negative reactivity is added to the core as it is inserted.
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| If one or more shutdown or part-length rods are not within limits, the affected rod(s) must be declared inoperable and the applicable Conditions and Required Actions of LCO 3.1.4 entered immediately.
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| This Required Action is based on the recognition that the shutdown and part-length rods are normally withdrawn beyond their insertion limits and are capable of being moved by their control rod drive mechanism.
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| Although the requirements of this LCO are not applicable during performance of the control rod exercise test, the inability to restore a control rod to within the limits of the LCO following rod exercising would be indicative of a problem affecting the OPERABILITY of the control rod. Therefore, entering the applicable Conditions and Required Actions of LCO 3.1.4 is appropriate since they provide the applicable compensatory measures commensurate with the inoperability of the control rod.
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| B.1 When Required Action A.1 cannot be met or completed within the required Completion Time, a controlled shutdown should be commenced. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.1.5-6 Revised 01/29/2020
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| INSERT Bases 3.1.5 Shutdown and Part-Length Rod Group Insertion Limits B 3.1.5 BASES SURVEILLANCE SR 3.1.5.1 REQUIREMENTS Verification that the shutdown and part-length rod groups are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown rods will be available to shut down the reactor, and the required SDM will be maintained following a reactor trip. Verification that the part-length rod groups are within their insertion limits ensures that they do not adversely affect power distribution requirements. This SR ensures that the shutdown and part-length rod groups are withdrawn before the regulating rods are withdrawn during a plant startup.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Section 5.1
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| : 2. FSAR, Section 14.2
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| : 3. FSAR, Section 14.6 Palisades Nuclear Plant B 3.1.5-7 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.1.6 Regulating Rod Group Position Limits B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Regulating Rod Group Position Limits BASES BACKGROUND The insertion limits of the regulating rod groups are initial assumptions in all safety analyses that assume full-length rod insertion upon reactor trip. The insertion limits directly affect core power distributions, assumptions of available SDM, and initial reactivity insertion rate. The applicable criteria for these reactivity and power distribution design requirements are contained in the Palisades Nuclear Plant design criteria (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2).
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| Limits on regulating rod group insertion have been established, and all regulating rod group positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking, ejected rod worth, reactivity insertion rate, and SDM limits are preserved.
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| The regulating rod groups operate with a predetermined amount of position overlap, in order to approximate a linear relation between rod worth and rod position (integral rod worth). The regulating rod groups are withdrawn and operate in a predetermined sequence. The group sequence and overlap limits are specified in the COLR.
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| The regulating rods are used for precise reactivity control of the reactor. The positions of the regulating rods are manually controlled.
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| They are capable of changing reactivity very quickly (compared to borating or diluting).
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| The power density at any point in the core must be limited to maintain specified acceptable fuel design limits, including limits that preserve the criteria specified in 10 CFR 50.46 (Ref. 2). Together, LCO 3.1.6; LCO 3.2.3, "QUADRANT POWER TILT (Tq)"; and LCO 3.2.4, "AXIAL SHAPE INDEX (ASI)," provide limits on control component operation and on monitored process variables to ensure the core operates within the linear heat rate (LCO 3.2.1, "Linear Heat Rate (LHR)") and FRT (LCO 3.2.2, "TOTAL RADIAL PEAKING FACTOR (FRT)) limits in the COLR.
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| Palisades Nuclear Plant B 3.1.6-1 Revised 01/29/2020
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| | |
| INSERT Bases 3.1.6 Regulating Rod Group Position Limits B 3.1.6 BASES BACKGROUND Operation within the LHR limits given in the COLR prevents power (continued) peaks that would exceed the Loss Of Coolant Accident (LOCA) limits derived by the Emergency Core Cooling System analysis. Operation within the FRT limits given in the COLR prevents Departure from Nucleate Boiling (DNB) during a loss of forced reactor coolant flow accident. In addition to the LHR and FRT limits, certain reactivity limits are preserved by regulating rod insertion limits. The regulating rod group insertion limits also restrict the ejected rod worth to the values assumed in the safety analysis and preserve the minimum required SDM in MODES 1 and 2.
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| The ejected rod case is limited to the reactivity worth for the highest worth rod ejected from the PDIL limit, thus limiting the maximum possible reactivity excursion.
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| The establishment of limiting safety system settings and LCOs requires that the expected long and short term behavior of the FRT be determined. The long term behavior relates to the variation of the steady state FRT with core burnup and is affected by the amount of rod insertion assumed, the portion of a burnup cycle over which such insertion is assumed, and the expected power level variation throughout the cycle. The short term behavior relates to transient perturbations to the steady state radial peaks, due to radial xenon redistribution. The magnitudes of such perturbations depend upon the expected use of the rods during anticipated power reductions and load maneuvering.
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| Analyses are performed, based on the expected mode of operation of the Nuclear Steam Supply System (base loaded, maneuvering, etc.).
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| The PDIL curve stated in the COLR dictates the acceptable regulating rod group positioning for anticipated power maneuvers and transient mitigation within the limits. The PDIL limitations stated in the COLR reflect the assumptions made in the safety analyses. This ensures that the FRT limits are not violated during power level maneuvering or transient mitigation.
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| The regulating rod group insertion and alignment limits are process variables that together characterize and control the three-dimensional power distribution of the reactor core. Additionally, the regulating rod group insertion limits control the reactivity that could be added in the event of a control rod ejection accident, and the shutdown and regulating bank insertion limits ensure the required SDM is maintained.
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| Palisades Nuclear Plant B 3.1.6-2 Revised 01/29/2020
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| INSERT Bases 3.1.6 Regulating Rod Group Position Limits B 3.1.6 BASES BACKGROUND Operation within the subject LCO limits will prevent fuel cladding (continued) failures that would breach the primary fission product barrier and release fission products to the reactor coolant in the event of a LOCA, loss of flow, ejected rod, or other accident requiring termination by a Reactor Protection System trip function.
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| APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation (Condition I) and anticipated operational occurrences (Condition II). The acceptance criteria for the regulating rod group position, ASI, and Tq LCOs are such as to preclude core power distributions from occurring that would violate the following fuel design criteria:
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| : a. During a large break LOCA, the peak cladding temperature must not exceed a limit of 2200°F, (Ref. 2);
| |
| : b. During a loss of forced reactor coolant flow accident, there must be at least a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition.
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| : c. During an ejected rod accident, the fission energy input to the fuel must not exceed 280 cal/gm (Ref. 3); and
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| : d. The rods must be capable of shutting down the reactor with a minimum required SDM, with the highest worth rod stuck fully withdrawn (Ref. 1).
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| Regulating rod group position, ASI, and Tq are process variables that together characterize and control the three-dimensional power distribution of the reactor core.
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| Fuel cladding damage does not occur when the core is operated outside these LCOs during normal operation. However, fuel cladding damage could result, should an accident occur with simultaneous violation of one or more of these LCOs. Changes in the power distribution can cause increased power peaking and corresponding increased local LHRs.
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| Palisades Nuclear Plant B 3.1.6-3 Revised 01/29/2020
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| INSERT Bases 3.1.6 Regulating Rod Group Position Limits B 3.1.6 BASES APPLICABLE The SDM requirement is ensured by limiting the regulating and SAFETY ANALYSES shutdown rod group insertion limits, so that the allowable inserted worth (continued) of the rods is such that sufficient reactivity is available to shut down the reactor to hot zero power. SDM assumes the maximum worth rod remains fully withdrawn upon trip (Ref. 4).
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| The most limiting SDM requirements for Mode 1 and 2 conditions at Beginning of Cycle (BOC) are determined by the requirements of several transients, e.g., Loss of Flow, etc. However, the most limiting SDM requirements for MODES 1 and 2 at End of Cycle (EOC) come from just one transient, Main Steam Line Break (MSLB). The requirements of the MSLB event at EOC for the full power and no load conditions are significantly larger than those of any other event at that time in cycle and, also, considerably larger than the most limiting requirements at BOC.
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| Although the most limiting SDM requirements at EOC are much larger than those at BOC, the available SDMs obtained via tripping the full-length control rods are substantially larger due to the much lower boron concentration at EOC. To verify that adequate SDMs are available throughout the cycle to satisfy the changing requirements, calculations are performed at both BOC and EOC. It has been determined that calculations at these two times in cycle are sufficient since the difference between available SDMs and the limiting SDM requirements are the smallest at these times in cycle. The measurement of full-length control rod bank worth performed as part of the Startup Testing Program demonstrates that the core has the expected shutdown capability. Consequently, adherence to LCO 3.1.5, Shutdown and Part-Length Rod Group Insertion Limits, and LCO 3.1.6 provides assurance that the available SDM at any time in cycle will exceed the limiting SDM requirements at that time in cycle.
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| Operation at the insertion limits or ASI limits may approach the maximum allowable linear heat generation rate or peaking factor, with the allowed Tq present. Operation at the insertion limit may also indicate the maximum ejected rod worth could be equal to the limiting value in fuel cycles that have sufficiently high ejected rod worth.
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| The regulating and shutdown rod insertion limits ensure that safety analyses assumptions for reactivity insertion rate, SDM, ejected rod worth, and peaking factors are preserved.
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| The regulating rod group position limits satisfy Criterion 2 of 10 CFR 50.36(c)(2).
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| Palisades Nuclear Plant B 3.1.6-4 Revised 01/29/2020
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| INSERT Bases 3.1.6 Regulating Rod Group Position Limits B 3.1.6 BASES LCO The limits on regulating rod group sequence, overlap, and physical insertion, as defined in the COLR, must be maintained because they serve the function of preserving power distribution, ensuring that the SDM is maintained, ensuring that ejected rod worth is maintained, and ensuring adequate negative reactivity insertion on trip. The overlap between regulating rod groups provides more uniform rates of reactivity insertion and withdrawal and is imposed to maintain acceptable power peaking during regulating rod group motion. For a control rod group to be considered above its insertion limit, all OPERABLE rods in that group, which are not misaligned, must be above the insertion limit (inoperable and misaligned rods are addressed by LCO 3.1.4).
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| The Power Dependent Insertion Limit (PDIL) alarm circuit is required to be OPERABLE for notification that the regulating rod groups are outside the required insertion limits. The Control Rod Out Of Sequence (CROOS) alarm circuit is required to be OPERABLE for notification that the rods are not within the required sequence and overlap limits. When the PDIL or the CROOS alarm circuit is inoperable, the verification of rod group positions is increased to ensure improper rod alignment is identified before unacceptable flux distribution occurs. The PDIL and CROOS alarms can be generated by either the synchro based Primary Indication Processor (PIP) node, or the reed switch based Secondary Position Indication (SPI) system since the SPI system, in conjunction with the host computer, is redundant to the PIP node in the task of control rod measurement, control rod monitoring and limit processing.
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| APPLICABILITY The regulating rod group sequence, overlap, and physical insertion limits shall be maintained with the reactor in MODES 1 and 2. These limits must be maintained, since they preserve the assumed power distribution, ejected rod worth, SDM, and reactivity rate insertion assumptions. Applicability in MODES 3, 4, and 5 is not required, since neither the power distribution nor ejected rod worth assumptions would be exceeded in these MODES. SDM is preserved in MODES 3, 4, and 5 by adjustments to the soluble boron concentration.
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| The Applicability has been modified by a Note indicating the LCO requirement is suspended while performing SR 3.1.4.3 (rod exercise test). Control rod exercising verifies the freedom of the rods to move, and requires the individual regulating rods to move below the LCO limits which could violate the LCO for their group.
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| Palisades Nuclear Plant B 3.1.6-5 Revised 01/29/2020
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| INSERT Bases 3.1.6 Regulating Rod Group Position Limits B 3.1.6 BASES ACTIONS LCOs 3.1.4, 3.1.5, and 3.1.6, and their ACTIONS were written to support each other. The combined intent is to assure the following:
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| : 1. There is adequate SDM available in withdrawn control rods to assure the reactor is shutdown by, and remains shutdown following, a reactor trip,
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| : 2. The control rod positioning does not cause unacceptable axial or radial flux peaking, and
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| : 3. The programmed rod withdrawal sequence and group overlap result in reactivity insertion rates within the assumptions of the Inadvertent Control Rod Bank Withdrawal Analyses.
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| The ACTIONS for rods that are mispositioned (misaligned or inserted beyond the limit) were written assuming that an OPERABLE rod discovered to be mispositioned would simply be re-positioned correctly.
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| While the associated Conditions would have to be entered, the rod could be re-positioned (thus exiting the LCO) without taking any other Required Action. A rod that remains mispositioned was assumed to be inoperable. The analyses account for operation with one (and only one) mispositioned rod (a dropped rod being the limiting case). With more than one mispositioned rod, the plant would be outside the bounds of the analyses and must be shutdown.
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| If a rod is discovered to be misaligned (ie, there is more than 8 inches between it and any other rod in its group, but all remaining rods in that group are within 8 inches of each other) Condition 3.1.4 C allows 2 hours to restore the rod alignment (thus exiting the LCO), perform SR 3 2.2.1 (verification that radial peaking is within limits), or reduce power to 75% RTP.
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| If one or more shutdown rods are inserted beyond the insertion limit, Condition 3.1.5 A is entered; the rods are declared inoperable and Condition 3.1.4 D (when one rod is immovable but trippable) or Condition 3.1.4 E (when a movable rod is inserted beyond its insertion limit, or when more than one rod is inoperable for any reason) must be entered.
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| If the rods can be moved, they should be withdrawn and all Conditions exited.
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| If one rod cannot be moved (but is still considered trippable),
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| operation may continue in accordance with Condition 3.1.4 D (and 3.1.4 C if it is misaligned).
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| Palisades Nuclear Plant B 3.1.6-6 Revised 01/29/2020
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| INSERT Bases 3.1.6 Regulating Rod Group Position Limits B 3.1.6 BASES ACTIONS If more than one rod cannot be moved, Condition 3.1.4 E must (continued) also be entered. The plant must be in MODE 3 in 6 hours in accordance with ACTION 3.1.4 E.1.
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| If one or more part length rods are inserted beyond the limit, Condition 3.1.5 A is entered; the rods are declared inoperable and Condition 3.1.4 E is entered (and 3.1.4 C if it is misaligned). Condition 3.1.4 D is not applicable to part-length rods since it only addresses full-length rods.
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| If the rods can be moved, they should be withdrawn and all Conditions exited.
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| If any part-length rods are inserted beyond the limit and cannot be moved, the plant must be placed in MODE 3 in 6 hours in accordance with ACTION 3.1.4 E.1.
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| If one or more OPERABLE regulating rods are inserted beyond the limit, Condition 3.1.6 A is entered; The rods must be restored to within limits (by rod withdrawal or power reduction) within two hours.
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| If a rod cannot be moved, it must be considered inoperable and Condition 3.1.4 D must be entered (and 3.1.4 C if it is misaligned).
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| Condition 3.1.4 D allows continued operation with one inoperable, but trippable, rod until the next reactor shutdown (MODE 3 entry). If more than one rod cannot be moved, Condition 3.1.4 E must be entered.
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| The plant must be in MODE 3 in 6 hours in accordance with ACTION 3.1.4 E.1.
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| The analyses do not account for the possibility of more then one rod failing to insert on a trip. While boron concentration might be adjusted to restore SHUTDOWN MARGIN, if two adjacent rods fail to insert that portion of the core could remain excessively reactive. Since the analyses must assume that one rod fails to insert, operation may not continue with a known untrippable rod. A shutdown would be required by Condition 3.1.4 E.
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| Palisades Nuclear Plant B 3.1.6-7 Revised 01/29/2020
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| INSERT Bases 3.1.6 Regulating Rod Group Position Limits B 3.1.6 BASES ACTIONS A.1 and A.2 (continued)
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| Operation beyond the insertion limit may result in a loss of SDM and excessive peaking factors. The insertion limit should not be violated during normal operation; this violation, however, may occur during transients when the operator is manually controlling the regulating rods in response to changing plant conditions.
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| When the regulating groups are inserted beyond the insertion limits, actions must be taken to either withdraw the regulating groups beyond the limits or to reduce THERMAL POWER to less than or equal to that allowed for the actual rod group position limit. Two hours provides a reasonable time to accomplish this, allowing the operator to deal with current plant conditions while limiting peaking factors to acceptable levels.
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| B.1 Operating outside the regulating rod group sequence and overlap limits specified in the COLR may result in excessive peaking factors. If the sequence and overlap limits are exceeded, the regulating rod groups must be restored to within the appropriate sequence and overlap. Two hours provides adequate time for the operator to restore the regulating rod group to within the appropriate sequence and overlap limits.
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| C.1 When the PDIL or the CROOS alarm circuit is inoperable, performing SR 3.1.6.1 once within 15 minutes following any rod motion ensures improper rod alignments are identified before unacceptable flux distributions occur.
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| D.1 When a Required Action cannot be completed within the required Completion Time, a controlled shutdown should be commenced. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.1.6-8 Revised 01/29/2020
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| INSERT Bases 3.1.6 Regulating Rod Group Position Limits B 3.1.6 BASES SURVEILLANCE SR 3.1.6.1 REQUIREMENTS With the PDIL alarm circuit OPERABLE, verification of each regulating rod group position is sufficient to detect rod positions that may approach the acceptable limits, and to provide the operator with time to undertake the Required Action(s) should the sequence or insertion limits be found to be exceeded.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.1.6.2 Demonstrating the PDIL alarm circuit OPERABLE verifies that the PDIL alarm circuit is functional. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.1.6.3 Demonstrating the CROOS alarm circuit OPERABLE verifies that the CROOS alarm circuit is functional. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Section 5.1
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| : 2. 10 CFR 50.46
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| : 3. FSAR, Section 14.16
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| : 4. FSAR, Section 14.4 Palisades Nuclear Plant B 3.1.6-9 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.1.7 STE B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Special Test Exceptions (STE)
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| BASES BACKGROUND The primary purpose of this STE is to permit relaxation of existing LCOs to allow the performance of certain PHYSICS TESTS. These tests are conducted to determine control rod worths, SHUTDOWN MARGIN (SDM), and specific reactor core characteristics.
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| Section XI of 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Processing Plants" (Ref. 1), requires that a test program be established to ensure that structures, systems, and components will perform satisfactorily in service. All functions necessary to ensure that specified design conditions are not exceeded during normal operation and anticipated operational occurrences must be tested. Testing is required as an integral part of the design, fabrication, construction, and operation of the power plant.
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| Requirements for notification of the NRC, for the purpose of conducting tests and experiments, are specified in 10 CFR 50.59, "Changes, tests, and experiments" (Ref. 2).
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| The key objectives of a test program are to (Ref. 3):
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| : a. Ensure that the facility has been adequately designed;
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| : b. Validate the analytical models used in design and analyses;
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| : c. Verify assumptions used for predicting plant response;
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| : d. Ensure that installation of equipment in the facility has been accomplished in accordance with design; and
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| : e. Verify that operating and emergency procedures are adequate.
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| To accomplish these objectives, testing is required during startup and low power operation after each shutdown that involved an alteration of the fuel assemblies in the reactor core. The PHYSICS TESTS requirements for reload fuel cycles ensure that the operating characteristics of the core are consistent with the design predictions, and that the core can be operated as designed.
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| Palisades Nuclear Plant B 3.1.7-1 Revised 01/29/2020
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| INSERT Bases 3.1.7 STE B 3.1.7 BASES BACKGROUND PHYSICS TESTS procedures are written and approved in accordance (continued) with the administrative processes for procedure controls. The procedures include all information necessary to permit a detailed execution of testing required to ensure that design intent is met.
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| PHYSICS TESTS are performed in accordance with these procedures and test results are approved prior to power escalation.
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| Examples of PHYSICS TESTS include determination of critical boron concentration, full-length control rod group and individual control rod worths, reactivity coefficients, flux symmetry, and core power distribution.
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| APPLICABLE It is acceptable to suspend certain LCOs for PHYSICS TESTS because SAFETY ANALYSES fuel damage criteria are not exceeded. Even if an accident occurs during a PHYSICS TEST with one or more LCOs suspended, fuel damage criteria are preserved because the limits on power distribution and shutdown capability are maintained during PHYSICS TESTS.
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| Requirements for reload fuel cycle PHYSICS TESTS are defined in ANSI/ANS-19.6.1-2005 (Ref. 4). Although these PHYSICS TESTS are generally accomplished within the limits of all LCOs, conditions may occur when one or more LCOs must be suspended to make completion of PHYSICS TESTS possible or practical. This is acceptable as long as the fuel design criteria are not violated. As long as the Linear Heat Rate (LHR) remains within its limit, fuel design criteria are preserved.
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| In this test, the following LCOs are suspended:
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| : a. LCO 3.1.4, "Control Rod Alignment";
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| : b. LCO 3.1.5, "Shutdown and Part-Length Rod Group Insertion Limits;
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| : c. LCO 3.1.6, "Regulating Rod Group Position Limits; and
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| : d. LCO 3.4.2, "PCS Minimum Temperature for Criticality.
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| This STE places limits on allowable THERMAL POWER during PHYSICS TESTS assuring the LHR and the Departure from Nucleate Boiling (DNB) parameters will be maintained within limits. It also places limits on the amount of control rod worth required to be available for reactivity control when control rod worth measurements are performed.
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| Palisades Nuclear Plant B 3.1.7-2 Revised 01/29/2020
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| INSERT Bases 3.1.7 STE B 3.1.7 BASES APPLICABLE SRs are conducted as necessary to ensure that reactor power and SAFETY ANALYSES shutdown capability remain within limits during PHYSICS TESTS.
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| (continued) Requiring 1% shutdown reactivity, based on predicted control rod worths, be available for trip insertion from the OPERABLE full-length control rod provides a high degree of assurance that shutdown capability is maintained for the most challenging postulated accident assuming all full-length control rods are inserted in the core. Since LCOs 3.1.5 and 3.1.6 are suspended, however, there is not the same degree of assurance during this test that the reactor would always be shut down if the highest worth full-length control rod was stuck out and calculational uncertainties or the estimated highest rod worth was not as expected (the single failure criterion is not met). This situation is judged acceptable, however, because specified acceptable fuel damage limits are still met. The risk of experiencing a stuck rod and subsequent criticality is reduced during this PHYSICS TEST exception by the requirement that 1% shutdown reactivity is available based on predicted control rod worths.
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| PHYSICS TESTS include measurement of core parameters or exercise of control components. Also involved are the shutdown and regulating rods, which affect power peaking and are required for shutdown of the reactor.
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| The limits for insertion of these rod groups are specified for each fuel cycle in the COLR.
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| As described in LCO 3.0.7, compliance with Special Test Exceptions LCOs is optional, and therefore no criteria of 10 CFR 50.36(c)(2) apply. Special Test Exception LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.
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| LCO This LCO relaxes the minimum primary coolant temperature at which the reactor may be made critical, permits individual full-length control rods and full-length control rod groups to be positioned outside of their normal alignment and insertion limits during the performance of PHYSICS TESTS such as those required to:
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| : a. Measure control rod worths;
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| : b. Measure control rod shadowing factors; and
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| : c. Measure temperature and power coefficients.
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| Palisades Nuclear Plant B 3.1.7-3 Revised 01/29/2020
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| INSERT Bases 3.1.7 STE B 3.1.7 BASES LCO This LCO specifies that a minimum amount of rod worth is immediately (continued) available for reactivity control when rod worth measurement tests are performed. This portion of the STE permits the periodic verification of the actual versus predicted control rod group worths.
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| The requirements of LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 may be suspended during the performance of PHYSICS TESTS, provided:
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| : a. THERMAL POWER is 2% RTP;
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| : b. 1% shutdown reactivity, based on predicted control rod worth, is available for trip insertion; and
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| : c. Tave is 500°F.
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| APPLICABILITY This LCO is applicable in MODE 2 because the reactor must be critical to perform the PHYSICS TESTS described in the LCO section.
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| ACTIONS A.1 If THERMAL POWER exceeds 2% RTP, THERMAL POWER must be reduced to restore the additional thermal margin provided by the reduction.
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| The 15 minute Completion Time ensures that prompt action shall be taken to reduce THERMAL POWER to within acceptable limits.
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| B.1 If < 1% shutdown reactivity is available for trip insertion, boration must be initiated promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components.
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| The operator should begin boration with the best source available for the plant conditions. Boration will be continued until 1% shutdown reactivity is achieved.
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| C.1 If the Tave requirement is not met, Tave must be restored. The 15 minutes Completion Time ensures that prompt action shall be taken to raise Tave within the required limit.
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| Palisades Nuclear Plant B 3.1.7-4 Revised 01/29/2020
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| INSERT Bases 3.1.7 STE B 3.1.7 BASES ACTIONS D.1 (continued)
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| If Required Actions of Condition A, Condition B, or Condition C cannot be completed within the required Completion Time, PHYSICS TESTS must be suspended within 1 hour. Allowing 1 hour for suspending PHYSICS TESTS allows the operator sufficient time to change any abnormal rod configuration back to within the limits of LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6, or to restore Primary Coolant System (PCS) temperature to within the limits of LCO 3.4.2.
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| SURVEILLANCE SR 3.1.7.1 REQUIREMENTS Verifying that THERMAL POWER is 2% RTP as specified in the PHYSICS TEST procedure and required by the safety analysis, ensures that adequate LHR and DNB parameter margins are maintained while LCOs are suspended. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.1.7.2 Verifying Tave 500°F during the PHYSICS TEST ensures that Tave remains in an analyzed range while the LCOs are suspended. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.1.7.3 Verification that 1% shutdown reactivity is available for trip insertion is performed by a reactivity balance calculation, considering the following reactivity effects:
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| : a. PCS boron concentration;
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| : b. Control rod group position;
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| : c. PCS average temperature;
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| : d. Fuel burnup based on gross thermal energy generation;
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| : e. Xenon concentration; and
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| : f. Isothermal Temperature Coefficient (ITC).
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| Palisades Nuclear Plant B 3.1.7-5 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.1.7 STE B 3.1.7 BASES SURVEILLANCE SR 3.1.7.3 REQUIREMENTS (continued) Using the ITC accounts for Doppler reactivity in this calculation because reactor power is maintained below 2% RTP, and for most of the PHYSIC TESTS below the point of adding heat the fuel temperature will be changing at the same rate as the PCS.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. 10 CFR 50, Appendix B, Section XI
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| : 2. 10 CFR 50.59
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| : 3. Regulatory Guide 1.68, Revision 2, August 1978
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| : 4. ANSI/ANS-19.6.1-2005, November 29, 2005 Palisades Nuclear Plant B 3.1.7-6 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.2.1 LHR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Linear Heat Rate (LHR)
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| BASES BACKGROUND The purpose of this LCO is to limit the core power distribution to the initial values assumed in the accident analyses. Operation within the limits imposed by this LCO either limits or prevents potential fuel cladding failures that could breach the primary fission product barrier and release fission products to the primary coolant in the event of a Loss Of Coolant Accident (LOCA), loss of flow accident, ejected control rod accident, or other postulated accident requiring termination by a Reactor Protection System trip function. This LCO limits the amount of damage to the fuel cladding during an accident by ensuring that the plant is operating within acceptable bounding conditions at the onset of a transient.
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| Methods of controlling the power distribution include:
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| : a. Using control rods to alter the axial power distribution;
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| : b. Decreasing control rod insertion by boration, thereby improving the radial power distribution; and
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| : c. Correcting off optimum conditions (e.g., a control rod drop or misoperation of the plant) that cause margin degradations.
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| The core power distribution is controlled so that, in conjunction with other core operating parameters (e.g., control rod insertion and alignment limits), the power distribution satisfies this LCO. The limiting safety system settings and this LCO are based on the accident analyses (Refs. 1 and 2), so that specified acceptable fuel design limits are not exceeded as a result of Anticipated Operational Occurrences (AOOs), and the limits of acceptable consequences are not exceeded for other postulated accidents.
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| Limiting power distribution changes over time also minimizes the xenon distribution changes, which is a significant factor in controlling the axial power distribution.
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| Power distribution is a product of multiple parameters, various combinations of which may produce acceptable power distributions.
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| Palisades Nuclear Plant B 3.2.1-1 Revised 01/29/2020
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| INSERT Bases 3.2.1 LHR B 3.2.1 BASES BACKGROUND The limits on LHR, TOTAL RADIAL PEAKING FACTOR (FRT),
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| (continued) QUADRANT POWER TILT (Tq), and AXIAL SHAPE INDEX (ASI),
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| which are obtained directly from the core reload analysis, ensure compliance with the safety limits on LHR and Departure from Nucleate Boiling Ratio (DNBR).
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| Either of the two core power distribution monitoring systems, the Incore Alarm portion of the Incore Monitoring System or the Excore Monitoring System, provides adequate monitoring of the core power distribution and is capable of verifying that the LHR is within its limits. The Incore Alarm System performs this function by continuously monitoring the local power at many points throughout the core and comparing the measurements to predetermined setpoints above which the limit on LHR could be exceeded. The Excore Monitoring System performs this function by providing comparison of the measured core ASI with predetermined ASI limits based on incore measurements. An Excore Monitoring System Allowable Power Level (APL), which may be less than RATED THERMAL POWER, and an additional restriction on Tq, are applied when using the Excore Monitoring System to ensure that the ASI limits adequately restrict the LHR to less than the limiting values.
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| In conjunction with the use of the Excore Monitoring System for monitoring LHR and in establishing ASI limits, the following assumptions are made:
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| : a. The control rod insertion limits of LCO 3.1.5, "Shutdown and Part-Length Rod Group Insertion Limits," and LCO 3.1.6, "Regulating Rod Group Position Limits," are satisfied;
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| : b. The additional Tq restriction of SR 3.2.1.6 is satisfied; and
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| : c. FRT, does not exceed the limits of LCO 3.2.2.
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| The limitations on the TOTAL RADIAL PEAKING FACTOR provided in the COLR ensure that the assumptions used in the analysis for establishing the LHR limits and Limiting Safety System Settings (LSSS) remain valid during operation at the various allowable control rod group insertion limits.
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| Palisades Nuclear Plant B 3.2.1-2 Revised 01/29/2020
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| INSERT Bases 3.2.1 LHR B 3.2.1 BASES BACKGROUND The Incore Monitoring System continuously provides a direct indication (continued) of the core power distribution. It also provides alarms that have been established for the individual incore detector segments, ensuring that the peak LHRs are maintained within the limits specified in the COLR.
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| The setpoints for these alarms include tolerances, set in conservative directions, for:
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| : a. A measurement calculational uncertainty factor (as identified in the COLR);
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| : b. An engineering uncertainty factor of 1.03; and
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| : c. A THERMAL POWER measurement uncertainty factor of 1.006 of 2565.4 MWt.
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| The measurement uncertainties associated with LHR and FRT are based on a statistical analysis performed on power distribution benchmarking results. The COLR includes the applicable measurement uncertainties for incore detector usage. The engineering and THERMAL POWER uncertainties are incorporated in the power distribution calculation performed by the fuel vendor.
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| The excore power distribution monitoring system consists of Power Range Channels 5 through 8. The power range channels monitor neutron flux from 0 to 125 percent full power. They are arranged symmetrically around the reactor core to provide information on the radial and axial flux distributions.
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| The power range detector assembly consists of two uncompensated ion chambers for each channel. One detector extends axially along the lower half of the core while the other, which is located directly above it, monitors flux from the upper half of the core. The DC current signal from each of the ion chambers is fed directly to the control room drawer assembly without pre-amplification. Each excore detector supplies data to a Thermal Margin Monitor (TMM). Each TMM uses these excore signals to calculate Axial Shape Index (ASI) on a continuous basis.
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| ASI can be defined as the compensated ratio of power developed in the upper and lower sections of the core. The TMM takes the excore detector signals and develops a power ratio (YE) that describes the distribution of neutron flux developed in the core by the formula:
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| YE = (L - U)/(L + U)
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| Where L is the lower excore segment flux, and U is the upper excore segment flux.
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| Palisades Nuclear Plant B 3.2.1-3 Revised 01/29/2020
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| INSERT Bases 3.2.1 LHR B 3.2.1 BASES BACKGROUND The excore detectors which are located within the concrete biological (continued) shield of the reactor must be compensated for the phenomenon of shape annealing. Shape annealing factors are developed to correct the excore readings for neutron attenuation from the core periphery to the excore detector locations. This accounts for any material that would cause neutron attenuation within the detector path such as: concrete, structural steel and so forth. This allows the excore detectors to represent an accurate measurement of the core power distribution.
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| Shape annealing has been found to be a linear relationship which can be correlated to the Axial Offset (AO) as determined by an Incore Detector System to the raw readings seen by the excore detectors.
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| Reactor Engineering has developed shape annealing factors for each individual Excore detector. The TMM uses the above calculated power ratio and the appropriate shape annealing factor to determine the ASI value for an individual excore detector channel.
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| APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation (Condition 1) or AOOs (Condition 2) (Ref. 3). The power distribution and control rod insertion and alignment LCOs preclude core power distributions that violate the following fuel design criteria:
| |
| : a. During a LOCA, peak cladding temperature must not exceed 2200°F (Ref. 4);
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| : b. During a loss of flow accident, there must be at least 95%
| |
| probability at the 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition (Ref. 3).
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| : c. During an ejected rod accident, the fission energy input to the fuel must not exceed 280 cal/gm; and
| |
| : d. The full-length control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3).
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| The power density at any point in the core must be limited to maintain the fuel design criteria (Ref. 4). This is accomplished by maintaining the power distribution and primary coolant conditions so that the peak LHR and DNB parameters are within operating limits supported by accident analyses (Ref. 1), with due regard for the correlations between measured quantities, the power distribution, and uncertainties in determining the power distribution.
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| Palisades Nuclear Plant B 3.2.1-4 Revised 01/29/2020
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| INSERT Bases 3.2.1 LHR B 3.2.1 BASES APPLICABLE Fuel cladding failure during a LOCA is limited by restricting the SAFETY ANALYSES maximum linear heat generation rate so that the peak cladding (continued) temperature does not exceed 2200°F (Ref. 4). High peak cladding temperatures are assumed to cause severe cladding failure by oxidation due to a Zircaloy water reaction.
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| The LCOs governing LHR, ASI, and the Primary Coolant System Operation ensure that these criteria are met as long as the core is operated within the LHR, ASI, FRT, and Tq limits. The latter are process variables that characterize the three dimensional power distribution of the reactor core. Operation within the limits for these variables ensures that their actual values are within the ranges used in the accident analyses.
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| Fuel cladding damage does not necessarily occur while the plant is operating at conditions outside the limits of these LCOs during normal operation. Fuel cladding damage could result, however, if an accident occurs from initial conditions outside the limits of these LCOs. The potential for fuel cladding damage exists because changes in the power distribution can cause increased power peaking and can correspondingly increase local LHR.
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| The Incore Monitoring System provides for monitoring of LHR, FRT, and QUADRANT POWER TILT to ensure that fuel design conditions and safety analysis assumptions are maintained. The Incore Monitoring System is also utilized to determine the target AXIAL OFFSET (AO) and to determine the Allowable Power Level (APL) when using the excore detectors.
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| The Excore Monitoring System provides for monitoring of ASI and QUADRANT POWER TILT to ensure that fuel design conditions and safety analysis assumptions are maintained.
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| LHR satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| LCO The power distribution LCO limits are based on correlations between power peaking and certain measured variables used as inputs to the LHR and DNBR operating limits. The power distribution LCO limits, except Tq, are provided in the COLR. The limitation on the LHR in the peak power fuel rod at the peak power elevation Z ensures that, in the event of a LOCA, the peak temperature of the fuel cladding does not exceed 2200°F.
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| Palisades Nuclear Plant B 3.2.1-5 Revised 01/29/2020
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| INSERT Bases 3.2.1 LHR B 3.2.1 BASES LCO The LCO requires that LHR be maintained within the limits specified in (continued) the COLR and either the Incore Alarm System or Excore Monitoring System be OPERABLE to monitor LHR. When using the Incore Alarm System, the LHR is not considered to be out of limits until there are four or more incore detectors simultaneously in alarm. When using the Excore Monitoring System, LHR is considered within limits when the conditions are acceptable for use of the Excore Monitoring System and the associated ASI and Tq limits specified in the SRs are met.
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| To be considered OPERABLE, the Incore Alarm System must have at least 90 of the 180 incore detectors OPERABLE and 2 incore detectors per axial level per core quadrant OPERABLE. In addition, the plant process computer must be OPERABLE and the required alarm setpoints entered into the plant computer. Only 36 of the 45 instrument locations are included in the Incore Alarm System Uncertainty Analysis (180 of the possible 215 detectors). Instrument locations 1, 4, 13, 34, 41, 42 and 45 are not included, and instrument locations 7 and 44 are used by the Reactor Vessel Level Monitoring System (RVLMS).
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| To be considered OPERABLE, the Excore Monitoring System must have been calibrated with OPERABLE incore detectors, the ASI must not have been out of limits for the last 24 hours, and THERMAL POWER must be less than the APL.
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| APPLICABILITY In MODE 1 with THERMAL POWER > 25% RTP, power distribution must be maintained within the limits assumed in the accident analysis to ensure that fuel damage does not result following an AOO. In MODE 1 with THERMAL POWER 25% RTP, and in other MODES, this LCO does not apply because there is not sufficient THERMAL POWER to require a limit on the core power distribution, and because ample thermal margin exists to ensure that the fuel integrity is not jeopardized and safety analysis assumptions remain valid.
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| ACTIONS A.1 There are three acceptable methods for verifying that LHR is within limits. The LCO requires monitoring by either an OPERABLE Incore Alarm System or an OPERABLE Excore Monitoring System. When both of the required systems are inoperable, Condition B allows for monitoring by taking manual readings of the incore detectors. Any of these three methods may indicate that the LHR is not within limits. With the LHR exceeding its limit, excessive fuel damage could occur following an accident. In this Condition, prompt action must be taken to restore the LHR to within the specified limits. One hour to restore the LHR to within its specified limits is reasonable and ensures that the core does not continue to operate in this Condition. The 1 hour Palisades Nuclear Plant B 3.2.1-6 Revised 01/29/2020
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| INSERT Bases 3.2.1 LHR B 3.2.1 BASES ACTIONS A.1 (Continued)
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| Completion Time also allows the operator sufficient time for evaluating core conditions and for initiating proper corrective actions.
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| ACTIONS B.1 and B.2 (continued)
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| With the Incore Alarm System inoperable for monitoring LHR and the Excore Monitoring System inoperable for monitoring LHR, THERMAL POWER must be reduced to 85% RTP within 2 hours. Operation at 85% RTP ensures that ample thermal margin is maintained. A 2 hour Completion Time is adequate to achieve the required plant condition without challenging plant systems. Additionally, with the Incore Alarm and Excore Monitoring Systems inoperable, LHR must be verified to be within limits within 4 hours, and every 2 hours thereafter by manually collecting incore detector readings at the terminal blocks in the control room utilizing a suitable signal detector. The manual readings shall be taken on a minimum of 10 individual detectors per quadrant (to include a total of 90 detectors in a 10 hour period). The time interval of 2 hours and the minimum of 10 detectors per quadrant are sufficient to maintain adequate surveillance of the power distribution to detect significant changes until the monitoring systems are returned to service.
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| As stated in SR 3.0.2, the 25% extension allowed by SR 3.0.2 may be applied to Required Actions whose Completion Time is stated as once per . . . however, the 25% extension does not apply to the initial performance of a Required Action with a periodic Completion Time that requires performance on a "once per. . ." basis. The 25% extension applies to each performance of the Required Action after the initial performance. Therefore, while Required Action 3.2.1 B.2 must be initially performed within 4 hours without any SR 3.0.2 extension, subsequent performances at the Once per 2 hours interval may utilize the 25% SR 3.0.2 extension.
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| C.1 If the Required Action and associated Completion Time are not met, THERMAL POWER must be reduced to 25% RTP. This reduced power level ensures that the core is operating within its thermal limits and places the core in a conservative condition. The allowed Completion Time of 4 hours is reasonable, based on operating experience, to reach 25% RPT from full power MODE 1 conditions in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.2.1-7 Revised 01/29/2020
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| INSERT Bases 3.2.1 LHR B 3.2.1 BASES SURVEILLANCE SR 3.2.1.1 REQUIREMENTS The Incore Alarm portion of the Incore Monitoring System provides continuous monitoring of LHR through the plant computer. The PIDAL computer program is used to generate alarm setpoints for the plant computer that are based on measured margin to allowed LHR. As the incore detectors are read by the plant computer, they are continuously compared to the alarm setpoints. If the Incore Alarm System LHR monitoring function is inoperable, excore detectors or manual recordings of the incore detector readings may be used to monitor LHR.
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| Periodically monitoring LHR ensures that the assumptions made in the Safety Analysis are maintained. This SR is modified by a Note that states that the SR is only required to be met when the Incore Alarm System is being used to monitor LHR. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.2.1.2 Continuous monitoring of the LHR is provided by the Incore Alarm System which provides adequate monitoring of the core power distribution and is capable of verifying that the LHR does not exceed its specified limits.
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| Performance of this SR verifies the Incore Alarm System can accurately monitor LHR by ensuring the alarm setpoints are based on a measured power distribution. Therefore, they are only applicable when the Incore Alarm System is being used to determine the LHR.
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| The alarm setpoints must be initially adjusted following each fuel loading prior to operation above 50% RTP, and periodically adjusted thereafter. The SR is modified by a Note which requires the SR to be met only when the Incore Alarm System is being used to determine LHR. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.2.1-8 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.2.1 LHR B 3.2.1 BASES SURVEILLANCE SR 3.2.1.3 REQUIREMENTS (continued) SR 3.2.1.3 requires, prior to initial use of the excore LHR monitoring function, verification that the absolute difference of the measured ASI and the target ASI has been 0.05 for each OPERABLE channel for the last 24 hours using the previous 24 hourly recorded values.
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| Performance of this SR verifies that plant conditions are acceptable for the Excore Monitoring System to accurately monitor the LHR (Ref. 5).
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| The prior to initial use verification identifies that there have been no significant power distribution anomalies while using other monitoring methods, e.g., the incore detectors, which may affected the ability of the excore detectors to monitor LHR.
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| The SR is modified by a Note that states that the SR is only required to be met when the Excore Monitoring System is being used to monitor LHR. Failure of this SR prevents the Excore Monitoring System from being considered OPERABLE for monitoring of LHR.
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| SR 3.2.1.4 SR 3.2.1.4 requires verification that THERMAL POWER is less than or equal to the Allowable Power Level (APL) which is limited to not more than 10% greater than the THERMAL POWER at which the APL was last determined. Performance of this SR also verifies that plant conditions are acceptable for the Excore Monitoring System to accurately monitor the LHR (Ref. 5). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| The SR is modified by a Note that states that the SR is only required to be met when the Excore Monitoring System is being used to monitor LHR. Failure of this SR prevents the Excore Monitoring System from being considered OPERABLE for monitoring of LHR.
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| Palisades Nuclear Plant B 3.2.1-9 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.2.1 LHR B 3.2.1 BASES SURVEILLANCE SR 3.2.1.5 REQUIREMENTS (continued) SR 3.2.1.5 requires verification that the absolute difference of the measured ASI and the target ASI is 0.05 every hour. This must be verified on at least 3 of the 4, 2 of the 3, or 2 of the 2 OPERABLE channels, whichever is the applicable case. However, any otherwise OPERABLE channel which indicates an absolute difference of > 0.05 must be considered out of limits. Performance of this SR verifies that plant conditions are acceptable for the Excore Monitoring System to be used to assure LHR is within limits (Ref. 5). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| The SR is modified by a Note that states that the SR is only required to be met when the Excore Monitoring System is being used to monitor LHR. Failure of this SR (when using an OPERABLE Excore Monitoring System) is a failure to verify that LHR is within limits and is therefore considered a failure to meet the LCO due to LHR not within limits as determined by the Excore Monitoring System.
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| SR 3.2.1.6 SR 3.2.1.6 requires verification that the QUADRANT POWER TILT is 0.03. Performance of this SR also verifies that plant conditions are acceptable for the Excore Monitoring System to be used to assure LHR is within limits (Ref. 5). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| The SR is modified by a Note that states that the SR is only required to be met when the Excore Monitoring System is being used to monitor LHR. Failure of this SR (when using an OPERABLE Excore Monitoring System) is a failure to verify that LHR is within limits and is therefore considered a failure to meet the LCO due to LHR not within limits as determined by the Excore Monitoring System.
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| Palisades Nuclear Plant B 3.2.1-10 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.2.1 LHR B 3.2.1 BASES REFERENCES 1. FSAR, Chapter 14
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| : 2. FSAR, Chapter 6
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| : 3. FSAR, Section 5.1
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| : 4. 10 CFR 50.46
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| : 5. Safety Evaluation Report for Palisades Nuclear Plant Operating License Amendment No. 68, Section 4, dated December 8, 1981 Palisades Nuclear Plant B 3.2.1-11 Revised 01/29/2020
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| INSERT Bases 3.2.2 Radial Peaking B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 TOTAL RADIAL PEAKING FACTOR (FRT)
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| BASES BACKGROUND The Background section of Bases B 3.2.1, Linear Hear Rate, is applicable to these Bases.
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| APPLICABLE The Applicable Safety Analyses section of Bases B 3.2.1 is applicable SAFETY ANALYSES to these Bases.
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| The TOTAL RADIAL PEAKING FACTOR satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| LCO The power distribution LCO limits are based on correlations between power peaking and certain measured variables used as inputs to the LHR and DNBR operating limits. The power distribution LCO limits, except Tq, are provided in the COLR.
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| The limitations on FRT are provided to ensure that assumptions used in the analysis for establishing DNB margin, LHR limit and the thermal margin/low pressure and variable high power trip setpoints remain valid during operation. Data from the incore detectors are used for determining the measured FRT.
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| APPLICABILITY In MODE 1 with THERMAL POWER > 25% RTP, power distribution must be maintained within the limits assumed in the accident analyses to ensure that fuel damage does not result following an AOO. In MODE 1 with THERMAL POWER 25% RTP, and in other MODES, this LCO does not apply because there is not sufficient THERMAL POWER to require a limit on the core power distribution, and because ample thermal margin exists to ensure that the fuel integrity is not jeopardized and safety analysis assumptions remain valid.
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| Palisades Nuclear Plant B 3.2.2-1 Revised 01/29/2020
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| INSERT Bases 3.2.2 Radial Peaking B 3.2.2 BASES ACTIONS A.1 If FRT exceeds its limits, FRT must be restored to within the limits as identified in the COLR within 6 hours. Restoration may be either by correcting the source of the peaking or by a reduction in THERMAL POWER. The THERMAL POWER typically necessary to achieve restoration is identified by the equation:
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| P = [1-3.33 ((FR/FL)-1)] (RTP)
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| Where FR is the measured value of FRT; and FL is the corresponding limit provided in the COLR. Operating at or below this power level, P, is typically sufficient to restore FRT within limits. If power reductions do not restore FRT to within limits within 6 hours, Condition B is applicable.
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| Six hours to restore FRT to within limit(s) is reasonable and ensures that the core does not continue to operate in this condition for an extended period. The 6 hour Completion Time also allows the operator sufficient time for evaluating core conditions and for initiating proper corrective actions.
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| B.1 If the Required Action and associated Completion Time are not met, THERMAL POWER must be reduced to 25% RTP. This reduced power level ensures that the core is operating within its thermal limits and places the core in a conservative condition. The allowed Completion Time of 4 hours is reasonable, based on operating experience, to reach 25% RTP from full power conditions in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.2.2-2 Revised 01/29/2020
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| INSERT Bases 3.2.2 Radial Peaking B 3.2.2 BASES SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The periodic Surveillance to determine FRT ensures that FRT remains within the range assumed in the analysis throughout the fuel cycle.
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| Determining FRT using the incore detectors after each fuel loading prior to the reactor exceeding 50% RTP ensures that the core is properly loaded.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES None Palisades Nuclear Plant B 3.2.2-3 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.2.3 Tq B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 QUADRANT POWER TILT (Tq)
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| BASES BACKGROUND The Background section for Bases B 3.2.1, Linear Heat Rate, is applicable to these Bases, with the following addition:
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| The power range monitoring system provides alarms when Tq exceeds predetermined values. The average of the four power range signals is developed by a single Comparitor Averager. Each power range channel compares its output signal to this average signal. Two channel deviation alarm bistables, set at different levels, are provided in each power range channel. The deviation alarms will annunciate when the associated channel signal is either above or below the average, however, only a signal above the average is of concern with regard to Tq.
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| APPLICABLE The Applicable Safety Analyses section of Bases B 3.2.1 is applicable SAFETY ANALYSES to these Bases.
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| The Tq satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| LCO The power distribution LCO limits are based on correlations between power peaking and the measured variables used as inputs to the LHR and DNBR operating limits. The power distribution LCO limits, except Tq, are provided in the COLR. The limits on Tq ensure that assumptions used in the analysis for establishing LHR limits and DNB margin remain valid during operation.
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| APPLICABILITY In MODE 1 with THERMAL POWER > 25% RTP, power distribution must be maintained within the limits assumed in accident analysis to ensure that fuel damage does not result following an AOO. In MODE 1 with THERMAL POWER 25% RTP, and in other MODES, this LCO does not apply because there is not sufficient THERMAL POWER to require a limit on core power distribution, and because ample thermal margin exists to ensure that the fuel integrity is not jeopardized and safety analysis assumptions remain valid.
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| Palisades Nuclear Plant B 3.2.3-1 Revised 01/29/2020
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| INSERT Bases 3.2.3 Tq B 3.2.3 BASES ACTIONS A.1 If the measured Tq is > 0.05, Tq must be restored within 2 hours or FRT must be determined to be within the limits of LCO 3.2.2, and determined to be within these limits every 8 hours thereafter, as long as Tq is out of limits. Two hours is sufficient time to allow the operator to reposition control rods, and significant radial xenon redistribution cannot occur within this time. The 8 hour Completion Time ensures changes in FRT can be identified before the limits of LCO 3.2.2 are exceeded.
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| As stated in SR 3.0.2, the 25% extension allowed by SR 3.0.2 may be applied to Required Actions whose Completion Time is stated as once per . . . however, the 25% extension does not apply to the initial performance of a Required Action with a periodic Completion Time that requires performance on a "once per. . ." basis. The 25% extension applies to each performance of the Required Action after the initial performance. Therefore, while Required Action 3.2.3 A.1 must be initially performed within 2 hours without any SR 3.0.2 extension, subsequent performances at the Once per 8 hours interval may utilize the 25% SR 3.0.2 extension.
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| B.1 With the measured Tq > 0.10, power must be reduced to < 50% RTP within 4 hours, FRT must be within specified limits to ensure that acceptable flux peaking factors are maintained as required by Condition A (which continues to be applicable). Based on operating experience, 4 hours is sufficient time for evaluation of these factors. If FRT is within limits, operation may proceed while attempts are made to restore Tq to within its limit. If the tilt is generated due to a control rod misalignment, continued operation at < 50% RTP allows for realignment; if the cause is other than control rod misalignment, continued operation may be necessary to discover the cause of the tilt. Reducing THERMAL POWER to < 50% RTP, and the more frequent measurement of FRT required by ACTION A.1, provide conservative protection from potential increased peaking due to xenon redistribution.
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| Palisades Nuclear Plant B 3.2.3-2 Revised 01/29/2020
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| INSERT Bases 3.2.3 Tq B 3.2.3 BASES ACTIONS C.1 (continued)
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| If Tq is > 0.15, or if Required Actions and associated Completion Times are not met, THERMAL POWER must be reduced to 25% RTP. This requirement ensures that the core is operating within its thermal limits and places the core in a conservative condition. Four hours is a reasonable time to reach 25% RTP in an orderly manner and without challenging plant systems.
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| SURVEILLANCE SR 3.2.3.1 REQUIREMENTS QUADRANT POWER TILT (Tq) is determined from excore detector readings which are calibrated using incore detector measurements (Ref. 1). Calibration factors are determined using incore measurements and an incore analysis computer program (Ref. 2). Each power range channel provides alarms if Tq exceeds its limits. Therefore, with all power range channels OPERABLE, this SR only requires verification that the channel deviation alarms do not indicate an excessive Tq. If the Excore Monitoring System Tq deviation alarm monitoring function is inoperable, excore detector readings or symmetric incore detector readings may be used to monitor Tq at 12 hour intervals. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Section 7.6.2.2
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| : 2. FSAR, Section 7.6.2.4 Palisades Nuclear Plant B 3.2.3-3 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.2.4 ASI B 3.2.4 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 AXIAL SHAPE INDEX (ASI)
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| BASES BACKGROUND The Background section for Bases B 3.2.1, Linear Heat Rate, is applicable to these Bases, with the following addition:
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| The Excore Monitoring System ASI alarm function consists of four channels. At least two channels of the ASI alarm function are necessary to verify that ASI is within limits. With one or more excore monitoring channels measured ASI differing from the incore measured AO by > 0.02 under steady state operating conditions, the ASI monitoring channel alarm setpoint may be adjusted to compensate for this deviation. This ensures that fuel design parameters can continue to be accurately monitored and not exceeded when the incore/excore alignment is not within normal tolerances. This may occur when the calibration cannot be performed or the alignment problem exists after the calibration.
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| APPLICABLE The Applicable Safety Analyses section for Bases B 3.2.1 is applicable SAFETY ANALYSES to these Bases.
