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| {{#Wiki_filter:Exeton Generation. 200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.90 NMP2L2562 November 19, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Nine Mile Point Nuclear Station, Unit 2 Renewed Facility Operating License No. NPF-69 NRC Docket No. 50-410 | | {{#Wiki_filter:}} |
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| ==SUBJECT:==
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| Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
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| In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (Exelon) is submitting a request for an amendment to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License No. NPF-69 for Nine Mile Point (NMP) Nuclear Station, Unit 2.
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| The proposed amendment would modify NMP, Unit 2 TS by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies."
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| The changes are consistent with NRC-approved Industry Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) change TSTF-425, Revision 3, (ADAMS Accession No. ML090850642). The Federal Register Notice published on July 6, 2009 (74 FR 31996), announced the availability of this TS improvement. provides a description of the proposed change, the requested confirmation of applicability, and plant-specific verifications. Attachment 2 provides documentation of Probabilistic Risk Assessment (PRA) technical adequacy. Attachments 3 provide the existing NMP Unit 2, mark-up TS pages to show the proposed changes. Attachments 4 provide the proposed NMP Unit 2 TS Bases changes. Attachment 5 provides a TSTF-425 versus NMP TS Cross-Reference. Attachment 6 provides the proposed No Significant Hazards Consideration.
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| License Amendment Request Adoption of TSTF-425, Rev. 3 November 19, 2014 Page 2 There are no regulatory commitments contained in this letter.
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| Exelon requests approval of the proposed license amendment by November 19, 2015, with the amendment being implemented within 120 days.
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| These proposed changes have been reviewed by the Plant Operations Review Committee and approved in accordance with Nuclear Safety Review Board procedures.
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| In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"
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| paragraph (b), Exelon is notifying the State of New York of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official If you should have any questions regarding this submittal, please contact Enrique Villar at 610-765-5736.
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| I declare under penalty of perjury that the foregoing is true and correct. Executed on the 1 9th of November 2014.
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| Respectfully, Jam'sBastow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1. Description and Assessment
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| : 2. Documentation of PRA Technical Adequacy
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| : 3. Proposed Technical Specification Page Changes - Unit 2
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| : 4. Proposed Technical Specification Bases Page Changes - Unit 2
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| : 5. TSTF-425 (NUREG-1433) vs. NMP Unit 2 Cross-Reference
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| : 6. Proposed No Significant Hazards Consideration cc: USNRC Region I Regional Administrator w/attachments II USNRC Senior Resident Inspector - NMP USNRC Project Manager, NRR - NMP A. L. Peterson, NYSERDA
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| ATTACHMENT 1 License Amendment Request Nine Mile Point Nuclear Station, Unit 2 Docket No. 50-410 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
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| Description and Assessment
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-410 Page 1 of 6 DESCRIPTION AND ASSESSMENT
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| ==1.0 DESCRIPTION==
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| The proposed amendment would modify the Nine Mile Point (NMP) Nuclear Station, Unit 2 Technical Specifications (TS) by relocating specific surveillance frequencies to a licensee-controlled program with the adoption of Technical Specification Task Force (TSTF) -425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b." Additionally, the change would add a new program, the Surveillance Frequency Control Program, to TS Section 5, Administrative Controls.
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| The changes are consistent with NRC-approved Industry/TSTF Standard Technical Specifications (STS) change TSTF-425, Revision 3, (ADAMS Accession No. ML090850642). The Federal Register Notice (FRN) published on July 6, 2009 (74 FR 31996), announced the availability of this TS improvement.
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| 2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation Exelon Generation Company, LLC (Exelon) has reviewed the Safety Evaluation dated July 6, 2009. This review included a review of the NRC staff's evaluation, TSTF-425, Revision 3, and the requirements specified in Nuclear Energy Institute (NEI) 04-10, Rev. 1 (ADAMS Accession No. ML071360456). includes Exelon's documentation with regard to Probabilistic Risk Assessment (PRA) technical adequacy consistent with the requirements of Regulatory Guide (RG) 1.200, Revision 1 (ADAMS Accession No. ML070240001), Section 4.2, and describes any PRA models without NRC-endorsed standards, including documentation of the quality characteristics of those models in accordance with RG 1.200.
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| Exelon has concluded that the justifications presented in the TSTF proposal and the Safety Evaluation prepared by the NRC staff is applicable to NMP Unit 2, and justify this amendment to incorporate the changes to the NMP TS.
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| 2.2 Optional Changes and Variations The proposed amendment is consistent with the STS changes described in TSTF-425, Revision 3; however, Exelon proposes variations or deviations from TSTF-425, as identified below.
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| : 1. Revised (i.e., clean) TS pages are not included in this amendment request given the number of TS pages affected, the straightforward nature of the proposed changes, and outstanding NMP amendment requests that will impact some of the same TS pages.
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| Providing only mark-ups of the proposed TS changes satisfies the requirements of 10 CFR 50.90 in that the mark-ups fully describe the changes desired. This is an administrative deviation from the NRC staff's model application dated July 6, 2009 (74 FR 31996) with no impact on the NRC staff's model safety evaluation published in the same FRN. As a result of this deviation, the contents and numbering of the attachments for this
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-410 Page 2 of 6 amendment request differ from the attachments specified in the NRC staff's model application. Mark-ups of the proposed TS changes are provided in Attachments 3 for NMP Unit 2. Additionally, mark-ups of the proposed changes to TS Bases pages are provided in Attachment 4 for NMP Unit 2. (NOTE: Some TS Bases pages provided do not contain any mark-ups. These pages are provided for completeness and for information purposes only.)
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| : 2. The definition of STAGGERED TEST BASIS is being retained in NMP Unit 2 TS Definition Section 1.1 since this terminology is mentioned in Administrative TS Section 5.5.13, "Control Room Habitability," which is not the subject of this amendment request and is not proposed to be changed. This is an administrative deviation from TSTF-425 with no impact on the NRC staff's model safety evaluation dated July 6, 2009 (74 FR 31996).
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| : 3. Attachment 5 provides a cross-reference between the NUREG-1433 Surveillances included in TSTF-425 versus the NMP Unit 2 Surveillances included in this amendment request. Attachment 5 includes a summary description of the referenced TSTF-425 Surveillances versus NMP Unit 2 TS Surveillances. This is provided for information purposes only and is not intended to be a verbatim description of the TS Surveillances.
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| This cross-reference highlights the following:
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| : a. NUREG-1433 Surveillances included in TSTF-425 and corresponding NMP Unit 2 Surveillances with identical surveillance numbers,
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| : b. NUREG-1433 Surveillances included in TSTF-425 and corresponding NMP Unit 2 Surveillances with differing surveillances numbers,
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| : c. NUREG-1433 Surveillances included in TSTF-425 that are not contained in the NMP Unit 2 TS, and
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| : d. NMP Unit 2 plant-specific Surveillances that are not contained in NUREG-1433, and therefore, are not included in the TSTF-425 mark-ups.
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| Concerning the above, NMP Unit 2 Surveillances that have surveillance numbers identical to the corresponding TSTF-425 NUREG-1433 Surveillances are not deviations from TSTF-425.
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| NMP Unit 2 surveillances with Surveillance numbers that differ from the corresponding TSTF-425 NUREG-1433 Surveillances are administrative deviations from TSTF-425 with no impact on the NRC staff's model safety evaluation dated July 6, 2009 (74 FR 31996).
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| For TSTF-425 NUREG-1433 Surveillances that are not contained in the NMP Unit 2 TS, the corresponding TSTF-425 NUREG-1433 mark-ups included in TSTF-425 for these Surveillances are not applicable to NMP Unit 2. This is an administrative deviation from TSTF-425 with no impact on the NRC staff's model safety evaluation dated July 6, 2009 (74 FR 31996).
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| For NMP Unit 2 plant-specific Surveillances that are not contained in TSTF-425 NUREG-1433, and therefore, are not included in the NUREG-1433 mark-ups provided in TSTF-425, Exelon has determined that the relocation of the frequencies for these NMP Unit 2 plant-specific Surveillances is consistent with TSTF-425, Revision 3, and with the NRC staff's model safety evaluation dated July 6, 2009 (74 FR 31996), including the scope exclusions identified in Section 1.0, "Introduction," of the model safety evaluation, since the plant-specific Surveillances involve fixed periodic frequencies. In accordance with
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-410 Page 3 of 6 TSTF-425, changes to the frequencies for these Surveillances would be controlled under the Surveillance Frequency Control Program (SFCP). The SFCP provides the necessary administrative controls to require that Surveillances related to testing, calibration and inspection are conducted at a frequency to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Changes to frequencies in the SFCP would be evaluated using the methodology and probabilistic risk guidelines contained in NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. ML071360456),
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| as approved by NRC letter dated September 19, 2007 (ADAMS Accession No. ML072570267). The NEI 04-10, Revision 1 methodology includes qualitative considerations, risk analyses, sensitivity studies and bounding analyses, as necessary, and recommended monitoring of the performance of Systems, Components, and Structures (SSCs) for which frequencies are changed to assure that reduced testing does not adversely impact the SSCs. In addition, the NEI 04-10, Revision 1 methodology satisfies the five key safety principles specified in RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decision-making: Technical Specifications," dated August 1998 (ADAMS Accession No. ML003740176), relative to changes in surveillance frequencies.
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| : 4. As discussed in Reference 1, Exelon proposes a variation from TSTF-425 that replaces text describing the basis for each frequency relocated to the SFCP. This variation is necessary because, independent of whether surveillance frequencies have been changed under the SFCP, the surveillance frequencies are not, in all cases, based on operating experience, equipment reliability or plant risk.
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| As required by proposed TS Section 5.5.16, "Surveillance Frequency Control Program,"
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| subsequent changes to the frequencies listed in the SFCP will be made in accordance with the NRC-endorsed methodology described in NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1. NEI 04-10 provides the methodology to identify, assess, implement, and monitor proposed changes to surveillance frequencies.