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| The ASI satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| LCO The power distribution LCO limits are based on correlations between power peaking and certain measured variables used as inputs to the LHR and DNBR operating limits. These power distribution LCO limits, except Tq, are provided in the COLR. The limitation on ASI ensures that the assumed axial power profiles used in the development of the inlet temperature LCO bound the measured axial power profile.
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| The limitation on ASI, along with the limitations of LCO 3.3.1, "Reactor Protective System Instrumentation," represents a conservative envelope of operating conditions consistent with the assumptions that have been analytically demonstrated adequate for maintaining an acceptable minimum DNBR throughout all AOOs. Operation of the core with conditions within the specified limits ensures that an acceptable minimum margin from DNB conditions is maintained in the event of any AOO.
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| Palisades Nuclear Plant B 3.2.4-1 Revised 01/29/2020
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| INSERT Bases 3.2.4 ASI B 3.2.4 BASES APPLICABILITY In MODE 1 with THERMAL POWER > 25% RTP, power distribution must be maintained within the limits assumed in the accident analyses to ensure that fuel damage does not result following an AOO. In MODE 1 with THERMAL POWER 25% RTP, and in other MODES, this LCO does not apply because there is not sufficient THERMAL POWER to require a limit on the core power distribution, and because ample thermal margin exists to ensure that the fuel integrity is not jeopardized and safety analysis assumptions remain valid.
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| ACTIONS A.1 Operating the core within ASI limits specified in the COLR and within the limits of LCO 3.3.1 ensures an acceptable margin for DNB and for maintaining local power density in the event of an AOO. Maintaining ASI within limits also ensures that the limits of 10 CFR 50.46 are not exceeded during accidents. The Required Actions to restore ASI must be completed within 2 hours to limit the duration the plant is operated outside the initial conditions assumed in the accident analyses. In addition, this Completion Time is sufficiently short that the xenon distribution in the core cannot change significantly.
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| B.1 If the Required Action and associated Completion Time are not met, core power must be reduced. Reducing THERMAL POWER to 25% RTP ensures that the core is operating farther from thermal limits and places the core in a conservative condition. Four hours is a reasonable amount of time, based on operating experience, to reduce THERMAL POWER to 25% RTP in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.2.4-2 Revised 01/29/2020
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| INSERT Bases 3.2.4 ASI B 3.2.4 BASES SURVEILLANCE SR 3.2.4.1 REQUIREMENTS Verifying that the ASI is within the limits specified in the COLR ensures that the core is not approaching DNB conditions. ASI is determined from excore detector readings which are calibrated using incore detector measurements (Ref. 1). Calibration factors are determined using incore measurements and an incore analysis computer program (Ref. 2). ASI is normally calculated and compared to the alarm setpoints continuously and automatically. Therefore, this SR only requires verification that alarms do not indicate an excessive ASI. If the Excore Monitoring System ASI Alarm function is inoperable, excore detector or incore indications may be used to monitor ASI. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Section 7.6.2.2
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| : 2. FSAR, Section 7.6.2.4 Palisades Nuclear Plant B 3.2.4-3 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 B 3.3 INSTRUMENTATION B 3.3.1 Reactor Protective System (RPS) Instrumentation BASES BACKGROUND The RPS initiates a reactor trip to protect against violating the acceptable fuel design limits and breaching the reactor coolant pressure boundary during Anticipated Operational Occurrences (AOOs). (As defined in 10 CFR 50, Appendix A, "Anticipated operational occurrences mean those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power.") By tripping the reactor, the RPS also assists the Engineered Safety Features (ESF) systems in mitigating accidents.
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| The protection and monitoring systems have been designed to ensure safe operation of the reactor. This is achieved by specifying Limiting Safety System Settings (LSSS) in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system parameters and equipment performance.
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| The LSSS, defined in this Specification as the Allowable Values, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits during Design Basis Accidents (DBAs).
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| During AOOs, which are those events expected to occur one or more times during the plant life, the acceptable limits are:
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| * The Departure from Nucleate Boiling Ratio (DNBR) shall be maintained above the Safety Limit (SL) value to prevent departure from nucleate boiling;
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| * Fuel centerline melting shall not occur; and
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| * The Primary Coolant System (PCS) pressure SL of 2750 psia shall not be exceeded.
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| Maintaining the parameters within the above values ensures that the offsite dose will be within applicable 10 CFR 50.67 (Ref. 6) and 10 CFR 100 (Ref. 2) criteria during AOOs.
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| Palisades Nuclear Plant B 3.3.1-1 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES BACKGROUND Accidents are events that are analyzed even though they are not (continued) expected to occur during the plant life. The acceptable limit during accidents is that the offsite dose shall be maintained within applicable 10 CFR 50.67 (Ref. 6) and 10 CFR 100 (Ref. 2) limits. Different accident categories allow a different fraction of these limits based on probability of occurrence. Meeting the acceptable dose limit for an accident category is considered having acceptable consequences for that event.
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| The RPS is segmented into four interconnected modules. These modules are:
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| * Measurement channels;
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| * RPS trip units;
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| * Matrix Logic; and
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| * Trip Initiation Logic.
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| This LCO addresses measurement channels and RPS trip units. It also addresses the automatic bypass removal feature for those trips with Zero Power Mode bypasses. The RPS Logic and Trip Initiation Logic are addressed in LCO 3.3.2, "Reactor Protective System (RPS) Logic and Trip Initiation." The role of the measurement channels, RPS trip units, and RPS Bypasses is discussed below.
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| Measurement Channels Measurement channels, consisting of pressure switches, field transmitters, or process sensors and associated instrumentation, provide a measurable electronic signal based upon the physical characteristics of the parameter being measured.
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| With the exception of High Startup Rate, which employs two instrument channels, and Loss of Load, which employs a single pressure sensor, four identical measurement channels with electrical and physical separation are provided for each parameter used in the direct generation of trip signals. These are designated channels A through D.
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| Some measurement channels provide input to more than one RPS trip unit within the same RPS channel. In addition, some measurement channels may also be used as inputs to Engineered Safety Features (ESF) bistables, and most provide indication in the control room.
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| Palisades Nuclear Plant B 3.3.1-2 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES BACKGROUND Measurement Channels (continued)
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| (continued)
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| In the case of High Startup Rate and Loss of Load, where fewer than four sensor channels are employed, the reactor trips provided are not relied upon by the plant safety analyses. The sensor channels do however, provide trip input signals to all four RPS channels.
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| When a channel monitoring a parameter exceeds a predetermined setpoint, indicating an abnormal condition, the bistable monitoring the parameter in that channel will trip. Tripping two or more channels of bistable trip units monitoring the same parameter de-energizes Matrix Logic, (addressed by LCO 3.3.2) which in turn de-energizes the Trip Initiation Logic. This causes all four DC clutch power supplies to de-energize, interrupting power to the control rod drive mechanism clutches, allowing the full length control rods to insert into the core.
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| For those trips relied upon in the safety analyses, three of the four measurement and trip unit channels can meet the redundancy and testability of GDC 21 in 10 CFR 50, Appendix A (Ref. 1). This LCO requires, however, that four channels be OPERABLE. The fourth channel provides additional flexibility by allowing one channel to be removed from service (trip channel bypassed) for maintenance or testing while still maintaining a minimum two-out-of-three logic.
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| Since no single failure will prevent a protective system actuation, this arrangement meets the requirements of IEEE Standard 279-1971 (Ref. 3).
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| Most of the RPS trips are generated by comparing a single measurement to a fixed bistable setpoint. Two trip Functions, Variable High Power Trip and Thermal Margin Low Pressure Trip, make use of more than one measurement to provide a trip.
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| The required RPS Trip Functions utilize the following input instrumentation:
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| * Variable High Power Trip (VHPT)
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| The VHPT uses Q Power as its input. Q Power is the higher of NI power from the power range NI drawer and primary calorimetric power (T power) based on PCS hot leg and cold leg temperatures. The measurement channels associated with the VHPT are the power range excore channels, and the PCS hot and cold leg temperature channels.
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| Palisades Nuclear Plant B 3.3.1-3 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES BACKGROUND Measurement Channels (continued)
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| * Variable High Power Trip (VHPT) (continued)
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| The Thermal Margin Monitors provide the complex signal processing necessary to calculate the TM/LP trip setpoint, VHPT trip setpoint and trip comparison, and Q Power calculation. On power decreases the VHPT setpoint tracks power levels downward so that it is always within a fixed increment above current power, subject to a minimum value.
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| On power increases, the trip setpoint remains fixed unless manually reset, at which point it increases to the new setpoint, a fixed increment above Q Power at the time of reset, subject to a maximum value. Thus, during power escalation, the trip setpoint must be repeatedly reset to avoid a reactor trip.
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| * High Startup Rate Trip The High Startup Rate trip uses the wide range Nuclear Instruments (NIs) to provide an input signal. There are only two wide range NI channels. The wide range channel signal processing electronics are physically mounted in RPS cabinet channels C (NI-1/3) and D (NI-2/4). Separate bistable trip units mounted within the NI-1/3 wide range channel drawer supply High Startup Rate trip signals to RPS channels A and C. Separate bistable trip units mounted within the NI-2/4 wide range channel drawer provide High Startup Rate trip signals to RPS channels B and D.
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| * Low Primary Coolant Flow Trip The Low Primary Coolant Flow Trip utilizes 16 flow measurement channels which monitor the differential pressure across the primary side of the steam generators. Each RPS channel, A, B, C, and D, receives a signal which is the sum of four differential pressure signals. This totalized signal is compared with a setpoint in the RPS Low Flow bistable trip unit for that RPS channel.
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| Palisades Nuclear Plant B 3.3.1-4 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES BACKGROUND Measurement Channels (continued)
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| (continued)
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| * Low Steam Generator Level Trips There are two separate Low Steam Generator Level trips, one for each steam generator. Each Low Steam Generator Level trip monitors four level measurement channels for the associated steam generator, one for each RPS channel.
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| * Low Steam Generator Pressure Trips There are also two separate Low Steam Generator Pressure trips, one for each steam generator. Each Low Steam Generator Pressure trip monitors four pressure measurement channels for the associated steam generator, one for each RPS channel.
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| * High Pressurizer Pressure Trip The High Pressurizer Pressure Trip monitors four pressurizer pressure channels, one for each RPS channel.
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| * Thermal Margin Low Pressure (TM/LP) Trip The TM/LP Trip utilizes bistable trip units. Each of these bistable trip units receives a calculated trip setpoint from the Thermal Margin Monitor (TMM) and compares it to the measured pressurizer pressure signal. The TM/LP setpoint is based on Q power (the higher of NI power from the power range NI drawer, or T power, based on PCS hot leg and cold leg temperatures) pressurizer pressure, PCS cold leg temperature, and Axial Shape Index. The TMM provide the complex signal processing necessary to calculate the TM/LP trip setpoint, TM/LP trip comparison signal, and Q Power.
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| Palisades Nuclear Plant B 3.3.1-5 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES BACKGROUND Measurement Channels (continued)
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| (continued)
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| * Loss of Load Trip The Loss of Load Trip is initiated by two-out-of-three logic from pressure switches in the turbine auto stop oil circuit that sense a turbine trip for input to all four RPS auxiliary trip units. The Loss of Load Trip is actuated by turbine auxiliary relays 305L and 305R.
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| Relay 305L provides input to RPS channels A and C; 305R to channels B and D. Relays 305L and 305R are energized on a turbine trip. Their inputs are the same as the inputs to the turbine solenoid trip valve, 20ET.
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| If a turbine trip is generated by loss of auto stop oil pressure, the auto stop oil pressure switches, by two-out-of-three logic, will actuate relays 305L and 305R and generate a reactor trip. If a turbine trip is generated by an input to the solenoid trip valve, relays 305L and 305R, which are wired in parallel, will also be actuated and will generate a reactor trip.
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| * Containment High Pressure Trip The Containment High Pressure Trip is actuated by four pressure switches, one for each RPS channel.
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| * Zero Power Mode Bypass Automatic Removal The Zero Power Bypass allows manually bypassing (i.e., disabling) four reactor trip functions, Low PCS Flow, Low SG A Pressure, Low SG B Pressure, and TM/LP (low PCS pressure),
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| when reactor power (as indicated by the wide range nuclear instrument channels) is below 10-4%. This bypassing is necessary to allow RPS testing and control rod drive mechanism testing when the reactor is shutdown and plant conditions would cause a reactor trip to be present.
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| The Zero Power Mode Bypass removal interlock uses the wide range nuclear instruments (NIs) as measurement channels.
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| There are only two wide range NI channels. Separate bistables are provided to actuate the bypass removal for each RPS channel. Bistables in the NI-1/3 channel provide the bypass removal function for RPS channels A and C; bistables in the NI-2/4 channel for RPS channels B and D.
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| Palisades Nuclear Plant B 3.3.1-6 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES BACKGROUND Several measurement instrument channels provide more than one (continued) required function. Those sensors shared for RPS and ESF functions are identified in Table B 3.3.1-1. That table provides a listing of those shared channels and the Specifications which they affect.
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| RPS Trip Units Two types of RPS trip units are used in the RPS cabinets; bistable trip units and auxiliary trip units:
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| A bistable trip unit receives a measured process signal from its instrument channel and compares it to a setpoint; the trip unit actuates three relays, with contacts in the Matrix Logic channels, when the measured signal is less conservative than the setpoint.
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| They also provide local trip indication and remote annunciation.
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| An auxiliary trip unit receives a digital input (contacts open or closed); the trip unit actuates three relays, with contacts in the Matrix Logic channels, when the digital input is received. They also provide local trip indication and remote annunciation.
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| Each RPS channel has four auxiliary trip units and seven bistable trip units.
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| The contacts from these trip unit relays are arranged into six coincidence matrices, comprising the Matrix Logic. If bistable trip units monitoring the same parameter in at least two channels trip, the Matrix Logic will generate a reactor trip (two-out-of-four logic).
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| Four of the RPS measurement channels provide contact outputs to the RPS, so the comparison of an analog input to a trip setpoint is not necessary. In these cases, the bistable trip unit is replaced with an auxiliary trip unit. The auxiliary trip units provide contact multiplication so the single input contact opening can provide multiple contact outputs to the coincidence logic as well as trip indication and annunciation.
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| Palisades Nuclear Plant B 3.3.1-7 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES BACKGROUND RPS Trip Units (continued)
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| (continued)
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| Trips employing auxiliary trip units include the VHPT, which receives contact inputs from the Thermal Margin Monitors; the High Startup Rate trip which employs contact inputs from bistables mounted in the two wide range drawers; the Loss of Load Trip which receives contact inputs from one of two auxiliary relays which are operated by two-out-of-three logic switches sensing turbine auto stop oil pressure; and the Containment High Pressure (CHP) trip, which employs containment pressure switch contacts.
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| There are four RPS trip units, designated as channels A through D, each channel having eleven trip units, one for each RPS Function. Trip unit output relays de-energize when a trip occurs.
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| All RPS Trip Functions, with the exception of the Loss of Load and CHP trips, generate a pretrip alarm as the trip setpoint is approached.
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| The Allowable Values are specified for each safety related RPS trip Function which is credited in the safety analysis. Nominal trip setpoints are specified in the plant procedures. The nominal setpoints are selected to ensure plant parameters do not exceed the Allowable Value if the instrument loop is performing as required. The methodology used to determine the nominal trip setpoints is also provided in plant documents. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value specified is more conservative than the analytical limit determined in the safety analysis in order to account for uncertainties appropriate to the trip Function. These uncertainties are addressed as described in plant documents. A channel is inoperable if its actual setpoint is not within its Allowable Value.
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| Setpoints in accordance with the Allowable Value will ensure that SLs of Chapter 2.0 are not violated during AOOs and the consequences of DBAs will be acceptable, providing the plant is operated from within the LCOs at the onset of the AOO or DBA and the equipment functions as designed.
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| Note that in the accompanying LCO 3.3.1, the Allowable Values of Table 3.3.1-1 are the LSSS.
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| Palisades Nuclear Plant B 3.3.1-8 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES BACKGROUND Reactor Protective System Bypasses (continued)
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| Three different types of trip bypass are utilized in the RPS, Operating Bypass, Zero Power Mode Bypass, and Trip Channel Bypass. The Operating Bypass or Zero Power Mode Bypass prevent the actuation of a trip unit or auxiliary trip unit; the Trip Channel Bypass prevents the trip unit output from affecting the Logic Matrix. A channel which is bypassed, other than as allowed by the Table 3.3.1-1 footnotes, cannot perform its specified safety function and must be considered to be inoperable.
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| Operating Bypasses The Operating Bypasses are initiated and removed automatically during startup and shutdown as power level changes. An Operating Bypass prevents the associated RPS auxiliary trip unit from receiving a trip signal from the associated measurement channel. With the bypass in place, neither the pre-trip alarm nor the trip will actuate if the measured parameter exceeds the set point. An annunciator is provided for each Operating Bypass. The RPS trips with Operating Bypasses are:
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| : a. High Startup Rate Trip bypass. The High Startup Rate trip is automatically bypassed when the associated wide range channel indicates below 1E-4% RTP, and when the associated power range excore channel indicates above 13% RTP. These bypasses are automatically removed between 1E-4% RTP and 13% RTP.
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| : b. Loss of Load bypass. The Loss of Load trip is automatically bypassed when the associated power range excore channel indicates below 17% RTP. The bypass is automatically removed when the channel indicates above the set point. The same power range excore channel bistable is used to bypass the High Startup Rate trip and the Loss of Load trip for that RPS channel.
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| Palisades Nuclear Plant B 3.3.1-9 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES BACKGROUND Operating Bypasses (continued)
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| (continued)
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| Each wide range channel contains two bistables set at 1E-4% RTP, one bistable unit for each associated RPS channel. Each of the two wide range channels affect the Operating Bypasses for two RPS channels; wide range channel NI-1/3 for RPS channels A and C, wide range channel NI-2/4 for RPS channels B and D. Each of the four power range excore channel affects the Operating Bypasses for the associated RPS channel. The power range excore channel bistables associated with the Operating Bypasses are set at a nominal 15%, and are required to actuate between 13% RTP and 17% RTP.
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| Zero Power Mode (ZPM) Bypass The ZPM Bypass is used when the plant is shut down and it is desired to raise the control rods for control rod drop testing with PCS flow, pressure or temperature too low for the RPS trips to be reset. ZPM bypasses may be manually initiated and removed when wide range power is below 1E-4% RTP, and are automatically removed if the associated wide range NI indicated power exceeds 1E-4% RTP. A ZPM bypass prevents the RPS trip unit from actuating if the measured parameter exceeds the set point. Operation of the pretrip alarm is unaffected by the zero power mode bypass. An annunciator indicates the presence of any ZPM bypass. The RPS trips with ZPM bypasses are:
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| : a. Low Primary Coolant System Flow.
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| : b. Low Steam Generator Pressure.
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| : c. Thermal Margin/Low Pressure.
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| The wide range NI channels provide contact closure permissive signals when indicated power is below 1E-4% RTP. The ZPM bypasses may then be manually initiated or removed by actuation of key-lock switches.
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| One key-lock switch located on each RPS cabinet controls the ZPM Bypass for the associated RPS trip channels. The bypass is automatically removed if the associated wide range NI indicated power exceeds 1E-4% RTP. The same wide range NI channel bistables that provide the ZPM Bypass permissive and removal signals also provide the high startup rate trip Operating Bypass actuation and removal.
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| Palisades Nuclear Plant B 3.3.1-10 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES BACKGROUND Trip Channel Bypass (continued)
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| A Trip Channel Bypass is used when it is desired to physically remove an individual trip unit from the system, or when calibration or servicing of a trip channel could cause an inadvertent trip. A trip Channel Bypass may be manually initiated or removed at any time by actuation of a key-lock switch. A Trip Channel Bypass prevents the trip unit output from affecting the RPS logic matrix. A light above the bypass switch indicates that the trip channel has been bypassed. Each RPS trip unit has an associated trip channel bypass:
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| The key-lock trip channel bypass switch is located above each trip unit.
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| The key cannot be removed when in the bypass position. Only one key for each trip parameter is provided, therefore the operator can bypass only one channel of a given parameter at a time. During the bypass condition, system logic changes from two-out-of-four to two-out-of-three channels required for trip.
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| APPLICABLE Each of the analyzed accidents and transients can be detected by one SAFETY ANALYSES or more RPS Functions. The accident analysis contained in Reference 4 takes credit for most RPS trip Functions. The High Startup Rate and Loss of Load Functions, which are not specifically credited in the accident analysis, are part of the NRC approved licensing basis for the plant, and are required to be operable in accordance with their respective LCO. The High Startup Rate and Loss of Load trips are purely equipment protective, and their use minimizes the potential for equipment damage.
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| The specific safety analyses applicable to each protective Function are identified below.
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| : 1. Variable High Power Trip (VHPT)
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| The VHPT provides reactor core protection against positive reactivity excursions.
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| The safety analysis assumes that this trip is OPERABLE to terminate excessive positive reactivity insertions during power operation and while shut down.
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| Palisades Nuclear Plant B 3.3.1-11 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES APPLICABLE 2. High Startup Rate Trip SAFETY ANALYSIS (continued) There are no safety analyses which take credit for functioning of the High Startup Rate Trip. The High Startup Rate trip is used to trip the reactor when excore wide range power indicates an excessive rate of change. The High Startup Rate trip minimizes transients for events such as a continuous control rod withdrawal or a boron dilution event from low power levels. The trip may be operationally bypassed when THERMAL POWER is
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| < 1E-4% RTP, when poor counting statistics may lead to erroneous indication. It may also be operationally bypassed at
| |
| > 13% RTP, where moderator temperature coefficient and fuel temperature coefficient make high rate of change of power unlikely.
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| There are only two wide range drawers, with each supplying contact input to auxiliary trip units in two RPS channels.
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| : 3. Low Primary Coolant System Flow Trip The Low PCS Flow trip provides DNB protection during events which suddenly reduce the PCS flow rate during power operation, such as loss of power to, or seizure of, a primary coolant pump.
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| Flow in each of the four PCS loops is determined from pressure drop from inlet to outlet of the SGs. The total PCS flow is determined, for the RPS flow channels, by summing the loop pressure drops across the SGs and correlating this pressure sum with the sum of SG differential pressures which exist at 100% flow (four pump operation at full power Tave). Full PCS flow is that flow which exists at RTP, at full power Tave, with four pumps operating.
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| 4, 5. Low Steam Generator Level Trip The Low Steam Generator Level trips are provided to trip the reactor in the event of excessive steam demand (to prevent overcooling the PCS) and loss of feedwater events (to prevent overpressurization of the PCS).
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| The Allowable Value assures that there will be sufficient water inventory in the SG at the time of trip to allow a safe and orderly plant shutdown and to prevent SG dryout assuming minimum AFW capacity.
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| Palisades Nuclear Plant B 3.3.1-12 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES APPLICABLE 4, 5. Low Steam Generator Level Trip (continued)
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| SAFETY ANALYSIS (continued) Each SG level is sensed by measuring the differential pressure in the upper portion of the downcomer annulus in the SG. These trips share four level sensing channels on each SG with the AFW actuation signal.
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| 6, 7. Low Steam Generator Pressure Trip The Low Steam Generator Pressure trip provides protection against an excessive rate of heat extraction from the steam generators, which would result in a rapid uncontrolled cooldown of the PCS. This trip provides a mitigation function in the event of an MSLB.
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| The Low SG Pressure channels are shared with the Low SG Pressure signals which isolate the steam and feedwater lines.
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| : 8. High Pressurizer Pressure Trip The High Pressurizer Pressure trip, in conjunction with pressurizer safety valves and Main Steam Safety Valves (MSSVs), provides protection against overpressure conditions in the PCS when at operating temperature. The safety analyses assume the High Pressurizer Pressure trip is OPERABLE during accidents and transients which suddenly reduce PCS cooling (e.g., Loss of Load, Main Steam Isolation Valve (MSIV) closure, etc.) or which suddenly increase reactor power (e.g., rod ejection accident).
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| The High Pressurizer Pressure trip shares four safety grade instrument channels with the TM/LP trip, Anticipated Transient Without Scram (ATWS) and PORV circuits, and the Pressurizer Low Pressure Safety Injection Signal.
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| Palisades Nuclear Plant B 3.3.1-13 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES APPLICABLE 9. Thermal Margin/Low Pressure (TM/LP) Trip SAFETY ANALYSIS (continued) The TM/LP trip is provided to prevent reactor operation when the DNBR is insufficient. The TM/LP trip protects against slow reactivity or temperature increases, and against pressure decreases.
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| The trip is initiated whenever the PCS pressure signal drops below a minimum value (Pmin) or a computed value (Pvar) as described below, whichever is higher.
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| The TM/LP trip uses Q Power, ASI, pressurizer pressure, and cold leg temperature (Tc) as inputs.
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| Q Power is the higher of core THERMAL POWER (T Power) or nuclear power. The T power uses hot leg and cold leg RTDs as inputs. Nuclear power uses the power range excore channels as inputs. Both the T and excore power signals have provisions for calibration by calorimetric calculations.
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| The ASI is calculated from the upper and lower power range excore detector signals, as explained in Section 1.1, Definitions.
| |
| The signal is corrected for the difference between the flux at the core periphery and the flux at the detectors.
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| The Tc value is the higher of the two cold leg signals.
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| The Low Pressurizer Pressure trip limit (Pvar)is calculated using the equations given in Table 3.3.1-2.
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| The calculated limit (Pvar) is then compared to a fixed Low Pressurizer Pressure trip limit (Pmin). The auctioneered highest of these signals becomes the trip limit (Ptrip). Ptrip is compared to the measured PCS pressure and a trip signal is generated when the measured pressure for that channel is less than or equal to Ptrip. A pre-trip alarm is also generated when P is less than or equal to the pre-trip setting, Ptrip + P.
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| The TM/LP trip setpoint is a complex function of these inputs and represents a minimum acceptable PCS pressure for the existing temperature and power conditions. It is compared to actual PCS pressure in the TM/LP trip unit.
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| Palisades Nuclear Plant B 3.3.1-14 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES APPLICABLE 10. Loss of Load Trip SAFETY ANALYSIS (continued) There are no safety analyses which take credit for functioning of the Loss of Load Trip.
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| The Loss of Load trip is provided to prevent lifting the pressurizer and main steam safety valves in the event of a turbine generator trip while at power. The trip is equipment protective. The safety analyses do not assume that this trip functions during any accident or transient. The Loss of Load trip uses two-out-of-three logic from pressure switches in the turbine auto stop oil circuit to sense a turbine trip for input to all four RPS auxiliary trip units.
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| : 11. Containment High Pressure Trip The Containment High Pressure trip provides a reactor trip in the event of a Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB). The Containment High Pressure trip shares sensors with the Containment High Pressure sensing logic for Safety Injection, Containment Isolation, and Containment Spray.
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| Each of these sensors has a single bellows which actuates two microswitches. One microswitch on each of four sensors provides an input to the RPS.
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| : 12. Zero Power Mode Bypass Removal The only RPS bypass considered in the safety analyses is the Zero Power Mode (ZPM) Bypass. The ZPM Bypass is used when the plant is shut down and it is desired to raise the control rods for control rod drop testing with PCS flow or temperature too low for the RPS Low PCS Flow, Low SG Pressure, or Thermal Margin/Low Pressure trips to be reset. ZPM bypasses are automatically removed if the wide range NI indicated power exceeds 1E-4% RTP.
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| Palisades Nuclear Plant B 3.3.1-15 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES APPLICABLE 12. Zero Power Mode Bypass Removal (continued)
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| SAFETY ANALYSIS (continued) The safety analyses take credit for automatic removal of the ZPM Bypass if reactor criticality due to a Continuous Control Rod Bank Withdrawal should occur with the affected trips bypassed and PCS flow, pressure, or temperature below the values at which the RPS could be reset. The ZPM Bypass would effectively be removed when the first wide range NI channel indication reached 1E-4% RTP. With the ZPM Bypass for two RPS channels removed, the RPS would trip on one of the un-bypassed trips.
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| This would prevent the reactor reaching an excessive power level.
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| If a reactor criticality due to a Continuous Control Rod Bank Withdrawal should occur when PCS flow, steam generator pressure, and PCS pressure (TM/LP) were above their trip setpoints, a trip would terminate the event when power increased to the minimum setting (nominally 30%) of the Variable High Power Trip. In this case, the monitored parameters are at or near their normal operational values, and a trip initiated at 30% RTP provides adequate protection.
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| The RPS design also includes automatic removal of the Operating Bypasses for the High Startup Rate and Loss of Load trips. The safety analyses do not assume functioning of either these trips or the automatic removal of their bypasses.
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| The RPS instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| LCO The LCO requires all instrumentation performing an RPS Function to be OPERABLE. Failure of the trip unit (including its output relays), any required portion of the associated instrument channel, or both, renders the affected channel(s) inoperable and reduces the reliability of the affected Functions. Failure of an automatic ZPM bypass removal channel may also impact the associated instrument channel(s) and reduce the reliability of the affected Functions.
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| Palisades Nuclear Plant B 3.3.1-16 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES LCO Actions allow Trip Channel Bypass of individual channels, but the (continued) bypassed channel must be considered to be inoperable. The bypass key used to bypass a single channel cannot be simultaneously used to bypass that same parameter in other channels. This interlock prevents operation with more than one channel of the same Function trip channel bypassed. The plant is normally restricted to 7 days in a trip channel bypass, or otherwise inoperable condition before either restoring the Function to four channel operation (two-out-of-four logic) or placing the channel in trip (one-out-of-three logic).
| |
| The Allowable Values are specified for each safety related RPS trip Function which is credited in the safety analysis. Nominal trip setpoints are specified in the plant procedures. The nominal setpoints are selected to ensure plant parameters do not exceed the Allowable Value if the instrument loop is performing as required. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value specified is more conservative than the analytical limit determined in the safety analysis in order to account for uncertainties appropriate to the trip Function.
| |
| These uncertainties are addressed as described in plant documents.
| |
| Neither Allowable Values nor setpoints are specified for the non-safety related RPS Trip Functions, since no safety analysis assumptions would be violated if they are not set at a particular value.
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| The following Bases for each trip Function identify the above RPS trip Function criteria items that are applicable to establish the trip Function OPERABILITY.
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| : 1. Variable High Power Trip (VHPT)
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| This LCO requires all four channels of the VHPT Function to be OPERABLE.
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| The Allowable Value is high enough to provide an operating envelope that prevents unnecessary VHPT trips during normal plant operations. The Allowable Value is low enough for the system to function adequately during reactivity addition events.
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| Palisades Nuclear Plant B 3.3.1-17 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES LCO 1. Variable High Power Trip (VHPT) (continued)
| |
| (continued)
| |
| The VHPT is designed to limit maximum reactor power to its maximum design and to terminate power excursions initiating at lower powers without power reaching this full power limit. During plant startup, the VHPT trip setpoint is initially at its minimum value, 30%. Below 30% RTP, the VHPT setpoint is not required to track with Q Power, i.e., be adjusted to within 15% RTP. It remains fixed until manually reset, at which point it increases to 15% above existing Q Power.
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| The maximum allowable setting of the VHPT is 109.4% RTP.
| |
| Adding to this the possible variation in trip setpoint due to calibration and instrument error, the maximum actual steady state power at which a trip would be actuated is 113.4%, which is the value assumed in the safety analysis.
| |
| : 2. High Startup Rate Trip This LCO requires four channels of High Startup Rate Trip Function to be OPERABLE in MODES 1 and 2.
| |
| The High Startup Rate trip serves as a backup to the administratively enforced startup rate limit. The Function is not credited in the accident analyses; therefore, no Allowable Value for the trip or operating bypass Functions is derived from analytical limits and none is specified.
| |
| The High Startup Rate Trip is required to be OPERABLE, in accordance with the LCO, even though the Trip Function is not credited in the accident analysis.
| |
| The four channels of the High Startup Rate trip are derived from two wide range NI signal processing drawers. Thus, a failure in one wide range channel could render two RPS channels inoperable. It is acceptable to continue operation in this condition because the High Startup Rate trip is not credited in any safety analyses.
| |
| The requirement for this trip Function is modified by a footnote, which allows the High Startup Rate trip to be bypassed when the wide range NI indicates below 10E-4% or when THERMAL POWER is above 13% RTP. If a High Startup Rate trip is bypassed when power is between these limits, it must be considered to be inoperable.
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| Palisades Nuclear Plant B 3.3.1-18 Revised 01/29/2020
| |
| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES LCO 3. Low Primary Coolant System Flow Trip (continued)
| |
| This LCO requires four channels of Low PCS Flow Trip Function to be OPERABLE.
| |
| This trip is set high enough to maintain fuel integrity during a loss of flow condition. The setting is low enough to allow for normal operating fluctuations from offsite power.
| |
| The Low PCS Flow trip setpoint of 95% of full PCS flow insures that the reactor cannot operate when the flow rate is less than 93% of the nominal value considering instrument errors. Full PCS flow is that flow which exists at RTP, at full power Tave, with four pumps operating.
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| The requirement for this trip Function is modified by a footnote, which allows use of the ZPM bypass when wide range power is below 1E-4% RTP. That bypass is automatically removed when the associated wide range channel indicates 1E-4% RTP. If a trip channel is bypassed when power is above 1E-4% RTP, it must be considered to be inoperable.
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| 4, 5. Low Steam Generator Level Trip This LCO requires four channels of Low Steam Generator Level Trip Function per steam generator to be OPERABLE.
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| The 25.9% Allowable Value assures that there is an adequate water inventory in the steam generators when the reactor is critical and is based upon narrow range instrumentation. The 25.9%
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| indicated level corresponds to the location of the feed ring.
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| 6, 7. Low Steam Generator Pressure Trip This LCO requires four channels of Low Steam Generator Pressure Trip Function per steam generator to be OPERABLE.
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| The Allowable Value of 500 psia is sufficiently below the full load operating value for steam pressure so as not to interfere with normal plant operation, but still high enough to provide the required protection in the event of excessive steam demand.
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| Since excessive steam demand causes the PCS to cool down, resulting in positive reactivity addition to the core, a reactor trip is required to offset that effect.
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| Palisades Nuclear Plant B 3.3.1-19 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES LCO 8. High Pressurizer Pressure Trip (continued)
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| This LCO requires four channels of High Pressurizer Pressure Trip Function to be OPERABLE.
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| The Allowable Value is set high enough to allow for pressure increases in the PCS during normal operation (i.e., plant transients) not indicative of an abnormal condition. The setting is below the lift setpoint of the pressurizer safety valves and low enough to initiate a reactor trip when an abnormal condition is indicated.
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| : 9. Thermal Margin/Low Pressure (TM/LP) Trip This LCO requires four channels of TM/LP Trip Function to be OPERABLE.
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| The TM/LP trip setpoints are derived from the core thermal limits through application of appropriate allowances for measurement uncertainties and processing errors. The allowances specifically account for instrument drift in both power and inlet temperatures, calorimetric power measurement, inlet temperature measurement, and primary system pressure measurement.
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| Other uncertainties including allowances for assembly power tilt, fuel pellet manufacturing tolerances, core flow measurement uncertainty and core bypass flow, inlet temperature measurement time delays, and ASI measurement, are included in the development of the TM/LP trip setpoint used in the accident analysis.
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| The requirement for this trip Function is modified by a footnote, which allows use of the ZPM bypass when wide range power is below 1E-4% RTP. That bypass is automatically removed when the associated wide range channel indicates 1E-4% RTP. If a trip channel is bypassed when power is above 1E-4% RTP, it must be considered to be inoperable.
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| Palisades Nuclear Plant B 3.3.1-20 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES LCO 10. Loss of Load Trip (continued)
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| The LCO requires four Loss of Load Trip Function channels to be OPERABLE in MODE 1 with THERMAL POWER 17% RTP.
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| The Loss of Load trip may be bypassed or be inoperable with THERMAL POWER < 17% RTP, since it is no longer needed to prevent lifting of the pressurizer safety valves or steam generator safety valves in the event of a Loss of Load. Loss of Load Trip unit must be considered inoperable if it is bypassed when THERMAL POWER is above 17% RTP.
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| This LCO requires four RPS Loss of Load auxiliary trip units, relays 305L and 305R, and pressure switches 63/AST-1, 63/AST-2, and 63/AST-3 to be OPERABLE. With those components OPERABLE, a turbine trip will generate a reactor trip.
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| The LCO does not require the various turbine trips, themselves, to be OPERABLE.
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| The Nuclear Steam Supply System and Steam Dump System are capable of accommodating the Loss of Load without requiring the use of the above equipment.
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| The Loss of Load Trip Function is not credited in the accident analysis; therefore, an Allowable Value for the trip cannot be derived from analytical limits, and is not specified.
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| The Loss of Load Trip is required to be OPERABLE, in accordance with the LCO, even though the Trip Function is not credited in the accident analysis.
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| : 11. Containment High Pressure Trip This LCO requires four channels of Containment High Pressure Trip Function to be OPERABLE.
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| The Allowable Value is high enough to allow for small pressure increases in containment expected during normal operation (i.e., plant heatup) that are not indicative of an abnormal condition.
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| The setting is low enough to initiate a reactor trip to prevent containment pressure from exceeding design pressure following a DBA and ensures the reactor is shutdown before initiation of safety injection and containment spray.
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| Palisades Nuclear Plant B 3.3.1-21 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES LCO 12. ZPM Bypass (continued)
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| The LCO requires that four channels of automatic Zero Power Mode (ZPM) Bypass removal instrumentation be OPERABLE.
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| Each channel of automatic ZPM Bypass removal includes a shared wide range NI channel, an actuating bistable in the wide range drawer, and a relay in the associated RPS cabinet. Wide Range NI channel 1/3 is shared between ZPM Bypass removal channels A and C; Wide Range NI channel 2/4, between ZPM Bypass removal channels B and D. An operable bypass removal channel must be capable of automatically removing the capability to bypass the affected RPS trip channels with the ZPM Bypass key switch at the proper setpoint.
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| APPLICABILITY This LCO requires all safety related trip functions to be OPERABLE in accordance with Table 3.3.1-1.
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| Those RPS trip Functions which are assumed in the safety analyses (all except High Startup Rate and Loss of Load), are required to be operable in MODES 1 and 2, and in MODES 3, 4, and 5 with more than one full-length control rod capable of being withdrawn and PCS boron concentration less than REFUELING BORON CONCENTRATION.
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| These trip Functions are not required while in MODES 3, 4, or 5, if PCS boron concentration is at REFUELING BORON CONCENTRATION, or when no more than one full-length control rod is capable of being withdrawn, because the RPS Function is already fulfilled. REFUELING BORON CONCENTRATION provides sufficient negative reactivity to assure the reactor remains subcritical regardless of control rod position, and the safety analyses assume that the highest worth withdrawn full-length control rod will fail to insert on a trip. Therefore, under these conditions, the safety analyses assumptions will be met without the RPS trip Function.
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| The High Startup Rate Trip Function is required to be OPERABLE in MODES 1 and 2, but may be bypassed when the associated wide range NI channel indicates below 1E-4% power, when poor counting statistics may lead to erroneous indication. In MODES 3, 4, 5, and 6, the High Startup Rate trip is not required to be OPERABLE. Wide range channels are required to be OPERABLE in MODES 3, 4, and 5, by LCO 3.3.9, "Neutron Flux Monitoring Channels," and in MODE 6, by LCO 3.9.2, Nuclear Instrumentation.
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| Palisades Nuclear Plant B 3.3.1-22 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES APPLICABILITY The High Startup Rate Trip Function is required to be OPERABLE in (continued) MODES 1 and 2, but may be bypassed when the associated wide range NI channel indicates below 1E-4% power, when poor counting statistics may lead to erroneous indication. In MODES 3, 4, 5, and 6, the High Startup Rate trip is not required to be OPERABLE. Wide range channels are required to be OPERABLE in MODES 3, 4, and 5, by LCO 3.3.9, "Neutron Flux Monitoring Channels," and in MODE 6, by LCO 3.9.2, Nuclear Instrumentation.
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| The Loss of Load trip is required to be OPERABLE with THERMAL POWER at or above 17% RTP. Below 17% RTP, the ADVs are capable of relieving the pressure due to a Loss of Load event without challenging other overpressure protection.
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| The trips are designed to take the reactor subcritical, maintaining the SLs during AOOs and assisting the ESF in providing acceptable consequences during accidents.
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| ACTIONS The most common causes of channel inoperability are outright failure of loop components or drift of those loop components which is sufficient to exceed the tolerance provided in the plant setpoint analysis. Loop component failures are typically identified by the actuation of alarms due to the channel failing to the safe condition, during CHANNEL CHECKS (when the instrument is compared to the redundant channels), or during the CHANNEL FUNCTIONAL TEST (when an automatic component might not respond properly). Typically, the drift of the loop components is found to be small and results in a delay of actuation rather than a total loss of function. Excessive loop component drift would, most likely, be identified during a CHANNEL CHECK (when the instrument is compared to the redundant channels) or during a CHANNEL CALIBRATION (when instrument loop components are checked against reference standards).
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| In the event a channel's trip setpoint is found nonconservative with respect to the Allowable Value, or the transmitter, instrument loop, signal processing electronics, or RPS bistable trip unit is found inoperable, all affected Functions provided by that channel must be declared inoperable, and the plant must enter the Condition for the particular protection Functions affected.
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| Palisades Nuclear Plant B 3.3.1-23 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES ACTIONS When the number of inoperable channels in a trip Function exceeds that (continued) specified in any related Condition associated with the same trip Function, then the plant is outside the safety analysis. Therefore, LCO 3.0.3 is immediately entered if applicable in the current MODE of operation.
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| A Note has been added to the ACTIONS to clarify the application of the Completion Time rules. The Conditions of this Specification may be entered independently for each Function. The Completion Times of each inoperable Function will be tracked separately for each Function, starting from the time the Condition was entered.
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| A.1 Condition A applies to the failure of a single channel in any required RPS Function, except High Startup Rate, Loss of Load, or ZPM Bypass Removal. (Condition A is modified by a Note stating that this Condition does not apply to the High Startup Rate, Loss of Load, or ZPM Bypass Removal Functions. The failure of one channel of those Functions is addressed by Conditions B, C, or D.)
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| If one RPS bistable trip unit or associated instrument channel is inoperable, operation is allowed to continue. Since the trip unit and associated instrument channel combine to perform the trip function, this Condition is also appropriate if both the trip unit and the associated instrument channel are inoperable. Though not required, the inoperable channel may be bypassed. The provision of four trip channels allows one channel to be bypassed (removed from service) during operations, placing the RPS in two-out-of-three coincidence logic. The failed channel must be restored to OPERABLE status or placed in trip within 7 days.
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| Required Action A.1 places the Function in a one-out-of-three configuration. In this configuration, common cause failure of dependent channels cannot prevent trip.
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| The Completion Time of 7 days is based on operating experience, which has demonstrated that a random failure of a second channel occurring during the 7 day period is a low probability event.
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| Palisades Nuclear Plant B 3.3.1-24 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES ACTIONS A.1 (continued)
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| (continued)
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| The Completion Time of 7 days is based on operating experience, which has demonstrated that a random failure of a second channel occurring during the 7 day period is a low probability event.
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| B.1 Condition B applies to the failure of a single High Startup Rate trip unit or associated instrument channel.
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| If one trip unit or associated instrument channel fails, it must be restored to OPERABLE status prior to entering MODE 2 from MODE 3. A shutdown provides the appropriate opportunity to repair the trip function and conduct the necessary testing. The Completion Time is based on the fact that the safety analyses take no credit for the functioning of this trip.
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| C.1 Condition C applies to the failure of a single Loss of Load or associated instrument channel.
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| If one trip unit or associated instrument channel fails, it must be restored to OPERABLE status prior to THERMAL POWER 17% RTP following a shutdown. If the plant is shutdown at the time the channel becomes inoperable, then the failed channel must be restored to OPERABLE status prior to THERMAL POWER 17% RTP. For this Completion Time, following a shutdown means this Required Action does not have to be completed until prior to THERMAL POWER 17% RTP for the first time after the plant has been in MODE 3 following entry into the Condition. The Completion Time trip assures that the plant will not be restarted with an inoperable Loss of Load trip channel.
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| Palisades Nuclear Plant B 3.3.1-25 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES ACTIONS D.1 and D.2 (continued)
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| Condition D applies when one or more automatic ZPM Bypass removal channels are inoperable. If the ZPM Bypass removal channel cannot be restored to OPERABLE status, the affected ZPM Bypasses must be immediately removed, or the bypassed RPS trip Function channels must be immediately declared to be inoperable. Unless additional circuit failures exist, the ZPM Bypass may be removed by placing the associated Zero Power Mode Bypass key operated switch in the normal position.
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| A trip channel which is actually bypassed, other than as allowed by the Table 3.3.1-1 footnotes, cannot perform its specified safety function and must immediately be declared to be inoperable.
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| E.1 and E.2 Condition E applies to the failure of two channels in any RPS Function, except ZPM Bypass Removal Function. (The failure of ZPM Bypass Removal Functions is addressed by Condition D.).
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| Condition E is modified by a Note stating that this Condition does not apply to the ZPM Bypass Removal Function.
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| Required Action E.1 provides for placing one inoperable channel in trip within the Completion Time of 1 hour. Though not required, the other inoperable channel may be (trip channel) bypassed.
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| Palisades Nuclear Plant B 3.3.1-26 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES ACTIONS E.1 and E.2 (continued)
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| (continued)
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| This Completion Time is sufficient to allow the operator to take all appropriate actions for the failed channels while ensuring that the risk involved in operating with the failed channels is acceptable. With one channel of protective instrumentation bypassed or inoperable in an untripped condition, the RPS is in a two-out-of-three logic for that function; but with another channel failed, the RPS may be operating in a two-out-of-two logic. This is outside the assumptions made in the analyses and should be corrected. To correct the problem, one of the inoperable channels is placed in trip. This places the RPS in a one-out-of-two for that function logic. If any of the other unbypassed channels for that function receives a trip signal, the reactor will trip.
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| Action E.2 is modified by a Note stating that this Action does not apply to (is not required for) the High Startup Rate and Loss of Load Functions.
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| One channel is required to be restored to OPERABLE status within 7 days for reasons similar to those stated under Condition A. After one channel is restored to OPERABLE status, the provisions of Condition A still apply to the remaining inoperable channel. Therefore, the channel that is still inoperable after completion of Required Action E.2 must be placed in trip if more than 7 days have elapsed since the initial channel failure.
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| F.1 The power range excore channels are used to generate the internal ASI signal used as an input to the TM/LP trip. They also provide input to the Thermal Margin Monitors for determination of the Q Power input for the TM/LP trip and the VHPT. If two power range excore channels cannot be restored to OPERABLE status, power is restricted or reduced during subsequent operations because of increased uncertainty associated with inoperable power range excore channels which provide input to those trips.
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| The Completion Time of 2 hours is adequate to reduce power in an orderly manner without challenging plant systems.
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| Palisades Nuclear Plant B 3.3.1-27 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES ACTIONS G.1, G.2.1, and G.2.2 (continued)
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| Condition G is entered when the Required Action and associated Completion Time of Condition A, B, C, D, E, or F are not met, or if the control room ambient air temperature exceeds 90°F.
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| If the control room ambient air temperature exceeds 90°F, all Thermal Margin Monitor channels are rendered inoperable because their operating temperature limit is exceeded. In this condition, or if the Required Actions and associated Completion Times are not met, the reactor must be placed in a condition in which the LCO does not apply.
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| To accomplish this, the plant must be placed in MODE 3, with no more than one full-length control rod capable of being withdrawn or with the PCS boron concentration at REFUELING BORON CONCENTRATION in 6 hours.
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| The Completion Time is reasonable, based on operating experience, for placing the plant in MODE 3 from full power conditions in an orderly manner and without challenging plant systems. The Completion Time is also reasonable to ensure that no more than one full-length control rod is capable of being withdrawn or that the PCS boron concentration is at REFUELING BORON CONCENTRATION.
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| SURVEILLANCE The SRs for any particular RPS Function are found in the SR column of REQUIREMENTS Table 3.3.1-1 for that Function. Most Functions are subject to CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION.
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| SR 3.3.1.1 Performance of the CHANNEL CHECK ensures that gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. Under most conditions, a CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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| Palisades Nuclear Plant B 3.3.1-28 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.1 (continued)
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| REQUIREMENTS (continued) Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limits.
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| The Containment High Pressure and Loss of Load channels are pressure switch actuated. As such, they have no associated control room indicator and do not require a CHANNEL CHECK.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.3.1.2 This SR verifies that the control room ambient air temperature is within the environmental qualification temperature limits for the most restrictive RPS components, which are the Thermal Margin Monitors. These monitors provide input to both the VHPT Function and the TM/LP Trip Function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.3.1.3 A calibration (heat balance) is performed when THERMAL POWER is 15%. The calibration consists of adjusting the "nuclear power calibrate" potentiometers to agree with the calorimetric calculation if the absolute difference is 1.5%. Nuclear power is adjusted via a potentiometer, or THERMAL POWER is adjusted via a Thermal Margin Monitor bias number, as necessary, in accordance with the calibration (heat balance) procedure. Performance of the calibration ensures that the two inputs to the Q power measurement are indicating accurately with respect to the much more accurate secondary calorimetric calculation.
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| Palisades Nuclear Plant B 3.3.1-29 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.3 (continued)
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| REQUIREMENTS (continued) The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| The Frequency is modified by a Note indicating this Surveillance must be performed within 12 hours after THERMAL POWER is 15% RTP.
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| The secondary calorimetric is inaccurate at lower power levels. The 12 hours allows time requirements for plant stabilization, data taking, and instrument calibration.
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| SR 3.3.1.4 It is necessary to calibrate the power range excore channel upper and lower subchannel amplifiers such that the measured ASI reflects the true core power distribution as determined by the incore detectors. ASI is utilized as an input to the TM/LP trip function where it is used to ensure that the measured axial power profiles are bounded by the axial power profiles used in the development of the Tinlet limitation of LCO 3.4.1. An adjustment of the excore channel is necessary only if reactor power is greater than 25% RTP and individual excore channel ASI differs from AXIAL OFFSET, as measured by the incores, outside the bounds of the following table:
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| Allowed Group 4 Group 4 Reactor Rods 128" withdrawn Rods <128" withdrawn Power 100% -0.020 (AO-ASI) 0.020 -0.040 (AO-ASI) 0.040
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| < 95 -0.033 (AO-ASI) 0.020 -0.053 (AO-ASI) 0.040
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| < 90 -0.046 (AO-ASI) 0.020 -0.066 (AO-ASI) 0.040
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| < 85 -0.060 (AO-ASI) 0.020 -0.080 (AO-ASI) 0.040
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| < 80 -0.120 (AO-ASI) 0.080 -0.140 (AO-ASI) 0.100
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| < 75 -0.120 (AO-ASI) 0.080 -0.140 (AO-ASI) 0.100
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| < 70 -0.120 (AO-ASI) 0.080 -0.140 (AO-ASI) 0.100
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| < 65 -0.120 (AO-ASI) 0.080 -0.140 (AO-ASI) 0.100
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| < 60 -0.160 (AO-ASI) 0.120 -0.180 (AO-ASI) 0.140
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| < 55 -0.160 (AO-ASI) 0.120 -0.180 (AO-ASI) 0.140
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| < 50 -0.160 (AO-ASI) 0.120 -0.180 (AO-ASI) 0.140
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| < 45 -0.160 (AO-ASI) 0.120 -0.180 (AO-ASI) 0.140
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| < 40 -0.160 (AO-ASI) 0.120 -0.180 (AO-ASI) 0.140
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| < 35 -0.160 (AO-ASI) 0.120 -0.180 (AO-ASI) 0.140
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| < 30 -0.160 (AO-ASI) 0.120 -0.180 (AO-ASI) 0.140
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| < 25 Below 25% RTP any AO/ASI difference is acceptable Table values determined with a conservative Pvar gamma constant of -9505.
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| Palisades Nuclear Plant B 3.3.1-30 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.4 (continued)
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| REQUIREMENTS (continued) Below 25% RTP any difference between ASI and AXIAL OFFSET is acceptable. A Note indicates the Surveillance is not required to have been performed until 12 hours after THERMAL POWER is 25% RTP.
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| Uncertainties in the excore and incore measurement process make it impractical to calibrate when THERMAL POWER is < 25% RTP. The 12 hours allows time for plant stabilization, data taking, and instrument calibration.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.3.1.5 A CHANNEL FUNCTIONAL TEST is performed on each RPS instrument channel, except Loss of Load and High Startup Rate, to ensure the entire channel will perform its intended function when needed. For the TM/LP Function, the constants associated with the Thermal Margin Monitors must be verified to be within tolerances.
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| A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
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| Any setpoint adjustment must be consistent with the assumptions of the current setpoint analysis.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.3.1-31 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.6 REQUIREMENTS (continued) A calibration check of the power range excore channels is performed using the internal test circuitry. This SR uses an internally generated test signal to check that the 0% and 50% levels read within limits for both the upper and lower detector, both on the analog meter and on the TMM screen. This check verifies that neither the zero point nor the amplifier gain adjustment have undergone excessive drift since the previous complete CHANNEL CALIBRATION.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.3.1.7 A CHANNEL FUNCTIONAL TEST on the Loss of Load and High Startup Rate channels is performed prior to a reactor startup to ensure the entire channel will perform its intended function.