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| NEI 04-10 identifies the need to address both quantitative and/or qualitative considerations when changing surveillance frequencies. As discussed in Section 4.0, Step 7, qualitative considerations include vendor-specified maintenance frequency, test intervals specified in applicable industry codes and standards, impact on defense-in-depth protection, and the existence of alternate testing of SSCs affected by the change. These qualitative considerations provide examples of instances where surveillance frequencies changed under the SFCP may not be based upon operating experience, equipment reliability, or plant risk.
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| As a result, Exelon's proposed variation from TSTF-425 provides wording that more accurately reflects the methodology described in NEI 04-10. However, in order to avoid future confusion regarding this issue, EGC will replace the TS Bases text insert proposed in Reference 1 (i.e., "The Frequency may be based on factors such as operating experience, equipment reliability, or plant risk, and is controlled under the Surveillance Frequency Control Program") with a revised insert that reads "The Surveillance Frequency is controlled under the Surveillance Frequency Control Program."
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| : 5. On July 2, 2009, NRC published in the FR, Notice of Availability of Technical Specification Improvement To Relocate Surveillance Frequencies to Licensee Control - Risk-Informed Technical Specification Task Force (RITSTF) Initiative 5b, Technical Specification Task
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-410 Page 4 of 6 Force-425, Revision 3. The FRN stated that each licensee applying for the changes proposed in TSTF-425 must include documentation regarding the PRA technical adequacy consistent with the guidance in Section 4.2 of RG 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment [PRA] Results for Risk-Informed Activities" (ADAMS Accession No. ML070240001).
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| Subsequent to NRC approval of NEI 04-10 and TSTF-425, RG 1.200, Revision 2, became effective in 2010. RG 1.200, Revision 2 endorses a revision to the PRA standard which combined new requirements for PRA models for external hazards, including fire and seismic hazards, with the existing internal events PRA standard. NEI 04-10, Rev. 1, does not require external hazards, including fire and seismic events, to be assessed quantitatively using PRA models, instead providing for qualitative and other quantitative methods to be used. Therefore, licensees adopting TSTF-425 and proposing to use these endorsed methods to assess external hazards would not submit any information on external hazard PRA models, since such models would not be used for this application.
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| At the July 10, 2014, TSTF/NRC public meeting, the NRC staff stated that licensees requesting adoption of TSTF-425 may address external events using either of two methods:
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| " A qualitative or quantitative evaluation of risk as described in NEI 04-10, Revision 1, or
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| * A quantitative evaluation using a PRA endorsed in RG 1.200, Revision 2.
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| The NRC staff stated that a licensee should describe in their application which method is being used to address external events.
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| Accordingly, Exelon proposed implementation of the Technical Specification Surveillance Frequency Control Program for NMP Unit 2 under TSTF-425 will follow the guidance in NEI 04-10, Revision 1, "Risk-informed Technical Specifications Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies," which was approved by the NRC on September 19, 2007 (ADAMS Accession No. ML072570267).
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| ==3.0 REGULATORY ANALYSIS==
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| 3.1 No Significant Hazards Consideration Exelon has reviewed the proposed No Significant Hazards Consideration (NSHC) determination published in the FRN dated July 6, 2009 (74 FR 31996). Exelon has concluded that the proposed NSHC presented in the FRN is applicable to NMP Unit 2 and is provided as to this amendment request, which satisfies the requirements of 10 CFR 50.91 (a).
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| 3.2 Applicable Regulatory Requirements A description of the proposed changes and their relationship to applicable regulatory requirements is provided in TSTF-425, Revision 3 (ADAMS Accession No. ML090850642) and the NRC's model safety evaluation published in the Notice of Availability, dated July 6, 2009 (74
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-410 Page 5 of 6 FR 31996). Exelon has concluded that the relationship of the proposed changes to the applicable regulatory requirements presented in the FRN is applicable to NMP Unit 2.
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| 3.3 Precedence This application is being made in accordance with the TSTF-425, Revision 3 (ADAMS Accession No. ML090850642). Exelon is not proposing significant variations or deviations from the TS changes described in TSTF 425 or in the content of the NRC staff's model safety evaluation published on July 6, 2009 (74 FR 31996). The NRC has previously approved amendments to the TS as part of the pilot process for TSTF-425, including but not limited to Amendment Nos.
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| 186 and 147 for Limerick Generating Station, Amendment No.276 for Oyster Creek Nuclear Power Station dated, September 27, 2010; Amendment Nos. 200 and 201 for Diablo Canyon Power Plant, Units 1 and 2, respectively, dated October 30, 2008; and Amendment Nos. 188 and 175 for South Texas Project, Units 1 and 2, respectively, dated October 31, 2008. The subject License Amendment Request proposes to relocate periodic surveillance frequencies to a licensee-controlled program and add a new program (the Surveillance Frequency Control Program) to the Administrative Controls section of TS in accordance with TSTF-425 and as discussed in the previously approved amendments.
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| 3.4 Conclusions In 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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| ==4.0 ENVIRONMENTAL CONSIDERATION==
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| Exelon has reviewed the environmental consideration included in the NRC staff's model safety evaluation published in the Federal Register on July 6, 2009 (74 FR 31996). Exelon has concluded that the staff's findings presented therein are applicable to NMP Unit 2, and the determination is hereby incorporated by reference for this application.
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-410 Page 6 of 6
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| ==5.0 REFERENCES==
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| : 1. TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b," March 18, 2009 (ADAMS Accession Number: ML090850642).
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| : 2. NRC Notice of Availability of Technical Specification Improvement to Relocate Surveillance Frequencies to Licensee Control - Risk-Informed Technical Specification Task Force (RITSTF) Initiative 5b, Technical Specification Task Force - 425, Revision 3, published on July 6, 2009 (74 FR 31996).
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| : 3. NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," April 2007 (ADAMS Accession Number:
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| ML071360456).
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| : 4. Regulatory Guide 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," January 2007 (ADAMS Accession Number: ML070240001).
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| : 5. Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decision making:
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| Technical Specifications," dated August 1998 (ADAMS Accession No. ML003740176).
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| ATTACHMENT 2 License Amendment Request Nine Mile Point Nuclear Station, Unit 2 Docket No. 50-410 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
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| Documentation of PRA Technical Adequacy
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-410 Page i ofi Documentation of PRA Technical Adequacy TABLE OF CONTENTS Section Page 2.1 O ve rv iew ........................................................................................................................ 1 2.2 Technical Adequacy of the PRA Model ................................................................... 2 2.2.1 Plant Changes Not Yet Incorporated Into The PRA Model .............................. 4 2.2.2 Applicability of Peer Review Findings and Observations ................................. 4 2.2.3 Consistency With Applicable PRA Standards ................................................. 5 2.2.4 Identification of Key Assumptions .................................................................... 9 2.3 External Events Considerations .............................................................................. 9 2.4 Su m m a ry ........................................................................................................................ 9 2.5 R eferences ................................................................................................................... 10
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-410 Page 1 of 10 Documentation of PRA Technical Adequacy 2.1 Overview The implementation of the Surveillance Frequency Control Program (also referred to as Technical Specifications Initiative 5b) at Nine Mile Point Nuclear Station, Unit 2 (NMP2) will follow the guidance provided in Nuclear Energy Institute (NEI) 04-10, Revision 1 [Ref. 1] in evaluating proposed Surveillance Test Interval (STI) also referred to as "surveillance frequency") changes.
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| The following steps of the risk-informed STI revision process are common to proposed changes to all STIs within the proposed licensee-controlled program.
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| * Each STI revision is reviewed to determine whether there are any commitments made to the NRC that may prohibit changing the interval. If there are no related commitments, or the commitments may be changed using a commitment change process based on NRC endorsed guidance, then evaluation of the STI revision would proceed. If a commitment exists and the commitment change process does not permit the change, then the STI revision would not be implemented.
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| * A qualitative analysis is performed for each STI revision that involves several considerations as explained in NEI 04-10, Revision 1.
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| * Each STI revision is reviewed by an Expert Panel, referred to as the Integrated Decision-making Panel (IDP), which is normally the same panel as is used for Maintenance Rule implementation, but with the addition of specialists with experience in surveillance tests and system or component reliability. If the IDP approves the STI revision, the change is documented and implemented, and available for audit by the NRC. If the IDP does not approve the STI revision, the STI value is left unchanged.
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| * Performance monitoring is conducted as recommended by the IDP. In some cases, no additional monitoring may be necessary beyond that already conducted under the Maintenance Rule. The performance monitoring helps to confirm that no failure mechanisms related to the revised test interval become important enough to alter the information provided for the justification of the interval changes.
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| " The IDP is responsible for periodic review of performance monitoring results. If it is determined that the time interval between successive performances of a surveillance test is a factor in the unsatisfactory performances of the surveillance, the IDP returns the STI back to the previously acceptable STI.
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| * In addition to the above steps, the Probabilistic Risk Assessment (PRA) is used when possible to quantify the effect of a proposed individual STI revision compared to acceptance criteria in NEI 04-10. The cumulative impact of all risk-informed STI revisions on relevant PRA models is compared to the risk acceptance criteria as delineated in NEI 04-10.
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| For those cases where the STI cannot be modeled in the plant PRA (or where a particular PRA model does not exist for a given hazard group), a qualitative or bounding analysis is performed to provide justification for the acceptability of the proposed test interval change.
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| The NEI 04-10 methodology endorses the guidance provided in Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.200, Revision 1 [Ref. 2], "An Approach for
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-410 Page 2 of 10 Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities." The guidance in RG-1.200 indicates that the following steps should be followed when performing PRA assessments (NOTE: Because of the broad scope of potential Initiative 5b applications and the fact that the risk assessment details will differ from application to application, each of the issues encompassed in Items 1 through 3 below will be covered with the preparation of each individual PRA assessment made in support of the individual STI interval requests. Item 3 satisfies one of the requirements of Section 4.2 of RG 1.200. The remaining requirements of Section 4.2 are addressed by Item 4 below.):
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| : 1. Identify the parts of the PRA used to support the application
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| - Structures, Systems and Components (SSCs), operational characteristics affected by the application and how these are implemented in the PRA model
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| - A definition of the acceptance criteria used for the application
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| : 2. Identify the scope of risk contributors addressed by the PRA model
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| - If not full scope (i.e., internal and external), identify appropriate compensatory measures or provide bounding arguments to address the risk contributors not addressed by the model.
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| : 3. Summarize the risk assessment methodology used to assess the risk of the application
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| - Include how the PRA model was modified to appropriately model the risk impact of the change request.
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| : 4. Demonstrate the Technical Adequacy of the PRA
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| - Identify plant changes (design or operational practices) that have been incorporated at the site, but are not yet in the PRA model and justify why the change does not impact the PRA results used to support the application. (addressed in each STI evaluation.)
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| - Document peer review findings and suggestions that are applicable to the parts of the PRA required for the application, and for those that have not yet been addressed justify why the significant contributors would not be impacted.
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| - Document that the parts of the PRA used in the decision are consistent with applicable standards endorsed by the RG (currently, in RG-1.200 Revision 1 this is just the internal events PRA standard). Provide justification to show that where specific requirements in the standard are not adequately met, it will not unduly impact the results.
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| - Identify key assumptions and approximations relevant to the results used in the decision-making process (addressed in each STI evaluation.)
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| The purpose of the remaining portion of this attachment is to address the requirements identified in Item 4 above.
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| 2.2 Technical Adequacy of the PRA Model The NM214A version of the NMP2 PRA model is the most recent evaluation of the Unit 2 risk profile at NMP2. The NMP2 PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-410 Page 3 of 10 model quantification process used for the NMP2 PRA is based on the event tree / fault tree methodology, which is a well-known methodology in the industry.
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| The NMP2 PRA Level 1 and Level 2 models were initially developed in response to NRC Generic Letter 88-20 Individual Plant Examination (IPE). Since the original IPE submittal, the NMP2 PRA has undergone several model revisions to incorporate improvements and maintain consistency with the as built, as operated plant.
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| A full update of the PRA model in accordance with RG 1.200 was completed and peer reviewed in July 2009. In December 2010 the NMP2 PRA was updated to credit modification ECP 000291 [Ref. 3] which allows firewater to be aligned to the DIV III Emergency Diesel Generator (i.e., 2EGS*EG2). As a result, CDF and LERF were decreased significantly.
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| In March 2012 the Nine Mile Point Unit 2 PRA was updated to reflect the changes associated with the Extended Power Uprate (EPU) project as well as changes associated with the instrument air dryer modifications implemented under EC20080024-000 [Ref. 4].
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| Exelon Generation Company, LLC (Exelon) employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating Exelon nuclear generation sites. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to the NMP2 PRA.
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| PRA Maintenance and Update The current NMP2 model of record (i.e., NM214A) was developed and maintained under the Constellation Energy Nuclear Group (CENG) processes and procedures prior to the adoption of the Exelon Management Model in August 2014. Thus, all future PRA model update and maintenance activities will be governed by the Exelon Risk Management procedures and processes. The Exelon Risk Management process ensures that the applicable PRA model is an accurate reflection of the as-built and as-operated plants. This process is defined in the Exelon Risk Management program, which consists of a governing procedure (i.e., ER-AA-600, "Risk Management") and subordinate implementation Training and Reference Manuals (T&RMs).
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| Exelon T&RM ER-AA-600-1015, "FPIE PRA Model Update," delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating Exelon nuclear generation sites. The overall Exelon Risk Management program, including ER-AA-600-1015, defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, industry operating experience, etc.), and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plants, the following activities are routinely performed:
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| * design changes and procedure changes are reviewed for their impact on the PRA model
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| * new engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model
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| * maintenance unavailabilities are captured, and their impact on CDF is trended
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-410 Page 4 of 10
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| * Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated approximately every four years In addition to these activities, Exelon risk management procedures provide the guidance for particular risk management maintenance activities. This guidance includes:
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| * Documentation of the PRA model, PRA products, and bases documents
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| * The approach for controlling electronic storage of Risk Management (RM) products including PRA update information, PRA models, and PRA applications
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| * Guidelines for updating the full power internal, fire and external events PRA models for Exelon Generation nuclear sites
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| * Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modifications) on SSCs within the scope of the Maintenance Rule (10 CFR 50.65(a)(4))
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| As indicated previously, RG 1.200 also requires that additional information be provided as part of the License Amendment Request submittal to demonstrate the technical adequacy of the PRA model used for the risk assessment. Each of these items (i.e., plant changes not yet incorporated in to the PRA model, relevant peer review findings, consistency with applicable PRA Standards, and the identification of key assumptions) will be discussed in turn.
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| 2.2.1 Plant Changes Not Yet Incorporated into the PRA Model The station PRA is representative of the station; however, during the period between PRA model updates, changes to the plant or its procedures are controlled using the Exelon Risk Management Configuration Control process. The majority of these changes do not necessitate any change to the PRA due to their relatively low impact on risk metrics. In order to ensure that that the PRA model accurately reflects the as built/as operated plant these changes are tracked in the Updating Requirements Evaluation (URE) database. This tool is used to track all changes that are identified as being potentially impacting on the PRA model and incorporate these changes in a timely manner.
| |
| As part of the PRA evaluation for each STI change request, a review of open items in the URE database for NMP2 is normally performed and an assessment of the impact on the results of the application is made prior to presenting the results of the risk analysis to the Independent Decision-making Panel (IDP). If a non-trivial impact is expected, then this may include the performance of additional sensitivity studies or PRA model changes to confirm the impact on the risk analysis.
| |
| 2.2.2 Applicability of Peer Review Findings and Observations A PRA model update to meet NRC RG 1.200 was completed in 2009 in accordance with NEI 05-04, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard [Ref. 5]. In updating the PRA, changes were made to the model to address most of the identified gaps, as well as to address other open URE items. This revision included a significant change in methodology, converting from RISKMAN Model U2PRA01B (Large Event Trees, Small Fault Trees) to CAFTA 08U2 Model Rev 00 (Small Event Trees, Large Fault
| |
| | |
| LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-410 Page 5 of 10 Trees). Following the peer review in July 2009, two additional updates were performed as noted above in December 2010 and March 2012..
| |
| 2.2.3 Consistency with Applicable PRA Standards As indicated above, a PRA ASME model update was completed in 2009. [Ref. 6] Subsequent updates were performed in 2010 and 2012 to incorporate plant design changes (supply firewater to the Division 3 Emergency Generator Cooling system, and a second update incorporated Extended Power Uprate (EPU) and design changes to the Instrument Air Dryer system).
| |
| The results of the Peer Review included 18 Findings which did not meet Capability Category II and 34 Suggestions that met Category II with areas for improvement. These Findings and Suggestions were addressed with changes to the model and/or documentation as required.
| |
| In conclusion, the latest NMP2 PRA model NM214A is the current model of record with six Peer Review Findings unresolved. These unresolved Findings are summarized in Table 2-1 along with an assessment of the impact for this application. All remaining gaps documented in the URE database including 10 unresolved Peer Review Suggestions have low impact on the PRA model or its ability to support a full range of PRA applications. The remaining gaps are documented in the URE database so that they can be tracked and their potential impacts accounted for in applications where appropriate.
| |
| | |
| LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-410 Page 6 of 10 TABLE 2-1 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD URE DESCRIPTION OF GAP APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO SRs t APPLICATION N22014-472 Demands from causes other than DA-C6 OPEN None surveillance tests were not included in DA-C7 Insignificant Impact - Defer for the collection of plant-specific data. consideration during next major update.
| |
| It was looked at during unit 1 update and (This F&O originated from SR DA-C6) considered again during unit 2 update. It is a little conservative and not considered significant to estimate using surveillance procedures.
| |
| N22014-474 F&O: 5-2: Routine system alignments IE-A6 OPEN - Insignificant None contributing to initiating event Wait for plant reliability model development.
| |
| frequencies are not included. Routine alignments are already included in the average initiating event frequency development. In addition, the addition of support system initiating event fault trees to the model (see F&O 2-16 and 4-4) adds some important alignments for these systems. It would be a significant effort to add the type of factors that are typically reserved for EOOS risk managment modeling such as 1/2 scram testing etc.
| |
| This will have to wait until a plant reliability program is developed (e.g., scram risk).
| |
| | |
| LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-410 Page 7 of 10 TABLE 2-1 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD URE DESCRIPTION OF GAP APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO SRs APPLICATION N22014-475 F&O 6-1: In some cases the assignment SR HR-G1 Open Insignificant No Impact of a conservative screening HEP value Section to 1 of HRA Notebook updated to may not have been appropriate given explicitly identify HEPs based on screening, the risk significance of the operator the basis for screening, and their action it represents. In particular, the importance.
| |
| use of a conservative screening value of 1 E-02 assigned to the HEP Detailed HRA will be considered in future ZHS05_HSROOMCOL. "Operator Fails updates as appropriate.
| |
| to open HPCS ROOM Doors and HVAC Duct" may not have been appropriate given the risk significance of the HPCS room cooling support system (This F&O originated from SR HR-G1)
| |
| N22014-476 F&O 6-5: The AS Notebook does not SR AS-Cl OPEN No Impact contain the event tree top event fault The final Post Peer Review issuance of the trees, which are necessary for AS Notebook will have all the understanding the accident sequence documentation in the AS Notebook as logic. suggested versus external (facilitates review etc).
| |
| N22014-473 F&O 2-16: This SR requires SR QU-D6 CLOSED: No Impact - SSC failure rates for identification of contributors to CDF. To Support System initiating event fault trees Initiating event contributions are satisfy Category II (and Ill) requires have been added to the model. The IE normalized to one year regardless including SSCs and operator actions Notebook refers to this. of the surveillance test frequency.
| |
| that contribute to IE frequencies. These SY.00 Notebook provides methodology.
| |
| are not included for NMP-2, so only Applicable SY notebooks develop the Category I has been met. models.
| |
| OPEN:
| |
| Equipment and operator contributions will be developed in QU Notebook.