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| A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests.
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| The High Startup Rate trip is actuated by either of the Wide Range Nuclear Instrument Startup Rate channels. NI-1/3 sends a trip signal to RPS channels A and C; NI-2/4 to channels B and D. Since each High Startup Rate channel would cause a trip on two RPS channels, the High Startup Rate trip is not tested when the reactor is critical.
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| The four Loss of Load Trip channels are all actuated by a single pressure switch monitoring turbine auto stop oil pressure which is not tested when the reactor is critical. Operating experience has shown that these components usually pass the Surveillance when performed at a Frequency of once per 7 days prior to each reactor startup.
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| Palisades Nuclear Plant B 3.3.1-32 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.8 REQUIREMENTS (continued) SR 3.3.1.8 performs a CHANNEL CALIBRATION.
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| CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor (except neutron detectors). The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be consistent with the setpoint analysis.
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| The bistable setpoints must be found to trip within the Allowable Values specified in the LCO and left set consistent with the assumptions of the setpoint analysis. The Variable High Power Trip setpoint shall be verified to reset properly at several indicated power levels during (simulated) power increases and power decreases.
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| The as-found and as-left values must also be recorded and reviewed for consistency with the assumptions of the setpoint analysis.
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| As part of the CHANNEL CALIBRATION of the wide range Nuclear Instrumentation, automatic removal of the ZPM Bypass for the Low PCS Flow, TM/LP must be verified to assure that these trips are available when required.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| This SR is modified by a Note which states that it is not necessary to calibrate neutron detectors because they are passive devices with minimal drift and because of the difficulty of simulating a meaningful signal. Slow changes in power range excore neutron detector sensitivity are compensated for by performing the calorimetric calibration (SR 3.3.1.3) and the calibration using the incore detectors (SR 3.3.1.4). Sudden changes in detector performance would be noted during the required CHANNEL CHECKS (SR 3.3.1.1).
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| Palisades Nuclear Plant B 3.3.1-33 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 21
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| : 2. 10 CFR 100
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| : 3. IEEE Standard 279-1971, April 5, 1972
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| : 4. FSAR, Chapter 14
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| : 5. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989
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| : 6. 10 CFR 50.67 Palisades Nuclear Plant B 3.3.1-34 Revised 01/29/2020
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| INSERT Bases 3.3.1 RPS Instrumentation B 3.3.1 Table B 3.3.1-1 (page 1 of 1)
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| Instruments Affecting Multiple Specifications Required Instrument Channels Affected Specifications Nuclear Instrumentation Source Range NI-1/3, Count Rate Indication @ C-150 Panel 3.3.8 (#1)
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| Source Range NI-1/3 & 2/4, Count Rate Signal 3.3.9 & 3.9.2 Wide Range NI-1/3 & 2/4, Flux Level 10-4 Bypass 3.3.1 (#3, 6, 7, 9, & 12)
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| Wide Range NI-1/3 & 2/4, Startup Rate 3.3.1 (#2)
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| Wide Range NI-1/3 & 2/4, Flux Level Indication @EC-06 Panel for 3.3.7 3.3.7 (#3) & 3.3.9 Power Range NI-5, 6, 7, & 8, Tq 3.2.1 & 3.2.3 Power Range NI-5, 6, 7, & 8, Q Power 3.3.1 (#1 & 9)
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| Power Range NI-5, 6, 7, & 8, ASI 3.3.1 (#9) & 3.2.1 & 3.2.4 Power Range NI-5, 6, 7, & 8, Loss of Load/High Startup Rate Bypass 3.3.1 (#2 & 10)
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| PCS T-Cold Instruments TT-0112CA, Temperature Signal (SPI T Power for PDIL Alarm Circuit) 3.1.6 TT-0112CA & 0122CA, Temperature Signal (C-150) 3.3.8 (#6 & 7)
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| TT-0122CB, Temperature Signal (PIP T Power for PDIL Alarm Circuit) 3.1.6 TT-0112CA & 0122CB, Temperature Signal (LTOP) 3.4.12.b.1 TT-0112CC & 0122CD (PTR-0112 & 0122) Temperature Indication 3.3.7 (#2)
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| TT-0112 & 0122 CC & CD, Temperature Signal (SMM) 3.3.7 (#5)
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| TT-0112 & 0122 CA, CB, CC, & CD, Temperature Signal (Q Power & TMM) 3.3.1 (#1 & 9) & 3.4.1.b PCS T-Hot Instruments TT-0112HA, Temperature Signal (SPI T Power for PDIL Alarm Circuit) 3.1.6 TT-0112HA & 0122HA, Temperature Signal (C-150) 3.3.8 (#4 & 5)
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| TT-0122HB, Temperature Signal (PIP T Power for PDIL Alarm Circuit) 3.1.6 TT-0112 & 0122 HC & HD, Temperature Signal (SMM) 3.3.7 (#5)
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| TT-0112HC & 0122HD (PTR-0112 & 0122) Temperature Indication 3.3.7 (#1)
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| TT-0112 & 0122 HA, HB, HC, & HD, Temperature Signal (Q Power & TMM) 3.3.1 (#1 & 9)
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| Thermal Margin Monitors PY-0102A, B, C, & D 3.3.1 (#1 & 9)
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| Pressurizer Pressure Instruments PT-0102A, B, C, & D, Pressure Signal (RPS & SIS) 3.3.1 (#8 & 9) &
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| 3.3.3 (#1.a & 7a)
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| PT-0104A & B, Pressure Signal (LTOP & SDC Interlock) 3.4.12.b.1 & 3.4.14 PT-0105A & B, Pressure Signal (WR Indication & LTOP) 3.3.7 (#5) & 3.4.12.b.1 PI-0110, Pressure Indication @ C-150 Panel 3.3.8 (#2)
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| SG Level Instruments LT-0751 & 0752 A, B, C, & D, Level Signal (RPS & AFAS) 3.3.1 (#4 & 5) &
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| 3.3.3 (#4.a & 4.b)
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| LI-0757 & 0758 A & B, Wide Range Level Indication 3.3.7 (#11 & 12)
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| LI-0757C & 0758C, Wide Range Level Indication @ C-150 Panel 3.3.8 (#10 & 11)
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| SG Pressure Instruments PT-0751 & 0752 A, B, C, & D, Pressure Signal (RPS & SG Isolation) 3.3.1 (#6 & 7) &
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| 3.3.3 (#2a, 2b, 7b, 7c)
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| PT-0751A and PT-0752A Pressure Signal (C-150/150A) 3.3.8 (#8 & 9)
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| PIC-0751 & 0752 C & D, Pressure Indication 3.3.7 (#13 & 14)
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| PI-0751E & 0752E, Pressure Indication @ C-150 Panel 3.3.8 (#8 & 9)
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| Containment Pressure Instruments PS-1801, 1802, 1803, & 1804, Switch Output (RPS) 3.3.1 (#11)
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| PS-1801, 1802A, 1803, & 1804A, Switch Output (ESF) 3.3.3 (#5.a)
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| PS-1801A, 1802, 1803A, & 1804, Switch Output (ESF) 3.3.3 (#5.b)
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| Note: The information provided in this table is intended for use as an aid to distinguish those instrument channels which provide more than one required function and to describe which specifications they affect. The information in this table should not be taken as inclusive for all instruments nor affected specifications.
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| Palisades Nuclear Plant B 3.3.1-35 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.2 RPS Logic and Trip Initiation B 3.3.2 B 3.3 INSTRUMENTATION B 3.3.2 Reactor Protective System (RPS) Logic and Trip Initiation BASES BACKGROUND The RPS initiates a reactor trip to protect against violating the acceptable fuel design limits and reactor coolant pressure boundary integrity during Anticipated Operational Occurrences (AOOs). (As defined in 10 CFR 50, Appendix A, "Anticipated operational occurrences mean those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power.") By tripping the reactor, the RPS also assists the Engineered Safety Features (ESF) systems in mitigating accidents.
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| The protection and monitoring systems have been designed to ensure safe operation of the reactor. This is achieved by specifying Limiting Safety System Settings (LSSS) in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system parameters and equipment performance.
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| The LSSS, defined in this Specification as the Allowable Value, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits during Design Basis Accidents (DBAs).
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| During AOOs, which are those events expected to occur one or more times during the plant life, the acceptable limits are:
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| * The Departure from Nucleate Boiling Ratio (DNBR) shall be maintained above the Safety Limit (SL) value to prevent departure from nucleate boiling;
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| * Fuel centerline melting shall not occur; and
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| * The Primary Coolant System (PCS) pressure SL of 2750 psia shall not be exceeded.
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| Maintaining the parameters within the above values ensures that the offsite dose will be within applicable 10 CFR 50.67 (Ref. 6) and 10 CFR 100 (Ref. 2) criteria during AOOs.
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| Palisades Nuclear Plant B 3.3.2-1 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.2 RPS Logic and Trip Initiation B 3.3.2 BASES BACKGROUND Accidents are events that are analyzed even though they are not (continued) expected to occur during the plant life. The acceptable limit during accidents is that the offsite dose shall be maintained within applicable 10 CFR 50.67 (Ref. 6) and 10 CFR 100 (Ref. 2) limits. Different accident categories allow a different fraction of these limits based on probability of occurrence. Meeting the acceptable dose limit for an accident category is considered having acceptable consequences for that event.
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| The RPS is segmented into four interconnected modules.
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| These modules are:
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| * Measurement channels (or pressure switches);
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| * Bistable trip units;
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| * Matrix Logic; and
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| * Trip Initiation Logic.
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| This LCO addresses the RPS Logic (Matrix Logic and Trip Initiation Logic), including Manual Trip capability. LCO 3.3.1, "Reactor Protective System (RPS) Instrumentation," provides a description of the role of the measurement channels and associated bistable trip units in the RPS.
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| The RPS Logic is summarized below:
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| RPS Logic The RPS Logic, consisting of Matrix Logic and Trip Initiation Logic, employs a scheme that provides a reactor trip when trip units in any two of the four channels sense the same input parameter trip. This is called a two-out-of-four trip logic. This logic and the clutch power supply configuration are shown in FSAR Figure 7-1 (Ref. 3).
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| Bistable trip unit relay contact outputs from the four channels are configured into six logic matrices. Each logic matrix checks for a coincident trip in the same parameter in two trip unit channels. The matrices are designated the AB, AC, AD, BC, BD, and CD matrices to reflect the bistable trip unit channels being monitored. Each logic matrix contains four normally energized matrix relays. When a coincidence is detected, consisting of a trip in the same Function in the two channels being monitored by the logic matrix, all four matrix relay coils de-energize.
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| Palisades Nuclear Plant B 3.3.2-2 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.2 RPS Logic and Trip Initiation B 3.3.2 BASES BACKGROUND RPS Logic (continued)
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| (continued)
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| The matrix relay contacts are arranged into trip paths, with one of the four matrix relays in each matrix opening contacts in one of the four trip paths. Each trip path provides power to one of the four normally energized clutch power supply M-contactors (M1, M2, M3, and M4).
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| The trip paths thus each have six contacts in series, one from each matrix, and perform a logical OR function, de-energizing the M-contactors if any one or more of the six logic matrices indicate a coincidence condition.
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| When a coincidence occurs in two RPS channels, all four matrix relays in the affected matrix de-energize. This in turn de-energizes all four M-contactors, which interrupt AC input power to the four clutch power supplies, allowing the full-length control rods to insert by gravity.
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| Manual reactor trip capability is afforded by two main control panel-mounted pushbuttons. One of these (on Control Panel CO-2) opens contacts in series with each of the four trip paths, de-energizing all M-contactors. The other pushbutton (on Control Panel CO-6) opens circuit breakers which provide AC input power to the M-contactor contacts and downstream clutch power supplies. Thus depressing either pushbutton will cause a reactor trip.
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| De-energizing the M-contactors removes AC power to the four clutch power supply inputs. Contacts from M-contactors M1 and M2 are in series with each other and in the AC power supply path to clutch power supplies PS1 and PS2 (these constitute a trip leg). M3 and M4 are similarly arranged with respect to clutch power supplies PS3 and PS4 (these constitute a second trip leg). Approximately half of the control rod clutches receive power from auctioneered clutch power supplies 1 and 3. The remaining control rod clutches receive clutch power from auctioneered clutch power supplies 2 and 4.
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| Matrix Logic refers to the matrix power supplies, trip channel bypass contacts, and interconnecting RPS cabinet matrix wiring between bistable and auxiliary trip unit relay contacts, including the matrix relays.
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| Contacts in the bistable and auxiliary trip units are excluded from the Matrix Logic definition, since they are addressed as part of the instrumentation channel.
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| The Trip Initiation Logic consists of the M-contactor isolation transformers, all interconnecting wiring, and the M-contactors.
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| Palisades Nuclear Plant B 3.3.2-3 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.2 RPS Logic and Trip Initiation B 3.3.2 BASES BACKGROUND RPS Logic (continued)
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| (continued)
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| Manual trip circuitry includes both manual reactor trip pushbuttons C0-2 and C0-6, and the interconnecting wiring necessary to effect deenergization of the clutch power supplies.
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| Neither the clutch power supplies nor the AC input power source to these supplies is considered as safety related. Operation may continue with one or two selective clutch power supplies de-energized.
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| It is possible to change the two-out-of-four RPS Logic to a two-out-of-three logic for a given input parameter in one channel at a time by Trip Channel Bypassing the RPS Trip unit output contacts in the Matrix Logic Ladder. Trip Channel Bypassing a trip unit effectively shorts the trip unit relay contacts in the three matrices associated with that channel. Thus, the bypassed trip units will function normally, producing normal channel trip indication and annunciation, but a reactor trip will not occur unless two additional channels indicate a trip condition. Trip Channel Bypassing can be simultaneously performed on any number of parameters in any number of channels, providing each parameter is bypassed in only one channel at a time. A single bypass key for each trip function interlock prevents simultaneous Trip Channel Bypassing of the same parameter in more than one channel.
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| Trip Channel Bypassing is normally employed during maintenance or testing.
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| Functional testing of the entire RPS, from trip unit input through the de-energizing of individual sets of clutch power supplies, can be performed either at power or during shutdown and is normally performed on a quarterly basis. FSAR Section 7.2 (Ref. 4) explains RPS testing in more detail.
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| APPLICABLE Reactor Protective System (RPS) Logic SAFETY ANALYSES The RPS Logic provides for automatic trip initiation to avoid exceeding the SLs during AOOs and to assist the ESF systems in ensuring acceptable consequences during accidents. All transients and accidents that call for a reactor trip assume the RPS Logic is functioning as designed.
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| Palisades Nuclear Plant B 3.3.2-4 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.2 RPS Logic and Trip Initiation B 3.3.2 BASES APPLICABLE Manual Trip SAFETY ANALYSIS (continued) There are no accident analyses that take credit for the Manual Trip; however, the Manual Trip is part of the RPS circuitry. It is used by the operator to shut down the reactor whenever any parameter is rapidly trending toward its trip setpoint. A Manual Trip accomplishes the same results as any one of the automatic trip Functions.
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| The RPS Logic and Trip Initiation satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| LCO Reactor Protective System (RPS) Logic Failures of individual trip unit relays and their contacts are addressed in LCO 3.3.1. This Specification addresses failures of the Matrix Logic not addressed in the above, such as the failure of matrix relay power supplies or the failure of the trip channel bypass contact in the bypass condition.
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| Loss of a single preferred AC bus will de-energize one of the two power supplies in each of three matrices. Because of power supply auctioneering, all four matrix relays will remain energized in each affected matrix.
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| Each of the four Trip Initiation Logic channels de-energizes one set of clutch power supplies if any of the six coincidence matrices de-energize their associated matrix relays. They thus perform a logical OR function.
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| Trip Initiation Logic channels 1 and 2 receive AC power from preferred AC bus Y-30. Trip Initiation Logic channels 3 and 4 receive AC input power from preferred AC bus Y-40. Because of clutch power supply output auctioneering, it is possible to de-energize either input bus without de-energizing control rod clutches.
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| : 1. Matrix Logic This LCO requires six channels of Matrix Logic to be OPERABLE in MODES 1 and 2, and in MODES 3, 4, and 5 when more than one full-length control rod is capable of being withdrawn and the PCS boron concentration is less than REFUELING BORON CONCENTRATION.
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| Palisades Nuclear Plant B 3.3.2-5 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.2 RPS Logic and Trip Initiation B 3.3.2 BASES LCO 2. Trip Initiation Logic (continued)
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| This LCO requires four channels of Trip Initiation Logic to be OPERABLE in MODES 1 and 2, and in MODES 3, 4, and 5 when more than one full-length control rod is capable of being withdrawn and the PCS boron concentration is less than REFUELING BORON CONCENTRATION.
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| : 3. Manual Trip The LCO requires both Manual Trip channels to be OPERABLE in MODES 1 and 2, and in MODES 3, 4, and 5 when more than one full-length control rod is capable of being withdrawn and the PCS boron concentration is less than REFUELING BORON CONCENTRATION.
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| Two independent pushbuttons are provided. Each pushbutton is considered to be a channel. Depressing either pushbutton interrupts power to all four clutch power supplies, tripping the reactor.
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| APPLICABILITY The RPS Matrix Logic, Trip Initiation Logic, and Manual Trip are required to be OPERABLE in MODES 1 and 2, and in MODES 3, 4, and 5 when more than one full-length control rod capable of being withdrawn and the PCS boron concentration is less than REFUELING BORON CONCENTRATION. This ensures the reactor can be tripped when necessary, but allows for maintenance and testing when the reactor trip is not needed.
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| In MODES 3, 4, and 5 with no more than one full-length control rod capable of being withdrawn or the PCS boron concentration at REFUELING BORON CONCENTRATION, these Functions do not have to be OPERABLE. However, LCO 3.3.9, Neutron Flux Monitoring Channels, does require neutron flux monitoring capability under these conditions.
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| Palisades Nuclear Plant B 3.3.2-6 Revised 01/29/2020
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| INSERT Bases 3.3.2 RPS Logic and Trip Initiation B 3.3.2 BASES ACTIONS When the number of inoperable channels in a trip Function exceeds that specified in any related Condition associated with the same trip Function, then the plant is outside the safety analysis. Therefore, LCO 3.0.3 is immediately entered if applicable in the current MODE of operation.
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| A.1 Condition A applies if one Matrix Logic channel is inoperable. The channel must be restored to OPERABLE status within 48 hours. The Completion Time of 48 hours provides the operator time to take appropriate actions and still ensures that any risk involved in operating with a failed channel is acceptable. Operating experience has demonstrated that the probability of a random failure of a second Matrix Logic channel is low during any given 48 hour interval. If the channel cannot be restored to OPERABLE status within 48 hours, Condition E is entered.
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| B.1 Condition B applies if one Trip Initiation Logic channel is inoperable.
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| The Required Action require de-energizing the affected clutch power supplies. This removes the need for the affected channel by performing its associated safety function. With the clutch power supplies associated with one initiation logic channel de-energized, the remaining two clutch power supplies prevent control rod clutches from de-energizing. The remaining clutch power supplies are in a one-out-of-two logic with respect to the remaining initiation logic channels in the clutch power supply path. This meets redundancy requirements, but testing on the OPERABLE channels cannot be performed without causing a reactor trip.
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| Required Action B.1 provides for de-energizing the affected clutch power supplies associated with the inoperable channel within a Completion Time of 1 hour.
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| Palisades Nuclear Plant B 3.3.2-7 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.2 RPS Logic and Trip Initiation B 3.3.2 BASES ACTIONS C.1 (continued)
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| Condition C applies to the failure of one Manual Trip channel. With one manual reactor trip channel inoperable operation may continue until the reactor is shut down for other reasons. Repair during operation is not required because one OPERABLE channel is all that is required for safe operation. No safety analyses assume operation of the Manual trip.
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| The Manual Trip channels are not testable without actually causing a reactor trip, so even if the difficulty were corrected, the post maintenance testing necessary to declare the channel OPERABLE could not be completed during operation. Because of this, the Required Action is to restore the inoperable channel to OPERABLE status prior to entering MODE 2 from MODE 3 during the next plant startup.
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| D.1 Condition D applies to the failure of both Trip Initiation Logic channels affecting the same trip leg. The affected control rod drive clutch power supplies must be de-energized immediately. With both channels inoperable, the RPS Function is lost if the affected clutch power supplies are not de-energized. Therefore, immediate action is required to de-energize the affected clutch power supplies. The immediate Completion Time is appropriate since there could be a loss of safety function if the associated clutch power supplies are not de-energized.
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| E.1, E.2.1 and E.2.2 Condition E is entered if Required Actions associated with Condition A, B, C, or D are not met within the required Completion Time or if for one or more Functions more than one Manual Trip, Matrix Logic, or Trip Initiation Logic channel is inoperable for reasons other than Condition D.
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| In Condition E the reactor must be placed in a MODE in which the LCO does not apply. The Completion Time of 6 hours to be in MODE 3 is reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.3.2-8 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.2 RPS Logic and Trip Initiation B 3.3.2 BASES ACTIONS E.1, E.2.1 and E.2.2 (continued)
| |
| (continued)
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| Required Actions E.2.1 and E.2.2 allow 6 hours to verify that no more than one full-length control rod is capable of being withdrawn or to verify that PCS boron concentration is at REFUELING BORON CONCENTRATION. The Completion Time is reasonable to place the plant in an operating condition in which the LCO does not apply.
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| SURVEILLANCE SR 3.3.2.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST on each RPS Logic channel is performed every 92 days to ensure the entire channel will perform its intended function when needed. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
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| This SR addresses the two tests associated with the RPS Logic: Matrix Logic and Trip Initiation Logic.
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| Matrix Logic Tests These tests are performed one matrix at a time. They verify that a coincidence in the two input channels for each Function removes power from the matrix relays. During testing, power is applied to the matrix relay test coils and prevents the matrix relay contacts from assuming their de-energized state. The Matrix Logic tests will detect any short circuits around the bistable contacts in the coincidence logic such as may be caused by faulty bistable relay or trip channel bypass contacts.
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| Palisades Nuclear Plant B 3.3.2-9 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.2 RPS Logic and Trip Initiation B 3.3.2 BASES SURVEILLANCE SR 3.3.2.1 (continued)
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| REQUIREMENTS (continued) Trip Initiation Logic Tests These tests are similar to the Matrix Logic tests, except that test power is withheld from one matrix relay at a time, allowing the initiation circuit to de-energize, de-energizing the affected set of clutch power supplies.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.3.2.2 A CHANNEL FUNCTIONAL TEST on the Manual Trip channels is performed prior to a reactor startup to ensure the entire channel will perform its intended function if required. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
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| The Manual Trip Function is not tested at power. However, the simplicity of this circuitry and the absence of drift concern makes this Frequency adequate. Additionally, operating experience has shown that these components usually pass the Surveillance when performed once within 7 days prior to each reactor startup.
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| REFERENCES 1. 10 CFR 50, Appendix A
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| : 2. 10 CFR 100
| |
| : 3. FSAR, Figure 7-1
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| : 4. FSAR, Section 7.2
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| : 5. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989
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| : 6. 10 CFR 50.67 Palisades Nuclear Plant B 3.3.2-10 Amendment No. 271 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 B 3.3 INSTRUMENTATION B 3.3.3 Engineered Safety Features (ESF) Instrumentation BASES BACKGROUND The ESF Instrumentation initiates necessary safety systems, based upon the values of selected plant parameters, to protect against violating core design limits and the Primary Coolant System (PCS) pressure boundary and to mitigate accidents.
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| The ESF circuitry generates the signals listed below when the monitored variables reach levels that are indicative of conditions requiring protective action. The inputs to each ESF actuation signal are also listed.
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| : 1. Safety Injection Signal (SIS).
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| : a. Containment High Pressure (CHP)
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| : b. Pressurizer Low Pressure
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| : 2. Steam Generator Low Pressure (SGLP);
| |
| : a. Steam Generator A Low Pressure
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| : b. Steam Generator B Low Pressure
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| : 3. Recirculation Actuation Signal (RAS);
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| : a. Safety Injection Refueling Water Tank (SIRWT) Low Level
| |
| : 4. Auxiliary Feedwater Actuation Signal (AFAS);
| |
| : a. Steam Generator A Low Level
| |
| : b. Steam Generator B Low Level
| |
| : 5. Containment High Pressure Signal (CHP);
| |
| : a. Containment High Pressure - Left Train
| |
| : b. Containment High Pressure - Right Train Palisades Nuclear Plant B 3.3.3-1 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES BACKGROUND 6. Containment High Radiation Signal (CHR);
| |
| (continued)
| |
| : a. Containment High Radiation
| |
| : 7. Automatic Bypass Removal
| |
| : a. Pressurizer Pressure Low Bypass
| |
| : b. Steam Generator A Low Pressure Bypass
| |
| : c. Steam Generator B Low Pressure Bypass In the above list of actuation signals, the CHP and RAS are derived from pressure and level switches, respectively.
| |
| Equipment actuated by each of the above signals is identified in the FSAR, Chapter 7. (Ref. 1).
| |
| The ESF circuitry, with the exception of RAS, employs two-out-of-four logic. Four independent measurement channels are provided for each function used to generate ESF actuation signals. When any two channels of the same function reach their setpoint, actuating relays are energized which, in turn, initiate the protective actions. Two separate and redundant trains of actuating relays, each powered from separate power supplies, are utilized. These separate relay trains operate redundant trains of ESF equipment.
| |
| RAS logic consists of output contacts of the relays actuated by the SIRWT level switches arranged in a "one-out-of-two taken twice" logic.
| |
| The contacts are arranged so that at least one low level signal powered from each station battery is required to initiate RAS. Loss of a single battery, therefore, cannot either cause or prevent RAS initiation.
| |
| The ESF logic circuitry contains the capability to manually block the SIS actuation logic and the SGLP action logic during normal plant shutdowns to avoid undesired actuation of the associated equipment.
| |
| In each case, when three of the four associated measurement channels are below the block setpoint, pressing a manual pushbutton will block the actuation signal for that train. If two of the four of the measurement channels increase above the block setpoint, the block will automatically be removed.
| |
| Palisades Nuclear Plant B 3.3.3-2 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES BACKGROUND (continued) The sensor subsystems, including individual channel actuation bistables, is addressed in this LCO. The actuation logic subsystems, manual actuation, and downstream components used to actuate the individual ESF components are addressed in LCO 3.3.4.
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| Measurement Channels Measurement channels, consisting of pressure switches, field transmitters, or process sensors and associated instrumentation, provide a measurable electronic signal based upon the physical characteristics of the parameter being measured.
| |
| Four identical measurement channels are provided for each parameter used in the generation of trip signals. These are designated Channels A through D. Measurement channels provide input to ESF bistables within the same ESF channel. In addition, some measurement channels may also be used as inputs to Reactor Protective System (RPS) bistables, and most provide indication in the control room.
| |
| When a channel monitoring a parameter indicates an abnormal condition, the bistable monitoring the parameter in that channel will trip.
| |
| In the case of RAS and CHP, the sensors are latching auxiliary relays from level and pressure switches, respectively, which do not develop an analog input to separate bistables. Tripping two or more channels monitoring the same parameter will actuate both channels of Actuation Logic of the associated ESF equipment.
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| Three of the four measurement and bistable channels are necessary to meet the redundancy and testability of GDC 21 in Appendix A to 10 CFR 50 (Ref. 2). The fourth channel provides additional flexibility by allowing one channel to be removed from service for maintenance or testing while still maintaining a minimum two-out-of-three logic.
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| Since no single failure will prevent a protective system actuation and no protective channel feeds a control channel, this arrangement meets the requirements of IEEE Standard 279 -1971 (Ref. 3).
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| Palisades Nuclear Plant B 3.3.3-3 Revised 01/29/2020
| |
| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES BACKGROUND Measurement Channels (continued)
| |
| (continued)
| |
| The ESF Actuation Functions are generated by comparing a single measurement to a fixed bistable setpoint. The ESF Actuation Functions utilize the following input instrumentation:
| |
| * Safety Injection Signal (SIS)
| |
| The Safety Injection Signal can be generated by any of three inputs: Pressurizer Low Pressure, Containment High Pressure, or Manual Actuation. Manual Actuation is addressed by LCO 3.3.4; Containment High Pressure is discussed below. Four instruments (channels A through D), monitor Pressurizer Pressure to develop the SIS actuation. Each of these instrument channels has two individually adjustable ESF bistable trip devices, one for the bypass removal circuit (discussed below) and one for SIS. Each ESF bistable trip device actuates two auxiliary relays, one for each actuation train. The output contacts from these auxiliary relays form the logic circuits addressed in LCO 3.3.4. The instrument channels associated with each Pressurizer Low Pressure SIS actuation bistable include the pressure measurement loop, the SIS actuation bistable, and the two auxiliary relays associated with that bistable. The bistables associated with automatic removal of the Pressurizer Low Pressure Bypass are discussed under Function 7.a, below.
| |
| * Low Steam Generator Pressure Signal (SGLP)
| |
| There are two separate Low Steam Generator Pressure signals, one for each steam generator. For each steam generator, four instruments (channels A through D) monitor pressure to develop the SGLP actuation. Each of these instrument channels has two individually adjustable ESF bistable trip devices, one for the bypass removal circuit (discussed below) and one for SGLP.
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| Each Steam SGLP bistable trip device actuates an auxiliary relay.
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| The output contacts from these auxiliary relays form the SGLP logic circuits addressed in LCO 3.3.4. The instrument channels associated with each Steam Generator Low Pressure Signal bistable include the pressure measurement loop, the SGLP actuation bistable, and the auxiliary relay associated with that bistable. The bistables associated with automatic removal of the SGLP Bypass are discussed under Function 7.a, below.
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| Palisades Nuclear Plant B 3.3.3-4 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES BACKGROUND Measurement Channels (continued)
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| (continued)
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| * Recirculation Actuation Signal (RAS)
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| There are four Safety Injection Refueling Water (SIRW) Tank level instruments used to develop the RAS signal. Each of these instrument channels actuates two auxiliary relays, one for each actuation train. The output contacts from these auxiliary relays form the logic circuits addressed in LCO 3.3.4. The SIRW Tank Low Level instrument channels associated with each RAS actuation bistable include the level instrument and the two auxiliary relays associated with that instrument.
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| * Auxiliary Feedwater Actuation Signal (AFAS)
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| There are two separate AFAS signals (AFAS channels A and B),
| |
| each one actuated on low level in either steam generator. For each steam generator, four level instruments (channels A through D) monitor level to develop the AFAS actuation signals. The output contacts from the bistables on these level channels form the AFAS logic circuits addressed in LCO 3.3.4. The instrument channels associated with each Steam Generator Low Level Signal bistable include the level measurement loop and the Low Level AFAS bistable.
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| * Containment High Pressure Actuation (CHP)
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| The Containment High Pressure signal is actuated by two sets of four pressure switches, one set for each train. The output contacts from these pressure switches form the CHP logic circuits addressed in LCO 3.3.4.
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| Palisades Nuclear Plant B 3.3.3-5 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES BACKGROUND Measurement Channels (continued)
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| (continued)
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| * Containment High Radiation Actuation (CHR)
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| The CHR signal can be generated by either of two inputs: High Radiation or Manual Actuation. Manual Actuation is addressed by LCO 3.3.4. Four radiation monitor instruments (channels A through D), monitor containment area radiation level to develop the CHR signal. Each CHR monitor bistable device actuates one auxiliary relay which has contacts in each CHR logic train addressed in LCO 3.3.4. The instrument channels associated with each CHR actuation bistable include the radiation monitor itself and the associated auxiliary relay.
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| * Automatic Bypass Removal Functions Pressurizer Low Pressure and Steam Generator Low Pressure logic circuits have the capability to be blocked to avoid undesired actuation when pressure is intentionally lowered during plant shutdowns. In each case these bypasses are automatically removed when the measured pressure exceeds the bypass permissive setpoint. The measurement channels which provide the bypass removal signal are the same channels which provide the actuation signal. Each of these pressure measurement channels has two bistables, one for actuation and one for the bypass removal Function. The pressurizer pressure channels include an auxiliary relay actuated by the bypass removal bistable.
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| The logic circuits for Automatic Bypass Removal Functions are addressed by LCO 3.3.4.
| |
| Several measurement instrument channels provide more than one required function. Those sensors shared for RPS and ESF functions are identified in Table B 3.3.1-1. That table provides a listing of those shared channels and the Specifications which they affect.
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| Palisades Nuclear Plant B 3.3.3-6 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES BACKGROUND Bistable Trip Units (continued)
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| There are four channels of bistables, designated A through D, for each ESF Function, one for each measurement channel. The bistables for all required Functions, except CHP and RAS, receive an analog input from the measurement device, compare the analog input to trip setpoints, and provide contact output to the Actuation Logic. CHP and RAS are actuated by pressure switches and level switches respectively.
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| The Allowable Values are specified for each safety related ESF trip Function which is credited in the safety analysis. Nominal trip setpoints are specified in the plant procedures. The nominal setpoints are selected to ensure plant parameters do not exceed the Allowable Value if the instrument loop is performing as required. The methodology used to determine the nominal trip setpoints is also provided in plant documents. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value specified is more conservative than the analytical limit determined in the safety analysis in order to account for uncertainties appropriate to the trip Function. These uncertainties are addressed as described in plant documents. A channel is inoperable if its actual setpoint is not within its Allowable Value.
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| Setpoints in accordance with the Allowable Value will ensure that Safety Limits of Chapter 2.0, "SAFETY LIMITS (SLs)," are not violated during Anticipated Operational Occurrences (AOOs) and that the consequences of Design Basis Accidents (DBAs) will be acceptable, providing the plant is operated from within the LCOs at the onset of the AOO or DBA and the equipment functions as designed. (As defined in 10 CFR 50, Appendix A, "Anticipated operational occurances mean those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power.")
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| ESF Instrument Channel Bypasses The only ESF instrument channels with built-in bypass capability are the Low SG Level AFAS bistables. Those bypasses are effected by a key operated switch, similar to the RPS Trip Channel Bypasses. A bypassed Low SG Level channel AFAS bistable cannot perform its specified function and must be considered inoperable.
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| Palisades Nuclear Plant B 3.3.3-7 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES BACKGROUND ESF Instrument Channel Bypasses (continued)
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| (continued)
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| While there are no other built-in provisions for instrument channel bypasses in the ESF design (bypassing any other channel output requires opening a circuit link, lifting a lead, or using a jumper), this LCO includes requirements for OPERABILITY of the instrument channels and bistables which provide input to the Automatic Bypass Removal Logic channels required by LCO 3.3.4, ESF Logic and Manual Initiation.
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| The Actuation Logic channels for Pressurizer Pressure and Steam Generator Low Pressure, however, have the ability to be manually bypassed when the associated pressure is below the range where automatic protection is required. These actuation logic channel bypasses may be manually initiated when three-out-of-four bypass permissive bistables indicate below their setpoint. When two-out-of-four of these bistables are above their bypass permissive setpoint, the actuation logic channel bypass is automatically removed. The bypass permissive bistables use the same four measurement channels as the blocked ESF function for their inputs.
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| APPLICABLE Each of the analyzed accidents can be detected by one or more ESF SAFETY ANALYSES Functions. One of the ESF Functions is the primary actuation signal for that accident. An ESF Function may be the primary actuation signal for more than one type of accident. An ESF Function may also be a secondary, or backup, actuation signal for one or more other accidents.
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| Functions not specifically credited in the accident analysis, serve as backups and are part of the NRC approved licensing basis for the plant.
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| Palisades Nuclear Plant B 3.3.3-8 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES APPLICABLE ESF protective Functions are as follows.
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| SAFETY ANALYSES (continued) 1. Safety Injection Signal (SIS)
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| The SIS ensures acceptable consequences during Loss of Coolant Accident (LOCA) events, including steam generator tube rupture, and Main Steam Line Breaks (MSLBs) or Feedwater Line Breaks (FWLBs) (inside containment). To provide the required protection, SIS is actuated by a CHP signal, or by two-out-of-four Pressurizer Low Pressure channels decreasing below the setpoint. SIS initiates the following actions:
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| : a. Start HPSI & LPSI pumps;
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| : b. Start component cooling water and service water pumps;
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| : c. Initiate service water valve operations;
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| : d. Initiate component cooling water valve operations;
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| : e. Start containment cooling fans (when coincident with a loss of offsite power);
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| : f. Enable Containment Spray Pump Start on CHP; and
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| : g. Initiate Safety Injection Valve operations.
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| Each SIS logic train is also actuated by a contact pair on one of the CHP initiation relays for the associated CHP train.
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| : 2. Steam Generator Low Pressure Signal (SGLP)
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| The SGLP ensures acceptable consequences during an MSLB or FWLB by isolating the steam generator if it indicates a low steam generator pressure. The SGLP concurrent with or following a reactor trip, minimizes the rate of heat extraction and subsequent cooldown of the PCS during these events.
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| Palisades Nuclear Plant B 3.3.3-9 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES APPLICABLE 2. Steam Generator Low Pressure Signal (SGLP) (continued)
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| SAFETY ANALYSES (continued) One SGLP circuit is provided for each SG. Each SGLP circuit is actuated by two-out-of-four pressure channels on the associated SG reaching their setpoint. SGLP initiates the following actions:
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| : a. Close the associated Feedwater Regulating valve and its bypass; and
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| : b. Close both Main Steam Isolation Valves.
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| : 3. Recirculation Actuation Signal At the end of the injection phase of a LOCA, the SIRWT will be nearly empty. Continued cooling must be provided by the ECCS to remove decay heat. The source of water for the ECCS pumps is automatically switched to the containment recirculation sump. Switchover from SIRWT to the containment sump must occur before the SIRWT empties to prevent damage to the ECCS pumps and a loss of core cooling capability. For similar reasons, switchover must not occur before there is sufficient water in the containment sump to support pump suction.
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| Furthermore, early switchover must not occur to ensure sufficient borated water is injected from the SIRWT to ensure the reactor remains shut down in the recirculation mode. An SIRWT Low Level signal initiates the RAS.
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| RAS initiates the following actions:
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| : a. Trip LPSI pumps (this trip can be manually bypassed);
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| : b. Switch HPSI and containment spray pump suction from SIRWT to Containment Sump by opening sump CVs and closing SIRWT CVs;
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| : c. Adjust cooling water to component cooling heat exchangers;
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| : d. Open HPSI subcooling valve CV-3071 if the associated HPSI pump is operating;
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| : e. After containment sump valve CV-3030 is opened, open HPSI subcooling valve CV-3070 if the associated HPSI pump is operating;
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| : f. Re-positions CV-3001 and CV-3002 to a predetermined throttled position.
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| : g. Close containment spray valve CV-3001 if containment sump valve CV-3030 does not open.
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| Palisades Nuclear Plant B 3.3.3-10 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES APPLICABLE 3 Recirculation Actuation Signal (continued)
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| SAFETY ANALYSES (continued) The RAS signal is actuated by separate sensors from those which provide tank level indication. The allowable range of 21" to 27" above the tank floor corresponds to 1.1% to 3.3% indicated level.
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| Typically the actual setting is near the midpoint of the allowable range.
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| 4 Auxiliary Feedwater Actuation Signal An AFAS initiates feedwater flow to both steam generators if a low level is indicated in either steam generator.
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| The AFAS maintains a steam generator heat sink during the following events:
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| * MSLB;
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| * FWLB;
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| * LOCA; and
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| * Loss of feedwater.
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| : 5. Containment High Pressure Signal (CHP)
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| The CHP signal closes all containment isolation valves not required for ESF operation and starts containment spray (if SIS enabled), ensuring acceptable consequences during LOCAs, control rod ejection events, MSLBs, or FWLBs (inside containment).
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| CHP is actuated by two-out-of-four pressure switches for the associated train reaching their setpoints. CHP initiates the following actions:
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| : a. Containment Spray;
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| : b. Safety Injection Signal;
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| : c. Main Feedwater Isolation; Palisades Nuclear Plant B 3.3.3-11 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES APPLICABLE 5. Containment High Pressure Signal (CHP) (continued)
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| SAFETY ANALYSIS (continued) d. Main Steam Line Isolation;
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| : e. Control Room HVAC Emergency Mode; and
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| : f. Containment Isolation Valve Closure.
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| : 6. Containment High Radiation Signal (CHR)
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| CHR is actuated by two-out-of-four radiation monitors exceeding their setpoints. CHR initiates the following actions to ensure acceptable consequences following a LOCA or control rod ejection event:
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| : a. Control Room HVAC Emergency Mode;
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| : b. Containment Isolation Valve Closure; and
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| : c. Block automatic starting of ECCS pump room sump pumps.
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| During refueling operations, separate switch-selectable radiation monitors initiate CHR, as addressed by LCO 3.3.6.
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| : 7. Automatic Bypass Removal Functions The logic circuitry provides automatic removal of the Pressurizer Pressure Low and Steam Generator Pressure Low actuation signal bypasses. There are no assumptions in the safety analyses which assume operation of these automatic bypass removal circuits, and no analyzed events result in conditions where the automatic removal would be required to mitigate the event. The automatic removal circuits are required to assure that logic circuit bypasses will not be overlooked during a plant startup.
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| The ESF Instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| Palisades Nuclear Plant B 3.3.3-12 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES LCO The LCO requires all channel components necessary to provide an ESF actuation to be OPERABLE.
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| The Bases for the LCO on ESF Functions are addressed below.
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| : 1. Safety Injection Signal (SIS)
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| This LCO requires four channels of SIS Pressurizer Low Pressure to be OPERABLE in MODES 1, 2, and 3.
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| The setpoint was chosen so as to be low enough to avoid actuation during plant operating transients, but to be high enough to be quickly actuated by a LOCA or MSLB. The settings include an uncertainty allowance which is consistent with the settings assumed in the MSLB analysis (which bounds the settings assumed in the LOCA analysis).
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| : 2. Steam Generator Low Pressure Signal (SGLP)
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| This LCO requires four channels of Steam Generator Low Pressure Instrumentation for each SG to be OPERABLE in MODES 1, 2, and 3. However, as indicated in Table 3.3.3-1, Note (a), the SGLP Function is not required to be OPERABLE in MODES 2 or 3 if all Main Steam Isolation Valves (MSIVs) are closed and deactivated and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated or isolated by closed manual valves.
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| The setpoint was chosen to be low enough to avoid actuation during plant operation, but be close enough to full power operating pressure to be actuated quickly in the event of a MSLB. The setting includes an uncertainty allowance which is consistent with the setting used in the Reference 4 analysis.
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| Each SGLP logic is made up of output contacts from four pressure bistables from the associated SG. When the logic circuit is satisfied, two relays are energized to actuate steam and feedwater line isolation.
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| Palisades Nuclear Plant B 3.3.3-13 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES LCO 2. Steam Generator Low Pressure Signal (SGLP) (continued)
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| (continued)
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| This LCO applies to failures in the four sensor subsystems, including sensors, bistables, and associated equipment. Failures in the actuation subsystems are considered Actuation Logic failures and are addressed in LCO 3.3.4.
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| : 3. Recirculation Actuation Signal (RAS)
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| This LCO requires four channels of SIRWT Low Level to be OPERABLE in MODES 1, 2, and 3.
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| The setpoint was chosen to provide adequate water in the containment sump for HPSI pump net positive suction head following an accident, but prevent the pumps from running dry during the switchover.
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| The upper limit on the Allowable Value for this trip is set low enough to ensure RAS does not initiate before sufficient water is transferred to the containment sump. Premature recirculation could impair the reactivity control Function of safety injection by limiting the amount of boron injection. Premature recirculation could also damage or disable the recirculation system if recirculation begins before the sump has enough water.
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| The lower limit on the SIRWT Low Level trip Allowable Value is high enough to transfer suction to the containment sump prior to emptying the SIRWT.
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| : 4. Auxiliary Feedwater Actuation Signal (AFAS)
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| The AFAS logic actuates AFW to each SG on a SG Low Level in either SG.
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| The Allowable Value was chosen to assure that AFW flow would be initiated while the SG could still act as a heat sink and steam source, and to assure that a reactor trip would not occur on low level without the actuation of AFW.
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| Palisades Nuclear Plant B 3.3.3-14 Revised 01/29/2020
| |
| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES LCO 4. Auxiliary Feedwater Actuation Signal (AFAS) (continued)
| |
| (continued)
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| This LCO requires four channels for each steam generator of Steam Generator Low Level to be OPERABLE in MODES 1, 2, and 3.
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| : 5. Containment High Pressure Signal (CHP)
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| This LCO requires four channels of CHP to be OPERABLE for each of the associated ESF trains (left and right) in MODES 1, 2, 3 and 4.
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| The setpoint was chosen so as to be high enough to avoid actuation by containment temperature or atmospheric pressure changes, but low enough to be quickly actuated by a LOCA or a MSLB in the containment.
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| : 6. Containment High Radiation Signal (CHR)
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| This LCO requires four channels of CHR to be OPERABLE in MODES 1, 2, 3, and 4.
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| The setpoint is based on the maximum primary coolant leakage to the containment atmosphere allowed by LCO 3.4.13 and the maximum activity allowed by LCO 3.4.16. N16 concentration reaches equilibrium in containment atmosphere due to its short half-life, but other activity was assumed to build up. At the end of a 24 hour leakage period the dose rate is approximately 20 R/h as seen by the area monitors. A large leak could cause the area dose rate to quickly exceed the 20 R/h setting and initiate CHR.
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| : 7. Automatic Bypass Removal The automatic bypass removal logic removes the bypasses which are used during plant shutdown periods, for Pressurizer Low Pressure and Steam Generator Low Pressure actuation signals.
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| The setpoints were chosen to be above the setpoint for the associated actuation signal, but well below the normal operating pressures.
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| Palisades Nuclear Plant B 3.3.3-15 Revised 01/29/2020
| |
| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES LCO 7. Automatic Bypass Removal (continued)
| |
| (continued)
| |
| This LCO requires four channels of Pressurizer Low Pressure bypass removal and four channels for each steam generator of Steam Generator Low Pressure bypass removal, to be OPERABLE in MODES 1, 2, and 3.
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| APPLICABILITY All ESF Functions are required to be OPERABLE in MODES 1, 2, and 3. In addition, Containment High Pressure and Containment High Radiation are required to be operable in MODE 4.
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| In MODES 1, 2, and 3 there is sufficient energy in the primary and secondary systems to warrant automatic ESF System responses to:
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| * Close the main steam isolation valves to preclude a positive reactivity addition and containment overpressure;
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| * Actuate AFW to preclude the loss of the steam generators as a heat sink (in the event the normal feedwater system is not available);
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| * Actuate ESF systems to prevent or limit the release of fission product radioactivity to the environment by isolating containment and limiting the containment pressure from exceeding the containment design pressure during a design basis LOCA or MSLB; and
| |
| * Actuate ESF systems to ensure sufficient borated inventory to permit adequate core cooling and reactivity control during a design basis LOCA or MSLB accident.
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| The CHP and CHR Functions are required to be OPERABLE in MODE 4 to limit leakage of radioactive material from containment and limit operator exposure during and following a DBA.
| |
| The SGLP Function is not required to be OPERABLE in MODES 2 and 3, if all MSIVs are closed and deactivated and all MFRVs and MFRV bypass valves are either closed and deactivated or isolated by closed manual valves, since the SGLP Function is not required to perform any safety functions under these conditions.
| |
| Palisades Nuclear Plant B 3.3.3-16 Revised 01/29/2020
| |
| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES APPLICABILITY In lower MODES, automatic actuation of ESF Functions is not required, (continued) because adequate time is available for plant operators to evaluate plant conditions and respond by manually operating the ESF components.
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| LCO 3.3.6 addresses automatic Refueling CHR isolation during CORE ALTERATIONS or during movement of irradiated fuel.
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| In MODES 5 and 6, ESFAS initiated systems are either reconfigured or disabled for shutdown cooling operation. Accidents in these MODES are slow to develop and would be mitigated by manual operation of individual components.
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| ACTIONS The most common causes of channel inoperability are outright failure of loop components or drift of those loop components which is sufficient to exceed the tolerance provided in the plant setpoint analysis. Loop component failures are typically identified by the actuation of alarms due to the channel failing to the safe condition, during CHANNEL CHECKS (when the instrument is compared to the redundant channels), or during the CHANNEL FUNCTIONAL TEST (when an automatic component might not respond properly). Typically, the drift of the loop components is found to be small and results in a delay of actuation rather than a total loss of function. Excessive loop component drift would, most likely, be identified during a CHANNEL CHECK (when the instrument is compared to the redundant channels) or during a CHANNEL CALIBRATION (when instrument loop components are checked against reference standards).
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| Typically, the drift is small and results in a delay of actuation rather than a total loss of function. Determination of setpoint drift is generally made during the performance of a CHANNEL FUNCTIONAL TEST when the process instrument is set up for adjustment to bring it to within specification. If the actual trip setpoint is not within the Allowable Value in Table 3.3.3-1, the channel is inoperable and the appropriate Condition(s) are entered.
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| In the event a channel's trip setpoint is found nonconservative with respect to the Allowable Value in Table 3.3.3-1, or the sensor, instrument loop, signal processing electronics, or ESF bistable is found inoperable, then all affected Functions provided by that channel must be declared inoperable and the plant must enter the Condition statement for the particular protection Function affected.
| |
| Palisades Nuclear Plant B 3.3.3-17 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES ACTIONS When the number of inoperable channels in a trip Function exceeds (continued) those specified in any related Condition associated with the same trip Function, then the plant is outside the safety analysis. Therefore, LCO 3.0.3 should be immediately entered if applicable in the current MODE of operation.
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| A Note has been added to clarify the application of the Completion Time rules. The Conditions of this Specification may be entered independently for each Function in Table 3.3.3-1. Completion Times for the inoperable channel of a Function will be tracked separately.
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| A.1 Condition A applies to the failure of a single bistable or associated instrumentation channel of one or more input parameters in each ESF Function except the RAS Function. Since the bistable and associated instrument channel combine to perform the actuation function, the Condition is also appropriate if both the bistable and associated instrument channel are inoperable.
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| ESF coincidence logic is normally two-out-of-four. If one ESF channel is inoperable, startup or power operation is allowed to continue as long as action is taken to restore the design level of redundancy.
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| If one ESF channel is inoperable, startup or power operation is allowed to continue, providing the inoperable channel actuation bistable is placed in trip within 7 days. The provision of four trip channels allows one channel to be inoperable in a non-trip condition up to the 7 day Completion Time allotted to place the channel in trip. Operating with one failed channel in a non-trip condition during operations, places the ESF Actuation Logic in a two-out-of-three coincidence logic.
| |
| If the failed channel cannot be restored to OPERABLE status in 7 days, the associated bistable is placed in a tripped condition. This places the function in a one-out-of-three configuration.
| |
| Palisades Nuclear Plant B 3.3.3-18 Revised 01/29/2020
| |
| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES ACTIONS A.1 (continued)
| |
| (continued)
| |
| In this configuration, common cause failure of the dependent channel cannot prevent ESF actuation. The 7 day Completion Time is based upon operating experience, which has demonstrated that a random failure of a second channel occurring during the 7 day period is a low probability event.
| |
| Condition A is modified by a Note which indicates it is not applicable to the SIRWT Low Level Function.
| |
| B.1 and B.2 Condition B applies to the failure of two channels in any of the ESF Functions except the RAS Function.
| |
| With two inoperable channels, one channel actuation device must be placed in trip within the 8 hour Completion Time. Eight hours is allowed for this action since it must be accomplished by a circuit modification, or by removing power from a circuit component. With one channel of protective instrumentation inoperable, the ESF Actuation Logic Function is in two-out-of-three logic, but with another channel inoperable the ESF may be operating with a two-out-of-two logic. This is outside the assumptions made in the analyses and should be corrected. To correct the problem, the second channel is placed in trip. This places the ESF in a one-out-of-two logic. If any of the other OPERABLE channels receives a trip signal, ESF actuation will occur.
| |
| One of the failed channels must be restored to OPERABLE status within 7 days, and the provisions of Condition A still applied to the remaining inoperable channel. Therefore, the channel that is still inoperable after completion of Required Action B.2 must be placed in trip if more than 7 days has elapsed since the channels initial failure.
| |
| Palisades Nuclear Plant B 3.3.3-19 Revised 01/29/2020
| |
| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES ACTIONS B.1 and B.2 (continued)
| |
| (continued)
| |
| Condition B is modified by a Note which indicates that it is not applicable to the SIRWT Low Level Function.
| |
| C.1 and C.2 Condition C applies to one RAS SIRWT Low Level channel inoperable.
| |
| The SIRWT low level circuitry is arranged in a "1-out-of-2 taken twice" logic rather than the more frequently used 2-out-of-4 logic. Therefore, Required Action C.1 differs from other ESF functions. With a bypassed SIRWT low level channel, an additional failure might disable automatic RAS, but would not initiate a premature RAS. With a tripped channel, an additional failure could cause a premature RAS, but would not disable the automatic RAS.
| |
| Since considerable time is available after initiation of SIS until RAS must be initiated, and since a premature RAS could damage the ESF pumps, it is preferable to bypass an inoperable channel and risk loss of automatic RAS than to trip a channel and risk a premature RAS.
| |
| The Completion Time of 8 hours allowed is reasonable because the Required Action involves a circuit modification.
| |
| Required Action C.2 requires that the inoperable channel be restored to OPERABLE status within 7 days. The Completion Time is reasonable based upon operating experience, which has demonstrated that a random failure of a second channel occurring during the 7 day period is a low probability event.
| |
| Palisades Nuclear Plant B 3.3.3-20 Revised 01/29/2020
| |
| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES ACTIONS D.1 and D.2 (continued)
| |
| If the Required Actions and associated Completion Times of Condition A, B, or C are not met for Functions 1, 2, 3, 4, or 7, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
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| E.1 and E.2 If the Required Actions and associated Completion Times of Condition A, B, or C are not met for Functions 5 or 6, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
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| SURVEILLANCE The SRs for any particular ESF Function are found in the SRs column REQUIREMENTS of Table 3.3.3-1 for that Function. Most functions are subject to CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION.
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| SR 3.3.3.1 A CHANNEL CHECK is performed once every 12 hours on each ESF input channel which is provided with an indicator to provide a qualitative assurance that the channel is working properly and that its readings are within limits. A CHANNEL CHECK is not performed on the CHP and SIRWT Low Level channels because they have no associated control room indicator.
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| Palisades Nuclear Plant B 3.3.3-21 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES SURVEILLANCE SR 3.3.3.1 (continued)
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| REQUIREMENTS (continued) Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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| Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off scale during times when Surveillance is required, the CHANNEL CHECK will only verify that they are off scale in the same direction.
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| Offscale low current loop channels are verified to be reading at the bottom of the range and not failed downscale.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.3.3-22 Amendment No. 271 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES SURVEILLANCE SR 3.3.3.2 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed to ensure the entire channel will perform its intended function when needed. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay.
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| This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests.
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| This test is required to be performed on ESF input channels provided with on-line testing capability. It is not required for the SIRWT Low Level channels since they have no built in test capability. The CHANNEL FUNCTIONAL TEST for SIRWT Low Level channels is performed as part of the required CHANNEL CALIBRATION.
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| The CHANNEL FUNCTIONAL TEST tests the individual channels using an analog test input to each bistable.
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| Any setpoint adjustment shall be consistent with the assumptions of the current setpoint analysis.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.3.3.3 CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive surveillances. CHANNEL CALIBRATIONS must be performed consistent with the setpoint analysis.
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| Palisades Nuclear Plant B 3.3.3-23 Amendment No. 271 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.3 ESF Instrumentation B 3.3.3 BASES SURVEILLANCE SR 3.3.3.3 (continued)
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| REQUIREMENTS (continued) The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the extension analysis. The requirements for this review are outlined in Reference 5.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Chapter 7
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| : 2. 10 CFR 50, Appendix A
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| : 3. IEEE Standard 279-1971
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| : 4. FSAR, Chapter 14
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| : 5. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989 Palisades Nuclear Plant B 3.3.3-24 Amendment No. 271 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.4 ESF Logic and Manual Initiation B 3.3.4 B 3.3 INSTRUMENTATION B 3.3.4 Engineered Safety Features (ESF) Logic and Manual Initiation BASES BACKGROUND The ESF Instrumentation initiates necessary safety systems, based upon the values of selected plant parameters, to protect against violating core design limits and the Primary Coolant System (PCS) pressure boundary and to mitigate accidents.
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| The ESF circuitry generates the following signals listed below when the monitored variables reach levels that are indicative of conditions requiring protective action. The inputs to each ESF Actuation Signal are also listed.
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| : 1. Safety Injection Signal (SIS);
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| : a. Containment High Pressure (CHP)
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| : b. Pressurizer Low Pressure
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| : 2. Steam Generator Low Pressure Signal (SGLP)
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| : a. Steam Generator A Low Pressure
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| : b. Steam Generator B Low Pressure
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| : 3. Recirculation Actuation Signal (RAS);
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| : a. Safety Injection Refueling Water Tank (SIRWT) Low Level
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| : 4. Auxiliary Feedwater Actuation Signal (AFAS)
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| : a. Steam Generator A Low Level
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| : b. Steam Generator B Low Level Palisades Nuclear Plant B 3.3.4-1 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.4 ESF Logic and Manual Initiation B 3.3.4 BASES BACKGROUND 5. Containment High Pressure Signal (CHP);
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| (continued)
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| : a. Containment High Pressure - Left Train
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| : b. Containment High Pressure - Right Train
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| : 6. Containment High Radiation Signal (CHR)
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| : a. Containment High Radiation In the above list of actuation signals, the CHP and RAS are derived from pressure and level switches, respectively.
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| Equipment actuated by each of the above signals is identified in the FSAR, Chapter 7 (Ref. 1).
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| The ESF circuitry, with the exception of RAS, employs two-out-of-four logic. Four independent measurement channels are provided for each function used to generate ESF actuation signals. When any two channels of the same function reach their setpoint, actuating relays initiate the protective actions. Two separate and redundant trains of actuating relays, each powered from separate power supplies, are utilized. These separate relay trains operate redundant trains of ESF equipment. The actuation relays are considered part of the actuation logic addressed by this LCO.
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| RAS logic consists of output contacts of the relays actuated by the SIRWT Low Level switches arranged in a "one-out-of-two taken twice" logic. The contacts are arranged so that at least one low level signal powered from each station battery is required to initiate RAS. Loss of a single battery, therefore, cannot either cause or prevent RAS initiation.
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| The sensor subsystem, including individual channel bistables, is addressed in LCO 3.3.3, Engineered Safety Features (ESF)
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| Instrumentation. This LCO addresses the actuation subsystem manual actuation, and downstream components used to actuate the individual ESF functions, as defined in the following section.
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| Palisades Nuclear Plant B 3.3.4-2 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.4 ESF Logic and Manual Initiation B 3.3.4 BASES BACKGROUND ESF Logic (continued)
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| Each of the six ESF actuation signals in Table 3.3.4-1 operates two trains of actuating relays. Each train is capable of initiating the ESF equipment to meet the minimum requirements to provide all functions necessary to operate the system associated with the plant's capability to cope with abnormal events.
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| The SGLP logic circuitry includes provisions such that the SGLP automatic actuation Function may be bypassed if three-out-of-four Steam Generator (SG) pressure channels are below a bypass permissive setpoint. Similarly, the SIS automatic actuation on Pressurizer Low Pressure may be bypassed when three-out-of-four channels are below a permissive setpoint. This actuation bypassing is performed when the ESF Functions are no longer required for protection. These actuation bypasses are enabled manually when the permissive conditions are satisfied.
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| All actuation bypasses are automatically removed when enabling conditions are no longer satisfied. If an SIS or SGLP automatic actuation channel is bypassed, other than as allowed by Table 3.3.4-1, the channel cannot perform its required safety function and must be considered to be inoperable.
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| Testing of a major portion of the ESF circuits is accomplished while the plant is at power. More extensive sequencer and load testing may be done with the reactor shut down. The test circuits are designed to test the redundant circuits separately such that the correct operation of each circuit may be verified by either equipment operation or by sequence lights.
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| Manual Initiation Manual ESF initiation capability is provided to permit the operator to manually actuate an ESF System when necessary.
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| Two control room mounted manual actuation switches are provided for SIS actuation, one for each train. Each SIS manual actuation switch affects one actuation channel, which actuates one train of SIS equipment.
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| There are no single manual controls provided to actuate CHP, however, CHP may be manually initiated using individual component controls.
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| Palisades Nuclear Plant B 3.3.4-3 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.4 ESF Logic and Manual Initiation B 3.3.4 BASES BACKGROUND Manual Initiation (continued)
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| (continued)
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| Two control room mounted manual actuation switches are provided for CHR actuation, each switch affects both actuation channels, which actuates both CHR trains.
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| There are no single manual controls provided to actuate SGLP, however, SGLP may be manually initiated using individual component controls.
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| RAS is actuated by manually actuating the circuit "Test" switch, however, RAS may also be manually initiated using individual component controls.
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| Manual actuation of AFW may be accomplished through pushbutton actuation of each AFAS channel or by use of individual pump and valve controls. Each automatic AFAS actuation channel starts the AFW pumps in their starting sequence (if P-8A fails to start, a P-8C start signal is generated, and if P-8C also fails to start, a P-8B start signal is generated) and opens the associated flow control valves.
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| APPLICABLE Each of the analyzed accidents can be detected by one or more ESF SAFETY ANALYSES Functions. One of the ESF Functions is the primary actuation signal for that accident. An ESF Function may be the primary actuation signal for more than one type of accident. An ESF Function may also be a secondary, or backup, actuation signal for one or more other accidents.
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| Functions such as Manual Initiation, not specifically credited in the accident analysis, serve as backups to Functions and are part of the NRC staff approved licensing basis for the plant.
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| The manual initiation is not required by the accident analysis. The ESF logic must function in all situations where the ESF function is required (as discussed in the Bases for LCO 3.3.3).
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| Each ESF Function and its associated safety analyses are discussed in the Applicable Safety Analyses section of the Bases for LCO 3.3.3, ESF Instrumentation.
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| The ESF satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| Palisades Nuclear Plant B 3.3.4-4 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.4 ESF Logic and Manual Initiation B 3.3.4 BASES LCO The LCO requires that all components necessary to provide an ESF actuation be OPERABLE.
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| The Bases for the LCO on ESF automatic actuation Functions are addressed in LCO 3.3.3. Those associated with the Manual Initiation or Actuation Logic are addressed below.
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| ESF Logic and Manual Initiation Functions are required to be OPERABLE in MODES 1, 2, and 3, or in MODES 1, 2, 3, and 4, as appropriate, when the associated automatic initiation channels addressed by LCO 3.3.3 are required.
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| : 1. Safety Injection Signal (SIS)
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| SIS is actuated by manual initiation, by a CHP signal, or by two-out-of-four Pressurizer Low Pressure channels decreasing below the setpoint. Each Manual Initiation channel consists of one pushbutton which directly starts the SIS actuation logic for the associated train. Each SIS logic train is also actuated by a contact pair on one of the CHP initiation relays for the associated CHP train.
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| : a. Manual Initiation This LCO requires two channels of SIS Manual Initiation to be OPERABLE.
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| : b. Actuation Logic This LCO requires two channels of SIS Actuation Logic to be OPERABLE. Failures in the actuation subsystems are addressed in this LCO.
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| : c. CHP Logic Trains The CHP initiation relay (5P-x) input to the SIS logic is considered part of the SIS logic. Two channels, one per SIS train, must be OPERABLE.
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| Palisades Nuclear Plant B 3.3.4-5 Revised 01/29/2020
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| INSERT Bases 3.3.4 ESF Logic and Manual Initiation B 3.3.4 BASES LCO 1. Safety Injection Signal (SIS) (continued)
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| (continued)
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| : d. Automatic Bypass Removal This LCO requires two channels of the automatic bypass removal logic for SIS Pressurizer Low Pressure to be OPERABLE. If an SIS automatic actuation channel is bypassed, other than as allowed by Table 3.3.4-1, the channel cannot perform its required safety function and must be considered to be inoperable.
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| As indicated by footnote (a), the Pressurizer Low Pressure logic train for each SIS train can be bypassed when three-out-of-four channels indicate below 1700 psia. This bypass prevents undesired actuation of SIS during a normal plant cooldown. The bypass signal is automatically removed when two-out-of-four channels exceed the setpoint, in accordance with the philosophy of removing bypasses when the enabling conditions are no longer satisfied.
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| The bypass permissive is set low enough so as not to be enabled during normal plant operation, but high enough to allow bypassing prior to reaching the trip setpoint.
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| : 2. Steam Generator Low Pressure Signal (SGLP)
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| : a. Manual Initiation This LCO requires two channels of SGLP Manual Initiation to be OPERABLE. As indicated by footnote (c), there is no manual control which actuates the SGLP logic circuits. The actuated components must be individually actuated using control room manual controls.
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| : b. Actuation Logic This LCO requires two channels of SGLP Actuation Logic to be OPERABLE, one for each SG.
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| Palisades Nuclear Plant B 3.3.4-6 Revised 01/29/2020
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| INSERT Bases 3.3.4 ESF Logic and Manual Initiation B 3.3.4 BASES LCO 2. Steam Generator Low Pressure Signal (SGLP) (continued)
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| (continued)
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| : c. Automatic Bypass Removal This LCO requires two channels, one for each SG, of the SGLP automatic bypass removal logic to be OPERABLE. If an SGLP automatic actuation channel is bypassed, other than as allowed by Table 3.3.4-1, the channel cannot perform its required safety function and must be considered to be inoperable.
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| As indicated by footnote (b), the SGLP from each SG may be bypassed when three-out-of-four channels indicate below 565 psia. This bypass prevents undesired actuation during a normal plant cooldown. The bypass signal is automatically removed when two-out-of-four channels exceed the setpoint, in accordance with the philosophy of removing bypasses when the enabling conditions are no longer satisfied.
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| The bypass permissive is set low enough so as not to be enabled during normal plant operation, but high enough to allow bypassing prior to reaching the trip setpoint.
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| : 3. Recirculation Actuation Signal (RAS)
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| : a. Manual Initiation This LCO requires two channels of RAS Manual Initiation to be OPERABLE. RAS is actuated by manually actuating the circuit "Test" switches.
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| : b. Actuation Logic This LCO requires two channels of RAS Actuation Logic to be OPERABLE.
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| Palisades Nuclear Plant B 3.3.4-7 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.4 ESF Logic and Manual Initiation B 3.3.4 BASES LCO 4. Auxiliary Feedwater Actuation Signal (AFAS)
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| (continued)
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| : a. Manual Initiation This LCO requires two channels of AFAS Manual Initiation to be OPERABLE. Each train of AFAS may be manually initiated with either of two sets of controls. Only one set of manual controls is required to be OPERABLE for each AFW train. One set of controls are the pushbuttons provided to actuate each train on the C-11 panel; the other set of controls are those manual controls provided on C-01 for each AFW pump and flow control valve.
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| : b. Actuation Logic This LCO requires two channels of AFAS Actuation Logic to be OPERABLE.
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| : 5. Containment High Pressure Signal (CHP)
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| : a. Manual Initiation As indicated by footnote (c), this LCO requires the manual controls necessary to actuate those valves and components actuated by an automatic CHP to be OPERABLE.
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| : b. Actuation Logic This LCO requires two channels of CHP Actuation Logic to be OPERABLE.
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| : 6. Containment High Radiation Signal (CHR)
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| : a. Manual Initiation This LCO requires two channels of CHR Manual Initiation to be OPERABLE. Pushbuttons are available for manual actuation of each CHR logic train.
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| Palisades Nuclear Plant B 3.3.4-8 Revised 01/29/2020
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| INSERT Bases 3.3.4 ESF Logic and Manual Initiation B 3.3.4 BASES LCO 6. Containment High Radiation Signal (CHR) (continued)
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| (continued)
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| : b. Actuation Logic This LCO requires two channels of CHR Actuation Logic to be OPERABLE.
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| APPLICABILITY ESF Functions are required to be OPERABLE in MODES 1, 2, and 3 or MODES 1, 2, 3, and 4 as specified in Table 3.3.4-1. In MODES 1, 2, and 3, there is sufficient energy in the primary and secondary systems to warrant automatic ESF System responses to:
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| * Close the MSIVs to preclude a positive reactivity addition and containment overpressure;
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| * Actuate AFW to preclude the loss of the steam generators as a heat sink (in the event the normal feedwater system is not available);
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| * Actuate ESF systems to prevent or limit the release of fission product radioactivity to the environment by isolating containment and limiting the containment pressure from exceeding the containment design pressure during a design basis LOCA or MSLB; and
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| * Actuate ESF systems to ensure sufficient borated inventory to permit adequate core cooling and reactivity control during a design basis LOCA or MSLB accident.
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| The CHP and CHR Functions are also required to be OPERABLE in MODE 4 to limit leakage of radioactive material from containment and limit operator exposure during and following a DBA.
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| The SGLP Function is not required to be OPERABLE in MODES 2 and 3, if all MSIVs are closed and deactivated and all MFRVs and MFRV bypass valves are either closed and deactivated or isolated by closed manual valves, since the SGLP Function is not required to perform any safety function under these conditions.
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| Palisades Nuclear Plant B 3.3.4-9 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.4 ESF Logic and Manual Initiation B 3.3.4 BASES APPLICABILITY In MODES 5 and 6, automatic actuation of ESF Functions is not (continued) required, because adequate time is available for plant operators to evaluate plant conditions and respond by manually operating the ESF components if required. In these MODES, ESF initiated systems are either reconfigured or disabled for shutdown cooling operation.
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| Accidents in these MODES are slow to develop and would be mitigated by manual operation of individual components.
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| ACTIONS When the number of inoperable channels in a trip Function exceeds those specified in any related Condition associated with the same trip Function, then the plant is outside the safety analysis. Therefore, LCO 3.0.3 should be immediately entered, if applicable in the current MODE of operation.
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| A Note has been added to the ACTIONS to clarify the application of the Completion Time rules. The Conditions of this Specification may be entered independently for each Function in Table 3.3.4-1 in the LCO.
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| Completion Times for the inoperable channel of a Function will be tracked separately.
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| A.1 Condition A applies to one Manual Initiation, Bypass Removal, or Actuation Logic channel inoperable. The channel must be restored to OPERABLE status to restore redundancy of the ESF Function. The 48 hour Completion Time is commensurate with the importance of avoiding the vulnerability of a single failure in the only remaining OPERABLE channel.
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| B.1 and B.2 If two Manual Initiation, Bypass Removal, or Actuation Logic channels are inoperable for Functions 1, 2, 3, or 4, or if the Required Action and associated Completion Time of Condition A cannot be met for Function 1, 2, 3, or 4, the reactor must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.3.4-10 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.4 ESF Logic and Manual Initiation B 3.3.4 BASES ACTIONS C.1 and C.2 (continued)
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| Condition C is entered when one or more Functions have two Manual Initiation or Actuation Logic channels inoperable for Functions 5 or 6, or when the Required Action and associated Completion Time of Condition A are not met for Functions 5 or 6. If Required Action A.1 cannot be met within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
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| SURVEILLANCE SR 3.3.4.1 REQUIREMENTS A SIS actuation functional test of each channel is performed using the installed control room test switches and test circuits for both with standby power and without standby power. When testing the with standby power circuits, proper operation of the SIS-X relays must be verified; when testing the without standby power circuits, proper operation of the DBA sequencer and the associated logic circuit must be verified. The test circuits are designed to block those SIS functions, such as injection of concentrated boric acid, which would interfere with plant operation.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.3.4.2 A CHANNEL FUNCTIONAL TEST of each AFAS Actuation Logic Channel is performed to ensure the channel will perform its intended function when needed. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests.
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| Palisades Nuclear Plant B 3.3.4-11 Amendment No. 271 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.4 ESF Logic and Manual Initiation B 3.3.4 BASES SURVEILLANCE SR 3.3.4.2 (continued)
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| REQUIREMENTS (continued) Instrumentation channel tests are addressed in LCO 3.3.3.
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| SR 3.3.4.2 addresses Actuation Logic tests of the AFAS using the installed test circuits.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.3.4.3 A CHANNEL FUNCTIONAL TEST is performed on the manual ESF initiation channels, Actuation Logic channels, and bypass removal channels for specified ESF Functions. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
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| This Surveillance verifies that the required channels will perform their intended functions when needed.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Chapter 7
| |
| : 2. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989 Palisades Nuclear Plant B 3.3.4-12 Amendment No. 271 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.5 DG - UV Start B 3.3.5 B 3.3 INSTRUMENTATION B 3.3.5 Diesel Generator (DG) - Undervoltage Start (UV Start)
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| BASES BACKGROUND The DGs provide a source of emergency power when offsite power is either unavailable or insufficiently stable to allow safe plant operation.
| |
| Undervoltage protection will generate a UV Start in the event a Loss of Voltage or Degraded Voltage condition occurs. There are two UV Start Functions for each 2.4 kV vital bus.
| |
| Undervoltage protection and load shedding features for safety-related buses at the 2,400 V and lower voltage levels are designed in accordance with 10 CFR 50, Appendix A, General Design Criterion 17 (Ref. 1) and the following features:
| |
| : 1. Two levels of automatic undervoltage protection from loss or degradation of offsite power sources are provided. The first level (loss of voltage) provides normal loss of voltage protection. The second level of protection (degraded voltage) has voltage and time delay set points selected for automatic trip of the offsite sources to protect safety-related equipment from sustained degraded voltage conditions at all bus voltage levels.
| |
| Coincidence logic is provided to preclude spurious trips.
| |
| : 2. The undervoltage protection system automatically prevents load shedding of the safety-related buses when the emergency generators are supplying power to the safeguards loads.
| |
| : 3. Control circuits for shedding of Class 1E and non-Class 1E loads during a Loss of Coolant Accident (LOCA) themselves are Class 1E or are separated electrically from the Class 1E portions.
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| Palisades Nuclear Plant B 3.3.5-1 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.5 DG - UV Start B 3.3.5 BASES BACKGROUND Description (continued)
| |
| Each 2,400 V Bus (1C and 1D) is equipped with two levels of undervoltage protection relays (Ref. 2). The first level (Loss of Voltage Function) relays 127-1 and 127-2 are set at approximately 77% of rated voltage with an inverse time relay. One of these relays measures voltage on each of the three phases. They protect against sudden loss of voltage as sensed on the corresponding bus using a three-out-of-three coincidence logic. The actuation of the associated auxiliary relays will trip the associated bus incoming circuit breakers, start its associated DG, initiate bus load shedding, and activate annunciators in the control room. The DG circuit breaker is closed automatically upon establishment of satisfactory voltage and frequency by the use of associated voltage sensing relay 127D-1 or 127D-2.
| |
| The second level of undervoltage protection (Degraded Voltage Function) relays 127-7 and 127-8 are set at approximately 92% of rated voltage, with one relay monitoring each of the three phases. These voltage sensing relays protect against sustained degraded voltage conditions on the corresponding bus using a three-out-of-three coincidence logic. These relays have an internal (built-in) 0.65 second time delay, after which the associated DG receives a start signal and annunciators in the control room are actuated. If the bus undervoltage condition exists for an additional six seconds (due to a six-second time delay relay), the associated bus incoming circuit breakers will be tripped and a bus load shed will be initiated.
| |
| Trip Setpoints The trip setpoints are based on the analytical limits discussed in References 3, 4, 5, 7, 9, and 10. The selection of these trip setpoints is such that adequate protection is provided when all sensor and processing time delays are taken into account. To allow for calibration tolerances, instrumentation uncertainties, and instrument drift, setpoints specified in SR 3.3.5.2 are conservatively adjusted with respect to the analytical limits. A detailed analysis of the degraded voltage protection is provided in References 3 and 4.
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| The specified setpoints will ensure that the consequences of accidents will be acceptable, providing the plant is operated from within the LCOs at the onset of the accident and the equipment functions as designed.
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| Palisades Nuclear Plant B 3.3.5-2 Revised 01/29/2020
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| | |
| INSERT Bases 3.3.5 DG - UV Start B 3.3.5 BASES APPLICABLE The DG - UV Start is required for Engineered Safety Features (ESF)
| |
| SAFETY ANALYSES systems to function in any accident with a loss of offsite power. Its design basis is that of the ESF Systems.
| |
| Accident analyses credit the loading of the DG based on a loss of offsite power during a LOCA. The diesel loading has been included in the delay time associated with each safety system component requiring DG supplied power following a loss of offsite power. This delay time includes contributions from the DG start, DG loading, and Safety Injection System component actuation.
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| The required channels of UV Start, in conjunction with the ESF systems powered from the DGs, provide plant protection in the event of any of the analyzed accidents discussed in Reference 6, in which a loss of offsite power is assumed. UV Start channels are required to meet the redundancy and testability requirements of GDC 21 in 10 CFR 50, Appendix A (Ref. 1).
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| The delay times assumed in the safety analysis for the ESF equipment include the 10 second DG start delay and the appropriate sequencing delay, if applicable. The response times for ESFAS actuated equipment include the appropriate DG loading and sequencing delay.
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| The DG - UV Start channels satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| LCO The LCO for the DG - UV Start requires that three channels per bus of each UV Start instrumentation Function be OPERABLE when the associated DG is required to be OPERABLE. The UV Start supports safety systems associated with ESF actuation.
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| The Bases for the trip setpoints are as follows:
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| The voltage trip setpoint is set low enough such that spurious trips of the offsite source due to operation of the undervoltage relays are not expected for any combination of plant loads and normal grid voltages.
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| Palisades Nuclear Plant B 3.3.5-3 Revised 01/29/2020
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| INSERT Bases 3.3.5 DG - UV Start B 3.3.5 BASES LCO This setpoint at the 2,400 V bus and reflected down to the 480 V buses (continued) has been verified through an analysis to be greater than the minimum allowable motor voltage (90% of nominal voltage). Motors are the most limiting equipment in the system. MCC contactor pickup and drop-out voltage is also adequate at the setpoint values. The analysis ensures that the distribution system is capable of starting and operating all safety-related equipment within the equipment voltage rating at the allowed source voltages. The power distribution system model used in the analysis has been verified by actual testing (Refs. 5 and 7).
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| The time delays involved will not cause any thermal damage as the setpoints are within voltage ranges for sustained operation. They are long enough to preclude trip of the offsite source caused by the starting of large motors and yet do not exceed the time limits of ESF actuation assumed in FSAR Chapter 14 (Ref. 6) and validated by Reference 8.
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| The time delays also will not result in failure of safety related equipment due to sustained degraded voltage conditions (Reference 9).
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| Calibration of the undervoltage relays verify that the time delays are sufficient to avoid spurious trips.
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| APPLICABILITY The DG - UV Start actuation Function is required to be OPERABLE whenever the associated DG is required to be OPERABLE per LCO 3.8.1, AC Sources - Operating, or LCO 3.8.2, AC Sources -
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| Shutdown, so that it can perform its function on a loss of power or degraded power to the vital bus.
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| ACTIONS A DG - UV Start channel is inoperable when it does not satisfy the OPERABILITY criteria for the channel's Function.
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| In the event a channel's trip setpoint is found nonconservative with respect to the specified setpoint, or the channel is found inoperable, then all affected Functions provided by that channel must be declared inoperable and the LCO Condition entered. The required channels are specified on a per DG basis.
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| Palisades Nuclear Plant B 3.3.5-4 Revised 01/29/2020
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| INSERT Bases 3.3.5 DG - UV Start B 3.3.5 BASES ACTIONS A.1 (continued)
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| Condition A applies if one or more of the three phase UV sensors or relay logic is inoperable for one or more Functions (Degraded Voltage or Loss of Voltage) per DG bus.
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| The affected DG must be declared inoperable and the appropriate Condition(s) entered. Because of the three-out-of-three logic in both the Loss of Voltage and Degraded Voltage Functions, the appropriate means of addressing channel failure is declaring the DG inoperable, and effecting repair in a manner consistent with other DG failures.
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| Required Action A.1 ensures that Required Actions for the affected DG inoperabilities are initiated. Depending upon plant MODE, the actions specified in LCO 3.8.1 or LCO 3.8.2, as applicable, are required immediately.
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| SURVEILLANCE SR 3.3.5.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST is performed on each UV Start logic channel to ensure that the logic channel will perform its intended function when needed. The Undervoltage sensing relays are tested by SR 3.3.5.2. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.3.5-5 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.5 DG - UV Start B 3.3.5 BASES SURVEILLANCE SR 3.3.5.2 REQUIREMENTS (continued) A CHANNEL CALIBRATION verifies the accuracy of each component within the instrument channel. This includes calibration of the undervoltage relays and demonstrates that the equipment falls within the specified operating characteristics defined by the manufacturer.
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| The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy.
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| The Degraded Voltage Function time delay setpoints reflect the voltage sensing relay nominal 0.65-second time delay, and the voltage sensing relay nominal 0.65-second time delay combined with the nominal six-second delay due to the external time delay relay.
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| CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the setpoint analysis.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. 10 CFR 50, Appendix A GDCs 17 and 21
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| : 2. FSAR, Section 8.6
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| : 3. Analysis EA-ELEC-VOLT-033
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| : 4. Analysis EA-ELEC-VOLT-034
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| : 5. Analysis EA-ELEC-EDSA-04
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| : 6. FSAR, Chapter 14
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| : 7. Analysis EA-ELEC-EDSA-03
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| : 8. Analysis A-NL-92-111
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| : 9. Analysis 0098-0189-CALC-001-PLP
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| : 10. Analysis EA-EC11464-01 Palisades Nuclear Plant B 3.3.5-6 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.6 Refueling CHR Instrumentation B 3.3.6 B 3.3 INSTRUMENTATION B 3.3.6 Refueling Containment High Radiation (CHR) Instrumentation BASES BACKGROUND This LCO addresses Refueling CHR actuation. When the Refueling CHR Monitors are enabled by their keylock switches, a CHR actuation may be automatically initiated by a signal from either of the Refueling CHR monitors or manually by actuation of either of the control room CHR Manual Initiate pushbuttons (pushing either Manual Initiate pushbutton will actuate both trains of CHR). A CHR signal initiates the following actions:
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| : a. Control Room HVAC Emergency Mode;
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| : b. Containment Isolation Valve Closure; and
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| : c. Block automatic starting of Engineered Safeguards pump room sump pumps.
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| The Refueling CHR signal provides automatic containment isolation valve closure during refueling operations, using two radiation monitors located in the refueling area of the containment (elevation 649 ft). The monitors are part of the plant area monitoring system and employ one-out-of-two logic for isolation. During normal operation these monitors are disconnected from the CHR relays and will not initiate a CHR signal. A switch is provided to connect the Refueling CHR monitors into the CHR actuation circuit, so that CHR actuation can be initiated by these monitors during refueling.
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| Palisades Nuclear Plant B 3.3.6-1 Revised 01/29/2020
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| INSERT Bases 3.3.6 Refueling CHR Instrumentation B 3.3.6 BASES BACKGROUND Each monitor actuates one train of CHR logic when containment (continued) radiation exceeds the setpoint. Two separate keylock switches, one per train, enable the Refueling CHR input to the CHR logic when switched to the "Refueling" position. Each Refueling CHR channel, associated keylock switch, and initiation circuit input to the CHR logic thus forms a one-out-of-one logic input to its associated CHR actuation logic train. The Refueling CHR isolation instrumentation is separate from the CHR instrumentation addressed in LCO 3.3.3, "ESF Instrumentation. However, the Refueling CHR Instrumentation does operate the same CHR actuation relays as the two-out-of-four CHR logic addressed in LCO 3.3.4. This LCO is not included in LCOs 3.3.3 and 3.3.4 because of the differences in APPLICABILITY and the single channel nature of the Refueling CHR input. The Refueling CHR signal performs the automatic containment isolation valve closure Function during refueling operations required by LCO 3.9.3, Containment Penetrations.
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| The Refueling CHR Instrumentation provides protection from release of radioactive gases and particulates from the containment in the event a fuel assembly should be severely damaged during handling.
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| The Refueling CHR Instrumentation will detect any abnormal radiation levels in the containment refueling area and will initiate purge valve closure to limit the release of radioactivity to the environment. The containment purge supply and exhaust valves are closed on a CHR signal when a high radiation level in containment is detected.
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| The Refueling CHR Instrumentation includes two independent, redundant actuation subsystems, as described above. Reference 1 describes the Refueling CHR circuitry.
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| Trip Setpoint No required setpoint is specified because these instruments are not assumed to function by any of the safety analyses. Typically, the instruments are set at about 25 mR/hr above expected background for planned operations (including movement of the reactor vessel head or internals).
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| Palisades Nuclear Plant B 3.3.6-2 Revised 01/29/2020
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| INSERT Bases 3.3.6 Refueling CHR Instrumentation B 3.3.6 BASES APPLICABLE The Refueling CHR Instrumentation isolates containment in the event SAFETY ANALYSES that area radiation exceeds an established level following a fuel handling accident. This ensures the radioactive materials are not released directly to the environment and significantly reduces the offsite doses from those calculated by the safety analyses, which do not credit containment isolation (Ref. 2). Either way, i.e., with or without containment isolation, the offsite doses remain within applicable 10 CFR 50.67 limits.
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| The Refueling CHR Instrumentation is not required by the fuel handling accident analyses to maintain offsite doses within applicable 10 CFR 50.67 limits, but containment isolation would provide a significant reduction of the resulting offsite doses. Therefore, the Refueling CHR Instrumentation satisfies the requirements of Criterion 4 of 10 CFR 50.36(c)(2).
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| LCO The LCO for the Refueling CHR Instrumentation requires that two channels of refueling CHR instrumentation and two channels of CHR manual initiation be OPERABLE, including the logic components necessary to initiate Refueling CHR Isolation. The CHR setpoint is chosen to be high enough to avoid inadvertent actuation in the event of normal background radiation fluctuations during fuel handling and movement of the reactor internals, but low enough to alarm and isolate the containment in the event of a Design Basis fuel handling accident.
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| APPLICABILITY In MODE 5 or 6, the Refueling CHR isolation of containment isolation valves is not normally required to be OPERABLE. However, during CORE ALTERATIONS or during movement of irradiated fuel within containment, there is the possibility of a fuel handling accident requiring containment isolation on high radiation in containment. Accordingly, the Refueling CHR Instrumentation must be OPERABLE during CORE ALTERATIONS and when moving any irradiated fuel in containment.
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| In MODES 1, 2, 3 and 4, both the Containment High Pressure (CHP) and CHR signals provide containment isolation as discussed in the Bases for LCO 3.3.3 and LCO 3.3.4.
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| Palisades Nuclear Plant B 3.3.6-3 Revised 01/29/2020
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| INSERT Bases 3.3.6 Refueling CHR Instrumentation B 3.3.6 BASES ACTIONS A.1, A.2.1, and A.2.2 Condition A applies to the failure of one Refueling CHR monitor channel, one CHR Manual Initiate channel, or one of each. The Required Action allows either initiation of a CHR signal by placing the inoperable channel in trip (which accomplishes the safety function of the inoperable channel), or suspension of CORE ALTERATIONS and movement of irradiated fuel assemblies within containment (which places the plant in a condition where the LCO does not apply). The Completion Time of 4 hours is acceptable because one additional channel of each Function remains operable during that period and the probability of an additional failure occurring during this period is very small.
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| The suspension of CORE ALTERATIONS and fuel movement shall not preclude completion of movement of a component to a safe position.
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| B.1 and B.2 Condition B applies when either no automatic Refueling CHR or no Manual CHR (or neither) is available. The Required Action is to immediately suspend CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. This places the plant in a condition where the LCO does not apply. The Completion Time is warranted on the basis that at least one containment isolation Function is completely lost.
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| The suspension of CORE ALTERATIONS and fuel movement shall not preclude completion of movement of a component to a safe position.
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| SURVEILLANCE SR 3.3.6.1 REQUIREMENTS Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
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| Palisades Nuclear Plant B 3.3.6-4 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.6 Refueling CHR Instrumentation B 3.3.6 BASES SURVEILLANCE SR 3.3.6.1 (continued)
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| REQUIREMENTS (continued) Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or actual differing radiation levels at the two detector locations. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.3.6.2 A CHANNEL FUNCTIONAL TEST is performed on each Refueling CHR channel to ensure the entire channel will perform its intended function.
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| A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.3.6-5 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.6 Refueling CHR Instrumentation B 3.3.6 BASES SURVEILLANCE SR 3.3.6.3 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each CHR Manual Initiation channel to ensure it will perform its intended function.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.3.6.4 A CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests.
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| No required setpoint is specified because these instruments are not assumed to function by any of the safety analyses.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Section 7.3
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| : 2. FSAR, Section 14.19 Palisades Nuclear Plant B 3.3.6-6 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.7 PAM Instrumentation B 3.3.7 B 3.3 INSTRUMENTATION B 3.3.7 Post Accident Monitoring (PAM) Instrumentation BASES BACKGROUND The primary purpose of the Post Accident Monitoring (PAM) instrumentation is to display plant variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions, for which no automatic control is provided, that are required for safety systems to accomplish their safety Functions for Design Basis Events.
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| The OPERABILITY of the PAM instrumentation ensures that there is sufficient information available on selected plant parameters to monitor and assess plant status and behavior following an accident.
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| The availability of PAM instrumentation is important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions can be determined. The required instruments are identified in FSAR Appendix 7C (Ref. 1) and address the recommendations of Regulatory Guide 1.97 (Ref. 2), as required by Supplement 1 to NUREG-0737, "TMI Action Items" (Ref. 3).
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| Type A variables are included in this LCO because they provide the primary information required to permit the control room operator to take specific manually controlled actions, for which no automatic control is provided, that are required for safety systems to accomplish their safety functions for Design Basis Accidents (DBAs).
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| Category I variables are the key variables deemed risk significant because they are needed to:
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| * Determine whether other systems important to safety are performing their intended functions;
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| * Provide information to the operators that will enable them to determine the potential for causing a gross breach of the barriers to radioactivity release; and
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| * Provide information regarding the release of radioactive materials to allow for early indication of the need to initiate action necessary to protect the public and for an estimate of the magnitude of any impending threat.
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| Palisades Nuclear Plant B 3.3.7-1 Revised 01/29/2020
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| INSERT Bases 3.3.7 PAM Instrumentation B 3.3.7 BASES BACKGROUND These key variables are identified in the plant specific Regulatory (continued) Guide 1.97 analyses (Ref. 1). This analysis identified the plant specific Type A and Category 1 variables and provided justification for deviating from the NRC proposed list of Category I variables.
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| The specific instrument Functions listed in Table 3.3.7-1 are discussed in the LCO Bases.
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| APPLICABLE The PAM instrumentation ensures the OPERABILITY of Regulatory SAFETY ANALYSES Guide 1.97 Type A variables, so that the control room operating staff can:
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| * Perform the diagnosis specified in the emergency operating procedures. These variables are restricted to preplanned actions for the primary success path of DBAs; and
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| * Take the specified, preplanned, manually controlled actions, for which no automatic control is provided, that are required for safety systems to accomplish their safety functions.
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| The PAM instrumentation also ensures OPERABILITY of Category I, non-Type A variables. This ensures the control room operating staff can:
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| * Determine whether systems important to safety are performing their intended functions;
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| * Determine the potential for causing a gross breach of the barriers to radioactivity release;
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| * Determine if a gross breach of a barrier has occurred; and
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| * Initiate action necessary to protect the public as well as to obtain an estimate of the magnitude of any impending threat.
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| Category I, non-Type A PAM instruments are retained in the Specification because they are intended to assist operators in minimizing the consequences of accidents. Therefore, these Category I variables are important in reducing public risk.
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| PAM instrumentation that satisfies the definition of Type A in Regulatory Guide 1.97 meets Criterion 3 of 10 CFR 50.36(c)(2).
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| Palisades Nuclear Plant B 3.3.7-2 Revised 01/29/2020
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| INSERT Bases 3.3.7 PAM Instrumentation B 3.3.7 BASES LCO LCO 3.3.7 requires at least two OPERABLE channels for all Functions except Containment Isolation Valve Position Indication. This is to ensure no single failure prevents the operators from being presented with the information necessary to determine the status of the plant and to bring the plant to, and maintain it in, a safe condition following that accident.
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| Furthermore, provision of at least two channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayed information.
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| For Containment Isolation Valve Position indication, the important information is the status of the containment penetrations. The LCO requires one position indication channel for each containment isolation valve listed in FSAR Appendix 7C (Ref. 1).
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| Listed below are discussions of the specified instrument Functions listed in Table 3.3.7-1. Component identifiers of the sensors, indicators, power supplies, displays, and recorders in each instrument loop are found in Reference 1.
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| 1, 2. Primary Coolant System (PCS) Hot and Cold Leg Temperature (wide range)
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| PCS wide range Hot and Cold Leg Temperatures are Type B, Category 1 variables provided for verification of core cooling and long term surveillance.
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| Reactor outlet temperature inputs to the PAM are provided by two wide range resistance elements and associated transmitters (one in each loop). The channels provide indication over a range of 50°F to 700°F.
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| : 3. Wide Range Neutron Flux Wide Range Neutron Flux indication is a Type B, Category 1 variable, and is provided to verify reactor shutdown.
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| : 4. Containment Floor Water Level (wide range)
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| Wide range Containment Floor Water Level is a Type B, Category 1 variable, and is provided for verification and long-term surveillance of PCS integrity.
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| Palisades Nuclear Plant B 3.3.7-3 Revised 01/29/2020
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| INSERT Bases 3.3.7 PAM Instrumentation B 3.3.7 BASES LCO 5. Subcooled Margin Monitor (continued)
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| The Subcooled Margin Monitor (SMM) is a Type A, Category 1 variable used to identify conditions, which require tripping of the primary coolant pumps and throttling of safety injection flows. Each SMM channel uses a number of PCS pressure and temperature inputs to determine the degree of PCS subcooling or superheat.
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| : 6. Pressurizer Level (Wide Range)
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| Pressurizer Level is a Type A, Category 1 variable, and is used to determine whether to terminate Safety Injection (SI), if still in progress, or to reinitiate SI if it has been stopped. Knowledge of pressurizer water level is also used to verify the plant conditions necessary to establish natural circulation in the PCS and to verify that the plant is maintained in a safe shutdown condition.
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| : 7. (Deleted)
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| : 8. Condensate Storage Tank (CST) Level CST Level is a Type D, Category 1 variable, and is provided to ensure water supply for AFW. The CST provides the safety grade water supply for the AFW System. Inventory is monitored by a 0 to 100% level indication. CST Level is displayed on a control room indicator. In addition, a control room annunciator alarms on low level.
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| The CST is the initial source of water for the AFW System. However, as the CST is depleted, manual operator action is necessary to replenish the CST.
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| Palisades Nuclear Plant B 3.3.7-4 Revised 01/29/2020
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| INSERT Bases 3.3.7 PAM Instrumentation B 3.3.7 BASES LCO 9. Primary Coolant System Pressure (wide range)
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| (continued)
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| PCS wide range pressure is a Type A, Category 1 variable provided for verification of core cooling and PCS integrity long-term surveillance.
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| Wide range PCS loop pressure is measured by pressure transmitters with a span of 0 psia to 3000 psig. Redundant monitoring capability is provided by two channels of instrumentation. Control room indications are provided on C12 and C02.
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| : 10. Containment Pressure (wide range)
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| Wide range Containment Pressure is a Type C, Category 1 variable, and is provided for verification of PCS and containment OPERABILITY.
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| It is also an input to decisions for initiating containment spray.
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| 11, 12. Steam Generator Water Level (wide range)
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| Wide range Steam Generator Water Level is a Type A, Category 1 variable, and is provided to monitor operation of decay heat removal via the steam generators. The steam generator level instrumentation covers a span extending from the tube sheet to the steam separators, with an indicated range of -140% to +150%. Redundant monitoring capability is provided by two channels of instrumentation for each SG.
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| Operator action for maintenance of heat removal is based on the control room indication of Steam Generator Water Level. The indication is used during a SG tube rupture to determine which SG has the ruptured tube. It is also used to determine when to initiate once through cooling on low water level.
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| 13, 14. SG Pressure Steam Generator Pressure is a Type A, Category 1 variable used in accident identification, including Loss of Coolant, and Steam Line Break. Redundant monitoring capability is provided by two channels of instrumentation for each SG.
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| Palisades Nuclear Plant B 3.3.7-5 Revised 01/29/2020
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| INSERT Bases 3.3.7 PAM Instrumentation B 3.3.7 BASES LCO 15. Containment Isolation Valve Position (continued)
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| Containment Isolation Valve (CIV) Position is a Type B, Category 1 variable, and is provided for verification of containment OPERABILITY.
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| CIV position is provided for verification of containment integrity. In the case of CIV position, the important information is the isolation status of the containment penetration. The LCO requires one channel of valve position indication in the control room to be OPERABLE for each CIV listed in FSAR Appendix 7C (Ref. 1). This is sufficient to redundantly verify the isolation status of each associated penetration via indicated status of the CIVs, and by knowledge of a passive (check) valve or a closed system boundary.
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| If a penetration flow path is isolated, position indication for the CIV(s) in the associated penetration flow path is not needed to determine status.
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| Therefore, as indicated in Note (a) the position indication for valves in an isolated penetration flow path is not required to be OPERABLE.
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| 16, 17, 18, 19. Core Exit Temperature Core Exit Temperature is a Type C, Category 1 variable, and is provided for verification and long term surveillance of core cooling.
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| Each Required Core Exit Thermocouple (CET) channel consists of a single environmentally qualified thermocouple.
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| The design of the Incore Instrumentation System includes a Type K (chromel alumel) thermocouple within each of the incore instrument detector assemblies.
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| The junction of each thermocouple is located above the core exit, inside the incore detector assembly guide tube, that supports and shields the incore instrument detector assembly string from flow forces in the outlet plenum region. These core exit thermocouples monitor the temperature of the reactor coolant as it exits the fuel assemblies.
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| The core exit thermocouples have a usable temperature range from 32°F to 2300°F, although accuracy is reduced at temperatures above 1800°F.
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| Palisades Nuclear Plant B 3.3.7-6 Revised 01/29/2020
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| INSERT Bases 3.3.7 PAM Instrumentation B 3.3.7 BASES LCO 20. Reactor Vessel Water Level (continued)
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| Reactor Vessel Water Level is monitored by the Reactor Vessel Level Monitoring System (RVLMS) and is a Type B, Category 1 variable provided for verification and long-term surveillance of core cooling.
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| The RVLMS provides a direct measurement of the collapsed liquid level above the fuel alignment plate. The collapsed level represents the amount of liquid mass that is in the reactor vessel above the core.
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| Measurement of the collapsed water level is selected because it is a direct indication of the water inventory. The collapsed level is obtained over the same temperature and pressure range as the saturation measurements, thereby encompassing all operating and accident conditions where it must function. Also, it functions during the recovery interval. Therefore, it is designed to survive the high steam temperature that may occur during the preceding core recovery interval.
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| The level range extends from the top of the vessel down to the top of the fuel alignment plate. A total of eight Heated Junction Thermocouple (HJTC) pairs are employed in each of the two RVLMS channels. Each pair consists of a heated junction TC and an unheated junction TC. The differential temperature at each HJTC pair provides discrete indication of uncovery at the HJTC pair location. This indication is displayed using LEDs in the control room. This provides the operator with adequate indication to track the progression of the accident and to detect the consequences of its mitigating actions or the functionality of automatic equipment.
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| A RVLMS channel consists of eight sensors in a probe. A channel is OPERABLE if four or more sensors, two or more of the upper four and two or more of the lower four, are OPERABLE.
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| : 21. Containment Area Radiation (high range)
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| High range Containment Area Radiation is a Type E, Category 1 variable, and is provided to monitor for the potential of significant radiation releases and to provide release assessment for use by operators in determining the need to invoke site emergency plans.
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| Palisades Nuclear Plant B 3.3.7-7 Revised 01/29/2020
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| INSERT Bases 3.3.7 PAM Instrumentation B 3.3.7 BASES APPLICABILITY The PAM instrumentation LCO is applicable in MODES 1, 2, and 3.
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| These variables are related to the diagnosis and preplanned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1, 2, and 3. In MODES 4, 5, and 6, plant conditions are such that the likelihood of an event occurring that would require PAM instrumentation is low; therefore, PAM instrumentation is not required to be OPERABLE in these MODES.
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| ACTIONS A note has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.7-1. The Completion Time(s) of the inoperable channel(s) of a Function will be tracked separately for each Function, starting from the time the Condition was entered for that Function.
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| A.1 When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days. The 30-day Completion Time is based on operating experience and takes into account the remaining OPERABLE channel, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.
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| B.1 This Required Action specifies initiation of actions in accordance with Specification 5.6.6, which requires a written report to be submitted to the Nuclear Regulatory Commission. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative Required Actions. This Required Action is appropriate in lieu of a shutdown requirement, given the likelihood of plant conditions that would require information provided by this instrumentation. Also, alternative Required Actions are identified before a loss of functional capability condition occurs.
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| C.1 When one or more Functions have two required channels inoperable (i.e., two channels inoperable in the same Function), one channel in the Function should be restored to OPERABLE status within 7 days. The Completion Time of 7 days is based on the relatively low probability of an event requiring PAM instrumentation operation and the availability of Palisades Nuclear Plant B 3.3.7-8 Revised 01/29/2020
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| INSERT Bases 3.3.7 PAM Instrumentation B 3.3.7 BASES ACTIONS alternate means to obtain the required information. Continuous (continued) operation with two required channels inoperable in a Function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAM instrumentation. Therefore, requiring restoration of one inoperable channel of the Function limits the risk that the PAM Function will be in a degraded condition should an accident occur.
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| D.1 Condition D is currently not used.
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| E.1 This Required Action directs entry into the appropriate Condition referenced in Table 3.3.7-1. The applicable Condition referenced in the Table is Function dependent. Each time Required Action C.1 is not met, and the associated Completion Time has expired, Condition E is entered for that channel and provides for transfer to the appropriate subsequent Condition.
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| F.1 and F.2 If the Required Action and associated Completion Time of Condition C is not met, and Table 3.3.7-1 directs entry into Condition F, the plant must be brought to a MODE in which the requirements of this LCO do not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 30 hours.
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| The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
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| G.1 Alternate means of monitoring Reactor Vessel Water Level and Containment Area Radiation have been developed and tested. These alternate means may be temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. If these alternate means are used, the Required Action is not to shut down the plant, but rather to follow the directions of Specification 5.6.6. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels.
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| Palisades Nuclear Plant B 3.3.7-9 Revised 01/29/2020
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| INSERT Bases 3.3.7 PAM Instrumentation B 3.3.7 BASES SURVEILLANCE A Note at the beginning of the Surveillance Requirements specifies that REQUIREMENTS the following SRs apply to each PAM instrumentation Function in Table 3.3.7-1.
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| SR 3.3.7.1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verify the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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| Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off scale during times when surveillance is required, the CHANNEL CHECK will only verify that they are off scale in the same direction. Off scale low current loop channels are verified to be reading at the bottom of the range and not failed downscale.
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| As indicated in the SR, a CHANNEL CHECK is only required for those channels which are normally energized.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.3.7-11 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.7 PAM Instrumentation B 3.3.7 BASES SURVEILLANCE SR 3.3.7.2 REQUIREMENTS (continued) A CHANNEL CALIBRATION is typically a complete check of the instrument channel including the sensor. Therefore, this SR is modified by a Note, which states that it is not necessary to calibrate neutron detectors because of the difficulty of simulating a meaningful signal.
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| Wide range and source range nuclear instrument channels are not calibrated to indicate the actual power level or the flux in the detector location. The circuitry is adjusted so that wide range and source range readings may be used to determine the approximate reactor flux level for comparative purposes. The Surveillance verifies the channel responds to the measured parameter within the necessary range and accuracy.
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| For the core exit thermocouples, a CHANNEL CALIBRATION is performed by substituting a known voltage for the thermocouple.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Appendix 7C, Regulatory Guide 1.97 Instrumentation
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| : 2. Regulatory Guide 1.97
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| : 3. NUREG-0737, Supplement 1 Palisades Nuclear Plant B 3.3.7-12 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.8 Alternate Shutdown System B 3.3.8 B 3.3 INSTRUMENTATION B 3.3.8 Alternate Shutdown System BASES BACKGROUND The Alternate Shutdown System provides the control room operator with sufficient instrumentation and controls to maintain the plant in a safe shutdown condition from a location other than the control room.
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| This capability is necessary to protect against the possibility that the control room becomes inaccessible. A safe shutdown condition is defined as MODE 3. With the plant in MODE 3, the Auxiliary Feedwater (AFW) System and the steam generator safety valves or the steam generator atmospheric dump valves can be used to remove core decay heat and meet all safety requirements. The long term supply of water for the AFW System and the ability to borate the Primary Coolant System (PCS) from outside the control room allow extended operation in MODE 3.
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| The Auxiliary Hot Shutdown Panels (C-150/C-150A) are located in the southwest electrical penetration room. These panels are comprised of two enclosures, the main enclosure C-150 and an auxiliary enclosure C-150A. The description below combines these two enclosures into one entity "Panel C-150."
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| Panel C-150 provides control of the AFW flow control valves and AFW turbine steam supply Valve. Indication of AFW flow, Steam Generator water level, pressurizer pressure, and pressurizer level are provided.
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| See FSAR Section 7.4 (Ref. 1) for operation via Panel C-150.
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| The instrumentation and equipment controls that are required are listed in Table 3.3.8-1.
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| Switches, which transfer control or instrument functions from the control room to the C-150 panel, alarm in the control room when the C-150 panel is selected.
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| APPLICABLE The Alternate Shutdown System is required to provide equipment at SAFETY ANALYSES appropriate locations outside the control room with a capability to maintain the plant in a safe condition in MODE 3.
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| The criteria governing the design and the specific system requirements of the Alternate Shutdown System are located in 10 CFR 50, Appendix A, GDC 19, and Appendix R (Ref. 2).
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| Palisadeas Nuclear B 3.3.8-1 Revised 01/29/2020
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| INSERT Bases 3.3.8 Alternate Shutdown System B 3.3.8 BASES APPLICABLE The Alternate Shutdown System has been identified as an important SAFETY ANALYSES contributor to the reduction of plant risk to accidents and, therefore, (continued) satisfies the requirements of Criterion 4 of 10 CFR 50.36(c)(2).
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| LCO The Alternate Shutdown System LCO provides the requirements for the OPERABILITY of one channel of the instrumentation and controls necessary to maintain the plant in MODE 3 from a location other than the control room. The instrumentation and controls required are listed in Table 3.3.8-1 in the accompanying LCO.