| |
| Update IE NB with correction factors
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| | |
| LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-410 Page 8 of 10 TABLE 2-1 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD URE DESCRIPTION OF GAP APPLICABLE CURRENT STATUS I COMMENT IMPORTANCE TO SRs APPLICATION 3-5 At the time of the Peer Review, various MU-Fl OPEN Documentation only - No Impact PRA documentation notebooks were not The final Post Peer Review issuance of all signed by performers, reviewers, or notebooks will be signed off as Rev 0.
| |
| approvers.
| |
| (This F&O originated from SR MU-Fl)
| |
| | |
| LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-410 Page 9 of 10 2.2.4 Identification of Key Assumptions The overall Initiative 5b process is a risk-informed process with the PRA model results providing one of the inputs to the IDP to determine if an STI change is warranted. The methodology recognizes that a key area of uncertainty for this application is the standby failure rate utilized in the determination of the STI extension impact. Therefore, the methodology requires the performance of selected sensitivity studies on the standby failure rate of the component(s) of interest for the STI assessment.
| |
| The results of the standby failure rate sensitivity study plus the results of any additional sensitivity studies identified during the performance of the reviews as outlined in 2.2.1 and 2.2.3 above (including a review of identified sources of uncertainty that were developed for NMP2 based on the EPRI 1016737 guidance [Ref. 7]) for each STI change assessment will be documented and included in the results of the risk analysis that goes to the IDP.
| |
| 2.3 External Events Considerations The NEI 04-10 methodology allows for STI change evaluations to be performed in the absence of quantifiable PRA models for all external hazards. For those cases where the STI cannot be modeled in the plant PRA (or where a particular PRA model does not exist for a given hazard group), a qualitative or bounding analysis is performed to provide justification for the acceptability of the proposed test interval change.
| |
| External hazards were evaluated in the NMP2 Individual Plant Examination of External Events (IPEEE) submittal in response to the NRC IPEEE Program (Generic Letter 88-20, Supplement 4)
| |
| [Ref. 8]. The IPEEE Program was a one-time review of external hazard risk and was limited in its purpose to the identification of potential plant vulnerabilities and the understanding of associated severe accident risks.
| |
| The results of the NMP2 IPEEE study are documented in the NMP2 IPEEE Main Report [Ref. 9].
| |
| The primary areas of external event evaluation at NMP2 were internal fire and seismic. The internal fire events were addressed by using the EPRI Fire Induced Vulnerability Evaluation (FIVE) methodology [Ref. 10] and the seismic evaluations were performed in accordance with the EPRI Seismic Margins Analysis (SMA) methodology [Ref. 11].
| |
| In addition to internal fires and seismic events, the NMP2 IPEEE analysis of high winds, floods, and other (HFO) external hazards was accomplished by reviewing the plant environs against regulatory requirements regarding these hazards. As such, these hazards were determined in the NMP2 IPEEE to be negligible contributors to overall plant risk.
| |
| The IPEEE Fire PRA model was also integrated with the NM214A internal events model as part of the 2009 update.
| |
| 2.4 Summary The NMP2 PRA technical capability evaluations and the maintenance and update processes described above provide a robust basis for concluding that the NMP2 PRA is suitable for use in risk-informed processes such as that proposed for the implementation of a Surveillance Frequency Control Program.
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| | |
| LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-410 Page 10 of 10 2.5 References
| |
| [1] Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, Industry Guidance Document, NEI 04-10, Revision 1, April 2007.
| |
| [2] Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities, Revision 1, January 2007.
| |
| [3] ECP-10-000291: ESR-10-000293 NMP-ESR (0000) -HPCS DIESEL SERVICE WATER PROJECT
| |
| [4] EC20080024-000: REPLACE INSTRUMENT AIR DRYERS 21AS-DRY1A AND 21AS-DRY1B
| |
| [5] NEI 05-04, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard
| |
| [6] American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, (ASME RA-S-2002), Addenda RA-Sc-2007, August 2007.
| |
| [7] Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, EPRI, Palo Alto, CA: December 2008 (Final). 1016737.
| |
| [8] NRC Generic Letter 88-20, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), Supplement 4," June 28, 1991.
| |
| [9] Nine Mile Point Unit 2 IPEEE, Niagara Mohawk Power Corporation, June 1995.
| |
| [10] Professional Loss Control, Inc., Fire-Induced Vulnerability Evaluation (FIVE) Methodology Plant Screening Guide, EPRI TR-100370, Electric Power Research Institute, Final Report, April 1992.
| |
| [11] NTS Engineering, et. al., A Method for Assessment of Nuclear Power Plant Seismic Margin, EPRI NP-6041, Electric Power Research Institute, Final Report, August 1991.
| |
| | |
| ATTACHMENT 3 License Amendment Request Nine Mile Point Nuclear Station, Unit 2 Docket No. 50-410 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
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| Proposed Technical Specification Page Changes - Unit 2 3.1.3-3 3.3.3.2-2 3.4.11-2 3.6.4.1-2 3.8.4-3 3.1.3-4 3.3.4.1-3 3.4.11-4 3.6.4.1-3 3.8.4-4 3.1.4-2 3.3.4.1-4 3.4.11-5 3.6.4.2-4 3.8.6-2 3.1.5-3 3.3.4.2-2 3.4.12-1 3.6.4.3-3 3.8.6-3 3.1.6-2 3.3.4.2-3 3.5.1-4 3.7.1-3 3.8.7-2 3.1.7-1 3.3.5.1-8 3.5.1-5 3.7.1-4 3.8.8-3 3.1.7-2 3.3.5.2-3 3.5.1-6 3.7.2-3 3.8.9-2 3.1.7-3 3.3.6.1-4 3.5.2-2 3.7.3-4 3.9.1-2 3.1.8-2 3.3.6.1-5 3.5.2-3 3.7.4-2 3.9.2-1 3.2.1-1 3.3.6.2-2 3.5.2-4 3.7.5-1 3.9.2-2 3.2.2-1 3.3.6.2-3 3.5.3-2 3.7.6-1 3.9.3-1 3.2.3-1 3.3.7.1-3 3.5.3-3 3.8.1-5 3.9.5-1 3.3.1.1-3 3.3.7.2-2 3.6.1.1-2 3.8.1-6 3.9.6-1 3.3.1.1-4 3.3.7.2-3 3.6.1.1-3 3.8.1-7 3.9.7-1 3.3.1.1-5 3.3.8.1-2 3.6.1.2-5 3.8.1-8 3.9.8-3 3.3.1.1-6 3.3.8.2-4 3.6.1.3-10 3.8.1-9 3.9.9-3 3.3.1.1-7 3.3.8.3-2 3.6.1.3-11 3.8.1-10 3.10.2-2 3.3.1.2-3 3.3.8.3-3 3.6.1.3-12 3.8.1-11 3.10.3-3 3.3.1.2-4 3.4.1-3 3.6.1.4-1 3.8.1-12 3.10.4-3 3.3.1.2-5 3.4.2-1 3.6.1.5-1 3.8.1-13 3.10.4-4 3.3.2.1-3 3.4.2-2 3.6.1.6-2 3.8.1-14 3.10.5-2 3.3.2.1-4 3.4.3-2 3.6.1.7-3 3.8.1-15 3.10.5-3 3.3.2.1-5 3.4.5-2 3.6.2.1-3 3.8.1-16 3.10.6-2 3.3.2.2-2 3.4.7-3 3.6.2.2-1 3.8.1-17 3.10.8-3 3.3.2.2-3 3.4.8-2 3.6.2.3-2 3.8.1-18 3.10.8-4 3.3.3.1-2 3.4.9-3 3.6.2.4-2 3.8.3-3 5.5-13 3.3.3.1-3 3.4.10-2 3.6.3.2-1 3.8.4-2
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| | |
| INSERT 1 In accordance with the Surveillance Frequency Control Program INSERT 2 5.5.14 Surveillance Frequency Control program This program provides controls for the 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 the Surveillance Requirements for which the Frequency is controlled by the program.
| |
| : b. Changes to the Frequency listed in the Surveillance Frequency Controlled Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequency," Revision 1.
| |
| : c. The provision of Surveillance Requirement 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.
| |
| | |
| Control Rod OPERABILITY 3.1.3 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. --------- NOTE -------------- D.1 Restore compliance 4 hours Not applicable when with BPWS.
| |
| THERMAL POWER
| |
| > 10% RTP. OR D.2 Restore control rod 4 hours Two or more inoperable to OPERABLE status.
| |
| control rods not in compliance with banked position withdrawal sequence (BPWS) and not separated by two or more OPERABLE control rods.
| |
| E. Required Action and E.1 Be in MODE 3. 12 hours associated Completion Time of Condition A, C, or D not met.
| |
| OR Nine or more control rods inoperable.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod. 24-e9Ur-JISET 1- (continued)
| |
| NMP2 3.1.3-3 Amendment 91
| |
| | |
| Control Rod OPERABILITY 3.1.3 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.1.3.2 Deleted SR 3.1.3.3 ---------------------- NOTE---------------
| |
| Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.
| |
| Insert each withdrawn control rod at least one notch.
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| SR 3.1.3.4 Verify each control rod scram time from In accordance fully withdrawn to notch position 05 is with
| |
| *7 seconds. SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)
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| NMP2 3.1.3-4 Amendment 94-, 130
| |
| | |
| Control Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY INSERT 1 SR 3.1.4.2 Verify, for a representative sample, each 209 days tested control rod scram time is within the eUF'I-ti'-e limits of Table 3.1.4-1 with reactor steam .peF.tieR--"
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| dome pressure > 800 psig. MODE 4 SR 3.1.4.3 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with declaring any reactor steam dome pressure. control rod OPERABLE after work on control rod or CRD System that could affect scram time SR 3.1.4.4 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with exceeding reactor steam dome pressure > 800 psig. 40% RTP after fuel movement within the affected core cell AND Prior to exceeding 40%
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| RTP after work on control rod or CRD System that could affect scram time NMP2 3.1.4-2 Amendment 94, 132
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| | |
| Control Rod Scram Accumulators 3.1.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 Declare the 1 hour associated control rod inoperable.
| |
| D. Required Action B.1 or D.1 --------- NOTE------
| |
| C.1 and associated Not applicable if all Completion Time not inoperable control met. rod scram accumulators are associated with fully inserted control rods.
| |
| Place the reactor Immediately mode switch in the shutdown position.