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| Equipment controls that are required by the alternative dedicated method of maintaining MODE 3 are as follows:
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| : 1. AFW flow control valves (CV-0727 and CV-0749); and
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| : 2. Turbine-driven AFW pump.
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| Instrumentation systems displayed on the Auxiliary Hot Shutdown Control Panel are:
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| : 1. Source range flux monitor;
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| : 2. AFW flow (HIC-0727 and HIC-0749C);
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| : 3. Pressurizer pressure;
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| : 4. Pressurizer level;
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| : 5. SG level and pressure;
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| : 6. Primary coolant temperatures (hot and cold legs);
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| : 7. Turbine-driven AFW pump low-suction pressure warning light; and
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| : 8. SIRW tank level.
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| A Function of an Alternate Shutdown System is OPERABLE if all instrument and control channels needed to support the remote shutdown Functions are OPERABLE.
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| Palisades Nuclear B 3.3.8-2 Revised 01/29/2020
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| INSERT Bases 3.3.8 Alternate Shutdown System B 3.3.8 BASES LCO The Alternate Shutdown System instrumentation and control circuits (continued) covered by this LCO do not need to be energized to be considered OPERABLE. This LCO is intended to ensure that the instrument and control circuits will be OPERABLE if plant conditions require that the Alternate Shutdown System be placed in operation.
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| Table 3.3.8-1 Indication Channel 1, Source Range Nuclear Instrumentation, uses the same detector and preamplifier as the control room channel. Optical isolation is provided between the control room and AHSDP (Alternate Hot Shut Down Panel) portions of the circuit.
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| When the control switches are changed to the "AHSDP" position, the detector and preamplifier is isolated from its normal power supply and connected into the AHSDP power supply.
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| Table 3.3.8-1 Indication Channels 2 and 12 are provided with their own pressure and level transmitter. The associated circuitry is energized when the AHSDP is energized.
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| The other Table 3.3.8-1 Indication Channels in Table 3.3.8-1 use a transmitter which also serves normal control room instrumentation.
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| When the control switches are changed to the "AHSDP" (Alternate Hot Shut Down Panel) position, the transmitter is isolated from its normal power supply and circuitry, and connected into the C-150 or C-150A panel circuit; control for AFW flow control valves CV-0727 and CV-0749 is also transferred to C-150. The transfer switches are alarmed in the control room.
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| APPLICABILITY The Alternate Shutdown System LCO is applicable in MODES 1, 2, and 3. This is required so that the plant can be maintained in MODE 3 for an extended period of time from a location other than the control room.
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| This LCO is not applicable in MODE 4, 5, or 6. In these MODES, the plant is already subcritical and in the condition of reduced PCS energy.
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| Under these conditions, considerable time is available to restore necessary instrument control Functions if control room instruments or control become unavailable.
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| Palisades Nuclear Plant B 3.3.8-3 Revised 01/29/2020
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| INSERT Bases 3.3.8 Alternate Shutdown System B 3.3.8 BASES ACTIONS A Note has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.8-1. The Completion Time of the inoperable channel of a Function will be tracked separately for each Function, starting from the time the Condition was entered for that Function.
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| A.1 Condition A addresses the situation where the required channels of the Remote Shutdown System are inoperable. This includes any Function listed in Table 3.3.8-1 as well as the control and transfer switches.
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| Required Action A.1 is to restore the channel to OPERABLE status within 30 days. This allows time to complete repairs on the failed channel. The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room.
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| B.1 and B.2 If the Required Action and associated Completion Time of Condition A are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.3.8-4 Revised 01/29/2020
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| INSERT Bases 3.3.8 Alternate Shutdown System B 3.3.8 BASES SURVEILLANCE SR 3.3.8.1 REQUIREMENTS This SR applies to the startup range neutron flux monitoring channel.
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| The CHANNEL FUNCTIONAL TEST consists of verifying proper response of the channel to the internal test signals, and verification that a detectable signal is available from the detector. After lengthy shutdown periods flux may be below the range of the channel indication. Signal verification with test equipment is acceptable.
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| The CHANNEL FUNCTIONAL TEST of the startup range neutron flux monitoring channel is performed once within 7 days prior to reactor startup. The Frequency is based on plant operating experience that demonstrates channel failure is rare.
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| SR 3.3.8.2 SR 3.3.8.2 verifies that each required Alternate Shutdown System transfer switch and control circuit performs its intended function. This verification is performed from AHSDPs C-150 and C-150A and locally, as appropriate. Operation of the equipment from the AHSDPs C-150 and C-150A is not necessary. The Surveillance can be satisfied by performance of a continuity check. This will ensure that if the control room becomes inaccessible, the plant can be maintained in MODE 3 from the auxiliary shutdown panel and the local control stations.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.3.8-5 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.8 Alternate Shutdown System B 3.3.8 BASES SURVEILLANCE SR 3.3.8.3 REQUIREMENTS (continued) A CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to the measured parameter within the necessary range and accuracy.
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| Performance of a CHANNEL CALIBRATION on Functions 1 through 15 ensures that the channels are operating accurately and within specified tolerances. This verification is performed from the AHSDPs and locally, as appropriate. A test of the AFW pump suction pressure alarm (Function 15) is included as part of its CHANNEL CALIBRATION. This will ensure that if the control room becomes inaccessible, the plant can be maintained in MODE 3 from the AHSDPs and local control stations.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| This SR is modified by two Notes. Note 1 states that the SR is not required for Functions 16, 17, and 18; Note 2 states that it is not necessary to calibrate neutron detectors because of the difficulty of simulating a meaningful signal. Wide range and source range nuclear instrument channels are not calibrated to indicate the actual power level or the flux in the detector location. The circuitry is adjusted so that wide range and source range readings may be used to determine the approximate reactor flux level for comparative purposes.
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| REFERENCES 1. FSAR, Section 7.4, Other Safety Related Protection, Control, and Display Systems
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| : 2. 10 CFR 50, Appendix A, GDC 19 and Appendix R.
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| Palisades Nuclear Plant B 3.3.8-6 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.9 Neutron Flux Monitoring Channels B 3.3.9 B 3.3 INSTRUMENTATION B 3.3.9 Neutron Flux Monitoring Channels BASES BACKGROUND The neutron flux monitoring channels consist of two combined source range/wide range channels, designated NI-1/3 and NI-2/4. The wide range portions, (NI-3 and NI-4) provide neutron flux power indication from < 1E-7% RTP to > 100% RTP. The source range portions, designated NI-1 and NI-2, provide source range indication over the range of 0.1 to 1E+5 cps.
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| This LCO addresses MODES 3, 4, and 5. In MODES 1 and 2, the neutron flux monitoring requirements are addressed by LCO 3.3.1, "Reactor Protective System (RPS) Instrumentation."
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| When the plant is shutdown, both neutron flux monitoring channels must be available to monitor neutron flux. If only one section of a neutron flux monitoring channel (source range or wide range) is functioning, the neutron flux monitoring channel may be considered OPERABLE if it is capable of detecting the existing reactor neutron flux.
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| In this application, the RPS channels need not be OPERABLE since the reactor trip Function is not required. By monitoring neutron flux, loss of SDM caused by boron dilution can be detected as an increase in flux. Two channels must be OPERABLE to provide single failure protection and to facilitate detection of channel failure by providing CHANNEL CHECK capability.
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| APPLICABLE The neutron flux monitoring channels are necessary to monitor core SAFETY ANALYSES reactivity changes. They are the primary means for detecting, and triggering operator actions to respond to, reactivity transients initiated from conditions in which the RPS is not required to be OPERABLE.
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| The neutron flux monitoring channel's LCO requirements support compliance with 10 CFR 50, Appendix A, GDC 13 (Ref. 1). The FSAR, Chapters 7 and 14 (Refs. 2 and 3, respectively), describes the specific neutron flux monitoring channel features that are critical to comply with the GDC.
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| Palisades Nuclear Plant B 3.3.9-1 Revised 01/29/2020
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| INSERT Bases 3.3.9 Neutron Flux Monitoring Channels B 3.3.9 BASES APPLICABLE The OPERABILITY of neutron flux monitoring channels is necessary to SAFETY ANALYSIS meet the assumptions of the safety analyses and provide for the (continued) detection of reduced SDM.
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| The neutron flux monitoring channels satisfy Criterion 4 of 10 CFR 50.36(c)(2).
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| LCO The LCO on the neutron flux monitoring channels ensures that adequate information is available to verify core reactivity conditions while shut down. The safety function of these instruments is to detect changes in core reactivity such as might occur from an inadvertent boron dilution.
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| Two neutron flux monitoring channels are required to be OPERABLE. If only one section of a neutron flux monitoring channel (source range or wide range) is functioning, the neutron flux monitoring channel may be considered OPERABLE if it is capable of detecting the existing reactor neutron flux. For example, with the source range count rate indicator functioning properly within its range, and in reasonable agreement with the other source range, a neutron flux monitor channel may be considered OPERABLE even though its wide range indicator is not functioning.
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| The source range nuclear instrumentation channels, NI-1 and NI-2, provide neutron flux coverage extending an additional one to two decades below the wide range channels for use during refueling, when neutron flux may be extremely low.
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| This LCO does not require OPERABILITY of the High Startup Rate Trip Function or the Zero Power Mode Bypass Removal Function. Those functions are addressed in LCO 3.3.1, RPS Instrumentation.
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| APPLICABILITY In MODES 3, 4, and 5, neutron flux monitoring channels must be OPERABLE to monitor core power for reactivity changes.
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| In MODES 1 and 2, neutron flux monitoring channels are addressed as part of the RPS in LCO 3.3.1.
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| The requirements for source range neutron flux monitoring in MODE 6 are addressed in LCO 3.9.2, "Nuclear Instrumentation."
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| a Palisades Nuclear Plant B 3.3.9-2 Revised 01/29/2020
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| INSERT Bases 3.3.9 Neutron Flux Monitoring Channels B 3.3.9 BASES ACTIONS A.1 and A.2 With one required channel inoperable, it may not be possible to perform a CHANNEL CHECK to verify that the other required channel is OPERABLE. Therefore, with one or more required channels inoperable, the neutron flux power monitoring Function cannot be reliably performed. Consequently, the Required Actions are the same for one required channel inoperable or more than one required channel inoperable. The absence of reliable neutron flux indication makes it difficult to ensure SDM is maintained. Required Action A.1, therefore, requires that all positive reactivity additions that are under operator control, such as boron dilution or PCS temperature changes, be halted immediately, preserving SDM.
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| SDM must be verified periodically to ensure that it is being maintained.
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| The initial Completion Time of 4 hours and once every 12 hours thereafter to perform SDM verification takes into consideration that Required Action A.1 eliminates many of the means by which SDM can be reduced. These Completion Times are also based on operating experience in performing the Required Actions and the fact that plant conditions will change slowly.
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| As stated in SR 3.0.2, the 25% extension allowed by SR 3.0.2 may be applied to Required Actions whose Completion Time is stated as once per . . . however, the 25% extension does not apply to the initial performance of a Required Action with a periodic Completion Time that requires performance on a "once per. . ." basis. The 25% extension applies to each performance of the Required Action after the initial performance. Therefore, while Required Action 3.3.9 A.2 must be initially performed within 4 hours without any SR 3.0.2 extension, subsequent performances at the Once per 12 hours interval may utilize the 25% SR 3.0.2 extension.
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| Palisades Nuclear Plant B 3.3.9-3 Revised 01/29/2020
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| INSERT Bases 3.3.9 Neutron Flux Monitoring Channels B 3.3.9 BASES SURVEILLANCE SR 3.3.9.1 REQUIREMENTS SR 3.3.9.1 is the performance of a CHANNEL CHECK on each required channel. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based upon the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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| Agreement criteria are determined by the plant staff and should be based on a combination of the channel instrument uncertainties including indication, and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limits. If the channels are within the criteria, it is an indication that the channels are OPERABLE.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.3.9-4 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.9 Neutron Flux Monitoring Channels B 3.3.9 BASES SURVEILLANCE SR 3.3.9.2 REQUIREMENTS (continued) SR 3.3.9.2 is the performance of a CHANNEL CALIBRATION. The Surveillance is a complete check and readjustment of the neutron flux channel from the preamplifier input through to the remote indicators.
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| This SR is modified by a Note which states that it is not necessary to calibrate neutron detectors because of the difficulty of simulating a meaningful signal. Wide range and source range nuclear instrument channels are not calibrated to indicate the actual power level or the flux in the detector location. The circuitry is adjusted so that wide range and source range readings may be used to determine the approximate reactor flux level for comparative purposes.
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| This LCO does not require the OPERABILITY of the High Startup Rate trip function or the Zero Power Mode Bypass removal function. The OPERABILITY of those functions does not have to be verified during performance of this SR. Those functions are addressed in LCO 3.3.1, RPS Instrumentation.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. 10 CFR 50, Appendix A, GDC 13
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| : 2. FSAR, Chapter 7
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| : 3. FSAR, Chapter 14 Palisades Nuclear Plant B 3.3.9-5 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.10 ESRV Instrumentation B 3.3.10 B 3.3 INSTRUMENTATION B 3.3.10 Engineered Safeguards Room Ventilation (ESRV) Instrumentation BASES BACKGROUND This LCO addresses the instrumentation which provides isolation of the ESRV System (Ref. 1). The ESRV Instrumentation high radiation signal provides automatic damper closure, using two radiation monitors.
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| One radiation monitor is located in the ventilation system duct work associated with each of the Engineered Safeguards (ES) pump rooms.
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| Upon detection of high radiation, the ESRV Instrumentation actuates isolation of the associated ES pump room by closing the dampers in the ventilation system inlet and discharge paths. Typically, high radiation would only be expected due to excessive leakage during the recirculation phase of operation following a Loss of Coolant Accident (LOCA). The ESRV System is addressed by LCO 3.7.13, Engineered Safeguards Room Ventilation (ESRV) Dampers.
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| APPLICABLE The ESRV Instrumentation isolates the ES pump rooms in the event of SAFETY ANALYSES high radiation in the pump rooms due to leakage during the recirculation phase. The analysis for a Maximum Hypothetical Accident (MHA) described in FSAR, Section 14.22 (Ref. 2), assumes a reduction factor in the potential radioactive releases from the ES pump rooms due to plateout following automatic isolation. However, no specific value is assumed in the MHA for the timing of actuation of the isolation. The results indicate that the potential MHA offsite doses would be less than applicable 10 CFR 50.67 limits.
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| The ESRV Instrumentation satisfies the requirements of Criterion 3 of 10 CFR 50.36(c)(2).
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| LCO The LCO for the ESRV Instrumentation requires both channels to be OPERABLE to initiate ES pump room isolation when high radiation exceeds the trip setpoint.
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| The ESRV Instrumentation Setpoint is specified as 2.2E+5 cpm. This setpoint is high enough to avoid inadvertent actuation in the event of normal background radiation fluctuations during testing, but low enough to isolate the ES pump room in the event of radiation levels indicative of a LOCA and excessive leakage during recirculation of primary coolant through the ES pump room.
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| Palisades Nuclear Plant B 3.3.10-1 Revised 01/29/2020
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| INSERT Bases 3.3.10 ESRV Instrumentation B 3.3.10 BASES APPLICABILITY The ESRV Instrumentation must be OPERABLE in MODES 1, 2, 3, and 4. In these MODES, the potential exists for an accident that could release fission product radioactivity into the primary coolant which could subsequently be released to the environment by leakage from the ES systems which are recirculating the coolant.
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| While in MODE 5 and in MODE 6, the ESRV Instrumentation need not be OPERABLE since the potential for radioactive releases is minimized and operator action is sufficient to ensure post accident offsite doses are maintained within the applicable 10 CFR 50.67 limits.
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| ACTIONS The most common causes of channel inoperability are outright failure of loop components or drift of those loop components which is sufficient to exceed the tolerance provided in the plant setpoint analysis. Loop component failures are typically identified by the actuation of alarms due to the channel failing to the safe condition, during CHANNEL CHECKS (when the instrument is compared to the redundant channels), or during the CHANNEL FUNCTIONAL TEST (when an automatic component might not respond properly). Typically, the drift of the loop components is found to be small and results in a delay of actuation rather than a total loss of function. Excessive loop component drift would, most likely, be identified during a CHANNEL CHECK (when the instrument is compared to the redundant channels) or during a CHANNEL CALIBRATION (when instrument loop components are checked against reference standards).
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| A Note has been added to the ACTIONS to clarify the application of the Completion Time rules. The Conditions of this Specification may be entered independently for each channel since each channel serves to isolate a different Engineered Safeguards Room. The Completion Times of each inoperable channel will be tracked separately, starting from the time the Condition was entered.
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| A.1 Condition A addresses the failure of one or both ESRV Instrumentation high radiation monitoring channels. Operation may continue as long as action is immediately initiated to isolate the ESRV System. With the inlet and exhaust dampers closed, the ESRV Instrumentation is no longer required since the potential pathway for radioactivity to escape to the environment has been removed.
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| Palisades Nuclear Plant B 3.3.10-2 Revised 01/29/2020
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| INSERT Bases 3.3.10 ESRV Instrumentation B 3.3.10 BASES ACTIONS A.1 (continued)
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| (continued)
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| The Completion Time for this Required Action is commensurate with the importance of maintaining the ES pump room atmosphere isolated from the outside environment when the ES pumps are circulating primary coolant.
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| SURVEILLANCE SR 3.3.10.1 REQUIREMENTS Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
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| Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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| Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limits.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.3.10-3 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.3.10 ESRV Instrumentation B 3.3.10 BASES SURVEILLANCE SR 3.3.10.2 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each ESRV Instrumentation channel to ensure the entire channel will perform its intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
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| Any setpoint adjustment must be consistent with the assumptions of the setpoint analyses.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.3.10.3 CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests.
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| CHANNEL CALIBRATIONS must be performed consistent with the setpoint analysis.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Section 7.4.5.2
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| : 2. FSAR, Section 14.22 Palisades Nuclear Plant B 3.3.10-4 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.1 PCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.1 PCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits BASES BACKGROUND These Bases address requirements for maintaining PCS pressure, temperature, and flow rate within limits assumed in the safety analyses.
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| The safety analyses (Ref. 1) of normal operating conditions and anticipated operational occurrences assume initial conditions within the normal steady state envelope. The limits placed on DNB related parameters ensure that these parameters, when appropriate measurement uncertainties are applied, will not be less conservative than were assumed in the analyses and thereby provide assurance that the minimum Departure from Nucleate Boiling Ratio (DNBR) will meet the required criteria for each of the transients analyzed.
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| Another set of limits on DNB related parameters is provided in Safety Limit (SL) 2.1.1, "Reactor Core Safety Limits." The restriction of the SLs prevent overheating of the fuel and cladding that would result in the release of fission products to the primary coolant. The limits of LCO 3.4.1, in combination with other LCOs, are designed to prevent violation of the reactor core SLs.
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| The LCO limits for minimum and maximum PCS pressures as measured at the pressurizer are consistent with operation within the nominal operating envelope and are bounded by those used as the initial pressures in the analyses.
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| The LCO limit for maximum PCS cold leg temperature is consistent with operation at steady state power levels and is bounded by those used as the initial temperatures in the analyses.
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| The LCO limits for minimum PCS flow rate is bounded by those used as the initial flow rates in the analyses. The PCS flow rate is not expected to vary during plant operation with all Primary Coolant Pumps running.
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| Palisades Nuclear Plant B 3.4.1-1 Revised 01/29/2020
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| INSERT Bases 3.4.1 PCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABLE The requirements of LCO 3.4.1 represent the initial conditions for SAFETY ANALYSES DNB limited transients analyzed in the safety analyses (Ref. 1). The safety analyses have shown that transients initiated from the limits of this LCO will meet the DNBR Safety Limit (SL 2.1.1). This is the acceptance limit for the PCS DNB parameters. Changes to the facility that could impact these parameters must be assessed for their impact on the DNBR criterion. The transients analyzed for include loss of coolant flow events and dropped or stuck control rod events. A key assumption for the analysis of these events is that the core power distribution is within the limits of LCO 3.1.6, "Regulating Rod Group Position Limits"; LCO 3.2.3, "Quadrant Power Tilt"; and LCO 3.2.4, "AXIAL SHAPE INDEX.
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| The PCS DNB limits satisfy Criterion 2 of 10 CFR 50.36(c)(2).
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| LCO This LCO specifies limits on the monitored process of variables PCS pressurizer pressure and PCS cold leg temperature, and the calculated value of PCS total flow rate to ensure that the core operates within the limits assumed for the plant safety analyses. These variables are contained in the COLR to provide operating and analysis flexibility from cycle to cycle. Operating within these limits will result in meeting the DNBR criterion in the event of a DNB limited transient.
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| The LCO numerical values for pressure and temperature specified in the COLR are given for the measurement location but have not been adjusted for instrument error. Plant specific limits of instrument error are established by the plant staff to meet the operational requirements of this LCO. Instrument errors and the PCS flow rate measurement error are applied to the LCO numerical values in the safety analysis.
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| APPLICABILITY In MODE 1, the limits on PCS pressurizer pressure, PCS cold leg temperature, and PCS flow rate must be maintained during steady state operation in order to ensure that DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In all other MODES, the power level is low enough so that DNBR is not a concern.
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| Palisades Nuclear Plant B 3.4.1-2 Revised 01/29/2020
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| INSERT Bases 3.4.1 PCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES ACTIONS A.1 Pressurizer pressure and cold leg temperature are controllable and measurable parameters. PCS flow rate is not a controllable parameter and is not expected to vary during steady state operation. With any of these parameters not within the LCO limits, action must be taken to restore the parameter.
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| The 2-hour Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause of the off normal condition, and to restore the readings within limits.
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| The Completion Time is based on plant operating experience.
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| B.1 If Required Action A.1 is not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours. In MODE 2, the reduced power condition eliminates the potential for violation of the accident analysis bounds.
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| Six hours is a reasonable time that permits the plant power to be reduced at an orderly rate without challenging plant systems.
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| SURVEILLANCE SR 3.4.1.1 and SR 3.4.1.2 REQUIREMENTS The Surveillance for monitoring pressurizer pressure and PCS cold leg temperature is performed using installed instrumentation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.4.1.3 Measurement of PCS total flow rate verifies that the actual PCS flow rate is within the bounds of the analyses. This verification may be performed by a calorimetric heat balance or other method.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. PCS flow rate must also be verified after plugging of each 10 or more steam generator tubes since plugging 10 or more tubes could result in an increase in PCS flow resistance.
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| Plugging less than 10 steam generator tubes will not have a significant impact on PCS flow resistance and, as such, does not require a verification of PCS flow rate.
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| Palisades Nuclear Plant B 3.4.1-3 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.1 PCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES SURVEILLANCE SR 3.4.1.3 (continued)
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| REQUIREMENTS The SR is modified by a Note that states the SR is only required to be performed 31 EFPD after THERMAL POWER is 90% RTP. The Note is necessary to allow measurement of the flow rate at normal operating conditions at power in MODE 1. The most common, and perhaps accurate, method used to perform the PCS total flow surveillance is by means of a primary to secondary heat balance (calorimetric) with the plant at or near full rated power. The most accurate results for such a test are obtained with the plant at or near full power when differential temperatures measured across the reactor are the greatest.
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| Consequently, the test should not be performed until reaching near full power (i.e., > 90% RTP) conditions. Similarly, test accuracy is also influenced by plant stability. In order for accurate results to be obtained, steady state plant conditions must exist to permit meaningful data to be gathered during the test. Typically, following an extended shutdown the secondary side of the plant will take up to several days to stabilize after power escalation. It is impracticable to perform a primary to secondary heat balance of the precision required for the PCS flow measurement until stabilization has been achieved. Furthermore, an integral part of the PCS flow heat balance involves the use of Ultrasonic Flow Measurement equipment for measuring steam generator feedwater flow. This equipment requires, stable plant operation at or near full power conditions before it can be used. As such, the Surveillance cannot be performed in MODE 2 or below, and will not yield accurate results if performed below 90% RTP.
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| REFERENCES 1. FSAR, Section 14.1 Palisades Nuclear Plant B 3.4.1-4 Revised 01/29/2020
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| INSERT Bases 3.4.2 PCS Minimum Temperature for Criticality B 3.4.2 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.2 PCS Minimum Temperature for Criticality BASES BACKGROUND Establishing the value for the minimum temperature for reactor criticality is based upon considerations for:
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| : a. Operation within the existing instrumentation ranges and accuracies;
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| : b. Operation within the bounds of the existing accident analyses; and
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| : c. Operation with the reactor vessel above its minimum nil ductility reference temperature when the reactor is critical.
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| The primary coolant moderator temperature coefficient used in core operating and accident analysis is typically defined for the normal operating temperature range (532°F to 570°F). The Reactor Protective System receives inputs from the narrow range hot leg and cold leg temperature instruments, which have a range of 515°F to 615°F. The PCS loop average temperature (Tave) is controlled using inputs of the same range. Nominal Tave for making the reactor critical is 532°F. Safety and operating analyses for lower than 525°F have not been made.
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| APPLICABLE There are no accident analyses that dictate the minimum temperature SAFETY ANALYSES for criticality, but existing transient analysis are bounding for operation at low power with cold leg temperatures of 525°F (Ref. 1).
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| The PCS minimum temperature for criticality satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| LCO The purpose of the LCO is to prevent criticality outside the normal operating regime (532°F to 570°F) and to prevent operation in an unanalyzed condition.
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| The LCO provides a reasonable distance between the hot zero power value of 532°F and the limit of 525°F. This allows adequate time to trend its approach and take corrective actions prior to exceeding the limit.
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| Palisades Nuclear Plant B 3.4.2-1 Revised 01/29/2020
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| INSERT Bases 3.4.2 PCS Minimum Temperature for Criticality B 3.4.2 BASES APPLICABILITY The reactor has been designed and analyzed to be critical in MODES 1 and 2 only and in accordance with this specification. Criticality is not permitted in any other MODE. Therefore, this LCO is applicable in MODE 1, and MODE 2 when Keff 1.0.
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| ACTIONS A.1 If Tave is below 525°F and cannot be restored in 30 minutes, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 2 with Keff < 1.0 within 30 minutes. Rapid reactor shutdown can be readily and practically achieved within a 30 minute period. The allowed time reflects the ability to perform this action and to maintain the plant within the analyzed range.
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| SURVEILLANCE SR 3.4.2.1 REQUIREMENTS PCS loop average temperature is required to be verified at or above 525°F. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Section 14.1.3 Palisades Nuclear Plant B 3.4.2-2 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.3 PCS P/T Limits B 3.4.3 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.3 PCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the PCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during PCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.
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| Figures 3.4.3-1 and 3.4.3-2 contain P/T limit curves for heatup, cooldown, and Inservice Leak and Hydrostatic (ISLH) testing, and data for the maximum rate of change of primary coolant temperature.
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| Each P/T limit curve defines an acceptable region for normal operation.
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| The P/T limit curves include an allowance to account for the fact that pressure is measured in the pressurizer rather than at the vessel beltline and to account for primary coolant pump discharge pressure. The use of the curves provides operational limits during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.
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| The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the Primary Coolant Pressure Boundary (PCPB). The vessel is the component most subject to brittle failure, and the LCO limits apply to the vessel.
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| 10 CFR 50, Appendix G (Ref. 2), requires the establishment of P/T limits for material fracture toughness requirements of the PCPB materials.
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| Reference 2 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests.
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| The neutron embrittlement effect on the material toughness is reflected by increasing the nil ductility reference temperature (RTNDT) as neutron fluence increases.
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| The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 4) and Appendix H of 10 CFR 50 (Ref. 5). The operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the requirements of 10 CFR 50, Appendix G (Ref. 2) .
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| Palisades Nuclear Plant B 3.4.3-1 Revised 01/29/2020
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| INSERT Bases 3.4.3 PCS P/T Limits B 3.4.3 BASES BACKGROUND A discussion of the methodology for the development of the P/T limit (continued) curves is provided in Reference 1 and Reference 7. The P/T limit curves were originally developed to be valid up to an accumulated reactor vessel wall fluence at the limiting circumferential weld of 2.192 x 1019 n/cm2 (E
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| >1.0 MeV). It was subsequently determined that this fluence would be reached prior to the operating license expiration date. In order to continue to use the existing P/T limit curves, an evaluation (Ref. 8) using more recently approved NRC methods was performed to demonstrate that the P/T limit curves are valid through the operating license expiration date, equivalent to 42.1 Effective Full Power Years (EFPY). This evaluation was performed using the adjusted RTNDT (ART) corresponding to the limiting beltline region material of the reactor vessel. The ART is defined as the sum of the initial reference temperature (RTNDT) of the material, the mean value for the adjustment in RTNDT caused by neutron irradiation, and a margin term to account for uncertainties in RTNDT, percent nickel, percent copper, neutron fluence and calculational procedures (Ref. 9).
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| The specific input parameters below were used to validate that the existing P/T limit curves are conservative through an applicability period of 42.1 EFPY. The input parameters are for the limiting reactor vessel material, which are the intermediate and lower shell axial welds 2-112 and 3-112.
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| : 1. A peak reactor vessel wall surface fluence of 2.161 x 1019 n/cm2 (E > 1.0 MeV)
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| : 2. ART values, at 1/4T = 252.7°F, and at 3/4T = 185.8°F
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| : 3. Initial RTNDT = -56 °F
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| : 4. Margin term = 65.5 °F The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.
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| Palisades Nuclear Plant B 3.4.3-2 Revised 01/29/2020
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| INSERT Bases 3.4.3 PCS P/T Limits B 3.4.3 BASES BACKGROUND The heatup curve represents a different set of restrictions than the (continued) cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal may alter the location of the tensile stress between the outer and inner walls.
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| The minimum temperature at which the reactor can be made critical, as required by Reference 2, shall be at least 40°F above the heatup curve or the cooldown curve and not less than the minimum permissible temperature for the ISLH testing. However, the criticality limit is not operationally limiting; a more restrictive limit exists in LCO 3.4.2, "PCS Minimum Temperature for Criticality, and LCO 3.1.7, Special Test Exceptions (STE)."
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| The consequence of violating the LCO limits is that the PCS has been operated under conditions that can result in brittle failure of the PCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the PCPB components.
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| The ASME Code, Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.
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| APPLICABLE The P/T limits are not derived from Design Basis Accident (DBA)
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| SAFETY ANALYSES Analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the PCPB, an unanalyzed condition. Reference 1 establishes the methodology for determining the P/T limits. Since the P/T limits are not derived from any DBA, there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.
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| The PCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2).
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| Palisades Nuclear Plant B 3.4.3-3 Revised 01/29/2020
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| INSERT Bases 3.4.3 PCS P/T Limits B 3.4.3 BASES LCO The two elements of this LCO are:
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| : a. The limit curves for heatup, cooldown, and ISLH testing; and
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| : b. Limits on the rate of change of temperature.
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| The LCO limits apply to all components of the PCS, except the pressurizer.
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| These limits define allowable operating regions and permit a large number of operating cycles while providing a wide margin to nonductile failure.
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| The limits for the rate of change of temperature control the thermal gradient through the vessel wall and are used as inputs for calculating the heatup, cooldown, and ISLH testing P/T limit curves. Additional cooldown rate restrictions were put in place due to the reactor vessel head nozzle repairs per Reference 7. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves.
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| Violating the LCO limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other PCPB components. The consequences depend on several factors, as follows:
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| : a. The severity of the departure from the allowable operating P/T regime or the severity of the rate of change of temperature;
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| : b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and
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| : c. The existences, sizes, and orientations of flaws in the vessel material.
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| APPLICABILITY The PCS P/T limits Specification provides a definition of acceptable operation for prevention of nonductile failure in accordance with 10 CFR 50, Appendix G (Ref. 2) and due to the reactor vessel nozzle repairs (Ref. 7). Although the P/T limits were developed to provide guidance for operation during heatup or cooldown (MODES 3, 4, and 5) or ISLH testing, their Applicability is at all times in keeping with the concern for nonductile failure. The additional cooldown rate restrictions for the reactor vessel nozzle repairs only apply when the reactor vessel head is on the reactor vessel. The limits do not apply to the pressurizer.
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| Palisades Nuclear Plant B 3.4.3-4 Revised 01/29/2020
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| INSERT Bases 3.4.3 PCS P/T Limits B 3.4.3 BASES APPLICABILITY During MODES 1 and 2, other Technical Specifications provide limits (continued) for operation that can be more restrictive than or can supplement these P/T limits. LCO 3.4.1, "PCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; LCO 3.4.2, "PCS Minimum Temperature for Criticality"; and Safety Limit 2.1, "Safety Limits," also provide operational restrictions for pressure and temperature and maximum pressure. Furthermore, MODES 1 and 2 are above the temperature range of concern for nonductile failure, and stress analyses have been performed for normal maneuvering profiles, such as power ascension or descent.
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| The actions of this LCO consider the premise that a violation of the limits occurred during normal plant maneuvering. Severe violations caused by abnormal transients, at times accompanied by equipment failures, may also require additional actions from emergency operating procedures.
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| ACTIONS A.1 and A.2 Operation outside the P/T limits must be corrected so that the PCPB is returned to a condition that has been verified by stress analyses.
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| The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.
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| Besides restoring operation to within limits, an evaluation is required to determine if PCS operation can continue. The evaluation must verify the PCPB integrity remains acceptable and must be completed before continuing operation. Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.
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| ASME Code, Section XI, Appendix E (Ref. 6), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.
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| The 72 hour Completion Time is reasonable to accomplish the evaluation.
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| The evaluation for a mild violation is possible within this time, but more severe violations may require special, event specific stress analyses or inspections. A favorable evaluation must be completed before continuing to operate.
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| Palisades Nuclear Plant B 3.4.3-5 Revised 01/29/2020
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| INSERT Bases 3.4.3 PCS P/T Limits B 3.4.3 BASES ACTIONS A.1 and A.2 (continued)
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| Condition A is modified by a Note requiring Required Action A.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the PCPB integrity.
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| B.1 and B.2 If a Required Action and associated Completion Time of Condition A are not met, the plant must be placed in a lower MODE because:
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| : a. The PCS remained in an unacceptable P/T region for an extended period of increased stress; or
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| : b. A sufficiently severe event caused entry into an unacceptable region.
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| Either possibility indicates a need for more careful examination of the event, best accomplished with the PCS at reduced pressure and temperature. With reduced pressure and temperature conditions, the possibility of propagation of undetected flaws is generally decreased.
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| Pressure and temperature are reduced by placing the plant in MODE 3 within 6 hours and in MODE 5 with PCS pressure < 270 psia within 36 hours.
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| The Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.4.3-6 Revised 01/29/2020
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| INSERT Bases 3.4.3 PCS P/T Limits B 3.4.3 BASES ACTIONS C.1 and C.2 (continued)
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| The actions of this LCO, anytime other than in MODE 1, 2, 3, or 4, consider the premise that a violation of the limits occurred during normal plant maneuvering. Severe violations caused by abnormal transients, at times accompanied by equipment failures, may also require additional actions from emergency operating procedures. Operation outside the P/T limits must be corrected so that the PCPB is returned to a condition that has been verified by stress analyses.
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| The Completion Time of "immediately" reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in a short period of time in a controlled manner.
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| Besides restoring operation to within limits, an evaluation is required to determine if PCS operation can continue. The evaluation must verify that the PCPB integrity remains acceptable and must be completed before continuing operation. Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.
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| ASME Code, Section XI, Appendix E (Ref. 6), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.
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| The Completion Time of prior to entering MODE 4 forces the evaluation prior to entering a MODE where temperature and pressure can be significantly increased. The evaluation for a mild violation is possible within several days, but more severe violations may require special, event specific stress analyses or inspections.
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| Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the PCPB integrity.
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| Palisades Nuclear Plant B 3.4.3-7 Revised 01/29/2020
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| INSERT Bases 3.4.3 PCS P/T Limits B 3.4.3 BASES SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the limits of Figure 3.4.3-1 and Figure 3.4.3-2 is required when PCS pressure and temperature conditions are undergoing planned changes. Calculation of the average hourly cooldown rate must consider changes in reactor vessel inlet temperature caused by initiating shutdown cooling, by starting primary coolant pumps with a temperature difference between the steam generator and PCS, or by stopping primary coolant pumps with shutdown cooling in service. The additional restrictions in Figure 3.4.3-2, required for the reactor vessel head nozzle repairs, use the average core exit temperature to provide the best indication available of the temperature of the head inside material temperature. This indication may be either the average of the core exit thermocouples or the vessel outlet temperature.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Surveillance for heatup and cooldown operations may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.
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| This SR is modified by a Note that requires this SR be performed only during PCS heatup and cooldown operations. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.
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| Palisades Nuclear Plant B 3.4.3-8 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.3 PCS P/T Limits B 3.4.3 BASES REFERENCES 1. Safety Evaluation for Palisades Nuclear Plant License Amendment No. 245, dated January 19, 2012
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| : 2. 10 CFR 50, Appendix G
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| : 3. Deleted
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| : 4. ASTM E 185-82, July 1982
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| : 5. 10 CFR 50, Appendix H
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| : 6. ASME, Boiler and Pressure Vessel Code, Section XI, Appendix E
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| : 7. Safety Evaluation for Palisades Nuclear Plant License Amendment No. 218, dated November 8, 2004
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| : 8. Engineering Analysis EA-EC27959-01, "Palisades Pressure-Temperature Limit Curves and Upper-Shelf Energy Evaluation,"
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| February 2012
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| : 9. Regulatory Guide 1.99, Revision 2, May 1988 Palisades Nuclear Plant B 3.4.3-9 Revised 01/29/2020
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| INSERT Bases 3.4.4 PCS Loops - MODES 1 and 2 B 3.4.4 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.4 PCS Loops - MODES 1 and 2 BASES BACKGROUND The primary function of the PCS is removal of the heat generated in the fuel due to the fission process and transfer of this heat, via the Steam Generators (SGs), to the secondary plant.
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| The secondary functions of the PCS include:
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| : a. Moderating the neutron energy level to the thermal state, to increase the probability of fission;
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| : b. Improving the neutron economy by acting as a reflector;
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| : c. Carrying the soluble neutron poison, boric acid;
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| : d. Providing a second barrier against fission product release to the environment; and
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| : e. Removing the heat generated in the fuel due to fission product decay following a plant shutdown.
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| The PCS configuration for heat transport uses two PCS loops. Each PCS loop contains an SG and two Primary Coolant Pumps (PCPs). A PCP is located in each of the two SG cold legs. The pump flow rate has been sized to provide core heat removal with appropriate margin to Departure from Nucleate Boiling (DNB) during power operation and for anticipated transients originating from power operation. This Specification requires two PCS loops with both PCPs in operation in each loop. The intent of the Specification is to require core heat removal with forced flow during power operation. Specifying two PCS loops provides the minimum necessary paths (two SGs) for heat removal.
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| APPLICABLE Safety analyses contain various assumptions for the Design Bases SAFETY ANALYSES Accident (DBA) initial conditions including PCS pressure, PCS temperature, reactor power level, core parameters, and safety system setpoints. The important aspect for this LCO is the primary coolant forced flow rate, which is represented by the number of PCS loops in service.
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| Palisades Nuclear Plant B 3.4.4-1 Revised 01/29/2020
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| INSERT Bases 3.4.4 PCS Loops - MODES 1 and 2 B 3.4.4 BASES APPLICABLE Both transient and steady state analyses have been performed to SAFETY ANALYSES establish the effect of flow on DNB. The transient or accident analysis (continued) for the plant has been performed assuming four PCPs are in operation. The majority of the plant safety analyses are based on initial conditions at high core power or zero power. The accident analyses that are of most importance to PCP operation are the Loss of Forced Primary Coolant Flow, Primary Coolant Pump Rotor Seizure and Uncontrolled Control Rod Withdrawal events (Ref. 1).
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| Steady state DNB analysis had been performed for the four pump combination. The steady state DNB analysis, which generates the pressure and temperature and Safety Limit (i.e., the Departure from Nucleate Boiling Ratio (DNBR) limit), assumes a maximum power level of 110.4% RTP. This is the design overpower condition for four pump operation. The 110.4% value is the accident analysis setpoint of the trip and is based on an analysis assumption that bounds possible instrumentation errors. The DNBR limit defines a locus of pressure and temperature points that result in a minimum DNBR greater than or equal to the critical heat flux correlation limit.
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| PCS Loops - MODES 1 and 2 satisfy Criteria 2 and 3 of 10 CFR 50.36(c)(2).
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| LCO The purpose of this LCO is to require adequate forced flow for core heat removal. Flow is represented by having both PCS loops with both PCPs in each loop in operation for removal of heat by the two SGs. To meet safety analysis acceptance criteria for DNB, four pumps are required at rated power.
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| Each OPERABLE loop consists of two PCPs providing forced flow for heat transport to an SG that is OPERABLE. SG, and hence PCS loop OPERABILITY with regards to SG water level is ensured by the Reactor Protection System (RPS) in MODES 1 and 2. A reactor trip places the plant in MODE 3 if any SG water level is 25.9% (narrow range) as sensed by the RPS. The minimum level to declare the SG OPERABLE is 25.9% (narrow range).
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| In MODES 1 and 2, the reactor can be critical and thus has the potential to produce maximum THERMAL POWER. Thus, to ensure that the assumptions of the accident analyses remain valid, all PCS loops are required to be in operation in these MODES to prevent DNB and core damage.
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| Palisades Nuclear Plant B 3.4.4-2 Revised 01/29/2020
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| INSERT Bases 3.4.4 PCS Loops - MODES 1 and 2 B 3.4.4 BASES APPLICABILITY The decay heat production rate is much lower than the full power heat rate. As such, the forced circulation flow and heat sink requirements are reduced for lower, noncritical MODES as indicated by the LCOs for MODES 3, 4, 5, and 6.
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| Operation in other MODES is covered by:
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| LCO 3.4.5, "PCS Loops-MODE 3";
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| LCO 3.4.6, "PCS Loops-MODE 4";
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| LCO 3.4.7, "PCS Loops-MODE 5, Loops Filled";
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| LCO 3.4.8, "PCS Loops-MODE 5, Loops Not Filled";
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| LCO 3.9.4, "Shutdown Cooling (SDC) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Shutdown Cooling (SDC) and Coolant Circulation-Low Water Level" (MODE 6).
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| ACTIONS A.1 If the requirements of the LCO are not met, the Required Action is to reduce power and bring the plant to MODE 3. This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits. It should be noted that the reactor will trip and place the plant in MODE 3 as soon as the RPS senses less than four PCPs operating.
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| The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems.
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| Palisades Nuclear Plant B 3.4.4-3 Revised 01/29/2020
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| INSERT Bases 3.4.4 PCS Loops - MODES 1 and 2 B 3.4.4 BASES SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This SR requires verification of the required number of loops in operation.
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| Verification may include indication of PCS flow, temperature, or pump status, which help to ensure that forced flow is providing heat removal while maintaining the margin to DNB. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Section 14.1 Palisades Nuclear Plant B 3.4.4-4 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.5 PCS Loops - MODE 3 B 3.4.5 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.5 PCS Loops - MODE 3 BASES BACKGROUND The primary function of the primary coolant in MODE 3 is removal of decay heat and transfer of this heat, via the Steam Generators (SGs), to the secondary plant fluid. The secondary function of the primary coolant is to act as a carrier for soluble neutron poison, boric acid.
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| In MODE 3, Primary Coolant Pumps (PCPs) are used to provide forced circulation heat removal during heatup and cooldown. The MODE 3 decay heat removal requirements are low enough that a single PCS loop with one PCP is sufficient to remove core decay heat. However, two PCS loops are required to be OPERABLE to provide redundant paths for decay heat removal. Any combination of OPERABLE PCPs and OPERABLE PCS loops can be used to fulfill the heat removal function.
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| Primary coolant natural circulation is not normally used but is sufficient for core cooling. However, natural circulation does not provide turbulent flow conditions. Therefore, boron reduction in natural circulation is prohibited because mixing to obtain a homogeneous concentration in all portions of the PCS cannot be ensured. Any combination of OPERABLE PCPs and OPERABLE PCS loops can be used to fulfill the mixing function.
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| APPLICABLE Failure to provide heat removal may result in challenges to a fission SAFETY ANALYSES product barrier. The PCS loops are part of the primary success path that functions or actuates to prevent or mitigate a Design Basis Accident or transient that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier.
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| PCS Loops - MODE 3 satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| LCO The purpose of this LCO is to require two PCS loops to be available for heat removal, thus providing redundancy. The LCO requires the two loops to be OPERABLE with the intent of requiring both SGs to be capable (> -84% water level) of transferring heat from the primary coolant at a controlled rate. Forced primary coolant flow is the required way to transport heat, although natural circulation flow provides adequate removal. A minimum of one running PCP meets the LCO requirement for one loop in operation.
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| Palisades Nuclear Plant B 3.4.5-1 Revised 01/29/2020
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| INSERT Bases 3.4.5 PCS Loops - MODE 3 B 3.4.5 BASES LCO Note 1 permits all PCPs to not be in operation 1 hour per 8 hour (continued) period. This means that natural circulation has been established using the SGs. The Note prohibits boron dilution when forced flow is stopped because an even concentration distribution cannot be ensured. Core outlet temperature is to be maintained at least 10°F below saturation temperature so that no vapor bubble may form and possibly cause a natural circulation flow obstruction. The response of the PCS without the PCPs depends on the core decay heat load and the length of time that the pumps are stopped. As decay heat diminishes, the effects on PCS temperature and pressure diminish. Without cooling by forced flow, higher heat loads will cause the reactor coolant temperature and pressure to increase at a rate proportional to the decay heat load. Because pressure can increase, the applicable system pressure limits (Pressure and Temperature (P/T) limits or Low Temperature Overpressure Protection (LTOP) limits) must be observed and forced flow or heat removal via the SGs must be re-established prior to reaching the pressure limit. The circumstances for stopping the PCPs are to be limited to situations where:
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| : a. Pressure and temperature increases can be maintained well within the allowable pressure (P/T limits and LTOP) and 10°F subcooling limits; or
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| : b. An alternate heat removal path through the SGs is in operation.
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| In MODE 3, it is sometimes necessary to stop all PCP forced circulation.
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| This is permitted to perform surveillance or startup testing, to perform the transition to and from SDC, or to avoid operation below the PCP minimum net positive suction head limit. The time period is acceptable because natural circulation is adequate for heat removal, or the reactor coolant temperature can be maintained subcooled and boron stratification affecting reactivity control is not expected.
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| Note 2 requires that one of the following conditions be satisfied before forced circulation (starting the first PCP) may be started:
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| : a. PCS cold leg temperature (Tc) is > 430°F;
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| : b. SG secondary temperature is equal to or less than the reactor inlet temperature (Tc);
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| : c. SG secondary temperature is < 100°F above Tc, and shutdown cooling is isolated from the PCS, and PCS heatup/cooldown rate is 10°F/hour; or Palisades Nuclear Plant B 3.4.5-2 Revised 01/29/2020
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| INSERT Bases 3.4.5 PCS Loops - MODE 3 B 3.4.5 BASES LCO d. SG secondary temperature is < 100 °F above Tc, and shutdown (continued) cooling is isolated from the PCS, and pressurizer level is 57%.
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| Satisfying any of the above conditions will preclude a large pressure surge in the PCS when the PCP is started. Energy additions from the steam generators could occur if a PCP was started when the steam generator secondary temperature is significantly above the PCS temperature. The maximum pressurizer level at which credit is taken for having a bubble (57%, which provides about 700 cubic feet of steam space) is based on engineering judgement and verified by LTOP analysis.
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| This level provides the same steam volume to dampen pressure transients as would be available at full power.
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| An OPERABLE PCS loop consists of any one (of the four) OPERABLE PCP and an SG that is OPERABLE and has the minimum water level specified in SR 3.4.5.2. A PCP is OPERABLE if it is capable of being powered and is able to provide forced flow if required.
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| APPLICABILITY In MODE 3, the heat load is lower than at power; therefore, one PCS loop in operation is adequate for transport and heat removal. A second PCS loop is required to be OPERABLE but is not required to be in operation for redundant heat removal capability.
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| Operation in other MODES is covered by:
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| LCO 3.4.4, "PCS Loops-MODES 1 and 2";
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| LCO 3.4.6, "PCS Loops-MODE 4";
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| LCO 3.4.7, "PCS Loops-MODE 5, Loops Filled";
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| LCO 3.4.8, "PCS Loops-MODE 5, Loops Not Filled";
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| LCO 3.9.4, "Shutdown Cooling (SDC) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Shutdown Cooling (SDC) and Coolant Circulation-Low Water Level" (MODE 6)
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| Palisades Nuclear Plant B 3.4.5-3 Revised 01/29/2020
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| INSERT Bases 3.4.5 PCS Loops - MODE 3 B 3.4.5 BASES (continued)
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| ACTIONS A.1 If one required PCS loop is inoperable, redundancy for forced flow heat removal is lost. The Required Action is restoration of the required PCS loop to OPERABLE status within a Completion Time of 72 hours. This time allowance is a justified period to be without the redundant, non-operating loop because a single loop in operation has a heat transfer capability greater than that needed to remove the decay heat produced in the reactor core. Mechanical system LCOs typically provide a 72 hour Completion Time under conditions when a required system can perform its required safety function, but may not be able to do so assuming an additional failure. When operating in accordance with the Required Actions of an LCO Condition, it is not necessary to be able to cope with an additional single failure.
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| B.1 If restoration is not possible within 72 hours, the plant must be placed in MODE 4 within 24 hours. In MODE 4, the plant may be placed on the SDC System. The Completion Time of 24 hours is compatible with required operation to achieve cooldown and depressurization from the existing plant conditions in an orderly manner and without challenging plant systems.
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| C.1 and C.2 If no PCS loop is in operation, except as provided in Note 1 in the LCO section, all operations involving a reduction of PCS boron concentration must be immediately suspended. This is necessary because boron dilution requires forced circulation for proper homogenization. Action to restore one PCS loop to OPERABLE status and operation shall be initiated immediately and continued until one PCS loop is restored to OPERABLE status and operation. The immediate Completion Times reflect the importance of maintaining operation for decay heat removal.
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| Palisades Nuclear Plant B 3.4.5-4 Revised 01/29/2020
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| INSERT Bases 3.4.5 PCS Loops - MODE 3 B 3.4.5 BASES (continued)
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| SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR requires verification that the required number of PCS loops are in operation. Verification include indication of PCS flow, temperature, and pump status, which help ensure that forced flow is providing heat removal and mixing of the soluble boric acid. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.4.5.2 This SR requires verification that the secondary side water level in each SG is -84% using the wide range level instrumentation. An adequate SG water level is required in order to have a heat sink for removal of the core decay heat from the primary coolant. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.4.5.3 Verification that the required PCP is OPERABLE ensures that the single failure criterion is met and that an additional PCS loop can be placed in operation, if needed, to maintain decay heat removal and primary coolant circulation. Verification is performed by verifying proper breaker alignment and power availability to the required PCP that is not in operation such that the PCP is capable of being started and providing forced PCS flow if needed. Proper breaker alignment and power availability means the breaker for the required PCP is racked-in and electrical power is available to energize the PCP motor. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES None Palisades Nuclear Plant B 3.4.5-5 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.6 PCS Loops - MODE 4 B 3.4.6 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.6 PCS Loops - MODE 4 BASES BACKGROUND In MODE 4, the primary function of the primary coolant is the removal of decay heat and transfer of this heat to the Steam Generators (SGs) or Shutdown Cooling (SDC) heat exchangers. The secondary function of the primary coolant is to act as a carrier for soluble neutron poison, boric acid.
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| In MODE 4, either Primary Coolant Pumps (PCPs) or SDC trains can be used for coolant circulation. The intent of this LCO is to provide forced flow from any one (of the four) PCP or one SDC train for decay heat removal and transport. The flow provided by one PCP loop or SDC train is adequate for heat removal. The other intent of this LCO is to require that two paths be available to provide redundancy for heat removal.
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| APPLICABLE The boron concentration must be uniform throughout the PCS volume SAFETY ANALYSES to prevent stratification of primary coolant at lower boron concentrations which could result in a reactivity insertion. Sufficient mixing of the primary coolant is assured if one PCP is in operation. PCS circulation is considered in the determination of the time available for mitigation of the inadvertent boron dilution event. By imposing a minimum flow through the reactor core of 2810 gpm, sufficient time is provided for the operator to terminate a boron dilution under asymmetric flow conditions. Due to its system configuration (i.e., no throttle valves) and large volumetric flow rate, a minimum flow rate is not imposed on the PCPs.
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| PCS Loops - MODE 4 satisfies Criterion 4 of 10 CFR 50.36(c)(2).
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| LCO The purpose of this LCO is to require that two loops or trains, PCS or SDC, be OPERABLE in MODE 4 and one of these loops or trains to be in operation. The LCO allows the two loops that are required to be OPERABLE to consist of any combination of PCS and SDC System loops. Any one PCS loop in operation, or SDC in operation with a flow 2810 gpm through the reactor core, provides enough flow to remove the decay heat from the core with forced circulation and provide sufficient mixing of the soluble boric acid. An additional loop or train is required to be OPERABLE to provide redundancy for heat removal.