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| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each control rod scram accumulator pressure is > 940 psig.
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| FINSERT I NMP2 3.1.5-3 Amendment 91
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| | |
| Rod Pattern Control 3.1.6 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Nine or more OPERABLE B.1 --------- NOTE-----
| |
| control rods not in RWM may be bypassed compliance with BPWS. as allowed by LCO 3.3.2.1.
| |
| Suspend withdrawal of Immediately control rods.
| |
| AND B.2 Place the reactor 1 hour mode switch in the shutdown position.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify all OPERABLE control rods comply 24-hehU-with BPWS.
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| JINSERT 1 NMP2 3.1.6-2 Amendment 91
| |
| | |
| SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem 7 days inoperable, to OPERABLE status.
| |
| B. Two SLC subsystems B.1 Restore one SLC 8 hours inoperable, subsystem to OPERABLE status.
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| C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium 24 h9'-s pentaborate solution is within the limits of Figure 3.1.7-1. / 1I IINSERT (continued)
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| NMP2 3.1.7-1 Amendment 94, 125
| |
| | |
| SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.2 Verify temperature of sodium pentaborate 24 h9Urs<
| |
| solution is > 700F.
| |
| SR 3.1.7.3 Verify temperature of pump suction piping 24-he-s <
| |
| up to the pump suction valve is _>700 F.
| |
| SR 3.1.7.4 Verify continuity of explosive charge. 31 daye(
| |
| SR 3.1.7.5 Verify the concentration of sodium pentaborate in solution is within the limits of Figure 3.1.7-1. AND Once within 24 hours after water or sodium pentaborate is added to solution AND Once within 24 hours after solution temperature is restored to
| |
| > 70OF SR 3.1.7.6 Verify each SLC subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the correct position.
| |
| (continued)
| |
| NMP2 3.1.7-2 Amendment 91
| |
| | |
| SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify each pump develops a flow rate In accordance
| |
| > 41.2 gpm at a discharge pressure with the
| |
| > 1327 psig. Inservice Testing Program SR 3.1.7.8 Verify flow through one SLC subsystem 2"1. mo.Mthc o.n.
| |
| 9 from pump into reactor pressure vessel. STAGGFEREDEI TErT BASIS INSERT 1 SR 3.1.7.9 Verify all heat traced piping between 24 mntM4hs--
| |
| storage tank and pump suction valve is unblocked. AND Once within 24 hours after piping temperature is restored to
| |
| > 70°F SR 3.1.7.10 Verify sodium pentaborate enrichment Prior to is > 92 atom percent B-10. addition to SLC tank NMP2 3.1.7-3 Amendment 91, 111, 117, 123, 4190, 143
| |
| | |
| SDV Vent and Drain Valves 3.1.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 NOTE- ----------------
| |
| Not required to be met on vent and drain valves closed during performance of SR 3.1.8.2.
| |
| Verify each SDV vent and drain valve is 31 days open.
| |
| SR 3.1.8.2 Cycle each SDV vent and drain valve to the fully closed and fully open position.
| |
| SR 3.1.8.3 Verify each SDV vent and drain valve:
| |
| : a. Closes in < 30 seconds after receipt of an actual or simulated scram signal; and
| |
| : b. Opens when the actual or simulated scram signal is reset.
| |
| NMP2 3.1.8-2 Amendment 91
| |
| | |
| APLHGR 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
| |
| LCO 3.2.1 All APLHGRs shall be less than or equal to the limits specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER _ 23% RTP.
| |
| ACTIONS COMPLETION REQUIRED ACTION TIME CONDITION TIME A. Any APLHGR not within A.1 Restore APLHGR(s) to 2 hours limits, within limits.
| |
| B. Required Action and B.1 Reduce THERMAL POWER 4 hours associated Completion Time not met.
| |
| to < 23% RTP. .1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal Once within to the limits specified in the COLR. 12 hours after
| |
| _ 23% RTP tANDINSERT 1 thereefefr NMP2 3.2.1-1 Amendment 9-1-, 140
| |
| | |
| MCPR 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)
| |
| LCO 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER >_23% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within A.1 Restore MCPR(s) to 2 hours limits. within limits.
| |
| B. Required Action and B.1 Reduce THERMAL POWER 4 hours associated Completion to < 23% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal Once within to the limits specified in the COLR. 12 hours after
| |
| _ 23% RTP (continued)
| |
| NMP2 3.2.2-1 Amendment 9-1-, 140
| |
| | |
| LHGR 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)
| |
| LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER > 23% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A.1 Restore LHGR(s) to 2 hours limits, within limits.
| |
| B. Required Action and B.1 Reduce THERMAL POWER 4 hours associated Completion to < 23% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to Once within the limits specified in the COLR. 12 hours after
| |
| Ž23% RTP AND FtNSERT 1 NMP2 3.2.3-1 Amendment 9-1-, 140
| |
| | |
| RPS Instrumentation 3.3.1.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME H. As required by H.1 Be in MODE 3. 12 hours Required Action D.1 and referenced in Table 3.3.1.1-1.
| |
| : 1. As required by 1.1 Initiate action to Immediately Required Action D.1 fully insert all and referenced in insertable control Table 3.3.1.1-1. rods in core cells containing one or more fuel assemblies.
| |
| SURVEILLANCE REQUIREMENTS
| |
| ----------------------------- NOTE --------------------------------
| |
| : 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. ....... -'* INSERT 1 j 4-SR 3.3.1.1.2 Perform CHANNEL CHECK. 24 hewFa4S -
| |
| (continued)
| |
| NMP2 3.3.1.1-3 Amendment 91, 92
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.3 --------------------- NOTE---------------
| |
| Not required to be performed until 12 hours after THERMAL POWER Ž_23% RTP.
| |
| Verify the absolute difference between 7. day the average power range monitor (APRM) channels and the calculated power
| |
| < 2% RTP while operating at _>23% RTP.
| |
| SR 3.3.1.1.4 --------------------- NOTE---------------
| |
| For Functions 1.a and 1.b, not required JINS"F to be performed when entering MODE 2 from I MODE 1 until 12 hours after entering MODE 2.
| |
| Perform CHANNEL FUNCTIONAL TEST. 7-day<
| |
| SR 3.3.1.1.5 Verify the source range monitor (SRM) and Prior to fully intermediate range monitor (IRM) channels withdrawing overlap. SRMs SR 3.3.1.1.6 --------------------- NOTE ---------------
| |
| Only required to be met during entry into MODE 2 from MODE 1.
| |
| Verify the IRM and APRM channels overlap.
| |
| (continued)
| |
| NMP2 3.3.1.1-4 Amendment 91 123, 140
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.7 Calibrate the local power range monitors. !000effeetey-full-pewe~
| |
| hewrs SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST. -edays <
| |
| SR 3.3.1.1.9 Calibrate the trip units. 92-days <
| |
| SR 3.3.1.1.10 -------------------- NOTES ---------------
| |
| : 1. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
| |
| : 2. For Function 2.e, the CHANNEL FUNCTIONAL TEST only requires toggling the appropriate outputs of the APRM.
| |
| Perform CHANNEL FUNCTIONAL TEST. I8A4dd. 4ys.--
| |
| SR 3.3.1.1.11 Perform CHANNEL CALIBRATION. I- < -Mcnth SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 24 Fnt <
| |
| (continued)
| |
| NMP2 3.3.1.1-5 Amendment 91, 92
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.13 --------------------NOTES ---------------
| |
| : 1. Neutron detectors are excluded.
| |
| : 2. For Functions 1.a and 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
| |
| : 3. For Function 2.e, the CHANNEL CALIBRATION only requires a verification of ORRM-Upscale setpoints in the APRM by the review of the "Show Parameters" display.
| |
| 7 INEýT1 Perform CHANNEL CALIBRATION. 24-weAth-,'
| |
| SR 3.3.1.1.14 Perform LOGIC SYSTEM FUNCTIONAL TEST. 2, m.Ath-R SR 3.3.1.1.15 Verify Turbine Stop Valve - Closure, and 24-meFhs <
| |
| Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are not bypassed when THERMAL POWER is _>26% RTP.
| |
| SR 3.3.1.1.16 Verify APRM OPRM-Upscale Function is not 24-menths <
| |
| bypassed when THERMAL POWER is Ž_26% RTP and recirculation drive flow is < 60% of rated recirculation drive flow.
| |
| (continued)
| |
| NMP2 3.3.1.1-6 Amendment 9-1-,92, 140 Corro.toedl bh, lotor of Juhl 214 2000
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.17 ------------------- NOTES --------------
| |
| : 1. Function 2.f digital electronics are excluded.
| |
| : 2. For Functions 3 and 4, the sensor response time may be assumed to be the design sensor response time.
| |
| : 3. For Function 5, "n" equals 4 channels for the purpose of determining the STAGGERED TEST BASIS Frequency. JINSERT 1
| |
| : 4. For Function 9, the RPS RESPONSE TIME is measured from start of turbine control valve fast closure.
| |
| Verify the RPS RESPONSE TIME is within 21 months on a limits. STAGGERED TEST BASIS NMP2 3.3.1.1-7 Amendment 91, 92
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS NOTE --------------------------------
| |
| Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified condition.
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.2.1 Perform CHANNEL CHECK. *FS SR 3.3.1.2.2 -------------------- NOTES ---------------
| |
| : 1. Only required to be met during CORE ALTERATIONS.
| |
| : 2. One SRM may be used to satisfy more than one of the following.
| |
| Verify an OPERABLE SRM detector is located in:
| |
| : a. The fueled region;
| |
| : b. The core quadrant where CORE ALTERATIONS are being performed when the associated SRM is included in the fueled region; and
| |
| : c. A core quadrant adjacent to where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region.
| |
| SR 3.3.1.2.3 Perform CHANNEL CHECK. 24-heurs (continued)
| |
| NMP2 3.3.1.2-3 Amendment 91
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY i
| |
| SR 3.3.1.2.4 --------------
| |
| NOTE---------------
| |
| Not required to be met with less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.
| |
| Verify count rate is: 12 Mum-h dU- GORE
| |
| ,ALTE-:RAT*'IONS-
| |
| : a. _>3.0 cps with a signal to noise ratio > 2:1; or
| |
| : b. > 1.3 cps with a signal to noise ratio >_5:1.
| |
| 4-SR 3.3.1.2.5 -NOTE The determination of signal to noise ratio is not required to be met with less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.
| |
| Perform CHANNEL FUNCTIONAL TEST and determination of signal to noise ratio.
| |
| i SR 3.3.1.2.6 ---------------
| |
| NOTE --------------
| |
| Not required to be performed until 12 hours after IRMs on Range 2 or below.
| |
| Perform CHANNEL FUNCTIONAL TEST and v1 days determination of signal to noise ratio.
| |
| (continued)
| |
| NMP2 3.3.1.2-4 Amendment 91
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.2.7 -------------------- NOTES ---------------
| |
| : 1. Neutron detectors are excluded.
| |
| : 2. Not required to be performed until 12 hours after IRMs on Range 2 or below.
| |
| Perform CHANNEL CALIBRATION. 22 4. M.thet,
| |
| [INSERT 1 NMP2 3.3.1.2-5 Amendment 91
| |
| | |
| Control Rod Block Instrumentation 3.3.2.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME E. One or more Reactor E.1 Suspend control rod Immediately Mode Switch - Shutdown withdrawal.
| |
| Position channels inoperable. AND E.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies.
| |
| SURVEILLANCE REQUIREMENTS
| |
| ---------------------------- NOTES -------------------------------
| |
| : 1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function.
| |
| : 2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.1.1 --------------------- NOTE---------------
| |
| Not required to be performed until 1 hour after any control rod is withdrawn at IINSERT 1 j
| |
| < 10% RTP in MODE 2.
| |
| T1 Perform CHANNEL FUNCTIONAL TEST. 2-,
| |
| (continued)
| |
| NMP2 3.3.2.1-3 Amendment 91
| |
| | |
| Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.1.2 --------------------- NOTE -----------------------------
| |
| Not required to be performed until 1 hour after THERMAL POWER is < 10% RTP in MODE 1.
| |
| Perform CHANNEL FUNCTIONAL TEST. 92 dayse
| |
| *1~
| |
| SR 3.3.2.1.3 Perform CHANNEL FUNCTIONAL TEST. 44 d-ays SR 3.3.2.1.4 --------------------- NOTE ---------------
| |
| Neutron detectors are excluded.
| |
| Verify the RBM:
| |
| : a. Low Power Range - Upscale Function is not bypassed when APRM Simulated Thermal Power is > 28% and < 63% RTP and a peripheral control rod is not selected.
| |
| : b. Intermediate Power Range - Upscale Function is not bypassed when APRM Simulated Thermal Power is > 63% and
| |
| < 83% RTP and a peripheral control rod is not selected.
| |
| : c. High Power Range - Upscale Function is not bypassed when APRM Simulated Thermal Power is > 83% RTP and a peripheral control rod is not selected.
| |
| 4-SR 3.3.2.1.5 Verify the RWM is not bypassed when THERMAL POWER is < 10% RTP.
| |
| SR 3.3.2.1.6 --------------------- NOTE ---------------
| |
| Not required to be performed until 1 hour after reactor mode switch is in the shutdown position.
| |
| Perform CHANNEL FUNCTIONAL TEST. 24'm..hse (continued)
| |
| NMP2 3.3.2.1-4 Amendment 9-1-, 123
| |
| | |
| Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.1.7 --------------------- NOTE -------------------- INSERT 1 Neutron detectors are excluded.
| |
| Perform CHANNEL CALIBRATION. 24e SR 3.3.2.1.8 Verify control rod sequences input to the Prior to RWM are in conformance with BPWS. declaring RWM OPERABLE following loading of sequence into RWM NMP2 3.3.2.1-5 Amendment 91
| |
| | |
| Feedwater System and Main Turbine High Water Level Trip Instrumentation 3.3.2.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 ------------ NOTE -------
| |
| associated Completion Only applicable if Time not met. inoperable channel is the result of an inoperable feedwater pump breaker.
| |
| Remove affected 4 hours feedwater pump(s) from service.
| |
| OR C.2 Reduce THERMAL POWER 4 hours to < 23% RTP.
| |
| SURVEILLANCE REQUIREMENTS
| |
| --------------------- NOTE --------------------------------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided feedwater system and main turbine high water level trip capability is maintained.
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.2.1 Perform CHANNEL CHECK. INSEART.
| |
| SR 3.3.2.2.2 Perform CHANNEL FUNCTIONAL TEST. 2- days (continued)
| |
| NMP2 3.3.2.2-2 Amendment &4-, 140
| |
| | |
| Feedwater System and Main Turbine High Water Level Trip Instrumentation 3.3.2.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY 4
| |
| SR 3.3.2.2.3 Perform CHANNEL CALIBRATION. The A24m*mr.h4-Allowable Value shall be < 203.8 inches.
| |
| "IINSERT 1TI SR 3.3.2.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST, 24-Fmenthe 4-including breaker and valve actuation.
| |
| NMP2 3.3.2.2-3 Amendment 91
| |
| | |
| PAM Instrumentation 3.3.3.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition C Table 3.3.3.1-1 for not met. the channel.
| |
| E. As required by E.1 Be in MODE 3. 12 hours Required Action D.1 and referenced in Table 3.3.3.1-1.
| |
| F. As required by F.1 Initiate action in Immediately Required Action D.1 accordance with and referenced in Specification 5.6.6.
| |
| Table 3.3.3.1-1.
| |
| SURVEILLANCE REQUIREMENTS
| |
| ---------------------------- NOTES -------------------------------
| |
| : 1. These SRs apply to each Function in Table 3.3.3.1-1, except where identified in the SR.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the other required channel in the associated Function is OPERABLE.
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.1.1 Perform CHANNEL CHECK. - ys (continued)
| |
| IINSERT 1 NMP2 3.3.3.1-2 Amendment 91
| |
| | |
| PAM Instrumentation 3.3.3.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.1.2 Perform CHANNEL CALIBRATION. 224 mR#nhs
| |
| [INSERT 1 NMP2 3.3.3.1-3 Amendment 9-, 124
| |
| | |
| Remote Shutdown System 3.3.3.2 SURVEILLANCE REQUIREMENTS NOTE--------------------------------
| |
| When an instrumentation channel is placed in an inoperable status solely for performance of Required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours.
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.2.1 Perform CHANNEL CHECK for each required ,1 ,days instrumentation channel that is normally energized.
| |
| SR 3.3.3.2.2 Verify each required control circuit and 24m.e.ths <- INSERT I transfer switch is capable of performing the intended functions.
| |
| SR 3.3.3.2.3 Perform CHANNEL CALIBRATION for each 24" m-eFh required instrumentation channel.
| |
| NMP2 3.3.3.2-2 Amendment 91
| |
| | |
| EOC-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS
| |
| ---------------------------- NOTE---------------------------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains EOC-RPT trip capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL FUNCTIONAL TEST. 2- days <
| |
| SR 3.3.4.1.2 Perform CHANNEL CALIBRATION. The 24- months <
| |
| Allowable Values shall be:
| |
| : a. TSV - Closure: < 7% closed; and
| |
| : b. TCV Fast Closure, Trip Oil Pressure - Low: Ž!465 psig. nINSERT 1 SR 3.3.4.1.3 Perform LOGIC SYSTEM FUNCTIONAL TEST, 24 mA-e,4h <
| |
| including breaker actuation.
| |
| SR 3.3.4.1.4 Verify TSV - Closure and TCV Fast Closure, 24 mneths <
| |
| Trip Oil Pressure - Low Functions are not bypassed when THERMAL POWER is > 26% RTP.
| |
| SR 3.3.4.1.5 --------------------- NOTE---------------
| |
| Breaker arc suppression time may be assumed from the most recent performance of SR 3.3.4.1.6.
| |
| Verify the EOC-RPT SYSTEM RESPONSE TIME 24 months on a is within limits. STAGGERED TEST
| |
| ,A*S, (continued)
| |
| NMP2 3.3.4.1-3 Amendment 9-, 140
| |
| | |
| EOC-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.4.1.6 Determine RPT breaker arc suppression 0- morthe time.
| |
| IINSERT I NMP2 3.3.4.1-4 Amendment 91
| |
| | |
| ATWS-RPT Instrumentation 3.3.4.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. One Function with B.1 Restore ATWS-RPT trip 72 hours ATWS-RPT trip capability.
| |
| capability not maintained.
| |
| C. Both Functions with C.1 Restore ATWS-RPT trip 1 hour ATWS-RPT trip capability for one capability not Function.
| |
| maintained.
| |
| D. Required Action and D.1 Remove the associated 6 hours associated Completion recirculation pump Time not met. breaker(s) from service.
| |
| OR 6 hours D.2 Be in MODE 2.
| |
| SURVEILLANCE REQUIREMENTS
| |
| ----------------------------- NOTE --------------------------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains ATWS-RPT trip capability.
| |
| NMP2 3.3.4.2-2 Amendment 91
| |
| | |
| ATWS-RPT Instrumentation 3.3.4.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST. 2-days <
| |
| SR 3.3.4.2.3 Calibrate the analog trip modules. 2 da<
| |
| SR 3.3.4.2.4 Verify, for the Reactor Vessel Steam Dome 24- months Pressure - High Function, the low frequency motor generator trip is not bypassed for > 29 seconds when THERMAL POWER is > 5% RTP.
| |
| SR 3.3.4.2.5 Perform CHANNEL CALIBRATION. The 24'monthe Allowable Values shall be:
| |
| INSERT 1
| |
| : a. Reactor Vessel Water Level - Low Low, Level 2: > 101.8 inches; and
| |
| : b. Reactor Vessel Steam Dome Pressure - High: < 1080 psig.