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| Palisades Nuclear Plant B 3.4.6-1 Revised 01/29/2020
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| INSERT Bases 3.4.6 PCS Loops - MODE 4 B 3.4.6 BASES LCO A SDC train may be considered OPERABLE (but not necessarily in (continued) operation) during re-alignment to, and when it is re-aligned for, LPSI service or for testing, if it is capable of being (locally or remotely) realigned to the SDC mode of operation and is not otherwise inoperable. Since SDC is a manually initiated system, it need not be considered inoperable solely because some additional manual valve realignments must be made in addition to the normal initiation actions. Because of the dual functions of the components that comprise the LPSI and shutdown cooling systems, the LPSI alignment may be preferred.
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| Note 1 permits all PCPs and SDC pumps to not be in operation 1 hour per 8 hour period. The Note prohibits boron dilution when forced flow is stopped because an even concentration distribution cannot be ensured. Core outlet temperature is to be maintained at least 10°F below saturation temperature so that no vapor bubble may form and possibly cause a natural circulation flow obstruction. The response of the PCS without the PCPs or SDC pumps depends on the core decay heat load and the length of time that the pumps are stopped. As decay heat diminishes, the effects on PCS temperature and pressure diminish. Without cooling by forced flow, higher heat loads will cause the primary coolant temperature and pressure to increase at a rate proportional to the decay heat load. Because pressure can increase, the applicable system pressure limits (Pressure and Temperature (P/T) limits or Low Temperature Overpressure Protection (LTOP) limits) must be observed and forced SDC flow or heat removal via the SGs must be re-established prior to reaching the pressure limit. The circumstances for stopping both PCPs or SDC pumps are to be limited to situations where:
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| : a. Pressure and temperature increases can be maintained well within the allowable pressure (P/T limits and LTOP) and 10°F subcooling limits; or
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| : b. An alternate heat removal path through the SGs is in operation.
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| In MODE 4, it is sometimes necessary to stop all PCPs or SDC forced circulation. This is permitted to change operation from one SDC train to the other, perform surveillance or startup testing, perform the transition to and from SDC, or to avoid operation below the PCP minimum net positive suction head limit. The time period is acceptable because natural circulation is acceptable for decay heat removal, the primary coolant temperature can be maintained subcooled, and boron stratification affecting reactivity control is not expected.
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| Palisades Nuclear Plant B 3.4.6-2 Revised 01/29/2020
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| INSERT Bases 3.4.6 PCS Loops - MODE 4 B 3.4.6 BASES LCO Note 2 requires that one of the following conditions be satisfied before (continued) forced circulation (starting the first PCP) may be started:
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| : a. SG secondary temperature is Tc;
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| : b. SG secondary temperature is < 100°F above Tc, and shutdown cooling is isolated from the PCS, and PCS heatup/cooldown rate is 10°F/hour; or
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| : c. SG secondary temperature is < 100°F above Tc, and shutdown cooling is isolated from the PCS, and pressurizer level is 57%.
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| Satisfying any of the above conditions will preclude a large pressure surge in the PCS when the PCP is started. Energy additions from the steam generators could occur if a PCP was started when the steam generator secondary temperature is significantly above the PCS temperature. The maximum pressurizer level at which credit is taken for having a bubble (57%, which provides about 700 cubic feet of steam space) is based on engineering judgement and verified by LTOP analysis.
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| This level provides the same steam volume to dampen pressure transients as would be available at full power.
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| Note 3 specifies a limitation on the simultaneous operation of primary coolant pumps P-50A and P-50B which allows the pressure limits in LCO 3.4.3, PCS Pressure and Temperature Limits, and LCO 3.4.12, Low Temperature Overpressure Protection System, to be higher than they would be without this limit. This is because the pressure in the reactor vessel downcomer region when primary coolant pumps P-50A and P-50B are operated simultaneously is higher than the pressure for other two primary coolant pump combinations.
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| An OPERABLE PCS loop consists of any one (of the four) OPERABLE PCP and an SG that has the minimum water level specified in SR 3.4.6.2 and is OPERABLE. PCPs are OPERABLE if they are capable of being powered and are able to provide forced flow through the reactor core.
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| An OPERABLE SDC train is composed of an OPERABLE SDC pump and an OPERABLE SDC heat exchanger. The two SDC heat exchangers operate as a single unit. A separate OPERABLE SDC heat exchanger is required for each OPERABLE SDC train. SDC pumps are OPERABLE if they are capable of being powered and are able to provide forced flow through the reactor core.
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| Palisades Nuclear Plant B 3.4.6-3 Revised 01/29/2020
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| INSERT Bases 3.4.6 PCS Loops - MODE 4 B 3.4.6 BASES (continued)
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| APPLICABILITY In MODE 4, this LCO applies because it is possible to remove core decay heat and to provide proper boron mixing with either the PCS loops and SGs, or the SDC System.
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| Operation in other MODES is covered by:
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| LCO 3.4.4, "PCS Loops-MODES 1 and 2";
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| LCO 3.4.5, "PCS Loops-MODE 3";
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| LCO 3.4.7, "PCS Loops-MODE 5, Loops Filled";
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| LCO 3.4.8, "PCS Loops-MODE 5, Loops Not Filled";
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| LCO 3.9.4, "Shutdown Cooling (SDC) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Shutdown Cooling (SDC) and Coolant Circulation-Low Water Level" (MODE 6).
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| ACTIONS A.1 If only one PCS loop is OPERABLE and in operation with no OPERABLE SDC trains, redundancy for heat removal is lost. Action must be initiated immediately to restore a second PCS loop or one SDC train to OPERABLE status. The immediate Completion Time reflects the importance of maintaining the availability of two paths for decay heat removal.
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| B.1 If only one SDC train is OPERABLE and in operation with no OPERABLE PCS loops, redundancy for heat removal is lost. The plant must be placed in MODE 5 within the next 24 hours. Placing the plant in MODE 5 is a conservative action with regard to decay heat removal. With only one SDC train OPERABLE, redundancy for decay heat removal is lost and, in the event of a loss of the remaining SDC train, it would be safer to initiate that loss from MODE 5 (< 200ºF) rather than MODE 4 (> 200ºF to < 300ºF). The Completion Time of 24 hours is reasonable, based on operating experience, to reach MODE 5 from MODE 4, with only one SDC train operating, in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.4.6-4 Revised 01/29/2020
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| INSERT Bases 3.4.6 PCS Loops - MODE 4 B 3.4.6 BASES ACTIONS C.1, C.2.1, and C.2.2 (continued)
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| If no PCS loops or SDC trains are OPERABLE, or no PCS loop is in operation and the SDC flow through the reactor core is < 2810 gpm, except during conditions permitted by Note 1 in the LCO section, all operations involving reduction of PCS boron concentration must be suspended. Action to restore one PCS loop or SDC train to OPERABLE status and operation shall be initiated immediately and continue until one loop or train is restored to operation and flow through the reactor core is restored to 2810 gpm. Boron dilution requires forced circulation for proper mixing, and the margin to criticality must not be reduced in this type of operation. The immediate Completion Times reflect the importance of decay heat removal.
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| SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires verification that one required loop or train is in operation.
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| This ensures forced flow is providing heat removal and mixing of the soluble boric acid. Verification may include flow rate (SDC only), or indication of flow, temperature, or pump status for the PCP. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.4.6.2 This SR requires verification of secondary side water level in the required SG(s) -84% using the wide range level instrumentation. An adequate SG water level is required in order to have a heat sink for removal of the core decay heat from the primary coolant. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.4.6-5 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.6 PCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE SR 3.4.6.3 REQUIREMENTS (continued) Verification that the required pump is OPERABLE ensures that an additional PCS loop or SDC train can be placed in operation, if needed to maintain decay heat removal and primary coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump that is not in operation such that the pump is capable of being started and providing forced PCS flow if needed. Proper breaker alignment and power availability means the breaker for the required pump is racked-in and electrical power is available to energize the pump motor The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES None Palisades Nuclear Plant B 3.4.6-6 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.7 PCS Loops - MODE 5, Loops Filled B 3.4.7 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.7 PCS Loops - MODE 5, Loops Filled BASES BACKGROUND In MODE 5 with the PCS loops filled, the primary function of the primary coolant is the removal of decay heat and transfer this heat either to the Steam Generator (SG) secondary side coolant via natural circulation (Ref.
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| : 1) or the Shutdown Cooling (SDC) heat exchangers. While the principal means for decay heat removal is via the SDC System, the SGs via natural circulation are specified as a backup means for redundancy.
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| Even though the SGs cannot produce steam in this MODE, they are capable of being a heat sink due to their large contained volume of secondary side water. If heatup of the PCS were to continue, the contained inventory of the SGs would be available to remove decay heat by producing steam. As long as the SG secondary side water is at a lower temperature than the primary coolant, heat transfer will occur. The rate of heat transfer is directly proportional to the temperature difference.
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| The secondary function of the primary coolant is to act as a carrier for soluble neutron poison, boric acid.
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| In MODE 5 with PCS loops filled, the SDC trains are the principal means for decay heat removal. The number of trains in operation can vary to suit the operational needs. The intent of this LCO is to provide forced flow from at least one SDC train for decay heat removal and transport.
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| The flow provided by one SDC train is adequate for decay heat removal.
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| The other intent of this LCO is to require that a second path be available to provide redundancy for decay heat removal.
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| The LCO provides for redundant paths of decay heat removal capability.
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| The first path can be an SDC train that must be OPERABLE and in operation. The second path can be another OPERABLE SDC train, or through the SGs, via natural circulation each having an adequate water level. Loops filled means the PCS loops are not blocked by dams and totally filled with coolant.
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| Palisades Nuclear Plant B 3.4.7-1 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.7 PCS Loops - MODE 5, Loops Filled B 3.4.7 BASES (continued)
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| APPLICABLE The boron concentration must be uniform throughout the PCS SAFETY ANALYSES volume to prevent stratification of primary coolant at lower boron concentrations which could result in a reactivity insertion. Sufficient mixing of the primary coolant is assured if one SDC pump is in operation.
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| PCS circulation is considered in the determination of the time available for mitigation of the inadvertent boron dilution event. By imposing a minimum flow through the reactor core of 2810 gpm, sufficient time is provided for the operator to terminate a boron dilution under asymmetric flow conditions.
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| PCS Loops - MODE 5 (Loops Filled) satisfies Criterion 4 of 10 CFR 50.36(c)(2).
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| LCO The purpose of this LCO is to require one SDC train be OPERABLE and in operation with either an additional SDC train OPERABLE or the secondary side water level of each SG -84%. SDC in operation with a flow through the reactor core 2810 gpm, provides enough flow to remove the decay heat from the core with forced circulation and provide sufficient mixing of the soluble boric acid. The second SDC train is normally maintained OPERABLE as a backup to the operating SDC train to provide redundant paths for decay heat removal. However, if the standby SDC train is not OPERABLE, a sufficient alternate method to provide redundant paths for decay heat removal is two SGs with their secondary side water levels -84%. Should the operating SDC train fail, the SGs could be used to remove the decay heat via natural circulation.
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| A SDC train may be considered OPERABLE (but not necessarily in operation) during re-alignment to, and when it is re-aligned for, LPSI service or for testing, if it is capable of being (locally or remotely) realigned to the SDC mode of operation and is not otherwise inoperable.
| |
| Since SDC is a manually initiated system, it need not be considered inoperable solely because some additional manual valve realignments must be made in addition to the normal initiation actions. Because of the dual functions of the components that comprise the LPSI and shutdown cooling systems, the LPSI alignment may be preferred.
| |
| Note 1 permits all SDC pumps to not be in operation 1 hour per 8 hour period. The Note prohibits boron dilution when forced flow is stopped because an even concentration distribution cannot be ensured. Core outlet temperature is to be maintained at least Palisades Nuclear Plant B 3.4.7-2 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.7 PCS Loops - MODE 5, Loops Filled B 3.4.7 BASES LCO 10°F below saturation temperature so that no vapor bubble may form (continued) and possibly cause a natural circulation flow obstruction. The response of the PCS without the SDC pumps depends on the core decay heat load and the length of time that the pumps are stopped.
| |
| As decay heat diminishes, the effects on PCS temperature and pressure diminish. Without cooling by forced flow, higher heat loads will cause the primary coolant temperature and pressure to increase at a rate proportional to the decay heat load. Because pressure can increase, the applicable system pressure limits (Pressure and Temperature (P/T) limits or Low Temperature Overpressure Protection (LTOP) limits) must be observed and forced SDC flow or heat removal via the SGs must be re-established prior to reaching the pressure limit.
| |
| In MODE 5 with loops filled, it is sometimes necessary to stop all SDC forced circulation. This is permitted to change operation from one SDC train to the other, perform surveillance or startup testing, perform the transition to and from the SDC, or to avoid operation below the PCP minimum net positive suction head limit. The time period is acceptable because natural circulation is acceptable for decay heat removal, the primary coolant temperature can be maintained subcooled, and boron stratification affecting reactivity control is not expected.
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| Note 2 allows both SDC trains to be inoperable for a period of up to 2 hours provided that one SDC train is in operation providing the required flow, the core outlet temperature is at least 10°F below the corresponding saturation temperature, and each SG secondary water level is 84%. This permits periodic surveillance tests or maintenance to be performed on the inoperable trains during the only time when such evolutions are safe and possible.
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| Note 3 requires that one of the following conditions be satisfied before forced circulation (starting the first PCP) may be started:
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| : a. SG secondary temperature is equal to or less than the reactor inlet temperature (Tc);
| |
| : b. SG secondary temperature is < 100°F above Tc, and shutdown cooling is isolated from the PCS, and PCS heatup/cooldown rate is 10°F/hour; or
| |
| : c. SG secondary temperature is < 100°F above Tc, and shutdown cooling is isolated from the PCS, and pressurizer level is 57%.
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| Palisades Nuclear Plant B 3.4.7-3 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.7 PCS Loops - MODE 5, Loops Filled B 3.4.7 BASES LCO Satisfying any of the above conditions will preclude a large pressure (continued) surge in the PCS when the PCP is started. Energy additions from the steam generators could occur if a PCP was started when the steam generator secondary temperature is significantly above the PCS temperature. The maximum pressurizer level at which credit is taken for having a bubble (57%, which provides about 700 cubic feet of steam space) is based on engineering judgement and verified by LTOP analysis.
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| This level provides the same steam volume to dampen pressure transients as would be available at full power.
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| Note 4 specifies a limitation on the simultaneous operation of primary coolant pumps P-50A and P-50B which allows the pressure limits in LCO 3.4.3, PCS Pressure and Temperature Limits, and LCO 3.4.12, Low Temperature Overpressure Protection System, to be higher than they would be without this limit.
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| Note 5 provides for an orderly transition from MODE 5 to MODE 4 during a planned heatup by permitting SDC trains to not be in operation when at least one PCP is in operation. This Note provides for the transition to MODE 4 where a PCP is permitted to be in operation and replaces the PCS circulation function provided by the SDC trains.
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| An OPERABLE SDC train is composed of an OPERABLE SDC pump and an OPERABLE SDC heat exchanger. The two SDC heat exchangers operate as a single unit. A separate OPERABLE SDC heat exchanger is required for each OPERABLE SDC train. SDC pumps are OPERABLE if they are capable of being powered and are able to provide forced flow through the reactor core.
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| An SG can perform as a heat sink via natural circulation when:
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| : a. SG has the minimum water level specified in SR 3.4.7.2.
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| : b. SG is OPERABLE.
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| : c. SG has available method of feedwater addition and a controllable path for steam release.
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| : d. Ability to pressurize and control pressure in the PCS.
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| If both SGs do not meet the above provisions, then LCO 3.4.7 item b (i.e.
| |
| the secondary side water level of each SG shall be -84%) is not met.
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| Palisades Nuclear Plant B 3.4.7-4 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.7 PCS Loops - MODE 5, Loops Filled B 3.4.7 BASES (continued)
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| APPLICABILITY In MODE 5 with PCS loops filled, this LCO requires forced circulation to remove decay heat from the core and to provide proper boron mixing.
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| One SDC train provides sufficient circulation for these purposes.
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| Operation in other MODES is covered by:
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| LCO 3.4.4, "PCS Loops-MODES 1 and 2";
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| LCO 3.4.5, "PCS Loops-MODE 3";
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| LCO 3.4.6, "PCS Loops-MODE 4";
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| LCO 3.4.8, "PCS Loops-MODE 5, Loops Not Filled";
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| LCO 3.9.4, "Shutdown Cooling (SDC) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Shutdown Cooling (SDC) and Coolant Circulation-Low Water Level" (MODE 6).
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| Palisades Nuclear Plant B 3.4.7-5 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.7 PCS Loops - MODE 5, Loops Filled B 3.4.7 BASES (continued)
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| ACTIONS A.1 and A.2 If one SDC train is inoperable and any SG has a secondary side water level < -84% (refer to LCO Bases section), redundancy for heat removal is lost. Action must be initiated immediately to restore a second SDC train to OPERABLE status or to restore the water level in the required SGs. Either Required Action A.1 or Required Action A.2 will restore redundant decay heat removal paths. The immediate Completion Times reflect the importance of maintaining the availability of two paths for decay heat removal.
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| B.1 and B.2 If no SDC trains are OPERABLE or SDC flow through the reactor core is
| |
| < 2810 gpm, except as permitted in Note 1, all operations involving the reduction of PCS boron concentration must be suspended. Action to restore one SDC train to OPERABLE status and operation shall be initiated immediately and continue until one train is restored to operation and flow through the reactor core is restored to 2810 gpm. Boron dilution requires forced circulation for proper mixing and the margin to criticality must not be reduced in this type of operation. The immediate Completion Times reflect the importance of maintaining operation for decay heat removal.
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| SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires verification that one SDC train is in operation.
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| Verification of the required flow rate ensures forced flow is providing heat removal and mixing of the soluble boric acid. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
| |
| Palisades Nuclear Plant B 3.4.7-6 Amendment No. 271 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.7 PCS Loops - MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE SR 3.4.7.2 REQUIREMENTS (continued) This SR requires verification of secondary side water level in the required SGs -84% using the wide range level instrumentation. An adequate SG water level is required in order to have a heat sink for removal of the core decay heat from the primary coolant. The Surveillance is required to be performed when the LCO requirement is being met by use of the SGs. If both SDC trains are OPERABLE, this SR is not needed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.4.7.3 Verification that the second SDC train is OPERABLE ensures that redundant paths for decay heat removal are available. The requirement also ensures that the additional train can be placed in operation, if needed, to maintain decay heat removal and primary coolant circulation.
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| Verification is performed by verifying proper breaker alignment and power available to the required pump that is not in operation such that the SDC pump is capable of being started and providing forced PCS flow if needed. Proper breaker alignment and power availability means the breaker for the required SDC pump is racked-in and electrical power is available to energize the SDC pump motor. The Surveillance is required to be performed when the LCO requirement is being met by one of two SDC trains, e.g., both SGs have < -84% water level. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. NRC Information Notice 95-35, Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation Palisades Nuclear Plant B 3.4.7-7 Amendment No. 271 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.8 PCS Loops - MODE 5, Loops Not Filled B 3.4.8 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.8 PCS Loops - MODE 5, Loops Not Filled BASES BACKGROUND In MODE 5 with the PCS loops not filled, the primary function of the primary coolant is the removal of decay heat and transfer of this heat to the Shutdown Cooling (SDC) heat exchangers. The Steam Generators (SGs) are not available as a heat sink when the loops are not filled. The secondary function of the primary coolant is to act as a carrier for the soluble neutron poison, boric acid. A loop is considered not filled if it has been drained so air has entered the loop which has not yet been removed.
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| In MODE 5 with loops not filled, only the SDC System can be used for coolant circulation. The number of trains in operation can vary to suit the operational needs. The intent of this LCO is to provide forced flow from at least one SDC train for decay heat removal and transport and to require that two paths be available to provide redundancy for heat removal.
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| APPLICABLE The boron concentration must be uniform throughout the PCS SAFETY ANALYSES volume to prevent stratification of primary coolant at lower boron concentrations which could result in a reactivity insertion. Sufficient mixing of the primary coolant is assured if one SDC pump is in operation.
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| PCS circulation is considered in the determination of the time available for mitigation of the inadvertent boron dilution event. By imposing a minimum flow through the reactor core of 2810 gpm, or a minimum flow through the reactor core 650 gpm with two of the three charging pumps incapable of reducing the boron concentration in the PCS below the minimum value necessary to maintain the required SHUTDOWN MARGIN, sufficient time is provided for the operator to terminate a boron dilution under asymmetric flow conditions.
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| PCS loops - MODE 5 (Loops Not Filled) satisfies Criterion 4 of 10 CFR 50.36(c)(2).
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| Palisades Nuclear Plant B 3.4.8-1 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.8 PCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES LCO The purpose of this LCO is to require a minimum of two SDC trains be OPERABLE and one of these trains be in operation. SDC in operation with a flow rate through the reactor core of 2810 gpm, or with a flow rate through the reactor core of 650 gpm with two of the three charging pumps incapable of reducing the boron concentration in the PCS below the minimum value necessary to maintain the required SHUTDOWN MARGIN, provides enough flow to remove the decay heat from the core with forced circulation and provide sufficient mixing of the soluble boric acid. The restriction on charging pump operations only applies to those cases where the potential exists to reduce the PCS boron concentration below minimum the boron concentration necessary to maintain the required SHUTDOWN MARGIN. It is not the intent of this LCO to restrict charging pump operations when the source of water to the pump suction is greater than or equal to the minimum boron concentration necessary to maintain the required SHUTDOWN MARGIN. An additional SDC train is required to be OPERABLE to meet the single failure criterion.
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| A SDC train may be considered OPERABLE (but not necessarily in operation) during re-alignment to, and when it is re-aligned for, LPSI service or for testing, if it is capable of being (locally or remotely) realigned to the SDC mode of operation and is not otherwise inoperable.
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| Since SDC is a manually initiated system, it need not be considered inoperable solely because some additional manual valve realignments must be made in addition to the normal initiation actions. Because of the dual functions of the components that comprise the LPSI and shutdown cooling systems, the LPSI alignment may be preferred.
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| Note 1 permits all SDC pumps to not be in operation for 1 hour. The Note prohibits boron dilution when forced flow is stopped because an even concentration distribution cannot be ensured. Core outlet temperature is to be maintained at least 10°F below saturation temperature so that no vapor bubble may form and possibly cause a flow obstruction. Operations which could drain the PCS and thereby cause a loss of, or failure to regain SDC capability are also prohibited.
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| In MODE 5 with loops not filled, it is sometimes necessary to stop all SDC forced circulation. This is permitted to change operation from one SDC train to the other, and to perform surveillance or startup testing. The time period is acceptable because the primary coolant will be maintained subcooled, and boron stratification affecting reactivity control is not expected.
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| Palisades Nuclear Plant B 3.4.8-2 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.8 PCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES LCO Note 2 allows one SDC train to be inoperable for a period of 2 hours (continued) provided that the other train is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable train during the only time when these tests are safe and possible.
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| An OPERABLE SDC train is composed of an OPERABLE SDC pump and an OPERABLE SDC heat exchanger. The two SDC heat exchangers operate as a single unit. A separate OPERABLE SDC heat exchanger is required for each OPERABLE SDC train. SDC pumps are OPERABLE if they are capable of being powered and are able to provide forced flow through the reactor core.
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| APPLICABILITY In MODE 5 with PCS loops not filled, this LCO requires forced circulation to remove decay heat from the core and to provide proper boron mixing.
| |
| One SDC train provides sufficient circulation for these purposes.
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| Operation in other MODES is covered by:
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| LCO 3.4.4, "PCS Loops-MODES 1 and 2";
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| LCO 3.4.5, "PCS Loops-MODE 3";
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| LCO 3.4.6, "PCS Loops-MODE 4";
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| LCO 3.4.7, "PCS Loops-MODE 5, Loops Filled";
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| LCO 3.9.4, "Shutdown Cooling (SDC) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Shutdown Cooling (SDC) and Coolant Circulation-Low Water Level" (MODE 6).
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| ACTIONS A.1 If one SDC train is inoperable, redundancy for heat removal is lost.
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| Action must be initiated immediately to restore a second train to OPERABLE status. The Completion Time reflects the importance of maintaining the availability of two paths for heat removal.
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| Palisades Nuclear Plant B 3.4.8-3 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.8 PCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES ACTIONS B.1 and B.2 (continued)
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| If no SDC trains are OPERABLE or SDC flow through the reactor core is not within limits, except as provided in Note 1, all operations involving the reduction of PCS boron concentration must be suspended. Action to restore one SDC train to OPERABLE status and operation shall be initiated immediately and continue until one train is restored to operation and flow through the reactor core is restored to within limits. Boron dilution requires forced circulation for proper mixing and the margin to criticality must not be reduced in this type of operation. The immediate Completion Time reflects the importance of maintaining operation for decay heat removal.
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| SURVEILLANCE SR 3.4.8.1 and SR 3.4.8.2 REQUIREMENTS These SRs require verification that one SDC train is in operation.
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| Verification of the required flow rate ensures forced circulation is providing heat removal and mixing of the soluble boric acid. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.4.8.1 and SR 3.4.8.2 are each modified by a Note to indicate the SR is only required to be met when complying with the applicable portion of the LCO. Therefore, it is only necessary to perform either SR 3.4.8.1, or SR 3.4.8.2 based on the method of compliance with the LCO.
| |
| Palisades Nuclear Plant B 3.4.8-4 Amendment No. 271 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.8 PCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES SURVEILLANCE SR 3.4.8.3 REQUIREMENTS (continued) This SR requires verification that two of the three charging pumps are incapable of reducing the boron concentration in the PCS below the minimum value necessary to maintain the required SHUTDOWN MARGIN. Making the charging pumps incapable reducing the boron concentration in the PCS may be accomplished by electrically disabling the pump motors, blocking potential dilution sources to the pump suction, or by isolating the pumps discharge flow path to the PCS. Verification may include visual inspection of the pumps configuration (e.g., pump breaker position or valve alignment), or the use of other administrative controls. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.4.8.3 is modified by a Note to indicate the SR is only required to be met when complying with LCO 3.4.8.b. When SDC flow through the reactor core is 2810 gpm, there is no restriction on charging pump operation.
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| SR 3.4.8.4 Verification that the required number of trains are OPERABLE ensures that redundant paths for heat removal are available and that additional trains can be placed in operation, if needed, to maintain decay heat removal and primary coolant circulation. Verification is performed by verifying proper breaker alignment and indicated power available to the required pump that is not in operation such that the SDC pump is capable of being started and providing forced PCS flow if needed. Proper breaker alignment and power availability means the breaker for the required SDC pump is racked-in and electrical power is available to energize the SDC pump motor. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES None Palisades Nuclear Plant B 3.4.8-5 Amendment No. 271 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.9 Pressurizer B 3.4.9 B 3.4 PRIMARY COOLANT SYSTEMS (PCS)
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| B 3.4.9 Pressurizer BASES BACKGROUND The pressurizer provides a point in the PCS where liquid and vapor are maintained in equilibrium under saturated conditions for pressure control purposes to prevent bulk boiling in the remainder of the PCS. Key functions include maintaining required primary system pressure during steady state operation and limiting the pressure changes caused by primary coolant thermal expansion and contraction during normal load transients.
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| The pressure control components addressed by this LCO include the pressurizer water level, required heaters capacity, and the emergency power supply to the heaters powered from electrical bus 1E. Pressurizer safety valves and pressurizer Power Operated Relief Valves (PORVs) are addressed by LCO 3.4.10, "Pressurizer Safety Valves," and LCO 3.4.11, "Pressurizer Power Operated Relief Valves (PORVs)," respectively.
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| The maximum water level limit has been established to ensure that a liquid to vapor interface exists to permit PCS pressure control, using the sprays and heaters during normal operation and proper pressure response for anticipated design basis transients. The water level limit serves two purposes:
| |
| : a. Pressure control during normal operation maintains subcooled reactor coolant in the loops and thus in the preferred state for heat transport; and
| |
| : b. By restricting the level to a maximum, expected transient primary coolant volume increases (pressurizer insurge) will not cause excessive level changes that could result in degraded ability for pressure control.
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| The maximum water level limit permits pressure control equipment to function as designed. The limit preserves the steam space during normal operation, thus, both sprays and heaters can operate to maintain the design operating pressure. The level limit also prevents filling the pressurizer (water solid) for anticipated design basis transients, thus ensuring that pressure relief devices (PORVs or pressurizer safety valves) can control pressure by steam relief rather than water relief. If the level limits were exceeded prior to a transient that creates a large pressurizer insurge volume leading to water relief, the maximum PCS pressure might exceed the Safety Limit of 2750 psia.
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| Palisades Nuclear Plant B 3.4.9-1 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.9 Pressurizer B 3.4.9 BASES BACKGROUND The requirement to have pressurizer heaters ensures that PCS (continued) pressure can be maintained. The pressurizer heaters maintain PCS pressure to keep the primary coolant subcooled. Inability to control PCS pressure during natural circulation flow could result in loss of single phase flow and decreased capability to remove core decay heat.
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| APPLICABLE In MODES 1, 2, and 3, the LCO requirement for a steam bubble is SAFETY ANALYSES reflected implicitly in the accident analyses. All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in the pressurizer. In making this assumption, the analyses neglect the small fraction of noncondensable gases normally present.
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| Safety analyses presented in the FSAR (Ref. 1) do not take credit for pressurizer heater operation; however, an implicit initial condition assumption of the safety analyses is that the PCS is operating at normal pressure.
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| Although the heaters are not specifically used in accident analysis, the need to maintain subcooling in the long term during loss of offsite power, as indicated in NUREG-0737, Clarification of TMI Action Plan Requirements, is the reason for their inclusion. The intent is to keep the primary coolant in a subcooled condition with natural circulation at hot, high pressure conditions for an undefined, but extended, time period after a loss of offsite power. While a loss of offsite power is a coincident occurrence assumed in the accident analyses, maintaining hot, high pressure conditions over an extended time period is not evaluated in the accident analyses.
| |
| The pressurizer satisfies Criterion 2 (for pressurizer water level) and Criterion 4 (for pressurizer heaters) of 10 CFR 50.36(c)(2).
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| Palisades Nuclear Plant B 3.4.9-2 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.9 Pressurizer B 3.4.9 BASES LCO The LCO requirement for the pressurizer to be OPERABLE with water level < 62.8% (hot full power pressurizer high level alarm setpoint) ensures that a steam bubble exists. Limiting the maximum operating water level preserves the steam space for pressure control. The LCO has been established to minimize the consequences of potential overpressure transients. Requiring the presence of a steam bubble is also consistent with analytical assumptions. During a plant heatup, the PCS is generally water solid in the lower temperature range of MODE 3.
| |
| Therefore, LCO 3.4.9.a has been modified by a Note which states that the pressurizer water level limit does not apply in MODE 3 until after a bubble has been established in the pressurizer and the pressurizer water level has been lowered to its normal operating band. The intent of this Note is to allow entry into the mode of Applicability during a plant heatup when the pressurizer water level is above the limit specified in the LCO. Once the normal pressurizer water level is established, compliance with the LCO must be met without reliance on the Note.
| |
| The LCO requires 375 kW of pressurizer heater capacity available from electrical bus 1D, and 375 kW of pressurizer heater capacity available from electrical bus 1E with the capability of being powered from an emergency power supply. In the event of a loss of offsite power, one half of the required heater capacity is normally connected to engineered safeguards bus 1D and can be manually controlled via a hand switch in the control room. This would provide sufficient heater capacity to establish and maintain natural circulation in a hot standby condition. To provide a redundant source of heater capacity should bus 1D become unavailable, methods and procedures have been established for manually connecting the required pressurizer heaters capacity, normally fed from electrical bus 1E, to engineered safeguards electrical bus 1C via a jumper cable. The amount of time required to make this connection (less than five hours) has been evaluated to assure that a 20°F subcooling margin, due to pressure decay, is not exceeded (Ref. 2).
| |
| The value of 375 kW is derived from the use of 30 heaters rated at approximately 12.5 kW each. The actual amount needed to maintain pressure is dependent on the ambient heat losses.
| |
| Palisades Nuclear Plant B 3.4.9-3 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.9 Pressurizer B 3.4.9 BASES APPLICABILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on PCS temperature resulting in the greatest effect on pressurizer level and PCS pressure control. Thus, the Applicability has been designated for MODES 1 and 2. The Applicability is also provided for MODE 3. The purpose is to prevent water solid PCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation. Although the requirements of LCO 3.4.12, Low Temperature Overpressure Protection (LTOP) System, ensures overpressure protection is provided in MODE 3 when the PCS cold leg temperature is < 430°F, the Applicability for the pressurizer is all inclusive of MODE 3 since the pressurizer heaters are required in all of MODE 3 to support plant operations. In MODES 4, 5, and 6, the pressurizer is no longer required and overpressure protection is provided by LTOP components specified in LCO 3.4.12.
| |
| In MODES 1, 2, and 3, there is the need to maintain the availability of pressurizer heaters capable of being powered from an emergency power supply. In the event of a loss of offsite power, the initial conditions of these MODES gives the greatest demand for maintaining the PCS in a hot pressurized condition with loop subcooling for an extended period.
| |
| For MODE 4, 5, or 6, it is not necessary to control pressure (by heaters) to ensure loop subcooling for heat transfer when the Shutdown Cooling System is in service and therefore the LCO is not applicable.
| |
| ACTIONS A.1 and A.2 With pressurizer water level not within the limit, action must be taken to restore the plant to operation within the bounds of the safety analyses.
| |
| To achieve this status, the plant must be brought to MODE 3, with the reactor tripped, within 6 hours and to MODE 4 within 30 hours. This takes the plant out of the applicable MODES and restores the plant to operation within the bounds of the safety analyses.
| |
| Six hours is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging plant systems. Further pressure and temperature reduction to MODE 4 brings the plant to a MODE where the LCO is not applicable. The 30 hour time to reach the nonapplicable MODE is reasonable based on operating experience for that evolution.
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| Palisades Nuclear Plant B 3.4.9-4 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.9 Pressurizer B 3.4.9 BASES ACTIONS B.1 (continued)
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| If < 375 kW of pressurizer heater capacity is available from either electrical bus 1D or electrical bus 1E, or the pressurizer heaters from electrical bus 1E are not capable of being powered from an emergency power supply, restoration is required within 72 hours. The Completion Time of 72 hours is reasonable considering that a demand caused by loss of offsite power would be unlikely in this period. Pressure control may be maintained during this time using the remaining available pressurizer heaters.
| |
| C.1 If <375 kW of pressurizer heater capacity is available from both electrical bus 1D and electrical bus 1E, or <375 kW of pressurizer heater capacity is available from electrical bus 1D and the pressurizer heaters from electrical bus 1E are not capable of being powered from an emergency power supply, restoration of either electrical bus pressurizer heaters to an OPERABLE status is required within 24 hours. This Condition is modified by a Note stating it is not applicable if the remaining electrical bus 1D or electrical bus 1E required pressurizer heaters are intentionally declared inoperable. The Condition does not apply to voluntary removal of redundant systems or components from service. The Condition is only applicable if either electrical bus 1D required pressurizer heaters or electrical bus 1E required pressurizer heaters are discovered to be inoperable, or if both electrical buses required pressurizer heaters are discovered to be inoperable at the same time. If both electrical buses required pressurizer heaters are inoperable, pressurizer heater capacity may not be available to maintain subcooling in the PCS loops during natural circulation cooldown following a loss of offsite power. The inoperability of both electrical buses required pressurizer heaters during the 24 hour Completion Time has been shown to be acceptable based on the infrequent use of the Required Action and the small incremental effect on plant risk (Ref. 3).
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| Palisades Nuclear Plant B 3.4.9-5 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.9 Pressurizer B 3.4.9 BASES ACTIONS D.1 and D.2 (continued)
| |
| If one or more of the electrical buses required pressurizer heaters cannot be restored to an OPERABLE status within the associated allowed Completion Times, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours and to MODE 4 within 30 hours. The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging safety systems. Similarly, the Completion Time of 30 hours is reasonable, based on operating experience, to reach MODE 4 from full power in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE SR 3.4.9.1 REQUIREMENTS This SR ensures that during steady state operation, pressurizer water level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. The Surveillance is performed by observing the indicated level. SR 3.4.9.1 is modified by a Note which states that verification of the pressurizer water level is not required to be met until 1 hour after a bubble has been established in the pressurizer and the pressurizer water level has been lowered to its normal operating band.
| |
| The intent of this Note is to prevent an SR 3.0.4 conflict by delaying the performance of this SR until after the water level in the pressurizer is within its normal operating band following a plant heatup. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
| |
| SR 3.4.9.2 The SR is satisfied when the power supplies are demonstrated to be capable of producing the minimum power and the capacity of the associated pressurizer heaters are verified to be 375 kW. (This may be done by testing the power supply output and by performing an electrical check on heater element continuity and resistance.) The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
| |
| Palisades Nuclear Plant B 3.4.9-6 Amendment No. 271 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.9 Pressurizer B 3.4.9 BASES SURVEILLANCE SR 3.4.9.3 REQUIREMENTS (continued) This SR only applies to the pressurizer heaters normally powered from electrical bus 1E since the pressurizer heaters powered from bus 1D are permanently connected to the engineered safeguards electrical system.
| |
| This SR confirms that the pressurizer heaters normally fed from electrical bus 1E are capable of being powered from electrical bus 1C by use of a jumper cable. It is not the intent of this SR to physically install the jumper cable, but to verify the necessary components are available for installation and to ensure the procedures and methods used to install the jumper cable are current. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
| |
| REFERENCES 1. FSAR, Chapter 14
| |
| : 2. FSAR, Section 4.3.7
| |
| : 3. WCAP-16125-NP-A, Justification for Risk-Informed Modifications to Selected Technical Specifications for Conditions Leading to Exigent Plant Shutdown, Revision 2, August 2010.
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| Palisades Nuclear Plant B 3.4.9-7 Amendment No. 271 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.10 Pressurizer Safety Valves B 3.4.10 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
| |
| B 3.4.10 Pressurizer Safety Valves BASES BACKGROUND The purpose of the three spring loaded pressurizer safety valves is to provide PCS overpressure protection. Operating in conjunction with the Reactor Protection System, three valves are used to ensure that the Safety Limit (SL) of 2750 psia is not exceeded for analyzed transients during operation in MODES 1 and 2 and portions of MODE 3. For the remainder of MODE 3, MODE 4, MODE 5, and MODE 6 with the reactor vessel head on, overpressure protection is provided by operating procedures and the LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System."
| |
| The self actuated pressurizer safety valves are designed in accordance with the requirements set forth in the American Society of Mechanical Engineering (ASME), Boiler and Pressure Vessel Code, Section III (Ref. 1). The required lift settings are given in Table 3.4.10-1 in the accompanying technical specification. The safety valves discharge steam from the pressurizer to a quench tank located in the containment.
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| The discharge flow is indicated by an increase in temperature downstream of the safety valves, acoustic monitors, and by an increase in the quench tank temperature and level.
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| The lift settings listed in Table 3.4.10-1 correspond to ambient conditions of the valves at nominal operating temperature and pressure.
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| This requires either that the valves be set hot or that a correlation between hot and cold settings be established.
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| The pressurizer safety valves are part of the primary success path and mitigate the effects of postulated accidents. OPERABILITY of the safety valves ensures that the PCS pressure will be limited to 110% of design pressure. The consequences of exceeding the ASME pressure limit (Ref. 1) could include damage to PCS components, increased leakage, or a requirement to perform additional stress analyses prior to resumption of reactor operation.
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| Palisades Nuclear Plant B 3.4.10-1 Revised 07/26/2017
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| INSERT Bases 3.4.10 Pressurizer Safety Valves B 3.4.10 BASES APPLICABLE All accident analyses in the FSAR that require safety valve actuation SAFETY ANALYSES assume operation of one or more pressurizer safety valves to limit increasing primary coolant pressure. The overpressure protection analysis assumes that the valves open at the high range of the lift setting including the allowable tolerance. The Loss of External Electrical Load incident and Loss of Normal Feedwater Flow incident are the two safety analyses events which rely on the pressurizer safety valves to mitigate an overpressurization of the PCS. The pressurizer safety valves must also accommodate pressurizer in surges that could occur from a Loss of Forced Primary Coolant Flow incident, and a Primary Pump Rotor Seizure incident. Single failure of a safety valve is neither assumed in the accident analysis nor required to be addressed by the ASME Code. Compliance with this specification is required to ensure that the accident analysis and design basis calculations remain valid.
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| The pressurizer safety valves satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| LCO The three pressurizer safety valves are set to open near the PCS design pressure (2500 psia) and within the ASME specified tolerance to avoid exceeding the maximum PCS design pressure SL, to maintain accident analysis assumptions, and to comply with ASME Code requirements. The nominal lift settings values listed in Table 3.4.10-1, plus an allowable tolerance of +/- 3%, establish the acceptable as-found pressure band for determining valve OPERABILITY. Following valve testing, an as-left tolerance of +/- 1% of the lift settings is imposed by SR 3.4.10.1 to account for setpoint drift during the surveillance interval.
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| The limit protected by this specification is the Primary Coolant Pressure Boundary (PCPB) SL of 110% of design pressure. The inoperability of any valve could result in exceeding the SL if a transient were to occur.
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| The consequences of exceeding the ASME pressure limit could include damage to one or more PCS components, increased leakage, or additional stress analysis being required prior to resumption of reactor operation.
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| Palisades Nuclear Plant B 3.4.10-2 Revised 07/26/2017
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| INSERT Bases 3.4.10 Pressurizer Safety Valves B 3.4.10 BASES APPLICABILITY In MODES 1 and 2, and portions of MODE 3 above the LTOP temperature, OPERABILITY of three valves is required because the combined capacity is required to keep primary coolant pressure below 110% of its design value during certain accidents. Portions of MODE 3 are conservatively included, although the listed accidents may not require three safety valves for protection.
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| The LCO is not applicable in MODE 3 when any PCS cold leg temperatures are < 430°F and MODES 4 and 5 because LTOP protection is provided. Overpressure protection is not required in MODE 6 with the reactor vessel head removed.
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| ACTIONS A.1 With one pressurizer safety valve inoperable, restoration must take place within 15 minutes. The Completion Time of 15 minutes reflects the importance of maintaining the PCS overpressure protection system.
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| An inoperable safety valve coincident with an PCS overpressure event could challenge the integrity of the PCPB.
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| B.1 and B.2 If the Required Action cannot be met within the required Completion Time or if two or more pressurizer safety valves are inoperable, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and at least one PCS cold leg temperature reduced to below 430°F within 12 hours. The 6 hours allowed is reasonable, based on operating experience, to reach MODE 3 from full power without challenging plant systems. Similarly, the 12 hours allowed is reasonable, based on operating experience, to reduce any PCS cold leg temperature < 430°F without challenging plant systems.
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| Below 430°F, overpressure protection is provided by LTOP. The change from MODE 1, 2, or 3 to MODE 3 with any PCS cold leg temperature < 430°F reduces the PCS energy (core power and pressure), lowers the potential for large pressurizer insurges, and thereby removes the need for overpressure protection by three pressurizer safety valves.
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| Palisades Nuclear Plant B 3.4.10-3 Revised 07/26/2017
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| INSERT Bases 3.4.10 Pressurizer Safety Valves B 3.4.10 BASES SURVEILLANCE SR 3.4.10.1 REQUIREMENTS SRs are specified in the INSERVICE TESTING PROGRAM.
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| Pressurizer safety valves are to be tested in accordance with the requirements of the ASME Code (Ref. 1), which provides the activities and the Frequency necessary to satisfy the SRs. No additional requirements are specified.
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| The pressurizer safety valve setpoint tolerance is +/- 3% for OPERABILITY; however, the valves are reset to within a tolerance of
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| +/- 1% during the Surveillance to allow for drift.
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| REFERENCES 1. ASME Code for Operation and Maintenance of Nuclear Power Plants.
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| Palisades Nuclear Plant B 3.4.10-4 Amendment No. 262 Revised 07/26/2017
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| INSERT Bases 3.4.11 Pressurizer PORVs B 3.4.11 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)
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| BASES BACKGROUND The pressurizer is equipped with two types of devices for pressure relief:
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| pressurizer safety valves and PORVs. The safety valves are addressed by LCO 3.4.10. The PORVs are solenoid-pilot operated relief valves which, when placed in the Auto position, automatically open at a specific set pressure when the pressurizer pressure increases and is automatically closed on decreasing pressure. The PORV may also be manually operated using controls installed in the control room.
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| A motor operated, normally closed, block valve is installed between the pressurizer and each PORV. The function of the block valve is to isolate the PORV. Block valve closure is accomplished manually using controls in the control room and may be used to isolate a leaking PORV to permit continued power operation. Most importantly, the block valve is used to isolate a stuck open PORV to isolate the resulting Loss Of Coolant Accident (LOCA). Closure terminates the PCS depressurization and coolant inventory loss.
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| The PORV, its block valve, and their respective controls are powered from safety class power supplies. Power supplies for the PORV are separate from those for the block valve. Power supply requirements are defined in NUREG-0737, Item II.G.1.
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| The primary purpose of this LCO is to ensure that the PORV and the block valve are operating correctly so the potential for a LOCA through the PORV pathway is minimized, or if a LOCA were to occur through a failed open PORV, the block valve could be manually operated to isolate the path.
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| In the event of an abnormal transient, the PORVs may be manually operated to depressurize the PCS as directed by the Emergency Operating Procedures. The PORVs may be used for depressurization when the pressurizer spray is not available, a condition that may be encountered during a loss of offsite power. Operators can manually open the PORVs to reduce PCS pressure in the event of a Steam Generator Tube Rupture (SGTR) with offsite power unavailable.
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| The PORVs may also be used for once-through core cooling in the case of multiple equipment failure events that are not within the design basis, such as a total loss of feedwater.
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| Palisades Nuclear Plant B 3.4.11-1 Revised 01/29/2020
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| INSERT Bases 3.4.11 Pressurizer PORVs B 3.4.11 BASES BACKGROUND If preferred during normal plant operation when PCS temperature is at (continued) or above 430°F and the PORV block valves are open, the PORVs may also function as an automatic overpressure device and limits challenges to the safety valves. Although the PORVs act as an overpressure device for operational purposes, safety analyses do not take credit for PORV actuation, but do take credit for the safety valves. Since the pressurizer safety valves provide the necessary automatic protection against excessive PCS pressure, automatic actuation of the PORVs is not required to be OPERABLE and the PORVs and their block valves are normally maintained in the closed position.
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| The PORVs also provide Low Temperature Overpressure Protection (LTOP) during heatup and cooldown. LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," addresses this function.
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| APPLICABLE The PORV small break LOCA break size is bounded by the spectrum SAFETY ANALYSES of piping breaks analyzed for plant licensing. The possibility of a small break LOCA through the PORV is reduced when the PORV flow path is OPERABLE and the PORV opening setpoint is established to be reasonably remote from expected transient challenges. The possibility is further minimized if the flow path is isolated.
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| Overpressure protection is provided by safety valves, and analyses do not take credit for the PORV opening for accident mitigation. However, technical findings and regulatory analysis discussed in NUREG-1316, Technical Findings and Regulatory Analysis Related to Generic Issue 70
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| - Evaluation of Power-Operated Relief Valve and Block Valve Reliability in PWR Nuclear Power Plants, have determined that maintaining the requirements for PORVs and block valves in the technical specifications can increase the reliability of these components and provide assurance they will function as required and that operating experience has shown these components to be important to public health and safety.
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| Pressurizer PORVs satisfy Criterion 4 of 10 CFR 50.36(c)(2).
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| Palisades Nuclear Plant B 3.4.11-2 Revised 01/29/2020
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| INSERT Bases 3.4.11 Pressurizer PORVs B 3.4.11 BASES LCO The LCO requires each PORV and its associated block valve to be OPERABLE. The block valve is required to be OPERABLE so it may be used to isolate the flow path of an inoperable PORV or, unisolate the flow path of an OPERABLE PORV. Thus, a block valve is considered OPERABLE if it is capable of being cycled in the open and close direction.
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| The PORV is required to be OPERABLE to provide PCS pressure control and maintain PCS integrity. For a PORV, OPERABILITY means the valve is capable of being cycled in the open and close direction.
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| APPLICABILITY With a PORV in the CLOSED position in MODES 1 and 2, and MODE 3 with all PCS cold leg temperatures 430°F, the PORV and its block valve are required to be OPERABLE to limit PCS leakage through the PORV flow path, and to be available for manual operation to mitigate abnormal transients which may be initiated from these MODES and condition.
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| With a PORV in the AUTO position in MODES 1 and 2, and MODE 3 with all PCS cold leg temperatures 430°F, the PORV and its block valve are required to be OPERABLE to limit the potential for a small break LOCA through the flow path. A likely cause for PORV small break LOCA is a result of pressure increase transients that cause the PORV to open.
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| Imbalances in the energy output of the core and heat removal by the secondary system can cause the PCS pressure to increase to the PORV opening setpoint. Pressure increase transients can occur any time the steam generators are used for heat removal. The most rapid increases will occur at higher operating power and pressure conditions of MODES 1 and 2. Pressure increases are less prominent in MODE 3 with PCS cold leg temperatures < 430°F because the core input energy is reduced, but the PCS pressure is high. Therefore, this LCO is applicable in MODES 1 and 2, and MODE 3 with all PCS cold leg temperatures 430°F.
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| The LCO is not applicable in MODE 3 with any PCS cold leg temperatures < 430°F when both pressure and core energy are decreased and the pressure surges become much less significant. The PORV setpoint is reduced for LTOP in MODE 3 when any PCS cold leg temperatures are < 430°F, and in MODES 4, 5, and MODE 6 with the reactor vessel head in place. LCO 3.4.12 addresses the PORV requirements in these MODES.
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| Palisades Nuclear Plant B 3.4.11-3 Revised 01/29/2020
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| INSERT Bases 3.4.11 Pressurizer PORVs B 3.4.11 BASES ACTIONS The ACTIONS are modified by a Note. The Note clarifies that each pressurizer PORV is treated as a separate entity, each with separate Completion Times (i.e., the Completion Time is on a component basis).
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| A.1 and A.2 If one PORV is inoperable it must either be isolated, by closing the associated block valve, or restored to OPERABLE status. The Completion Time of 1 hour is reasonable based on the small potential that the PORVs will be required to function during this time period and provides the operator adequate time to correct the situation. If the inoperable valve cannot be restored to OPERABLE status, it must be isolated within the specified time. Because there is at least one PORV that remains OPERABLE, an additional 72 hours is provided to restore the inoperable PORV to OPERABLE status. Mechanical system LCOs typically provide a 72 hour Completion Time under conditions when a required system can perform its required safety function, but may not be able to do so assuming an additional failure. When operating in accordance with the Required Actions of an LCO Condition, it is not necessary to be able to cope with an additional single failure B.1 and B.2 If one block valve is inoperable, then it must be restored to OPERABLE status, or the associated PORV placed in manual control. Placing a PORV in manual control is accomplished by placing the PORV hand switch in the CLOSE position. The prime importance for the capability to close the block valve is to isolate a stuck open PORV. Therefore, if the block valve cannot be restored to OPERABLE status within 1 hour, the Required Action is to place the PORV in manual control to preclude its automatic opening for an overpressure event and to avoid the potential for a stuck open PORV at a time that the block valve is inoperable. The Completion Time of 1 hour is reasonable based on the small potential that the PORVs will be required to function during this time period and provides the operator time to correct the situation. Because at least one PORV remains OPERABLE, the operator is permitted a Completion Time of 72 hours to restore the inoperable block valve to OPERABLE status.
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| Palisades Nuclear Plant B 3.4.11-4 Revised 01/29/2020
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| INSERT Bases 3.4.11 Pressurizer PORVs B 3.4.11 BASES ACTIONS B.1 and B.2 (continued)
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| Mechanical system LCOs typically provide a 72 hour Completion Time under conditions when a required system can perform its required safety function, but may not be able to do so assuming an additional failure.
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| When operating in accordance with the Required Actions of an LCO Condition, it is not necessary to be able to cope with an additional single failure.
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| The time allowed to restore the block valve is based upon the Completion Time for restoring an inoperable PORV in Condition A since the PORVs are not capable of automatically mitigating an overpressure event when placed in manual control. If the block valve is restored within the Completion Time of 72 hours, the PORV is restored to OPERABLE status.
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| C.1 and C.2 If more than one PORV is inoperable, it is necessary to either restore at least one valve within the Completion Time of 1 hour or isolate the flow path by closing the associated block valves and restoring at least one PORV to OPERABLE status within 2 hours. The Completion Time of 1 hour is reasonable based on the small potential that the PORVs will be required to function during this time period, and provides the operator time to correct the situation. If one PORV is restored and one PORV remains inoperable, then the plant will be in Condition A with the time clock started at the original declaration of having two PORVs inoperable.
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| D.1 and D.2 If two block valves are inoperable, it is necessary to either restore the block valves within the Completion Time of 1 hour or place the associated PORVs in manual control and restore at least one block valve to OPERABLE status within 2 hours and the remaining block valve in 72 hours. The Completion Time of 1 hour to either restore the block valves or place the associated PORVs in manual control is reasonable based on the small potential that the PORVs will be required to function during this time period, and provides the operator time to correct the situation.
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| Palisades Nuclear Plant B 3.4.11-5 Revised 01/29/2020
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| INSERT Bases 3.4.11 Pressurizer PORVs B 3.4.11 BASES ACTIONS E.1 (continued)
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| If the Required Actions and associated Completion Times are not met, then the plant must be brought to a stable condition which minimizes the potential for transients affecting the PCS. The plant must be brought to at least MODE 3 within 6 hours. With one or two PORVs or block valves inoperable, exiting the MODE of Applicability (i.e., MODE 3 with any PCS cold leg temperature < 430°F) may not be desirable since below 430°F the PORVs and their associated block valves are required to support LTOP operations (LCO 3.4.12). Although LCO 3.0.4 would allow entry into LCO 3.4.12, reducing PCS temperature below 430°F may not be prudent since below 430°F the PORVs are credited in the safety analysis to protect the PCS from an inadvertent overpressure event. At or above 430°F, the PORVs are not credited in the safety analysis and thus have no safety function. If practical, the inoperable PORVs or block valves should be restored to an OPERABLE status while the PCS is above 430°F to avoid entering a plant condition where the PORVs are required for LTOP. If necessary, LCO 3.0.4 would allow the plant to be placed in MODE 5 to facilitate repairs. In this plant condition, overpressure protection may be provided by establishing the required vent path specified in LCO 3.4.12.
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| The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging safety systems. In MODE 3 with any PCS cold leg temperature < 430°F, and MODES 4 and 5 and MODE 6 with the reactor vessel head on, maintaining PORV OPERABILITY is required by LCO 3.4.12.
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| Palisades Nuclear Plant B 3.4.11-6 Revised 01/29/2020
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| INSERT Bases 3.4.11 Pressurizer PORVs B 3.4.11 BASES SURVEILLANCE SR 3.4.11.1 REQUIREMENTS Block valve cycling verifies that it can be opened and closed if necessary.
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| The basis for the Frequency of prior to entering MODE 4 from MODE 5 if not performed in the previous 92 days reflects the importance of not routinely cycling the block valves during the period when the PCS is pressurized since this practice may result in the associated PORV being opened by the increase inlet pressure to the PORV. The 92 days portion of the Frequency is consistent with the testing frequency stipulated by ASME Section XI as modified by the Cold Shutdown Testing Basis used in support of the second 120 month interval of the Inservice Valve Testing Program which only requires the block valves to be cycled during Cold Shutdown conditions. If the block valve is closed to isolate a PORV that is capable of being manually cycled, the OPERABILITY of the block valve is of importance because opening the block valve is necessary to permit the PORV to be used for manual control of primary coolant pressure. If a block valve is open and its associated PORV was stuck open, the OPERABILITY of the block valve is of importance because closing the block valve is necessary to isolate the stuck opened PORV.
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| SR 3.4.11.2 SR 3.4.11.2 requires complete cycling of each PORV. PORV cycling demonstrates its function and is performed when the PCS temperature is
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| > 200°F. Stroke testing of the PORVs above 200°F is desirable since it closer simulates the temperature and pressure environmental effects on the valves and thus represents a better test condition for assessing PORV performance under normal plant conditions. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES None Palisades Nuclear Plant B 3.4.11-7 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.12 LTOP System B 3.4.12 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.12 Low Temperature Overpressure Protection (LTOP) System BASES BACKGROUND The LTOP System controls PCS pressure at low temperatures so the integrity of the Primary Coolant Pressure Boundary (PCPB) is not compromised by violating the Pressure and Temperature (P/T) limits of 10 CFR 50, Appendix G (Ref. 1). The reactor vessel is the limiting PCPB component requiring such protection. LCO 3.4.3, "PCS Pressure and Temperature (P/T) Limits," provides the allowable combinations for operational pressure and temperature during cooldown, shutdown, and heatup to keep from violating the Reference 1 requirements during the LTOP MODES.
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| The toughness of the reactor vessel material decreases at low temperatures. As the vessel neutron exposure accumulates, the material toughness decreases and becomes less resistant to pressure stress at low temperatures (Ref. 2). PCS pressure, therefore, is maintained low at low temperatures and is increased only as temperature is increased.
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| The potential for vessel overpressurization is most acute when the PCS is water solid, which occurs only while shutdown. Under that condition, a pressure fluctuation can occur more quickly than an operator can react to relieve the condition. Exceeding the PCS P/T limits by a significant amount could cause brittle fracture of the reactor vessel. LCO 3.4.3 requires administrative control of PCS pressure and temperature during heatup and cooldown to prevent exceeding the P/T limits.
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| This LCO provides PCS overpressure protection by limiting coolant injection capability and requiring adequate pressure relief capacity.
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| Limiting coolant injection capability requires all High Pressure Safety Injection (HPSI) pumps be incapable of injection into the PCS when any PCS cold leg temperature is < 300°F. The pressure relief capacity requires either two OPERABLE redundant Power Operated Relief Valves (PORVs) or the PCS depressurized and a PCS vent of sufficient size.
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| One PORV or the PCS vent is the overpressure protection device that acts to terminate an increasing pressure event.
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| Palisades Nuclear Plant B 3.4.12-1 Revised 01/29/2020
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| INSERT Bases 3.4.12 LTOP System B 3.4.12 BASES BACKGROUND With limited coolant injection capability, the ability to provide core (continued) coolant addition is restricted. The LCO does not require the chemical and volume control system to be deactivated or the Safety Injection Signals (SIS) blocked. Due to the lower pressures in the LTOP MODES and the expected core decay heat levels, the chemical and volume control system can provide adequate flow via the makeup control valve. If conditions require the use of an HPSI pump for makeup in the event of loss of inventory, then a pump can be made available through manual actions.
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| The LTOP System for pressure relief consists of two PORVs with temperature dependent lift settings or a PCS vent of sufficient size.
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| Two PORVs are required for redundancy. One PORV has adequate relieving capability to prevent overpressurization for the allowed coolant injection capability.
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| PORV Requirements As designed for the LTOP System, an open signal is generated for each PORV if the PCS pressure approaches a limit determined by the LTOP actuation logic. The actuation logic monitors PCS pressure and cold leg temperature to determine when the LTOP overpressure setting is approached. If the indicated pressure meets or exceeds the calculated value, a PORV is opened.
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| The LCO presents the PORV setpoints for LTOP by specifying Figure 3.4.12-1, LTOP Setpoint Limit. Having the setpoints of both valves within the limits of the LCO ensures the P/T limits will not be exceeded in any analyzed event.
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| When a PORV is opened in an increasing pressure transient, the release of coolant causes the pressure increase to slow and reverse. As the PORV releases coolant, the system pressure decreases until a reset pressure is reached and the valve closed. The pressure continues to decrease below the reset pressure as the valve closes.
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| Palisades Nuclear Plant B 3.4.12-2 Revised 01/29/2020
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| INSERT Bases 3.4.12 LTOP System B 3.4.12 BASES BACKGROUND PCS Vent Requirements (continued)
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| Once the PCS is depressurized, a vent exposed to the containment atmosphere will maintain the PCS at containment ambient pressure in an PCS overpressure transient if the relieving requirements of the transient do not exceed the capabilities of the vent. Thus, the vent path must be capable of relieving the flow resulting from the limiting LTOP mass injection or heatup transient and maintaining pressure below the P/T limits. The required vent capacity may be provided by one or more vent paths.
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| Reference 3 has determined that any vent path capable of relieving 167 gpm at a PCS pressure of 315 psia is acceptable. The 167 gpm flow rate is based on an assumed charging imbalance due to interruption of letdown flow with three charging pumps operating, a 40°F per hour PCS heatup rate, a 60°F per hour pressurizer heatup rate, and an initially depressurized and vented PCS. Neither HPSI pump nor Primary Coolant Pump (PCP) starts need to be assumed with the PCS initially depressurized, because LCO 3.4.12 requires both HPSI pumps to be incapable of injection into the PCS and LCO 3.4.7, PCS Loops-MODE 5, Loops Filled, places restrictions on starting a PCP.
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| The pressure relieving ability of a vent path depends not only upon the area of the vent opening, but also upon the configuration of the piping connecting the vent opening to the PCS. A long, or restrictive piping connection may prevent a larger vent opening from providing adequate flow, while a smaller opening immediately adjacent to the PCS could be adequate. The areas of multiple vent paths cannot simply be added to determine the necessary vent area.
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| The following vent path examples are acceptable:
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| : 1. Removal of a steam generator primary manway;
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| : 2. Removal of the pressurizer manway;
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| : 3. Removal of a PORV or pressurizer safety valve;
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| : 4. Both PORVs and associated block valves open; and
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| : 5. Opening of both PCS vent valves MV-PC514 and MV-PC515.
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| Palisades Nuclear Plant B 3.4.12-3 Revised 01/29/2020
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| INSERT Bases 3.4.12 LTOP System B 3.4.12 BASES BACKGROUND Reference 4 determined that venting the PCS through MV-PC514 and (continued) MV-PC515 provided adequate flow area. The other listed examples provide greater flow areas with less piping restriction and are therefore acceptable. Other vent paths shown to provide adequate capacity could also be used. The vent path(s) must be above the level of reactor coolant, to prevent draining the PCS.
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| One open PORV provides sufficient flow area to prevent excessive PCS pressure. However, if the PORVs are elected as the vent path, both valves must be used to meet the single failure criterion, since the PORVs are held open against spring pressure by energizing the operating solenoid.
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| When the shutdown cooling system is in service with MO-3015 and MO-3016 open, additional overpressure protection is provided by the relief valves on the shutdown cooling system. References 5 and 6 show that this relief capacity will prevent the PCS pressure from exceeding its pressure limits during any of the above mentioned events.
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| APPLICABLE Safety analyses (Ref. 7) demonstrate that the reactor vessel is SAFETY ANALYSES adequately protected against exceeding the Reference 1 P/T limits during shutdown. In MODES 1 and 2, and in MODE 3 with all PCS cold leg temperature at or exceeding 430°F, the pressurizer safety valves prevent PCS pressure from exceeding the Reference 1 limits. Below 430°F, overpressure prevention falls to the OPERABLE PORVs or to a depressurized PCS and a sufficiently sized PCS vent. Each of these means has a limited overpressure relief capability.
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| The actual temperature at which the pressure in the P/T limit curve falls below the pressurizer safety valve setpoint increases as the reactor vessel material toughness decreases due to neutron embrittlement. Each time the P/T limit curves are revised, the LTOP System should be re-evaluated to ensure its functional requirements can still be satisfied using the PORV method or the depressurized and vented PCS condition.
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| Reference 3 contains the acceptance limits that satisfy the LTOP requirements. When originally generated, the validity period for the LTOP Setpoint Limit curve in Figure 3.4.12-1, which is based on the Reference 3 analysis, ended prior to the operating license expiration date. A subsequent analysis was performed (Ref. 9) which demonstrated that the current LTOP Setpoint Limit curve is valid through the operating license expiration date, equivalent to 42.1 effective full power years of operation.
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| Any change to the PCS must be evaluated against these analyses to determine the impact of the change on the LTOP acceptance limits.
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| Palisades Nuclear Plant B 3.4.12-4 Revised 01/29/2020
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| INSERT Bases 3.4.12 LTOP System B 3.4.12 BASES APPLICABLE Transients that are capable of overpressurizing the PCS are SAFETY ANALYSES categorized as either mass injection or heatup transients (continued)
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| Mass Injection Type Transients
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| : a. Inadvertent safety injection; or
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| : b. Charging/letdown flow mismatch.
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| Heatup Type Transients
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| : a. Inadvertent actuation of pressurizer heaters;
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| : b. Loss of Shutdown Cooling (SDC); or
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| : c. PCP startup with temperature asymmetry within the PCS or between the PCS and steam generators.
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| Rendering both HPSI pumps incapable of injection is required during the LTOP MODES to ensure that mass injection transients beyond the capability of the LTOP overpressure protection system, do not occur. The Reference 3 analyses demonstrate that either one PORV or the PCS vent can maintain PCS pressure below limits when three charging pump are actuated. Thus, the LCO prohibits the operation of both HPSI pumps and does not place any restrictions on charging pump operation.
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| Fracture mechanics analyses were used to establish the applicable temperature range for the LTOP LCO as below 430°F. At and above this temperature, the pressurizer safety valves provide the reactor vessel pressure protection. The pressure-temperature limit curves and LTOP curve are based on reactor vessel material properties which change over time due to radiation embrittlement. These curves are valid for the period of time corresponding to the reactor vessel material condition which was assumed when the curves were generated. At the time the curves were developed, they were based on being valid up to a neutron irradiation accumulation equal to 2.192 x 1019 n (neutrons)/cm2 (Ref. 3). The vessel materials in the current curve analysis (Ref. 9) were assumed to have a neutron irradiation accumulation equal to 42.1 effective full power years of operation. The current analysis determined an LTOP enable temperature that is bounded by the LTOP LCO.
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| Palisades Nuclear Plant B 3.4.12-5 Revised 01/29/2020
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| INSERT Bases 3.4.12 LTOP System B 3.4.12 BASES APPLICABLE PORV Performance SAFETY ANALYSES (continued) The fracture mechanics analyses show that the vessel is protected when the PORVs are set to open at or below the setpoint curve specified in Figure 3.4.12-1 of the accompanying LCO. The setpoint is derived by modeling the performance of the LTOP System, assuming the limiting allowed LTOP transient. The valve qualification process considered pressure overshoot and undershoot beyond the PORV opening and closing setpoints, resulting from signal processing and valve stroke times.
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| The PORV setpoints at or below the derived limit ensure the Reference 1 limits will be met.
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| The PORV setpoints will be re-evaluated for compliance when the P/T limits are revised. The P/T limits are periodically modified as the reactor vessel material toughness decreases due to embrittlement caused by neutron irradiation. Revised P/T limits are determined using neutron fluence projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3 discuss these examinations.
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| The PORVs are considered active components. Thus, the failure of one PORV represents the worst case, single active failure.
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| PCS Vent Performance With the PCS depressurized, analyses show the required vent size is capable of mitigating the limiting allowed LTOP overpressure transient. In that event, this size vent maintains PCS pressure less than the maximum PCS pressure on the P/T limit curve.
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| The PCS vent is passive and is not subject to active failure.
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| LTOP System satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| Palisades Nuclear Plant B 3.4.12-6 Revised 01/29/2020
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| INSERT Bases 3.4.12 LTOP System B 3.4.12 BASES LCO This LCO is required to ensure that the LTOP System is OPERABLE.
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| The LTOP System is OPERABLE when both HPSI pumps are incapable of injecting into the PCS and pressure relief capabilities are OPERABLE.
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| Violation of this LCO could lead to the loss of low temperature overpressure mitigation and violation of the Reference 1 limits as a result of an operational transient.
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| To limit the coolant injection capability, LCO 3.4.12.a requires both HPSI pumps be incapable of injecting into the PCS. LCO 3.4.12.a is modified by two Notes. Note 1 only requires both HPSI pumps to be incapable of injecting into the PCS when any PCS cold leg temperature is < 300°F.
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| When all PCS cold leg temperatures are 300°F, a start of both HPSI pumps in conjunction with a charging/letdown imbalance will not cause the PCS pressure to exceed the 10 CFR 50 Appendix G limits. Thus, a restriction on HPSI pump operation when all PCS cold leg temperatures are 300°F is not required. Note 2 is provided to assure that this LCO does not cause hesitation in the use of a HPSI pump for PCS makeup if it is needed due to a loss of shutdown cooling or a loss of PCS inventory.
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| The elements of the LCO that provide overpressure mitigation through pressure relief are:
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| : a. Two OPERABLE PORVs; or
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| : b. The PCS depressurized and vented.
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| A PORV is OPERABLE for LTOP when its block valve is open, its lift setpoint is set consistent with Figure 3.4.12-1 in the accompanying LCO and testing has proven its ability to open at that setpoint, and motive power is available to the valve and its control circuit.
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| A PCS vent is OPERABLE when open with an area capable of relieving 167 gpm at a PCS pressure of 315 psia.
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| Each of these methods of overpressure prevention is capable of mitigating the limiting LTOP transient.
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| Palisades Nuclear Plant B 3.4.12-7 Revised 01/29/2020
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| INSERT Bases 3.4.12 LTOP System B 3.4.12 BASES APPLICABILITY This LCO is applicable in MODE 3 when the temperature of any PCS cold leg is < 430°F, in MODES 4 and 5, and in MODE 6 when the reactor vessel head is on. The pressurizer safety valves provide overpressure protection that meets the Reference 1 P/T limits at and above 430°F.
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| When the reactor vessel head is off, overpressurization cannot occur.
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| LCO 3.4.3 provides the operational P/T limits for all MODES.
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| LCO 3.4.10, "Pressurizer Safety Valves," requires the OPERABILITY of the pressurizer safety valves that provide overpressure protection during MODES 1 and 2, and MODE 3 with all PCS cold leg temperatures 430°F.
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| Low temperature overpressure prevention is most critical during shutdown when the PCS is water solid, and a mass addition or a heatup transient can cause a very rapid increase in PCS pressure with little or no time available for operator action to mitigate the event.
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| ACTIONS A Note prohibits the application of LCO 3.0.4.b to inoperable PORVs used for LTOP. There is an increased risk associated with entering MODE 4 from MODE 5 with PORVs used for LTOP inoperable and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
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| A.1 With one or two HPSI pumps capable of injecting into the PCS, overpressurization is possible.
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| The immediate Completion Time to initiate actions to restore restricted coolant injection capability to the PCS reflects the importance of maintaining overpressure protection of the PCS.
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| Palisades Nuclear Plant B 3.4.12-8 Revised 01/29/2020
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| INSERT Bases 3.4.12 LTOP System B 3.4.12 BASES ACTIONS B.1 (continued)
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| With one required PORV inoperable and pressurizer water level 57%,
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| the required PORV must be restored to OPERABLE status within a Completion Time of 7 days. Two valves are required to meet the LCO requirement and to provide low temperature overpressure mitigation while withstanding a single failure of an active component.
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| The Completion Time is based on only one PORV being required to mitigate an overpressure transient, the likelihood of an active failure of the remaining valve path during this time period being very low, and that a steam bubble exists in the pressurizer. Since the pressure response to a transient is greater if the pressurizer steam space is small or if the PCS is solid, the Completion Time for restoration of a PORV flow path to service is shorter. The maximum pressurizer level at which credit can be taken for having a bubble (57%, which provides about 700 cubic feet of steam space) is based on judgment rather than by analysis. This level provides the same steam volume to dampen pressure transients as would be available at full power. This steam volume provides time for operator action (if the PORVs failed to operate) in the interval between an inadvertent SIS and PCS pressure reaching the 10 CFR 50, Appendix G pressure limit. The time available for action would depend upon the existing pressure and temperature when the inadvertent SIS occurred.
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| C.1 The consequences of operational events that will overpressurize the PCS are more severe at lower temperature (Ref. 8). With the pressurizer water level > 57%, less steam volume is available to dampen pressure increases resulting from an inadvertent mass injection or heatup transients. Thus, with one required PORV inoperable and the pressurizer water level > 57%, the Completion Time to restore the required PORV to OPERABLE status is 24 hours.
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| The 24 hour Completion Time to restore the required PORV to OPERABLE status when the pressurizer water level is > 57%, which usually occurs in MODE 5 or in MODE 6 when the vessel head is on, is a reasonable amount of time to investigate and repair PORV failures without a lengthy period with only one PORV OPERABLE to protect against overpressure events.
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| Palisades Nuclear Plant B 3.4.12-9 Revised 01/29/2020
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| INSERT Bases 3.4.12 LTOP System B 3.4.12 BASES ACTIONS D.1 (continued)
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| If two required PORVs are inoperable, or if the Required Actions and the associated Completion Times are not met, or if the LTOP System is inoperable for any reason other than Condition A, B, or C, the PCS must be depressurized and a vent established within 8 hours. The vent must be sized to provide a relieving capability of 167 gpm at a pressure of 315 psia which ensures the flow capacity is greater than that required for the worst case mass injection transient reasonable during the applicable MODES. This action protects the PCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel.
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| The Completion Time of 8 hours to depressurize and vent the PCS is based on the time required to place the plant in this condition and the relatively low probability of an overpressure event during this time period due to operator attention and administrative requirements.
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| SURVEILLANCE SR 3.4.12.1 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass injection capability, both HPSI pumps are verified to be incapable of injecting into the PCS. The HPSI pumps are rendered incapable of injecting into the PCS by means that assure that a single event cannot cause overpressurization of the PCS due to operation of the pump. Typical methods for accomplishing this are by pulling the HPSI pump breaker control power fuses, racking out the HPSI pump motor circuit breaker, or closing the manual discharge valve.
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| SR 3.4.12.1 is modified by a Note which only requires the SR to be met when complying with LCO 3.4.12.a. When all PCS cold leg temperature are 300°F, a start of both HPSI pumps in conjunction with a charging/letdown imbalance will not cause the PCS pressure to exceed the 10 CFR 50 Appendix G limits. Thus, this SR is only required when any PCS cold leg temperature is reduced to less than 300°F.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.4.12-10 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.12 LTOP System B 3.4.12 BASES SURVEILLANCE SR 3.4.12.2 REQUIREMENTS (continued) SR 3.4.12.2 requires a verification that the required PCS vent, capable of relieving 167 gpm at a PCS pressure of 315 psia, is OPERABLE by verifying its open condition.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| The passive vent arrangement must only be open to be OPERABLE.
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| This Surveillance need only be performed if vent valves are being used to satisfy the requirements of this LCO. This Surveillance does not need to be performed for vent paths relying on the removal of a steam generator primary manway cover, pressurizer manway cover, safety valve or PORV since their position is adequately addressed using administrative controls and the inadvertent reinstallation of these components is unlikely. The Frequencies consider operating experience with mispositioning of unlocked and locked vent valves, respectively.
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| SR 3.4.12.3 The PORV block valve must be verified open to provide the flow path for each required PORV to perform its function when actuated. The valve can be remotely verified open in the main control room.
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| The block valve is a remotely controlled, motor operated valve. The power to the valve motor operator is not required to be removed, and the manual actuator is not required locked in the inactive position. Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure event.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.4.12-11 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.12 LTOP System B 3.4.12 BASES SURVEILLANCE SR 3.4.12.4 REQUIREMENTS (continued) A successful CHANNEL FUNCITONAL TEST of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. PORV actuation could depressurize the PCS and is not required. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| A Note has been added indicating this SR is required to be performed 12 hours after decreasing any PCS cold leg temperature to < 430°F. This Note allows a discrete period of time to perform the required test without delaying entry into the MODE of Applicability for LTOP. This option may be exercised in cases where an unplanned shutdown below 430°F is necessary as a result of a Required Action specifying a plant shutdown, or other plant evolutions requiring an expedited cooldown of the plant.
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| The test must be performed within 12 hours after entering the LTOP MODES.
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| SR 3.4.12.5 Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is required to adjust the entire channel so that it responds and the valve opens within the required LTOP range and with accuracy to known input.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.4.12-12 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.12 LTOP System B 3.4.12 BASES REFERENCES 1. 10 CFR 50, Appendix G
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| : 2. Generic Letter 88-11
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| : 3. CPC Engineering Analysis, EA-A-PAL-92-095-01
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| : 4. CPC Engineering Analysis, EA-TCD-90-01
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| : 5. CPC Engineering Analysis, EA-E-PAL-89-040-1
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| : 6. CPC Corrective Action Document, A-PAL-91-011
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| : 7. FSAR, Section 7.4
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| : 8. Generic Letter 90-06
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| : 9. Engineering Analysis EA-EC27959-01, "Palisades Pressure-Temperature Limit Curves and Upper-Shelf Energy Evaluation,"
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| February 2012 Palisades Nuclear Plant B 3.4.12-13 Revised 01/29/2020
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| INSERT Bases 3.4.13 PCS Operational LEAKAGE B 3.4.13 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.13 PCS Operational LEAKAGE BASES BACKGROUND Components that contain or transport primary coolant to or from the reactor core make up the PCS. Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the PCS.
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| During plant life, the joint and valve interfaces can produce varying amounts of PCS LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the PCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE.
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| The Palisades Nuclear Plant design criteria (Ref. 1) require means for detecting and, to the extent practical, identifying the source of PCS LEAKAGE.
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| The safety significance of PCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring primary coolant LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur detrimental to the safety of the facility and the public.
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| A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with PCS LEAKAGE detection.
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| This LCO deals with protection of the Primary Coolant Pressure Boundary (PCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analysis radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a Loss Of Coolant Accident (LOCA).
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| Palisades Nuclear Plant B 3.4.13-1 Revised 01/29/2020
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| INSERT Bases 3.4.13 PCS Operational LEAKAGE B 3.4.13 BASES BACKGROUND As defined in 10 CFR 50.2, the PCPB includes all those pressure-(continued) containing components, such as the reactor pressure vessel, piping, pumps, and valves, which are:
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| (1) Part of the primary coolant system, or (2) Connected to the primary coolant system, up to and including any and all of the following:
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| (i) The outermost containment isolation valve in system piping which penetrates the containment, (ii) The second of two valves normally closed during normal reactor operation in system piping which does not penetrate the containment, (iii) The pressurizer safety valves and PORVs.
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| APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses do not SAFETY ANALYSES address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for all events resulting in a discharge of steam from the steam generators to the atmosphere assumes that primary to secondary LEAKAGE from all steam generators (SGs) is 0.3 gpm or increases to 0.3 gpm as a result of accident induced conditions. The LCO requirement to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gallons per day is significantly less than the conditions assumed in the safety analysis.
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| Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a Main Steam Line Break (MSLB), Steam Generator Tube Rupture (SGTR) and the Control Rod Ejection (CRE) accident analyses. The leakage contaminates the secondary fluid.
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| The FSAR (Ref. 2 and 5) analysis for SGTR assumes the contaminated secondary fluid is released via the Main Steam Safety Valves and Atmospheric Dump Valves. The 0.3 gpm primary to secondary LEAKAGE safety analysis assumption is inconsequential, relative to the dose contribution from the affected SG.
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| The MSLB (Ref 3 and 5) is more limiting than SGTR for site radiation releases. The safety analysis for the MSLB accident assumes the entire 0.3 gpm primary to secondary LEAKAGE is through the affected steam generator as an initial condition.
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| The CRE (Ref 4 and 5) accident with primary fluid release through the Palisades Nuclear Plant B 3.4.13-2 Revised 01/29/2020
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| INSERT Bases 3.4.13 PCS Operational LEAKAGE B 3.4.13 BASES APPLICABLE Atmospheric Dump Valves is the less limiting event for site radiation SAFETY ANALYSES releases. The safety analysis for the CRE accident assumes 0.3 gpm (continued) primary to secondary LEAKAGE in one steam generator as an initial condition.
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| The dose consequences resulting from the SGTR, MSLB and CRE accidents are within applicable 10 CFR 50.67 limits and meet the requirements of Appendix A of 10 CFR 50 (GDC 19).
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| PCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| LCO PCS operational LEAKAGE shall be limited to:
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| : a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE from within the PCPB is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in increased LEAKAGE. Violation of this LCO could result in continued degradation of the PCPB.
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| LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.
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| As defined in Section 1.0, pressure boundary LEAKAGE is LEAKAGE (except SG LEAKAGE) through a nonisolable fault in an PCS component body, pipe wall, or vessel wall.
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| : b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE from within the PCPB is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period.
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| Violation of this LCO could result in continued degradation of the PCPB, if the LEAKAGE is from the pressure boundary.
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| : c. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE from within the PCPB is allowed because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the PCS makeup system. Identified LEAKAGE includes LEAKAGE to the containment from specifically located sources which is known not to adversely affect the OPERABILITY of required leakage detection systems, but does not include pressure boundary LEAKAGE or controlled Primary Coolant Pump (PCP) seal leakoff to the Volume Control Tank (a normal function Palisades Nuclear Plant B 3.4.13-3 Revised 01/29/2020
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| INSERT Bases 3.4.13 PCS Operational LEAKAGE B 3.4.13 BASES LCO c. Identified LEAKAGE (continued)
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| (continued) not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.
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| LCO 3.4.14, "PCS Pressure Isolation Valve (PIV) Leakage,"
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| measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in PCS LEAKAGE when the other is leaktight. If both valves leak and result in a loss of mass from the PCS, the loss must be included in the allowable identified LEAKAGE.
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| : d. Primary to Secondary LEAKAGE Through Any One SG The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 6). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day. The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.
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| APPLICABILITY In MODES 1, 2, 3, and 4, the potential for PCPB LEAKAGE is greatest when the PCS is pressurized.
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| In MODES 5 and 6, LEAKAGE limits are not required because the primary coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.
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| ACTIONS A.1 Unidentified LEAKAGE or identified LEAKAGE in excess of the LCO limits must be reduced to within limits within 4 hours. This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the PCPB.
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| Palisades Nuclear Plant B 3.4.13-4 Revised 01/29/2020
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| INSERT Bases 3.4.13 PCS Operational LEAKAGE B 3.4.13 BASES ACTIONS B.1 and B.2 (continued)
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| If any pressure boundary LEAKAGE from within the PCPB exists or primary to secondary LEAKAGE is not within limit, or if unidentified or identified LEAKAGE cannot be reduced to within limits within 4 hours, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. The reactor must be brought to MODE 3 within 6 hours and to MODE 5 within 36 hours. This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.
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| The allowed Completion Times are reasonable, based on operating experience, to reach the required conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 5, the pressure stresses acting on the PCPB are much lower, and further deterioration is much less likely.
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| SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying PCS LEAKAGE to be within the LCO limits ensures the integrity of the PCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an PCS water inventory balance.
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| The PCS water inventory balance must be performed with the reactor at steady state operating conditions. The Surveillance is modified by two Notes. Note 1 states that the SR is not required to be performed in MODES 3 and 4, until 12 hours of steady state operation have elapsed.
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| Steady state operation is required to perform a proper water inventory balance; calculations during maneuvering are not useful and a Note requires the Surveillance to be met only when steady state is established. For PCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable PCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and PCP seal leakoff.
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| An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level.
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| These leakage detection systems are specified in LCO 3.4.15, "PCS Leakage Detection Instrumentation."
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| Palisades Nuclear Plant B 3.4.13-5 Revised 01/29/2020
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| INSERT Bases 3.4.13 PCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE SR 3.4.13.1 (continued)
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| REQUIREMENTS (continued) Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.
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| tA Note under the Frequency column states that this SR is required to be performed during steady state operation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.4.13.2 This SR verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.17, Steam Generator Tube Integrity, should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 7. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.
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| The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.
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| The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 7). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.4.13-6 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.13 PCS Operational LEAKAGE B 3.4.13 BASES (continued)
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| REFERENCES 1. FSAR, Section 5.1.5
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| : 2. FSAR, Section 14.15
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| : 3. FSAR, Section 14.14
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| : 4. FSAR, Section 14.16
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| : 5. FSAR, Section 14.24
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| : 6. NEI 97-06, Steam Generator Program Guidelines
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| : 7. EPRI, Pressurized Water Reactor Primary-to-Secondary Leak Guidelines Palisades Nuclear Plant B 3.4.13-7 Revised 01/29/2020
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| INSERT Bases 3.4.14 PCS PIV Leakage B 3.4.14 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.14 PCS Pressure Isolation Valve (PIV) Leakage BASES BACKGROUND The Reactor Safety Study (RSS), WASH-1400 (Ref. 1), identified a special class of Loss of Coolant Accidents (LOCAs) where the accident is initiated by the failure of check valves which separate the high pressure Primary Coolant System (PCS) from lower pressure systems connected to the PCS. This check valve failure could cause overpressurization and rupture of the lower pressure piping and result in a LOCA that bypasses containment. With the containment bypassed, the leakage would not be available for recirculation and when the Safety Injection Refueling Water Tank (SIRWT) emptied core cooling would be lost. This event has become known as Event V.
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| When pressure isolation is provided by two in-series check valves and failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are important to safety, they should be tested periodically to ensure low probability of gross failure. Periodic examination of check valves must be undertaken to verify that each valve is seated properly and functioning as a pressure isolation device. The testing will reduce the overall risk of an inter-system LOCA. The testing may be accomplished by direct volumetric leakage measurement or by other equivalent means capable of demonstrating that leakage limits are not exceeded. The PCS PIV LCO allows PCS high pressure operation when leakage through these valves exists in amounts that do not compromise safety. The PIV leakage limit applies to each individual valve. Leakage through both PIVs in series in a line must be included as part of the identified LEAKAGE, governed by LCO 3.4.13, "PCS Operational LEAKAGE." This is true during operation only when the loss of PCS mass through two valves in series is determined by a water inventory balance (SR 3.4.13.1).
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| A known component of the identified LEAKAGE before operation begins is the least of the two individual leakage rates determined for leaking series PIVs during the required surveillance testing; leakage measured through one PIV in a line is not PCS operational LEAKAGE if the other is leaktight.
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| Palisades Nuclear Plant B 3.4.14-1 Revised 01/29/2020
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| | |
| INSERT Bases 3.4.14 PCS PIV Leakage B 3.4.14 BASES BACKGROUND Although this specification provides a limit on allowable PIV leakage (continued) rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems. Therefore, this specification also addresses the potential for overpressurization of the low pressure piping in the Shutdown Cooling (SDC) system caused by the inadvertent opening of the SDC suction valves (MO-3015 and MO-3016) when the PCS pressure is above the design pressure of the SDC System. The leakage limit is an indication that the PIVs between the PCS and the connecting systems are degraded or degrading. PIV leakage or inadvertent valve positioning could lead to overpressure of the low pressure piping or components. Failure consequences could be a LOCA outside of containment, which is an unanalyzed condition that could degrade the ability for low pressure injection.
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| PIVs are provided to isolate the PCS from the following systems:
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| : a. Shutdown Cooling System; and
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| : b. Safety Injection System.
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| The PIVs which are required to be leak tested are listed in Table B 3.4.14-1.
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| Violation of this LCO could result in overpressurization of a low pressure system and the loss of the integrity of a fission product barrier.
| |
| APPLICABLE Reference 1 identified potential intersystem LOCAs as a significant SAFETY ANALYSES contributor to the risk of core melt. The dominant accident sequence in the intersystem LOCA category is the failure of low pressure piping outside of containment. The accident is the result of a postulated failure of the PIVs, which are part of the Primary Coolant Pressure Boundary (PCPB), and the subsequent pressurization of the lower pressure piping downstream of the PIVs from the PCS.
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| Overpressurization failure of the lower pressure piping would result in a LOCA outside containment and subsequent risk of core melt.
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| Reference 2 evaluated various PIV configurations, leakage testing of the valves, and operational changes to determine the effect on the probability of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA.
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| PCS PIV leakage satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| Palisades Nuclear Plant B 3.4.14-2 Revised 01/29/2020
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| INSERT Bases 3.4.14 PCS PIV Leakage B 3.4.14 BASES LCO PCS PIV leakage is identified LEAKAGE into closed systems connected to the PCS. Isolation valve leakage is usually on the order of drops per minute. Leakage that increases significantly suggests that corrective action must be taken. The PIVs which are required to be leak tested are listed in Table 3.4.14-1.
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| The LCO PIV leakage limit is a maximum of 5 gpm. Reference 3 permits leakage testing at a lower pressure differential than that between maximum PCS pressure and the normal pressure of the connected system during PCS operation (the maximum pressure differential). The observed leakage rate must be corrected to the maximum pressure differential, assuming leakage is directly proportional to the square root of pressure differential.
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| The LCO also requires the SDC suction valve interlocks to be OPERABLE in order to prevent the inadvertent opening of the SDC suction valves when PCS pressure is above the 300 psig design pressure of the SDC suction piping. When PCS pressure is > 280 psia as sensed by the pressurizer narrow range pressure channels, an inhibit signal is placed on the control circuit for the SDC suction valves which prevents the valves from opening and thus avoiding a potential overpressurization event of the SDC piping. For the SDC suction valve interlocks to be OPERABLE, two channels of pressurizer narrow range pressure instruments must be capable of providing an open inhibit signal to their respective isolation valve.
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| APPLICABILITY In MODES 1, 2, 3, and 4, this LCO applies because the PIV leakage potential is greatest when the PCS is pressurized. In MODE 4, the requirements of this LCO are not required when in, or during the transition to or from, the SDC mode of operation since these evolutions are performed when PCS pressure is less than the limiting design pressure of the systems addressed by this specification.
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| In MODES 5 and 6, leakage limits are not provided because the lower primary coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment.
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| Palisades Nuclear Plant B 3.4.14-3 Revised 01/29/2020
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| INSERT Bases 3.4.14 PCS PIV Leakage B 3.4.14 BASES ACTIONS The ACTIONS are modified by two Notes. Note 1 is added to provide clarification that each flow path allows separate entry into a Condition.
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| This is allowed based on the functional independence of the flow path.
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| Note 2 requires an evaluation of affected systems if a PIV is inoperable. The leakage may have affected system operability or isolation of a leaking flow path with an alternate valve may have degraded the ability of the interconnected system to perform its safety function.
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| A.1 and A.2 Required Action A.1 requires that isolation with one valve must be performed within 4 hours whenever one or more flow paths with leakage from one or more PIVs is not within limits. Four hours provides time to reduce leakage in excess of the allowable limit or to isolate the flow path if leakage cannot be reduced while restricting operation with leaking isolation valves. Required Action A.1 is modified by a Note stating that the valves used for isolation must meet the same leakage requirement as the PIVs and must be in the PCPB or the high pressure portion of the system.
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| The 72 hour Completion Time after exceeding the limit allows for the restoration of the leaking PIV to OPERABLE status. This time frame considers the time required to complete this action and the low probability of a second valve failing during this period.
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| B.1 and B.2 If leakage cannot be reduced or if the affected system can not be isolated within the specified Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours and to MODE 5 within 36 hours. This action reduces the leakage and also reduces the potential for a LOCA outside the containment. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.4.14-4 Revised 01/29/2020
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| INSERT Bases 3.4.14 PCS PIV Leakage B 3.4.14 BASES ACTIONS C.1 (continued)
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| The inoperability of the SDC suction valve interlocks renders the SDC suction isolation valves incapable of preventing an inadvertent opening of the valves at PCS pressures in excess of the SDC systems design pressure. If the SDC suction valve interlocks are inoperable, operation may continue as long as the suction penetration is closed by at least one closed deactivated valve within 4 hours. This action accomplishes the purpose of the interlock. The 4 hour Completion Time provides time to accomplish the action and restricts operation with an inoperable interlock.
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| SURVEILLANCE SR 3.4.14.1 REQUIREMENTS Performance of leakage testing on each PCS PIV or isolation valve used to satisfy Required Action A.1 is required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of up to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure condition.
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| For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves.
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| If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.
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| Testing is to be performed every 9 months whenever the plant has been in MODE 5 for 7 days or more, but may be extended up to a maximum of 18 months, a typical refueling cycle, if the plant does not go into MODE 5 for at least 7 days. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| The leakage limit is to be met at the PCS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.
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| Palisades Nuclear Plant B 3.4.14-5 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.14 PCS PIV Leakage B 3.4.14 BASES SURVEILLANCE SR 3.4.14.1 (continued)
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| REQUIREMENTS SR 3.4.14.1 is modified by three Notes. Note 1 states that the SR is only required to be performed in MODES 1 and 2. Entry into MODES 3 and 4 is allowed to establish the necessary differential pressure and stable conditions to allow performance of this surveillance.
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| Note 2 further restricts the PIV leakage rate acceptance criteria by limiting the reduction in margin between the measured leakage rate and the maximum permissible leakage rate by 50% or greater.
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| Reductions in margin by 50% or greater may be indicative of PIV degradation and warrant inspection or additional testing. Thus, leakage rates less than 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
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| Note 3 limits the minimum test differential pressure to 150 psid during performance of PIV leakage testing.
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| SR 3.4.14.2 Verifying that the SDC suction valve interlocks are OPERABLE ensures that PCS pressure will not pressurize the SDC system beyond 125% of its design pressure of 300 psig. The interlock setpoint that prevents the valves from being opened is set so the actual PCS pressure must be < 280 psia to open the valves. This setpoint ensures the SDC design pressure will not be exceeded and the SDC relief valves will not lift. The narrow range pressure transmitters that provide the SDC suction valve interlocks are sensed from the pressurizer. Due to the elevation differences between these narrow range pressure transmitter calibration points and the SDC suction piping, the pressure in the SDC suction piping will be higher than the indicated pressurizer pressure. Due to this pressure difference, the SDC suction valve interlocks are conservatively set at or below 280 psia to ensure that the 300 psig (315 psia) design pressure of the suction piping is not exceeded. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.4.14-6 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.14 PCS PIV Leakage B 3.4.14 BASES SURVEILLANCE SR 3.4.14.3 REQUIREMENTS (continued) This SR requires a verification that the four Low Pressure Safety Injection (LPSI) check valves (CK-3103, CK-3118, CK-3133 and CK-3148) in the SDC flow path reclose after stopping SDC flow.
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| Performance of this SR is necessary to ensure the LPSI check valves are closed to prevent overpressurization of the LPSI subsystem from the High Pressure Safety Injection (HPSI) subsystem.
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| Overpressurization of the LPSI piping could occur if the LPSI check valves were not closed upon the receipt of a Safety Injection Signal and PCS pressure remained relatively high (e.g., during a small break LOCA). In this case, the higher pressure water from the discharge of the HPSI pumps could cause the lower pressure LPSI piping to exceed its design pressure. This event could result in a loss of emergency core cooling water outside containment which reduces the volume of water available for recirculation from the containment sump (Ref. 4).
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| SR 3.4.14.3 is required to be performed on a Frequency of prior to entering MODE 2 whenever the LPSI check valves have been used for SDC. This ensures the LPSI check valves are closed whenever they have been opened for SDC operations prior to a reactor startup. The SR is modified by a Note which states that the surveillance is only required to be performed in MODES 1 and 2. Thus, entry into MODES 3 and 4 is allowed to establish the necessary differential pressure and to establish stable conditions to allow performance of this surveillance.
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| REFERENCES 1. WASH-1400 (NUREG-75/014), Appendix V, October 1975
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| : 2. NUREG-0677, May 1980
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| : 3. ASME Code for Operation and Maintenance of Nuclear Power Plants.
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| : 4. Letter from Consumers Power Company to D.M. Crutchfield (NRC) Requesting a Change to the Palisades Plant Technical Specification, dated July 29, 1982 Palisades Nuclear Plant B 3.4.14-7 Revised 01/29/2020
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| INSERT Bases 3.4.14 PCS PIV Leakage B 3.4.14 BASES TABLE B 3.4.14-1 (page 1 of 1)
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| Required PCS Pressure Isolation Valves System Valve No.
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| High Pressure Safety Injection Loop 1A, Cold Leg CK - 3101 CK - 3104 Loop 1B, Cold Leg CK - 3116 CK - 3119 Loop 2A, Cold Leg CK - 3131 CK - 3134 Loop 2B, Cold Leg CK - 3146 CK - 3149 Low Pressure Safety Injection Loop 1A, Cold Leg CK - 3103 Loop 1B, Cold Leg CK - 3118 Loop 2A, Cold Leg CK - 3133 Loop 2B, Cold Leg CK - 3148 Palisades Nuclear Plant B 3.4.14-8 Revised 01/29/2020
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| INSERT Bases 3.4.15 PCS Leakage Detection Instrumentation B 3.4.15 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.15 PCS Leakage Detection Instrumentation BASES BACKGROUND The Palisades Nuclear Plant design criteria (Ref. 1) require means for detecting and, to the extent practical, identifying the location of the source of PCS LEAKAGE.
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| Leakage detection instrumentation must have the capability to detect significant Primary Coolant Pressure Boundary (PCPB) degradation as soon after occurrence as practical to minimize the potential for propagation to a gross failure. Thus, an early indication or warning signal is necessary to permit proper evaluation of all unidentified LEAKAGE.
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| Industry practice has shown that water flow changes of 0.5 gpm to 1.0 gpm can readily be detected in contained volumes by monitoring changes in water level, in flow rate, or in the operating frequency of a pump. The containment sump, which is used to collect unidentified LEAKAGE, is instrumented with level transmitters providing sump level indication in the control room. The sensitivity of these instruments is acceptable for detecting increases in unidentified LEAKAGE.
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| The primary coolant contains radioactivity that, when released to the containment, can be detected by radiation monitoring instrumentation.
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| Primary coolant radioactivity levels will be low during initial reactor startup and for a few weeks thereafter until activated corrosion products have been formed and fission products appear from fuel element cladding contamination or cladding defects. An instrument sensitivity capable of detecting a 100 cm3/min leak in 45 minutes based on 1%
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| failed fuel is practical for the leakage detection instrument (Ref. 2).
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| Radioactivity detection is included for monitoring gaseous activities because of its sensitivity to PCS LEAKAGE.
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| An increase in humidity of the containment atmosphere would indicate release of water vapor to the containment. Humidity detectors are capable of detecting a 10% change in humidity which would result from approximately 150 gallons of primary water leakage (Ref. 2).
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| Palisades Nuclear Plant B 3.4.15-1 Revised 01/29/2020
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| INSERT Bases 3.4.15 PCS Leakage Detection Instrumentation B 3.4.15 BASES BACKGROUND Since the humidity level is influenced by several factors, a quantitative (continued) evaluation of an indicated leakage rate by this means may be questionable and should be compared to observed increases in liquid flow into or from the containment sump and condensate flow from the containment air coolers. Humidity level monitoring is considered most useful as an indirect indication to alert the operator to a potential problem.
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| The containment air cooler design includes a sump with a drain, a liquid level switch, and an overflow path. Normally, very little water will be condensed from the containment atmosphere and the small amount of condensate will easily flow out through the sump drain. If flow to the sump is greater than 20 gpm, the level in the sump will rise to the liquid level switch (approximately 6 inches from the bottom of the sump) and triggers an alarm in the control room. Excessive flow to the sump is indicative of a service water leak, steam leak, or a primary coolant system leak. A steam leak or primary coolant leak would be accompanied by an increase in the containment atmosphere humidity which would be detected by the containment humidity sensors and displayed in the control room. Since excessive containment air cooler drainage may be attributed to causes other than PCS LEAKAGE, an evaluation of PCS LEAKAGE should be confirmed using diverse instrumentation required by this specification.
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| Air temperature and pressure monitoring methods may also be used to infer unidentified LEAKAGE to the containment. Containment temperature and pressure fluctuate during plant operation, but a rise above the normally indicated range of values may indicate PCS LEAKAGE into the containment. The relevance of temperature and pressure measurements are affected by containment free volume and, for temperature, detector location. Alarm signals can be valuable in recognizing rapid and sizable leakage to the containment.
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| Temperature and pressure monitors are not required by this LCO.
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| Palisades Nuclear Plant B 3.4.15-2 Revised 01/29/2020
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| INSERT Bases 3.4.15 PCS Leakage Detection Instrumentation B 3.4.15 BASES APPLICABLE The need to evaluate the severity of an alarm or an indication is SAFETY ANALYSES important to the operators, and the ability to compare and verify with indications from other systems is necessary. The system sensitivities are described in the FSAR (Ref. 2). Multiple instrument locations are utilized, if needed, to ensure the transport delay time of the LEAKAGE from its source to an instrument location is acceptable.
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| The safety significance of PCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring PCS LEAKAGE into the containment area are necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE provides quantitative information to the operators, allowing them to take corrective action should leakage occur detrimental to the safety of the facility and the public.
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| PCS leakage detection instrumentation satisfies Criterion 1 of 10 CFR 50.36(c)(2).
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| LCO One method of protecting against large PCS LEAKAGE is based on the ability of instruments to rapidly detect extremely small leaks. This LCO requires instruments of diverse monitoring principles to be OPERABLE to provide a high degree of confidence that extremely small leaks are detected in time to allow actions to place the plant in a safe condition when PCS LEAKAGE indicates possible PCPB degradation.
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| The LCO is satisfied when monitors of diverse measurement means are available. Thus, a combination which includes one instrument channel from each of any three of the following; containment sump level indication, gaseous activity monitor, containment air cooler condensate level switch, or containment humidity monitor provides an acceptable minimum. For the containment air cooler condensate level switch only an operating containment air cooler may be relied upon to fulfill the LCO requirements for an OPERABLE leakage detection instrument.
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| APPLICABILITY Because of elevated PCS temperature and pressure in MODES 1, 2, 3, and 4, PCS leakage detection instrumentation is required to be OPERABLE.
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| In MODE 5 or 6, the temperature is 200°F and pressure is maintained low or at atmospheric pressure.
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| Palisades Nuclear Plant B 3.4.15-3 Revised 01/29/2020
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| INSERT Bases 3.4.15 PCS Leakage Detection Instrumentation B 3.4.15 BASES APPLICABILITY Since the temperatures and pressures are far lower than those for (continued) MODES 1, 2, 3, and 4, the likelihood of leakage and crack propagation is much smaller. Therefore, the requirements of this LCO are not applicable in MODES 5 and 6.
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| ACTIONS A.1 and A.2 If one or two required leak detection instrument channels are inoperable, a periodic surveillance for PCS water inventory balance, SR 3.4.13.1, must be performed at an increased frequency of 24 hours to provide information that is adequate to detect leakage.
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| As stated in SR 3.0.2, the 25% extension allowed by SR 3.0.2 may be applied to Required Actions whose Completion Time is stated as once per . . . however, the 25% extension does not apply to the initial performance of a Required Action with a periodic Completion Time that requires performance on a "once per. . ." basis. The 25% extension applies to each performance of the Required Action after the initial performance. Therefore, while Required Action 3.4.15 A.1 must be initially performed within 24 hours without any SR 3.0.2 extension, subsequent performances may utilize the 25% SR 3.0.2 extension.
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| Restoration of the required instrument channels to an OPERABLE status is required to regain the function in a Completion Time of 30 days after the instruments failure. This time is acceptable considering the frequency and adequacy of the PCS water inventory balance required by Required Action A.1.
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| B.1 and B.2 If the Required Action cannot be met within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.4.15-4 Revised 01/29/2020
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| INSERT Bases 3.4.15 PCS Leakage Detection Instrumentation B 3.4.15 BASES ACTIONS C.1 (continued)
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| If all required monitors are inoperable, no automatic means of monitoring leakage are available and immediate plant shutdown in accordance with LCO 3.0.3 is required.
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| SURVEILLANCE SR 3.4.15.1, SR 3.4.15.2, and SR 3.4.15.3 REQUIREMENTS These SRs require the performance of a CHANNEL CHECK for each required containment sump level indicator, containment atmosphere gaseous activity monitor, and containment atmosphere humidity monitor. The check gives reasonable confidence the channel is operating properly. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.4.15.4 SR 3.4.15.4 requires the performance of a CHANNEL FUNCTIONAL TEST of the required containment air cooler condensate level switch.
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| Since this instrumentation does not include control room indication of flow rate, a CHANNEL CHECK is not possible. The test ensures that the level switch can perform its function in the desired manner. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.4.15-5 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.15 PCS Leakage Detection Instrumentation B 3.4.15 BASES SURVEILLANCE SR 3.4.15.5, SR 3.4.15.6, and SR 3.4.15.7 REQUIREMENTS (continued) These SRs require the performance of a CHANNEL CALIBRATION for each required containment sump level, containment atmosphere gaseous activity, and containment atmosphere humidity channel. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Section 5.1.5
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| : 2. FSAR, Sections 4.7 and 6.3 Palisades Nuclear Plant B 3.4.15-6 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.16 PCS Specific Activity B 3.4.16 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.16 PCS Specific Activity BASES BACKGROUND 10 CFR 100.11 and 10 CFR 50.67 specify the maximum dose an individual at the site boundary can receive for 2 hours during an accident.
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| The limits on specific activity ensure that the doses are held within applicable limits during analyzed transients and accidents.
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| The PCS specific activity LCO limits the allowable concentration level of radionuclides in the primary coolant. The LCO limits are established to minimize the offsite radioactivity dose consequences in the event of a Steam Generator Tube Rupture (SGTR) or other accident.
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| The LCO contains specific activity limits for both DOSE EQUIVALENT I-131 and gross specific activity. The allowable levels are intended to limit the 2 hour dose at the site boundary to within applicable dose guideline limits. The limits in the LCO were standardized based on parametric evaluations of offsite radioactivity dose consequences for typical site locations.
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| The parametric evaluations showed the potential offsite dose levels for a SGTR accident were an appropriately small fraction of the 10 CFR 100 dose guideline limits. Each evaluation assumes a broad range of site applicable atmospheric dispersion factors.
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| APPLICABLE The LCO limits on the specific activity of the primary coolant ensure SAFETY ANALYSES that the resulting offsite doses will not exceed applicable limits following a SGTR or other accident. The SGTR safety analysis (Ref. 1) assumes the specific activity of the primary coolant at the LCO limits and an existing primary coolant Steam Generator (SG) tube leakage rate of 0.3 gpm.
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| The analysis also assumes a reactor trip and a turbine trip at the same time as the SGTR event.
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| The analysis for the SGTR accident is an input to the acceptance limits for PCS specific activity. Reference to this analysis is used to assess changes to the facility that could affect PCS specific activity as they relate to the acceptance limits.
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| Palisades Nuclear Plant B 3.4.16-1 Revised 01/29/2020
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| INSERT Bases 3.4.16 PCS Specific Activity B 3.4.16 BASES APPLICABLE The rise in pressure in the ruptured SG causes radioactive SAFETY ANALYSES contaminated steam to discharge to the atmosphere through the (continued) atmospheric dump valves or the main steam safety valves. The atmospheric discharge stops when the affected SG is isolated below approximately 525°F. The unaffected SG removes core decay heat by venting steam until Shutdown Cooling conditions are reached.