| |
| SR 3.3.4.2.6 Perform LOGIC SYSTEM FUNCTIONAL TEST, 24- onthe including breaker actuation.
| |
| NMP2 3.3.4.2-3 Amendment 91
| |
| | |
| ECCS Instrumentation 3.3.5.1 SURVEILLANCE REQUIREMENTS
| |
| ---------------------------- NOTES --------------------
| |
| : 1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Functions 3.e, 3.g, 3.h, and 3.i; and (b) for up to 6 hours for Functions other than 3.e, 3.g, 3.h, and 3.i, provided the associated Function or the redundant Function maintains ECCS initiation capability.
| |
| SURVEILLANCE FREQUENCY 4
| |
| SR 3.3.5.1.1 Perform CHANNEL CHECK. 12 h9.rS <
| |
| SR 3.3.5.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days 4 -
| |
| SR 3.3.5.1.3 Calibrate the trip unit. 92 days <-
| |
| INSERT 1 SR 3.3.5.1.4 Perform CHANNEL CALIBRATION. 92-days SR 3.3.5.1.5 SR 3.3.5.1.6 Perform CHANNEL CALIBRATION.
| |
| Perform LOGIC SYSTEM FUNCTIONAL TEST.
| |
| 24 -me4h 24 -Ieieths E
| |
| <- I NMP2 3.3.5.1-8 Amendment 91
| |
| | |
| RCIC System Instrumentation 3.3.5.2 SURVEILLANCE REQUIREMENTS
| |
| ---------------------------- NOTES --------------------- --------------------
| |
| : 1. Refer to Table 3.3.5.2-1 to determine which SRs apply for each RCIC Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Functions 4 and 5; and (b) for up to 6 hours for Functions 1, 2, and 3 provided the associated Function maintains RCIC initiation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.5.2.1 Perform CHANNEL CHECK.
| |
| SR 3.3.5.2.2 Perform CHANNEL FUNCTIONAL TEST.
| |
| SR 3.3.5.2.3 Calibrate the trip units. 92-days SR 3.3.5.2.4 Perform CHANNEL CALIBRATION.
| |
| SR 3.3.5.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST.
| |
| NMP2 3.3.5.2-3 Amendment 91
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS
| |
| ----------------------------- NOTES-------------------------------
| |
| : 1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment Isolation Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains isolation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.6.1.1 Perform CHANNEL CHECK. !2 hn'-;t-SR 3.3.6.1.2 --------------------- NOTE---------------
| |
| Only required to be met when the Allowable Value is adjusted in accordance with Table 3.3.6.1-1, footnote (b).
| |
| Verify the actual ambient temperature 12 h9uFG<
| |
| reading for all OPERABLE channels is
| |
| -> Tamb.
| |
| SR 3.3.6.1.3 Perform CHANNEL FUNCTIONAL TEST. 2days <
| |
| SR 3.3.6.1.4 Calibrate the trip unit. 92--s<
| |
| SR 3.3.6.1.5 Perform CHANNEL CALIBRATION. 24', cnths <-
| |
| SR 3.3.6.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST. 224me.4he-(continued)
| |
| NMP2 3.3.6.1-4 Amendment 91
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.6.1.7 --------------------- NOTE ---------------
| |
| The sensor response time may be assumed to be the design sensor response time.
| |
| Verify the ISOLATION SYSTEM RESPONSE TIME 24 months on a is within limits. STAGGERED TEST BAS TI JINSERT1 NMP2 3.3.6.1-5 Amendment 91
| |
| | |
| Secondary Containment Isolation Instrumentation 3.3.6.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.1.2 Declare associated 1 hour secondary containment isolation valves inoperable.
| |
| AND C.2.1 Place the associated 1 hour standby gas treatment (SGT) subsystem in operation.
| |
| OR C.2.2 Declare associated 1 hour SGT subsystem inoperable.
| |
| SURVEILLANCE REQUIREMENTS
| |
| ---------------------------- NOTES -------------------------------
| |
| : 1. Refer to Table 3.3.6.2-1 to determine which SRs apply for each Secondary Containment Isolation Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains isolation capability.
| |
| NMP2 3.3.6.2-2 Amendment 91
| |
| | |
| Secondary Containment Isolation Instrumentation 3.3.6.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.6.2.2 Perform CHANNEL FUNCTIONAL TEST. 2--days,,,,-
| |
| SR 3.3.6.2.3 Calibrate the trip unit. 92 dRYG*--, - INSERT 1 SR 3.3.6.2.4 Perform CHANNEL CALIBRATION. 24 m..t SR 3.3.6.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24-math-NMP2 3.3.6.2-3 Amendment 91
| |
| | |
| CREF System Instrumentation 3.3.7.1 SURVEILLANCE REQUIREMENTS
| |
| ----------------------------- NOTES -------------------------------
| |
| : 1. Refer to Table 3.3.7.1-1 to determine which SRs apply for each Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains CREF initiation capability.
| |
| NMP2 3.3.7.1-3 Amendment 91
| |
| | |
| Mechanical Vacuum Pump Isolation Instrumentation 3.3.7.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Isolate the 12 hours associated Completion associated mechanical Time not met. vacuum pump(s).
| |
| OR C.2 Remove the associated 12 hours mechanical vacuum pump breaker(s) from service.
| |
| OR C.3 Isolate the main 12 hours steam lines.
| |
| OR C.4 Be in MODE 3. 12 hours SURVEILLANCE REQUIREMENTS
| |
| ----------------------------- NOTE --------------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided mechanical vacuum pump isolation capability is maintained.
| |
| SURVEILLANCE FREQUENCY SR 3.3.7.2.1 Perform CHANNEL CHECK.
| |
| 12 hcrz< IINSERT 1 SR 3.3.7.2.2 Perform CHANNEL FUNCTIONAL TEST. F2-dmys (continued)
| |
| NMP2 3.3.7.2-2 Amendment 91
| |
| | |
| Mechanical Vacuum Pump Isolation Instrumentation 3.3.7.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.7.2.3 Perform CHANNEL CALIBRATION. The Allowable Value shall be < 3.6 x full power background. 2INSERT 1 J.
| |
| SR 3.3.7.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST :24-e~
| |
| including isolation valve and mechanical vacuum pump breakers actuation.
| |
| NMP2 3.3.7.2-3 Amendment 91
| |
| | |
| LOP Instrumentation 3.3.8.1 SURVEILLANCE REQUIREMENTS
| |
| ----------------------------- NOTES --------------------
| |
| : 1. Refer to Table 3.3.8.1-1 to determine which SRs apply for each LOP Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 2 hours provided the associated Function maintains LOP initiation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.1.1 Perform CHANNEL FUNCTIONAL TEST. 31 days SR 3.3.8.1.2 Perform CHANNEL CALIBRATION. 24 ..... ,.. IINSERT I i
| |
| SR 3.3.8.1.3 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 m.e NMP2 3.3.8.1-2 Amendment 91
| |
| | |
| RPS Electric Power Monitoring - Logic 3.3.8.2 SURVEILLANCE REQUIREMENTS
| |
| ----------------------------- NOTE --------------------------------
| |
| When an RPS electric power monitoring assembly is placed in an inoperable status solely for performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours provided the other RPS electric power monitoring assembly for the associated RPS logic bus maintains trip capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.2.1 Perform CHANNEL FUNCTIONAL TEST.
| |
| SR 3.3.8.2.2 Perform CHANNEL CALIBRATION. The Allowable Values shall be:
| |
| : a. Overvoltage (with time delay set to
| |
| < 2.5 seconds)
| |
| Bus A*< 130.2 V Bus B < 129.8 V
| |
| : b. Undervoltage (with time delay set to
| |
| < 2.5 seconds)
| |
| Bus A 115.5 V Bus B 114.2 V
| |
| : c. Underfrequency (with time delay set to
| |
| < 2.5 seconds)
| |
| Bus A 57.5 Hz Bus B > 57.5 Hz SR 3.3.8.2.3 Perform a system functional test. 24-meRths NMP2 3.3.8.2-4 Amendment 04, 99
| |
| | |
| RPS Electric Power Monitoring - Scram Solenoids 3.3.8.3 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Initiate action to Immediately associated Completion fully insert all Time of Condition A or insertable control B not met in MODE 5 rods in core cells with any control rod containing one or withdrawn from a core more fuel assemblies.
| |
| cell containing one or more fuel assemblies.
| |
| SURVEILLANCE REQUIREMENTS
| |
| -------------------------------- NOTE --------------------------------
| |
| When an RPS electric power monitoring assembly is placed in an inoperable status solely for performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours provided the other RPS electric power monitoring assembly for the associated RPS scram solenoid bus maintains trip capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.3.1 Perform CHANNEL FUNCTIONAL TEST. I. "4g'.
| |
| (feRtnlned)
| |
| INSERT 11 NMP2 3.3.8.3-2 Amendment 91
| |
| | |
| RPS Electric Power Monitoring - Scram Solenoids 3.3.8.3 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.3.2 Perform CHANNEL CALIBRATION. The 4.4-....
| |
| Allowable Values shall be:
| |
| : a. Overvoltage (with time delay set to
| |
| < 2.5 seconds)
| |
| Bus A*< 127.6 V Bus B *_127.6 V INSERT 1
| |
| : b. Undervoltage (with time delay set to
| |
| < 2.5 seconds)
| |
| Bus A 113.0 V Bus B 113.6 V
| |
| : c. Underfrequency (with time delay set to
| |
| < 2.5 seconds)
| |
| Bus A > 57.5 Hz Bus B 57.5 Hz SR 3.3.8.3.3 Perform a system functional test. 24-meMthi NMP2 3.3.8.3-3 Amendment Q-1-, 99
| |
| | |
| Recirculation Loops Operating 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 --------------------- NOTE -----------------------------
| |
| Not required to be performed until 24 hours after both recirculation loops are in operation.