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| The safety analysis shows the radiological consequences of a SGTR accident are within applicable 10 CFR 50.67 limits. Operation with iodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not exceed the limit of 40 µCi/gm for more than 48 hours.
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| This is acceptable because of the low probability of a SGTR accident occurring during the established 48 hour time limit. The occurrence of a SGTR accident at these permissible levels could increase the site boundary dose levels.
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| PCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2).
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| LCO The specific iodine activity is limited to 1.0 Ci/gm DOSE EQUIVALENT I-131, and the gross specific activity in the primary coolant is limited to the number of Ci/gm equal to 100 divided by (average disintegration energy). The limit on DOSE EQUIVALENT I-131 ensures the offsite doses during an accident remains within applicable 10 CFR 50.67 limits.
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| The SGTR accident analysis (Ref. 1) shows that the 2 hour site boundary dose levels are within acceptable limits. Violation of the LCO may result in primary coolant radioactivity levels that could, in the event of an SGTR or other accident, lead to site boundary doses that exceed the applicable 10 CFR 50.67 limits.
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| Palisades Nuclear Plant B 3.4.16-2 Revised 01/29/2020
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| INSERT Bases 3.4.16 PCS Specific Activity B 3.4.16 BASES APPLICABILITY In MODES 1 and 2, and in MODE 3 with PCS average temperature 500°F, operation within the LCO limits for DOSE EQUIVALENT I-131 and gross specific activity is necessary to contain the potential consequences of an SGTR or other accident to within applicable 10 CFR 50.67 limits.
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| For operation in MODE 3 with PCS average temperature < 500°F, and in MODES 4 and 5, the release of radioactivity in the event of an SGTR is unlikely since the saturation pressure corresponding to the primary coolant temperature is below the lift settings of the atmospheric dump valves and main steam safety valves.
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| ACTIONS A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S) while relying on the ACTIONS.
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| A.1 and A.2 With the DOSE EQUIVALENT I-131 greater than the LCO limit, samples at intervals of 4 hours must be taken to demonstrate the limit 40 µCi/gm is not exceeded. The Completion Time of 4 hours is required to obtain and analyze a sample.
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| As stated in SR 3.0.2, the 25% extension allowed by SR 3.0.2 may be applied to Required Actions whose Completion Time is stated as once per . .
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| . however, the 25% extension does not apply to the initial performance of a Required Action with a periodic Completion Time that requires performance on a "once per. . ." basis. The 25% extension applies to each performance of the Required Action after the initial performance. Therefore, while Required Action 3.4.16 A.1 must be initially performed within 4 hours without any SR 3.0.2 extension, subsequent performances may utilize the 25% SR 3.0.2 extension.
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| Sampling must continue for trending. The DOSE EQUIVALENT I-131 must be restored to within limits within 48 hours.
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| The Completion Time of 48 hours is required if the limit violation resulted from normal iodine spiking.
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| A Note to the Required Actions of Condition A excludes the MODE change restriction of LCO 3.0.4. This exception allows entry into the applicable MODE(S) while relying on the ACTIONS even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to, power operation.
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| Palisades Nuclear Plant B 3.4.16-3 Revised 01/29/2020
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| INSERT Bases 3.4.16 PCS Specific Activity B 3.4.16 BASES ACTIONS B.1 (continued)
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| If a Required Action and associated Completion Time of Condition A is not met or if the DOSE EQUIVALENT I-131 is 40 µCi/gm or above, or with the gross specific activity in excess of the allowed limit, the plant must be placed in a MODE in which the requirement does not apply.
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| The change within 6 hours to MODE 3 with PCS average temperature
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| < 500°F lowers the saturation pressure of the primary coolant below the setpoints of the main steam safety valves and prevents venting the SG to the environment in an SGTR event. The allowed Completion Time of 6 hours is required to reach MODE 3 below 500°F from full power conditions and without challenging plant systems.
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| SURVEILLANCE SR 3.4.16.1 REQUIREMENTS The Surveillance requires performing a gamma isotopic analysis as a measure of the gross specific activity of the primary coolant. While basically a quantitative measure of radionuclides with half lives longer than 15 minutes, excluding iodines, this measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in gross specific activity.
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| Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The Surveillance is applicable in MODES 1 and 2, and in MODE 3 with PCS average temperature at least 500°F. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.4.16-4 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.16 PCS Specific Activity B 3.4.16 BASES SURVEILLANCE SR 3.4.16.2 REQUIREMENTS (continued) This Surveillance is performed to ensure iodine remains within limits during normal operation and following fast power changes when fuel failure is more apt to occur. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency, between 2 hours and 6 hours after any power change of 15% RTP within a 1 hour period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results. If any (may be more than one) power change 15% RTP occurs within a 1 hour period, then more than one sample may be required to ensure that an iodine peak sample is obtained between the 2 and 6 hour Frequency requirement. This SR is modified by a Note which states that the SR is only required to be performed in MODE 1. Entrance into a lower MODE does not preclude completion of this surveillance.
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| SR 3.4.16.3 A radiochemical analysis for determination is required with the plant operating in MODE 1 equilibrium conditions. The determination directly relates to the LCO and is required to verify plant operation within the specified gross activity LCO limit. The analysis for is a measurement of the average energies per disintegration for isotopes with half lives longer than 15 minutes, excluding iodines. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| This SR has been modified by a Note that indicates sampling is required to be performed within 31 days after 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 48 hours. This ensures the radioactive materials are at equilibrium so the analysis for is representative and not skewed by a crud burst or other similar abnormal event.
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| REFERENCES 1. FSAR, Section 14.15 Palisades Nuclear Plant B 3.4.16-5 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.4.17 SG Tube Integrity B 3.4.17 B 3.4 PRIMARY COOLANT SYSTEM (PCS)
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| B 3.4.17 Steam Generator (SG) Tube Integrity BASES BACKGROUND Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers.
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| The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the primary coolant pressure boundary (PCPB) and, as such, are relied on to maintain the primary systems pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the PCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the PCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.4, PCS Loops - MODES 1 and 2, LCO 3.4.5, PCS Loops - MODE 3, LCO 3.4.6, PCS Loops - MODE 4, and LCO 3.4.7, PCS Loops - MODE 5, Loops Filled.
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| SG tube integrity means that the tubes are capable of performing their intended PCPB safety function consistent with the licensing basis, including applicable regulatory requirements.
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| Steam generator tubing is subject to a variety of degradation mechanisms. Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively.
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| The SG performance criteria are used to manage SG tube degradation.
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| Specification 5.5.8, Steam Generator (SG) Program, requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 5.5.8, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. The SG performance criteria are described in Specification 5.5.8. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.
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| The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1).
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| Palisades Nuclear Plant B 3.4.17-1 Amendment No. 226 Revision 04/14/2016
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| INSERT Bases 3.4.17 SG Tube Integrity B 3.4.17 BASES APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate that bounds the operational LEAKAGE rate limits in LCO 3.4.13, PCS Operational LEAKAGE, plus the leakage rate associated with a double-ended rupture of a single tube.
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| The accident analysis for a SGTR assumes the contaminated secondary fluid is released to the atmosphere via the Main Steam Safety Valves and Atmospheric Dump Valves.
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| The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from all SGs of 0.3 gpm or is assumed to increase to 0.3 gpm as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I-131 is assumed to be equal to the LCO 3.4.16, PCS Specific Activity, limits.
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| For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the applicable limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3), 10 CFR 50.67 (Ref. 7) or the NRC approved licensing basis (e.g., a small fraction of these limits).
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| Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
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| LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.
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| During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity.
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| In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall , between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet.
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| The tube-to-tubesheet weld is not considered part of the tube.
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| A SG tube has tube integrity when it satisfies the SG performance criteria.
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| The SG performance criteria are defined in Specification 5.5.8, Steam Generator Program, and describe acceptable SG tube performance.
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| The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.
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| Palisades Nuclear Plant B 3.4.17-2 Amendment No. 226 Revised 04/14/2016
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| INSERT Bases 3.4.17 SG Tube Integrity B 3.4.17 BASES LCO There are three SG performance criteria: structural integrity, accident (continued) induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.
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| The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation. Tube collapse is defined as, For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero. The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term significant is defined as An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established. For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis.
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| The division between primary and secondary classifications will be based on detailed analysis and/or testing.
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| Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code, Section III, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification.
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| This includes safety factors and applicable design basis loads based on ASME Code, Section III, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).
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| The accident induced leakage performance criterion ensures that the primary to secondary LEAKAGE caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceed 0.3 gpm per SG. The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.
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| Palisades Nuclear Plant B 3.4.17-3 Amendment No. 226 Revision 04/14/2016
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| INSERT Bases 3.4.17 SG Tube Integrity B 3.4.17 BASES LCO The operational LEAKAGE performance criterion provides an observable (continued) indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13, PCS Operational LEAKAGE, and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.
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| APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.
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| PCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.
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| ACTIONS The ACTIONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube. Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated Required Actions.
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| A.1 and A.2 Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.17.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is Palisades Nuclear Plant B 3.4.17-4 Amendment No. 226 Revised 04/14/2016
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| INSERT Bases 3.4.17 SG Tube Integrity B 3.4.17 BASES ACTIONS A.1 and A.2 (continued)
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| (continued) discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Condition B applies.
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| A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.
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| If the evaluation determines that the affected tube(s) have tube integrity, Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged prior to entering MODE 4 following the next refueling outage or SG inspection.
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| This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.
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| B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours and MODE 5 within 36 hours.
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| The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.
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| SURVEILLANCE SR 3.4.17.1 REQUIREMENTS During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.
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| During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the as found condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.
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| Palisades Nuclear Plant B 3.4.17-5 Amendment No. 226 Revised 04/14/2016
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| INSERT Bases 3.4.17 SG Tube Integrity B 3.4.17 BASES SURVEILLANCE SR 3.4.17.1 (continued)
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| REQUIREMENTS (continued) The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation.
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| Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.
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| The Steam Generator Program defines the Frequency of SR 3.4.17.1.
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| The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.5.8 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.
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| SR 3.4.17.2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging.
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| The tube repair criteria delineated in Specification 5.5.8 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.
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| The Frequency of prior to entering MODE 4 following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.
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| Palisades Nuclear Plant B 3.4.17-6 Amendment No. 226 Revised 04/14/2016
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| INSERT Bases 3.4.17 SG Tube Integrity B 3.4.17 BASES REFERENCES 1. NEI 97-06, Steam Generator Program Guidelines
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| : 2. 10 CFR 50 Appendix A, GDC 19
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| : 3. 10 CFR 100
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| : 4. ASME Boiler and Pressure Vessel Code, Section III, Subsection NB
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| : 5. Draft Regulatory Guide 1.121, Basis for Plugging Degraded Steam Generator Tubes, August 1976
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| : 6. EPRI, Pressurized Water Reactor Steam Generator Examination Guidelines
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| : 7. 10 CFR 50.67 Palisades Nuclear Plant B 3.4.17-7 Amendment No. 226 Revised 04/14/2016
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| INSERT Bases 3.5.1 SITs B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
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| B 3.5.1 Safety Injection Tanks (SITs)
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| BASES BACKGROUND The functions of the four SITs are to supply water to the reactor vessel during the blowdown phase of a Loss of Coolant Accident (LOCA), to provide inventory to help accomplish the refill phase that follows thereafter, and to provide Primary Coolant System (PCS) makeup for a small break LOCA.
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| The blowdown phase of a LOCA is the initial period of the transient during which the PCS departs from equilibrium conditions, and heat from fission product decay, hot internals, and the vessel continues to be transferred to the primary coolant. The blowdown phase of the transient ends when the PCS pressure falls to a value approaching that of the containment atmosphere.
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| The refill phase of a LOCA follows immediately after the primary coolant inventory has vacated the core through steam flashing and ejection out through the break. The core is essentially in adiabatic heatup. The balance of the SITs' inventory is then available to help fill voids in the lower plenum and reactor vessel downcomer to establish a recovery level at the bottom of the core and ongoing reflood of the core with the addition of Safety Injection (SI) water.
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| The SITs are pressure vessels partially filled with borated water and pressurized with nitrogen gas (Ref. 2). The SITs are passive components, since no operator or control action is required for them to perform their function. Internal tank pressure and elevation head are sufficient to discharge the contents to the PCS, if PCS pressure decreases below the SIT pressure.
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| Each SIT is piped into one PCS cold leg via the injection lines utilized by the High Pressure Safety Injection and Low Pressure Safety Injection (HPSI and LPSI) systems. Each SIT is isolated from the PCS by a motor operated isolation valve and two check valves in series. The motor operated isolation valves are normally open, with power removed from the valve motor to prevent inadvertent closure prior to or during an accident.
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| Palisades Nuclear Plant B 3.5.1-1 Revised 01/29/2020
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| INSERT Bases 3.5.1 SITs B 3.5.1 BASES BACKGROUND The SIT gas and water volumes, gas pressure, tank elevation, and (continued) outlet pipe size are selected to allow three of the four SITs to partially recover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that three SITs are adequate for this function is consistent with the LOCA assumption that the entire contents of one SIT will be lost via the break during the blowdown phase of a LOCA.
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| APPLICABLE The SITs are credited in both the large and small break LOCA SAFETY ANALYSES analyses at full power (Ref. 1). These are the Design Basis Accidents (DBAs) that establish the acceptance limits for the SITs. Reference to the analyses for these DBAs is used to assess changes to the SITs as they relate to the acceptance limits.
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| In performing the LOCA calculations, conservative assumptions are made concerning the availability of SI flow. These assumptions include signal generation time, equipment starting times, and delivery time due to system piping. In the early stages of a LOCA with a loss of offsite power, the SITs provide the sole source of makeup water to the PCS.
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| (The assumption of a loss of offsite power is required by regulations.)
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| This is because the LPSI pumps and HPSI pumps cannot deliver flow until the Diesel Generators (DGs) start, come to rated speed, and go through their timed loading sequence. In cold leg breaks, the entire contents of one SIT are assumed to be lost through the break during the blowdown and reflood phases.
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| The limiting large break LOCA is a double ended guillotine cold leg break at the discharge of the primary coolant pump. During this event, the SITs discharge to the PCS as soon as PCS pressure decreases to below SIT pressure. No operator action is assumed during the blowdown stage of a large break LOCA.
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| The worst case small break LOCA assumes a time delay before pumped flow reaches the core. For the larger range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated initially by the SITs, with pumped flow then providing continued cooling.
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| As break size decreases, the SITs and HPSI pumps both play a part in terminating the rise in clad temperature. As break size continues to decrease, the role of the SITs continues to decrease until they are not required, and the HPSI pumps become solely responsible for terminating the temperature increase.
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| Palisades Nuclear Plant B 3.5.1-2 Revised 01/29/2020
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| INSERT Bases 3.5.1 SITs B 3.5.1 BASES APPLICABLE This LCO helps to ensure that the following acceptance criteria, SAFETY ANALYSES established by 10 CFR 50.46 for the ECCS, will be met following a (continued) LOCA:
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| : a. Maximum fuel element cladding temperature is 2200°F;
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| : b. Maximum cladding oxidation is 0.17 times the total cladding thickness before oxidation;
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| : c. Maximum hydrogen generation from a zirconium water reaction is 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; and
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| : d. The core is maintained in a coolable geometry.
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| Since the SITs discharge during the blowdown phase of a LOCA, they do not contribute to the long term cooling requirements of 10 CFR 50.46.
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| Since the SITs are passive components, single active failures are not applicable to their operation. The SIT isolation valves, however, are not single failure proof; therefore, whenever the valves are open, power is removed from their operators and the switch is key locked open.
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| These precautions ensure that the SITs are available during an accident. With power supplied to the valves, a single active failure could result in a valve closure, which would render one SIT unavailable for injection. If the contents of a second SIT is lost through the break, only the contents of two SITs would reach the core. Since the only active failure that could affect the SITs would be the closure of a motor operated outlet valve, the requirement to remove power from these eliminates this failure mode.
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| The minimum volume requirement for the SITs ensures that three SITs can provide adequate inventory to reflood the core and downcomer following a LOCA. The downcomer then remains flooded until the HPSI and LPSI systems start to deliver flow.
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| The maximum volume limit is based on maintaining an adequate gas volume to ensure proper injection and the ability of the SITs to fully discharge, as well as limiting the maximum boron inventory in the SITs.
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| Palisades Nuclear Plant B 3.5.1-3 Revised 01/29/2020
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| INSERT Bases 3.5.1 SITs B 3.5.1 BASES APPLICABLE The minimum SIT volume of 1040 ft3 and the maximum SIT volume of SAFETY ANALYSES 1176 ft3 correspond to a level of 174 inches and 200 inches, (continued) respectively. Each SIT is equipped with two float type level switches which activate control room alarms on high and low level. To allow for instrument inaccuracy, the low SIT level switch alarm is set at 176 inches and the high SIT alarm is set at 198 inches. As a backup to the SIT level switches and to facilitate operator use, level indication is also provided by a differential pressure transmitter which displays in percent tank level. The narrow indicating range of the differential pressure transmitter contains high and low alarms. The high level alarm trips at a slightly lower level than the high level switch and the low level alarm trips at a slightly higher level than the low level switch to alert the operator they are approaching the technical specification values.
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| The minimum nitrogen cover pressure requirement ensures that the contained gas volume will generate discharge flow rates during injection that are consistent with those assumed in the safety analyses.
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| A minimum pressure of 200 psig is used in the analyses. Each of the four SITs is equipped with two pressure switches and one pressure transmitter. The pressure switches activate separate control room alarms. One pressure switch provides a high pressure alarm and the other provides a low pressure alarm. The pressure transmitter provides a display of tank pressure and a common high/low pressure alarm. The low pressure alarms from the pressure switch and pressure transmitter are set sufficiently above the 200 psig value used in the safety analysis to provide margin for instrument inaccuracies. The high pressure alarms from the pressure switch and pressure transmitter are set well below the 250 psig tank design pressure and sufficiently above the normal operating pressure to avoid nuisance alarms.
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| The 1720 ppm limit for minimum boron concentration was established to ensure that, following a LOCA with a minimum level in the SITs, the reactor will remain subcritical in the cold condition following mixing of the SITs, Safety Injection Refueling Water Tank and PCS water volumes. Small break LOCAs assume that all full-length control rods are inserted, except for the control rod of highest worth, which is withdrawn from the core. Large break LOCA analyses assume that all full-length control rods remain withdrawn until the blowdown phase is over. For large break LOCAs, the initial reactor shutdown is accomplished by void formation. The most limiting case occurs at beginning of core life.
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| Palisades Nuclear Plant B 3.5.1-4 Revised 01/29/2020
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| INSERT Bases 3.5.1 SITs B 3.5.1 BASES APPLICABLE The maximum boron limit of 2500 ppm in the SITs is based on boron SAFETY ANALYSES precipitation in the core following a LOCA. With the reactor vessel at (continued) saturated conditions, the core dissipates heat by boiling. Because of this boiling phenomenon in the core, the boric acid concentration will increase in this region. If allowed to proceed in this manner, a point will be reached where boron precipitation will occur in the core. Post LOCA emergency procedures direct the operator to establish simultaneous hot and cold leg injection to prevent this condition by establishing a forced flow path through the core regardless of break location. These procedures are based on the minimum time in which precipitation could occur, assuming that maximum boron concentrations exist in the borated water sources used for injection following a LOCA. Boron concentrations in the SITs in excess of the limit could result in precipitation earlier than assumed in the analysis.
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| The SITs satisfy Criterion 3 of 10 CFR 50.36(c)(2).
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| LCO The LCO establishes the minimum conditions required to ensure that the SITs are available to accomplish their core cooling safety function following a LOCA. Four SITs are required to be OPERABLE to ensure that 100% of the contents of three of the SITs will reach the core during a LOCA.
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| This is consistent with the assumption that the contents of one tank spill through the break. If the contents of fewer than three tanks are injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 could be violated.
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| For an SIT to be considered OPERABLE, the isolation valve must be fully open, with power to the valve operator removed, and the limits established in the SR for contained volume, boron concentration, and nitrogen cover pressure must be met.
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| APPLICABILITY In MODES 1 and 2 the SIT OPERABILITY requirements are based on an assumption of full power operation. Although cooling requirements decrease as power decreases, the SITs are required to be OPERABLE during the MODES when the reactor is critical.
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| In MODE 3 and below, the rate of PCS blowdown is such that the ECCS pumps can provide adequate injection to ensure that peak clad temperature remains below the 10 CFR 50.46 limit of 2200°F.
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| Palisades Nuclear Plant B 3.5.1-5 Revised 01/29/2020
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| INSERT Bases 3.5.1 SITs B 3.5.1 BASES APPLICABILITY In MODES 3, 4, 5, and 6, the SIT motor operated isolation valves may (continued) be closed to isolate the SITs from the PCS. This allows PCS cooldown and depressurization without discharging the SITs into the PCS or requiring depressurization of the SITs.
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| ACTIONS A.1 If the boron concentration of one SIT is not within limits, it must be returned to within the limits within 72 hours. In this condition, the ability to maintain subcriticality or minimum boron precipitation time may be reduced, but the reduced concentration effects on core subcriticality during reflood are minor. Boiling of the ECCS water in the core during reflood concentrates the boron in the saturated liquid that remains in the core. In addition, the volume of the SIT is still available for injection.
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| Since the boron requirements are based on the average boron concentration of the total volume of three SITs, the consequences are less severe than they would be if an SIT were not available for injection.
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| Thus, 72 hours is allowed to return the boron concentration to within limits.
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| The combination of redundant level and pressure instrumentation for any single SIT provides sufficient information so that it is not worthwhile to always attempt to correct drift associated with one instrument, with the resulting radiation exposures during entry into containment, as there is sufficient time to repair one in the event that a second one became inoperable. Because these instruments do not initiate a safety action, it is reasonable to extend the allowable outage time for them. While technically inoperable, the SIT will be available to fulfill its safety function during this time, and, thus, this Completion Time results in a negligible increase in risk.
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| B.1 If one SIT is inoperable, for reasons other than boron concentration or the inability to verify level or pressure, the SIT must be returned to OPERABLE status within 24 hours. In this Condition, the required contents of three SITs cannot be assumed to reach the core during a LOCA as is assumed in the safety analysis.
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| CE-NPSD-994 (Ref. 3) provides a series of deterministic and probabilistic findings that support the 24 hour Completion Time as having no affect on risk as compared to shorter periods for restoring the SIT to OPERABLE status.
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| Palisades Nuclear Plant B 3.5.1-6 Revised 01/29/2020
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| INSERT Bases 3.5.1 SITs B 3.5.1 BASES ACTIONS C.1 (continued)
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| If the SIT cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power condition in an orderly manner and without challenging plant systems.
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| D.1 If more than one SIT is inoperable, the plant is in a condition outside the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.
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| SURVEILLANCE SR 3.5.1.1 REQUIREMENTS Verification that each SIT isolation valve is fully open, as indicated in the control room, ensures that SITs are available for injection and ensures timely discovery if a valve should be partially closed. If an isolation valve is not fully open, the rate of injection to the PCS would be reduced. Although a motor operated valve should not change position with power removed, a closed valve could result in not meeting accident analysis assumptions. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.5.1.2 and SR 3.5.1.3 SIT borated water volume and nitrogen cover pressure should be verified to be within specified limits in order to ensure adequate injection during a LOCA. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.5.1-7 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.5.1 SITs B 3.5.1 BASES SURVEILLANCE SR 3.5.1.4 REQUIREMENTS (continued) The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.5.1.5 Verification that power is removed from each SIT isolation valve operator ensures that an active failure could not result in the undetected closure of an SIT motor operated isolation valve. If this were to occur, only two SITs would be available for injection, given a single failure coincident with a LOCA. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Section 14.17
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| : 2. FSAR, Chapter 6.1
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| : 3. CE-NPSD-994, "CEOG Joint Applications Report for Safety Injection Tank AOT/STI Extension," May 1995 Palisades Nuclear Plant B 3.5.1-8 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.5.2 ECCS - Operating B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
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| B 3.5.2 ECCS - Operating BASES BACKGROUND The function of the ECCS is to provide core cooling and negative reactivity to ensure that the reactor core is protected after any of the following accidents:
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| : a. Loss of Coolant Accident (LOCA);
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| : b. Control Rod Ejection accident;
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| : c. Loss of secondary coolant accident, including a Main Steam Line Break (MSLB) or Loss of Normal Feedwater; and
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| : d. Steam Generator Tube Rupture (SGTR).
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| The addition of negative reactivity is designed primarily for the loss of secondary coolant accident where primary cooldown could add enough positive reactivity to achieve criticality and return to significant power.
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| There are two phases of ECCS operation: injection and recirculation.
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| In the injection phase, all injection is initially added to the Primary Coolant System (PCS) via the cold legs. After the Safety Injection Refueling Water Tank (SIRWT) has been depleted, the recirculation phase is entered as the ECCS suction is automatically transferred to the containment sump.
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| Two suitably redundant, 100% capacity trains are provided. Each train consists of a High Pressure Safety Injection (HPSI) and Low Pressure Safety Injection (LPSI) subsystem. In MODES 1 and 2, and in MODE 3 with PCS temperature 325°F, both trains must be OPERABLE. This ensures that 100% of the core cooling requirements can be provided in the event of a single active failure.
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| Palisades Nuclear Plant B 3.5.2-1 Revised 01/29/2020
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| INSERT Bases 3.5.2 ECCS - Operating B 3.5.2 BASES BACKGROUND Each train of a Safety Injection Signal (SIS) actuates LPSI flow by (continued) starting one LPSI pump and opening two LPSI loop injection valves.
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| Each train of an SIS actuates HPSI flow by starting one HPSI pump, opening the four associated HPSI loop injection valves, and closing the pressure control valves associated with each Safety Injection Tank. In addition, each train of a SIS will provide a confirmatory open signal to the normally open Component Cooling Water valves which supply seal and bearing cooling to the LPSI, HPSI, and Containment Spray pumps.
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| The safety analyses assume that one only train of safety injection is available to mitigate an accident. While operating under the provisions of an ACTION, an additional single failure need not be assumed in assuring that a loss of function has not occurred. Therefore, the LPSI flow assumed in the safety analyses can be met if there is an OPERABLE LPSI flow path from the SIRWT to any two PCS loops.
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| The HPSI flow assumed in the safety analyses can be met if there is an OPERABLE HPSI flow path from the SIRWT to each cold leg. In each case, an OPERABLE flow path must include an OPERABLE pump and an OPERABLE injection valve.
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| A suction header supplies water from the SIRWT or the containment sump to the ECCS pumps. Separate piping supplies each train. The discharge headers from each HPSI pump divide into four supply lines after entering the containment, one feeding each PCS cold leg. The discharge headers from each LPSI pump combine to supply a common header which divides into four supply lines after entering containment, one feeding each PCS cold leg.
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| The hot-leg injection piping connects the HPSI Train 1 header and the HPSI Train 2 header to the PCS hot-leg. For long term core cooling after a large LOCA, Hot-leg injection is used to assure that for a large cold-leg PCS break, net core flushing flow can be maintained and excessive boric acid concentration in the core which could result in eventual precipitation and core flow blockage will be prevented. Within a few hours after a LOCA, if shutdown cooling is not in operation, the operator initiates simultaneous hot-leg and cold-leg injection. Hot-leg injection motor-operated valve throttle position and installed flow orifices cause HPSI flows to be split approximately equally between hot- and cold-leg injection paths.
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| Palisades Nuclear Plant B 3.5.2-2 Revised 01/29/2020
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| INSERT Bases 3.5.2 ECCS - Operating B 3.5.2 BASES BACKGROUND Motor operated valves are set to maximize the LPSI flow to the PCS.
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| (continued) This flow balance directs sufficient flow to the core to meet the analysis assumptions following a LOCA in one of the PCS cold legs.
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| For LOCAs coincident with a loss of off-site power that are too small to initially depressurize the PCS below the shutoff head of the HPSI pumps, the core cooling function is provided by the Steam Generators (SGs) until the PCS pressure decreases below the HPSI pump shutoff head.
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| During low temperature conditions in the PCS, limitations are placed on the maximum number of HPSI pumps that may be OPERABLE. Refer to the Bases for LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," for the basis of these requirements.
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| During a large break LOCA, PCS pressure could decrease to
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| < 200 psia in < 20 seconds. The ECCS systems are actuated upon receipt of an SIS. If offsite power is available, the safeguard loads start immediately. If offsite power is not available, all loads will be shed at the time the diesel generators receive an automatic start signal. With load shedding completed, the diesel generator breakers will close automatically when generator voltage approaches a normal operating value. Closing of the breakers will reset the load shedding signals and start the sequencer. The sequencers will initiate operation of the engineered safeguard equipment required for the accident. The time delay associated with diesel starting, sequenced loading, and pump starting determines the time before pumped flow is available to the core following a LOCA.
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| The active ECCS components, along with the passive Safety Injection Tanks (SITs) and the Safety Injection Refueling Water Tank (SIRWT),
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| covered in LCO 3.5.1, "Safety Injection Tanks (SITs)," and LCO 3.5.4, "Safety Injection Refueling Water Tank (SIRWT)," provide the cooling water necessary to meet the Palisades Nuclear Plant design criteria (Ref. 1).
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| Palisades Nuclear Plant B 3.5.2-3 Revised 01/29/2020
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| INSERT Bases 3.5.2 ECCS - Operating B 3.5.2 BASES APPLICABLE The LCO helps to ensure that the following acceptance criteria, SAFETY ANALYSES established by 10 CFR 50.46 for ECCSs, will be met following a LOCA:
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| : a. Maximum fuel element cladding temperature is 2200°F;
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| : b. Maximum cladding oxidation is 0.17 times the total cladding thickness before oxidation;
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| : c. Maximum hydrogen generation from a zirconium water reaction is 0.01 times the hypothetical amount generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react;
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| : d. Core is maintained in a coolable geometry; and
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| : e. Adequate long term core cooling capability is maintained.
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| The LCO also limits the potential for a post trip return to power following an MSLB event.
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| Both a HPSI and a LPSI subsystem are assumed to be OPERABLE in the large break LOCA analysis at full power (Ref. 2). This analysis establishes a minimum required runout flow for the HPSI and LPSI pumps, as well as the maximum required response time for their actuation. The HPSI pump is also credited in the small break LOCA analysis. This analysis establishes the flow and discharge head requirements at the design point for the HPSI pump. The SGTR and MSLB accident analyses also credit the HPSI pumps, but are not limiting in their design.
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| The large break LOCA event with a loss of offsite power and a single failure (disabling one ECCS train) establishes the OPERABILITY requirements for the ECCS. During the blowdown stage of a LOCA, the PCS depressurizes as primary coolant is ejected through the break into the containment. The nuclear reaction is terminated either by moderator voiding (during large breaks) or control rod insertion (during small breaks).
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| Following depressurization, emergency cooling water is injected into the cold legs, flows into the downcomer, fills the lower plenum, and refloods the core.
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| On smaller breaks, PCS pressure will stabilize at a value dependent upon break size, heat load, and injection flow. The smaller the break, the higher this equilibrium pressure. In all LOCA analyses, injection flow is not credited until PCS pressure drops below the shutoff head of the HPSI pumps.
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| Palisades Nuclear Plant B 3.5.2-4 Revised 01/29/2020
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| INSERT Bases 3.5.2 ECCS - Operating B 3.5.2 BASES APPLICABLE The LCO ensures that an ECCS train will deliver sufficient water to SAFETY ANALYSES match decay heat boiloff rates soon enough to minimize core damage (continued) for a large LOCA. It also ensures that the HPSI pump will deliver sufficient water during a small break LOCA and provide sufficient boron to limit the return to power following an MSLB event. For smaller LOCAs, PCS inventory decreases until the PCS can be depressurized below the HPSI pumps' shutoff head. During this period of a small break LOCA, the SGs continue to serve as the heat sink providing core cooling.
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| ECCS - Operating satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| LCO In MODES 1 and 2, and in MODE 3 with PCS temperature 325°F, two independent (and redundant) ECCS trains are required to ensure that sufficient ECCS flow is available, assuming there is a single failure affecting either train. Additionally, individual components within the ECCS trains may be called upon to mitigate the consequences of other transients and accidents.
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| An ECCS train consists of an HPSI subsystem and a LPSI subsystem.
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| In addition, each train includes the piping, instruments, and controls to ensure the availability of an OPERABLE flow path capable of taking suction from the SIRWT on an SIS and automatically transferring suction to the containment sump upon a Recirculation Actuation Signal (RAS).
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| During an event requiring ECCS actuation, a flow path is provided to ensure an abundant supply of water from the SIRWT to the PCS, via the HPSI and LPSI pumps and their respective supply headers, to each of the four cold leg injection nozzles is available. During the recirculation phase, a flow path is provided from the containment sump to the PCS via the HPSI pumps. For worst case conditions, the containment building water level alone is not sufficient to assure adequate Net Positive Suction Head (NPSH) for the HPSI pumps.
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| Therefore, to obtain adequate NPSH, a portion of the Containment Spray (CS) pump discharge flow is diverted from downstream of the shutdown cooling heat exchangers to the suction of the HPSI pumps at recirculation during a large break LOCA. In this configuration, the CS pumps and shutdown cooling heat exchangers provide a support function for HPSI flow path OPERABILITY. The OPERABILITY requirements for the CS pumps and shutdown cooling heat exchangers are addressed in LCO 3.6.6, Containment Cooling Systems. Support system OPERABILITY is addressed by LCO 3.0.6.
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| The flow path for each train must maintain its designed independence to ensure that no single active failure can disable both ECCS trains.
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| Palisades Nuclear Plant B 3.5.2-5 Revised 01/29/2020
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| INSERT Bases 3.5.2 ECCS - Operating B 3.5.2 BASES APPLICABILITY In MODES 1 and 2, and in MODE 3 with PCS temperature 325°F, the ECCS OPERABILITY requirements for the limiting Design Basis Accident (DBA) large break LOCA are based on full power operation.
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| Although reduced power would not require the same level of performance, the accident analysis does not provide for reduced cooling requirements in the lower MODES. The HPSI pump performance is based on the small break LOCA, which establishes the pump performance curve and has less dependence on power. The requirements of MODE 2 and MODE 3 with PCS temperature 325°F, are bounded by the MODE 1 analysis.
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| The ECCS functional requirements of MODE 3, with PCS temperature
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| < 325°F, and MODE 4 are described in LCO 3.5.3, "ECCS - Shutdown."
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| In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "PCS Loops -
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| MODE 5, Loops Filled," and LCO 3.4.8, "PCS Loops - MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.4, "Shutdown Cooling (SDC) and Coolant Circulation - High Water Level," and LCO 3.9.5, "Shutdown Cooling (SDC) and Coolant Circulation - Low Water Level."
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| ACTIONS A.1 Condition A is applicable whenever one LPSI subsystem is inoperable.
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| With one LPSI subsystem inoperable, action must be taken to restore OPERABLE status within 7 days. In this condition, the remaining OPERABLE ECCS train is adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure to the remaining LPSI subsystem could result in loss of ECCS function. The 7 day Completion Time is reasonable to perform corrective maintenance on the inoperable LPSI subsystem. While mechanical system LCOs typically provide a 72 hour Completion Time, this 7 day Completion Time is based on the findings of the deterministic and probabilistic analysis in Reference 5. Reference 5 concluded that extending the Completion Time to 7 days for an inoperable LPSI subsystem provides plant operational flexibility while simultaneously reducing overall plant risk. This is because the risks incurred by having the LPSI subsystem unavailable for a longer time at power will be substantially offset by the benefits associated with avoiding unnecessary plant transitions and by reducing risk during plant shutdown operations.
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| Palisades Nuclear Plant B 3.5.2-6 Revised 01/29/2020
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| INSERT Bases 3.5.2 ECCS - Operating B 3.5.2 BASES ACTIONS B.1 (continued)
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| Condition B is applicable whenever one or more ECCS trains is inoperable for reasons other than one inoperable LPSI subsystem.
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| Action B.1 requires restoration of both ECCS trains, (HPSI and LPSI) to OPERABLE status within 72 hours. The 72 hour Completion Time is based on an NRC study (Ref. 3), assuming that at least 100% of the required ECCS flow (that assumed in the safety analyses) is available.
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| If less than 100% of the required ECCS flow is available, Condition D must also be entered.
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| Mechanical system LCOs typically provide a 72 hour Completion Time under conditions when a required system can perform its required safety function, but may not be able to do so assuming an additional failure. When operating in accordance with the Required Actions of an LCO Condition, it is not necessary to be able to cope with an additional single failure.
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| The ECCS can provide one hundred percent of the required ECCS flow following the occurrence of any single active failure. Therefore, the ECCS function can be met during conditions when those components which could be deactivated by a single active failure are known to be inoperable. Under that condition, however, the ability to provide the function after the occurrence of an additional failure cannot be guaranteed. Therefore, continued operation with one or more trains inoperable is allowed only for a limited time.
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| C.1 and C.2 Condition C is applicable when the Required Actions of Condition A or B cannot be completed within the required Completion Time. Either Condition A or B is applicable whenever one or more ECCS trains is inoperable. Therefore, when Condition C is applicable, either Condition A or B is also applicable. Being in Conditions A or B, and Condition C concurrently maintains both Completion Time clocks for instances where equipment repair allows exit from Condition C while the plant is still within the applicable conditions of the LCO.
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| If the inoperable ECCS trains cannot be restored to OPERABLE status within the required Completion Times of Condition A and B, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and PCS temperature reduce to < 325°F within 24 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power in an orderly manner and without challenging plant systems.
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| Palisades Nuclear Plant B 3.5.2-7 Revised 01/29/2020
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| INSERT Bases 3.5.2 ECCS - Operating B 3.5.2 BASES ACTIONS D.1 (continued)
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| Condition D is applicable with one or more trains inoperable when there is less than 100% of the required ECCS flow available. Either Condition A or B is applicable whenever one or more ECCS trains is inoperable.
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| Therefore, when this Condition is applicable, either Condition A or B is also applicable. Being in Conditions A or B, and Condition D concurrently maintains both Completion Time clocks for instances where equipment repair allows exit from Condition D (and LCO 3.0.3) while the plant is still within the applicable conditions of the LCO.
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| One hundred percent of the required ECCS flow can be provided by one OPERABLE HPSI subsystem and one OPERABLE LPSI subsystem. The required LPSI flow (that assumed in the safety analyses) is available if there is an OPERABLE LPSI flow path from the SIRWT to any two PCS loops. Shutdown cooling flow control valve, CV-3006 must be full open. The required HPSI flow (that assumed in the safety analyses) is available if there is an OPERABLE HPSI flow path from the SIRWT to each PCS loop (having less than all four PCS loop flowpaths may be acceptable if verified against current safety analyses). A Containment Spray Pump and a sub-cooled suction valve must be available to support each OPERABLE HPSI pump. In each case, an OPERABLE flow path must include an OPERABLE pump and OPERABLE loop injection valves.
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| Reference 4 describes situations in which one component, such as the shutdown cooling flow control valve, CV-3006, can disable both ECCS trains. With one or more components inoperable, such that 100% of the required ECCS flow (that assumed in the safety analyses) is not available, the facility is in a condition outside the accident safety analyses.
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| With less than 100% of the required ECCS flow available, the plant is in a condition outside the assumptions of the safety analyses. Therefore, LCO 3.0.3 must be entered immediately.
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| Palisades Nuclear Plant B 3.5.2-8 Revised 01/29/2020
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| INSERT Bases 3.5.2 ECCS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the PCS is maintained. CV-3027 and CV-3056 are stop valves in the minimum recirculation flow path for the ECCS pumps.
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| If either of these valves were closed when the PCS pressure was above the shutoff head of the ECCS pumps, the pumps could be damaged by running with insufficient flow and thus render both ECCS trains inoperable.
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| Placing HS-3027A and HS-3027B for CV-3027, and HS-3056A and HS-3056B for CV-3056, in the open position ensures that the valves cannot be inadvertently misaligned or change position as the result of an active failure. These valves are of the type described in Reference 4, which can disable the function of both ECCS trains and invalidate the accident analysis. CV-3027 and CV-3056 are capable of being closed from the control room since the SIRWT must be isolated from the containment during the recirculation phase of a LOCA. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve automatically repositions within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.5.2-9 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.5.2 ECCS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.3 REQUIREMENTS (Continued) SR 3.5.2.3 verifies CV-3006 is in the open position and that its air supply is isolated. CV-3006 is the shutdown cooling flow control valve located in the common LPSI flow path. The valve must be verified in the full open position to support the low pressure injection flow assumptions used in the accident analyses. The inadvertent misposition of this valve could result in a loss of low pressure injection flow and thus invalidate these flow assumptions. CV-3006 is designed to be held open by spring force and closed by air pressure. To ensure the valve cannot be inadvertently misaligned or change position as the result of a hot short in the control circuit, the air supply to CV-3006 is isolated. Isolation of the air supply to CV-3006 is acceptable since the valve does not require automatic repositioning during an accident.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| SR 3.5.2.4 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller damage or other hydraulic component problems is required by the ASME Code. This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant safety analysis. SRs are specified in the INSERVICE TESTING PROGRAM of the ASME Code. The ASME Code provides the activities and Frequencies necessary to satisfy the requirements.
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| SR 3.5.2.5, SR 3.5.2.6, and SR 3.5.2.7 These SRs demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated actuation signal, i.e., on an SIS or RAS, that each ECCS pump starts on receipt of an actual or simulated actuation signal, i.e., on an SIS, and that the LPSI pumps stop on receipt of an actual or simulated actuation signal, i.e., on an RAS. RAS opens the HPSI subcooling valve CV-3071, if the associated HPSI pump is operating. After the containment sump valve CV-3030 opens from RAS, HPSI subcooling valve CV-3070 will open, if the associated HPSI pump is operating. RAS will re-position CV-3001 and CV-3002 to a predetermined throttled position. RAS will close Palisades Nuclear Plant B 3.5.2-10 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.5.2 ECCS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.5, SR 3.5.2.6, and SR 3.5.2.7 REQUIREMENTS (continued) containment spray valve CV-3001, if containment sump valve CV-3030 does not open. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The actuation logic is tested as part of the Engineered Safety Feature (ESF) testing, and equipment performance is monitored as part of the INSERVICE TESTING PROGRAM.
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| SR 3.5.2.8 The HPSI Hot Leg Injection motor operated valves and the LPSI loop injection valves have position switches which are set at other than the full open position. This surveillance verifies that these position switches are set properly.
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| The HPSI Hot leg injection valves are manually opened during the post-LOCA long term cooling phase to admit HPSI injection flow to the PCS hot leg. The open position limit switch on each HPSI hot leg isolation valves is set to establish a predetermined flow split between the HPSI injection entering the PCS hot leg and cold legs.
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| The LPSI loop injection MOVs open automatically on a SIS signal. The open position limit switch on each LPSI loop injection valve is set to establish the maximum possible flow through that valve. The design of these valves is such that excessive turbulence is developed in the valve body when the valve disk is at the full open position. Stopping the valve travel at slightly less than full open reduces the turbulence and results in increased flow. Verifying that the position stops are properly set ensures that a single low pressure safety injection subsystem is capable of delivering the flow rate required in the safety analysis.
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Palisades Nuclear Plant B 3.5.2-11 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.5.2 ECCS - Operating B 3.5.2 BASES SR 3.5.2.9 Periodic inspection of the ECCS containment sump passive strainer assemblies ensures that the post-LOCA recirculation flowpath to the ECCS train containment sump suction inlets is unrestricted. Periodic inspection of the containment sump entrance pathways, which include containment sump passive strainer assemblies, containment sump downcomer debris screens, containment floor drain debris screens, containment sump vent debris screens, and reactor cavity corium plug bottom cup support assemblies, ensures that the containment sump stays in proper operating condition. The migration of LOCA-generated debris larger than the strainer perforation diameter through the two one-inch reactor cavity drain line corium plugs is not considered to be credible. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| REFERENCES 1. FSAR, Section 5.1
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| : 2. FSAR, Section 14.17
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| : 3. NRC Memorandum to V. Stello, Jr., from R. L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975
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| : 4. IE Information Notice No. 87-01, January 6, 1987
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| : 5. CE-NPSD-994, "CEOG Joint Applications Report for Safety Injection Tank AOT/STI Extension," May 1995 Palisades Nuclear Plant B 3.5.2-12 Amendment No. 271 Revised 01/29/2020
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| INSERT Bases 3.5.3 ECCS - Shutdown B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
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| B 3.5.3 ECCS - Shutdown BASES BACKGROUND The Background section for Bases B 3.5.2, "ECCS - Operating," is applicable to these Bases, with the following modifications.
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| In MODE 3 with Primary Coolant System (PCS) temperature < 325°F and in MODE 4, an ECCS train is defined as one Low Pressure Safety Injection (LPSI) train. The LPSI flow path consists of piping, valves, and pumps that enable water from the Safety Injection Refueling Water Tank (SIRWT), and subsequently the containment sump, to be injected into the PCS following a Loss of Coolant Accident (LOCA).
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| APPLICABLE In Mode 3 with PCS temperature < 325°F and in Mode 4 the normal SAFETY ANALYSES compliment of ECCS components is reduced from that which is available during operations above Mode 3 with PCS temperature 325°F. The acceptability for the reduced ECCS operational requirements is based on engineering judgement rather than specific analysis and considers such factors as the reduced probability that a LOCA will occur, and the reduced energy stored in the fuel. The reduction in ECCS operational requirements include:
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| : 1) Isolation of the Safety Injection Tanks (SITs) since PCS pressure is expected to be reduced below the SIT injection pressure,
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| : 2) Reliance on manual safety injection initiation since the automatic Safety Injection Signal (SIS) is not required by the technical specifications below 300°F,
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| : 3) Rendering the High Pressure Safety Injection (HPSI) pumps incapable of injecting into the PCS. The HPSI pumps are rendered incapable of injecting into the PCS in accordance with the requirements of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System". This action assures that a single mass addition event initiated at a pressure within the limits of LCO 3.4.12 cannot cause the PCS pressure to exceed the 10 CFR 50 Appendix G limit.
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| Palisades Nuclear Plant B 3.5.3-1 Revised 07/22/2002
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| INSERT Bases 3.5.3 ECCS - Shutdown B 3.5.3 BASES APPLICABLE At a PCS temperature of 325°F the maximum allowed PCS pressure SAFETY ANALYSES corresponds to the LTOP setpoint limit which is approximately 800 psia.
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| (continued) Below 800 psia postulated piping flaws of critical size are considered unlikely since normal operation at 2060 psia serves as a proof test against ruptures. In addition, since the reactor has been shutdown for a period of time, the decay heat and sensible heat levels are greatly reduced from the full power case.
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| Although a pipe break in the PCS pressure boundary is considered unlikely, break sizes larger and smaller than approximately 0.1 ft2 are considered separately when analyzing ECCS response.
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| For breaks larger than approximately 0.1 ft2, the event is characterized by a very rapid depressurization of the PCS to near the containment pressure. Due to the reduced temperature and pressure of the PCS, the time to complete blowdown is extended from that assumed in the full power case. During this time, the fuel is cooled by the flow through the core towards the break. Automatic safety injection actuation is not assumed to occur since the pressurizer pressure SIS may be bypassed below 1700 psig. Therefore, operator action is relied upon to initiate ECCS flow. Indication that would alert the operator that a LOCA had occurred include; a loss of pressurizer level, rapid decrease in PCS pressure, increase in containment pressure, and containment high radiation alarm. Since the saturation pressure for 325°F is approximately 100 psia, the LPSI pumps are capable of providing the required heat removal function. When the OPERABLE LPSI pump is being used to fulfill the shutdown cooling function, the PCS pressure is
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| < 300 psia. As such, the rate of PCS blowdown is reduced providing some time to manually realign the OPERABLE LPSI pump to the ECCS mode of operation.
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| For breaks smaller than approximately 0.1 ft2, the event is characterized by a slow depressurization of the PCS and a relatively long time for the PCS level to drop below the tops of the hot legs. In MODE 3 with PCS temperature < 325°F and in the upper range of MODE 4 before shutdown cooling is established, the spectrum of smaller break sizes are more limiting than larger breaks in terms of ECCS performance since the PCS could stay above the shutoff head of the LPSI pumps.
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| For these break sizes, sufficient time, well in excess of the recommended 10 minutes attributed for manual operator action, is available to either initiate once through cooling using the PORVs, or by re-establishing HPSI pump injection capability. In either case, the core remains covered and the criteria of 10 CFR 50.46 preserved.
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| ECCS - Shutdown satisfies Criterion 3 of 10 CFR 50.36(c)(2).
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| Palisades Nuclear Plant B 3.5.3-2 Revised 07/22/2002
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| INSERT Bases 3.5.3 ECCS - Shutdown B 3.5.3 BASES LCO In MODE 3 with PCS temperature < 325°F and in MODE 4, an ECCS train is comprised of a single LPSI train. Each LPSI train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the SIRWT and transferring suction to the containment sump.
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| During an event requiring ECCS actuation, a flow path is required to supply water from the SIRWT to the PCS via one LPSI pump and at least one supply header to a cold leg injection nozzle. In the long term, this flow path may be switched to take its supply from the containment sump.
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| With PCS temperature < 325°F, one LPSI pump is acceptable without single failure consideration, based on the stable reactivity condition of the reactor and the limited core cooling requirements. The High Pressure Safety Injection (HPSI) pumps may therefore be released from the ECCS train requirements. With PCS temperature < 300°F, both HPSI pumps must be rendered incapable of injection into the PCS in accordance with LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System."
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| The LCO is further modified by a Note that allows a LPSI train to be considered OPERABLE during alignment and operation for shutdown cooling, if capable of being manually realigned (remote or local) to the ECCS mode of operation and not otherwise inoperable. This allows operation of a LPSI pump in the shutdown cooling mode.
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| APPLICABILITY In MODES 1 and 2, and in MODE 3 with PCS temperature 325°F, the OPERABILITY requirements for ECCS are covered by LCO 3.5.2.
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| In MODE 3 with PCS temperature < 325°F and in MODE 4, one OPERABLE ECCS train is acceptable without single failure consideration, based on the stable reactivity condition of the reactor and the limited core cooling requirements.
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| Palisades Nuclear Plant B 3.5.3-3 Revised 07/22/2002
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| INSERT Bases 3.5.3 ECCS - Shutdown B 3.5.3 BASES APPLICABILITY In MODES 5 and 6, plant conditions are such that the probability of an (continued) event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "PCS Loops -
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| MODE 5, Loops Filled," and LCO 3.4.8, "PCS Loops - MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.4, "Shutdown Cooling (SDC) and Coolant Circulation - High Water Level," and LCO 3.9.5, "Shutdown Cooling (SDC) and Coolant Circulation - Low Water Level."
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| ACTIONS A.1 With no LPSI train OPERABLE, the plant is not prepared to respond to a loss of coolant accident. Action must be initiated Immediately to restore at least one LPSI train to OPERABLE status. The Immediate Completion Time reflects the importance of maintaining an OPERABLE LPSI train and ensures that prompt action is taken to restore the required cooling capacity.
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| SURVEILLANCE SR 3.5.3.1 REQUIREMENTS The applicable Surveillance descriptions from Bases 3.5.2 apply.
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| REFERENCES The applicable references from Bases 3.5.2 apply.
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| Palisades Nuclear Plant B 3.5.3-4 Revised 07/22/2002
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| INSERT Bases 3.5.4 SIRWT B 3.5.4 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
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| B 3.5.4 Safety Injection Refueling Water Tank (SIRWT)
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| BASES BACKGROUND The SIRWT supports the ECCS and the Containment Spray System by providing a source of borated water for Engineered Safety Feature (ESF) pump operation.
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| The SIRWT supplies two ECCS trains by separate, redundant supply headers. Each header also supplies one train of the Containment Spray System. An air operated isolation valve is provided in each header which isolates the SIRWT from the ECCS after the ESF pump suction has been transferred to the containment sump following depletion of the SIRWT during a Loss of Coolant Accident (LOCA).
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| A separate header is used to supply the Chemical and Volume Control System (CVCS) from the SIRWT. Use of a single SIRWT to supply both trains of the ECCS and Containment Spray System is acceptable since the SIRWT is a passive component, and passive failures are not assumed to occur concurrently with any Design Basis Event during the injection phase of an accident. Not all the water stored in the SIRWT is available for injection following a LOCA; the location of the ESF pump suction piping in the SIRWT will result in some portion of the stored volume being unavailable.
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| The High Pressure Safety Injection (HPSI) and Low Pressure Safety Injection (LPSI) pumps are provided with recirculation lines that ensure each pump can maintain minimum flow requirements when operating at shutoff head conditions. These lines discharge back to the SIRWT, which vents to the atmosphere. When the suction for the ESF pumps is transferred to the containment sump, the recirculation path must be isolated to prevent is a release of the containment sump contents to the SIRWT. If not isolated, this flow path could result in a release of contaminants to the atmosphere and the eventual loss of suction head for the ESF pumps.
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| Palisades Nuclear Plant B 3.5.4-1 Revised 01/29/2020
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| INSERT Bases 3.5.4 SIRWT B 3.5.4 BASES BACKGROUND This LCO ensures that:
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| (continued)
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| : a. The SIRWT contains sufficient borated water to support ESF pump operation during}}
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