| |
| Verify jet pump loop flow mismatch with 24-her-s -
| |
| both recirculation loops in operation is:
| |
| : a. < 10% of rated core flow when operating at an effective core flow \ 'I- INSERT 1
| |
| < 70% of rated core flow; and
| |
| : b. < 5% of rated core flow when operating at an effective core flow _>70% of rated core flow.
| |
| NMIP2 3.4.1-3 Amendment Q-1-, 92
| |
| | |
| FCVs 3.4.2 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 Flow Control Valves (FCVs)
| |
| LCO 3.4.2 A recirculation loop FCV shall be OPERABLE in each operating recirculation loop.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS NOTE Separate Condition entry is allowed for each FCV.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or two required A.1 Lock up the FCV. 4 hours FCVs inoperable.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 Verify each FCV fails "as is" on loss of 2.-.thsi..
| |
| hydraulic pressure at the hydraulic unit.
| |
| (continued)
| |
| JINSERT I NMP2 3.4.2-1 Amendment 91
| |
| | |
| FCVs 3.4.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.4.2.2 Verify average rate of each FCV movement is:
| |
| : a. < 11% of stroke per second for opening; and
| |
| : b. < 11% of stroke per second for closing.
| |
| NMP2 3.4.2-2 Amendment 91
| |
| | |
| Jet Pumps 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 --------------------- NOTES ---------------
| |
| : 1. Not required to be performed until 4 hours after associated recirculation loop is in operation.
| |
| : 2. Not required to be performed until 24 hours after > 23% RTP.
| |
| Verify at least two of the following 24- hGF n
| |
| criteria (a, b, and c) are satisfied for each operating recirculation loop:
| |
| : a. Jet pump loop flow versus flow control valve position differs by < 10% from established patterns.
| |
| : b. Jet pump loop flow versus recirculation loop drive flow differs by < 10% from established patterns.
| |
| : c. Each jet pump diffuser to lower plenum differential pressure differs by < 20%
| |
| from established patterns.
| |
| NMP2 3.4.3-2 Amendment 9-1-, 140
| |
| | |
| RCS Operational LEAKAGE 3.4.5 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND or B not met.
| |
| C.2 Be in MODE 4. 36 hours OR Pressure boundary LEAKAGE exists.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify RCS unidentified and identified 12 heurs LEAKAGE and unidentified LEAKAGE increase are within limits.
| |
| IINSERT I1 NMP2 3.4.5-2 Amendment 91
| |
| | |
| RCS Leakage Detection Instrumentation 3.4.7 SURVEILLANCE REQUIREMENTS
| |
| ---------------------- NOTE -----------------------------------------------------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the other required leakage detection instrumentation is OPERABLE.
| |
| SURVEILLANCE FREQUENCY i
| |
| SR 3.4.7.1 Perform CHANNEL CHECK of required drywell atmospheric monitoring system.
| |
| SR 3.4.7.2 Perform CHANNEL FUNCTIONAL TEST of the . d-as drywell floor drain tank fill rate monitoring system.
| |
| SR 3.4.7.3 Perform source check of required drywell .1 -days<
| |
| atmospheric monitoring system.
| |
| SR 3.4.7.4 Perform CHANNEL FUNCTIONAL TEST of required 184 days <
| |
| drywell atmospheric monitoring system.
| |
| SR 3.4.7.5 Perform CHANNEL CALIBRATION of required 24-meths leakage detection instrumentation.
| |
| NMP2 3.4.7-3 Amendment 94, 139
| |
| | |
| RCS Specific Activity 3.4.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2.2.1 Be in MODE 3. 12 hours AND B.2.2.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 ------------------- NOTE ----------------
| |
| Only required to be performed in MODE 1.
| |
| Verify reactor coolant DOSE 7-a EQUIVALENT 1-131 specific activity is
| |
| < 0.2 [tCi/gm.
| |
| [INS-ERT 1I NMP2 3.4.8-2 Amendment 91
| |
| | |
| RHR Shutdown Cooling System - Hot Shutdown 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 ----------------------- NOTE---------------
| |
| Not required to be met until 2 hours after reactor steam dome pressure is less than the RHR cut-in permissive pressure.
| |
| Verify one RHR shutdown cooling subsystem or recirculation pump is operating.
| |
| INSERT 1 NMP2 3.4.9-3 Amendment 91
| |
| | |
| RHR Shutdown Cooling System - Cold Shutdown 3.4.10 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. No RHR shutdown B.1 Verify reactor 1 hour from cooling subsystem in coolant circulating discovery of no operation. by an alternate reactor coolant method. circulation AND AND No recirculation pump in operation. Once per 12 hours thereafter AND B.2 Monitor reactor Once per hour coolant temperature and pressure.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify one RHR shutdown cooling subsystem '..h9Urs or recirculation pump is operating. I JINSERT 1 NMP2 3.4.10-2 Amendment 91
| |
| | |
| RCS P/T Limits 3.4.11 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. ------- NOTE -----------------
| |
| Required Action C.2 C.1 Initiate action to Immediately shall be completed if restore parameter(s) this Condition is to within limits.
| |
| entered.
| |
| AND Requirements of the C.2 Determine RCS is Prior to LCO not met in other acceptable for entering MODE 2 than MODES 1, 2, operation. or 3 and 3.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 --------------------- NOTE ---------------
| |
| Only required to be performed during RCS heatup and cooldown operations, and RCS system leakage and hydrostatic testing. JINSERT 1 Verify:
| |
| : a. RCS pressure and RCS temperature are within the applicable limits specified in the PTLR;
| |
| : b. RCS heatup and cooldown rates are within limits specified in the PTLR; and
| |
| : c. RCS temperature change during system leakage and hydrostatic testing is maintained within limits specified in the PTLR.
| |
| (continued)
| |
| NMP2 3.4.11-2 Amendment 94, 145
| |
| | |
| RCS P/T Limits 3.4.11 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.4.11.5 --------------------- NOTE ---------------
| |
| Only required to be met in single loop operation with THERMAL POWER < 30% RTP or the operating jet pump loop flow < 50%
| |
| rated jet pump loop flow.
| |
| Verify the difference between the bottom Once within 15 head coolant temperature and the RPV minutes prior coolant temperature is within limits specified in the to an increase PTLR. in THERMAL POWER or an increase in jet pump loop flow SR 3.4.11.6 --------------------- NOTE ---------------
| |
| Only required to be met in single loop operation when the idle recirculation loop is not isolated from-the RPV, and with THERMAL POWER < 30% RTP or the operating jet pump loop flow < 50% rated jet pump loop flow.
| |
| Verify the difference between the reactor Once within 15 coolant temperature in the recirculation minutes prior loop not in operation and the RPV coolant to an increase temperature is within limits specified in the PTLR. in THERMAL POWER or an increase in jet pump loop flow SR 3.4.11.7 --------------------- NOTE ---------------
| |
| Only required to be performed when tensioning the reactor vessel head bolting studs. JINSERT 1 Verify reactor vessel flange and head 3.. . ......
| |
| flange temperatures are within limits specified in the PTLR.
| |
| (continued)
| |
| NMP2 3.4.11-4 Amendment 94,145
| |
| | |
| RCS P/T Limits 3.4.11 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.4.11.8 --------------------- NOTE ----------------------------
| |
| Not required to be performed until 30 minutes after RCS temperature 5 80°F in MODE 4. INSERT 1 Verify reactor vessel flange and head 30 mntes flange temperatures are within limits specified in the PTLR.
| |
| SR 3.4.11.9 --------------------- NOTE ---------------
| |
| Not required to be performed until 12 hours after RCS temperature < 90'F in MODE 4. INSERT 1 Verify reactor vessel flange and head !2 hours flange temperatures are within limits specified in the PTLR.
| |
| NMP2 3.4.11-5 Amendment 94, 145
| |
| | |
| Reactor Steam Dome Pressure 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Reactor Steam Dome Pressure LCO 3.4.12 The reactor steam dome pressure shall be < 1035 psig.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes pressure not within dome pressure to limit, within limit.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify reactor steam dome pressure is
| |
| * 1035 psig.
| |
| [INSERT I NMP2 3.4.12-1 Amendment 91
| |
| | |
| ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
| |
| -4.
| |
| SR 3.5.1.1 Verify, for each ECCS injection/spray subsystem, the piping is filled with water from the pump discharge valve to the injection valve.
| |
| SR 3.5.1.2 ---------------
| |
| NOTE --------------
| |
| Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the residual heat removal cut-in permissive pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.
| |
| Verify each ECCS injection/spray subsystem manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
| |
| i SR 3.5.1.3 Verify:
| |
| : a. For each ADS nitrogen receiver discharge header, the pressure is > 160 psig; and
| |
| : b. For each ADS nitrogen receiver tank, the pressure is > 334 psig.
| |
| (continued)
| |
| NMP2 3.5.1-4 Amendment 91
| |
| | |
| ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY 4-SR 3.5.1.4 Verify each ECCS pump develops the In accordance specified flow rate with the specified with the developed head. Inservice Testing Program TOTAL SYSTEM FLOW RATE DEVELOPED HEAD LPCS _ 6350 gpm _ 284 psid LPCS A, B _ 7450 gpm _ 127 psid LPCI C > 7450 gpm _ 140 psid HPCS _ 6350 gpm _ 327 psid SR 3.5.1.5 ---------------------- NOTE ---------------
| |
| Vessel injection/spray may be excluded.
| |
| Verify each ECCS injection/spray subsystem 24 meRthS actuates on an actual or simulated automatic initiation signal.
| |
| SR 3.5.1.6 --------------------- NOTE ---------------- -IINSERT 1 I Valve actuation may be excluded.
| |
| Verify the ADS actuates on an actual or 24 Fnentl4s <
| |
| simulated automatic initiation signal.
| |
| SR 3.5.1.7 --------------------- NOTE ----------------
| |
| Not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test.
| |
| Verify each required ADS valve actuator strokes when manually actuated. S-TAGGtGE7RED TE ST RAS'S feF eaGh (continued)
| |
| NMP2 3.5.1-5 Amendment 91
| |
| | |
| ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.1.8 ---------------------- NOTE ----------------
| |
| Instrumentation response time may be assumed to be the design instrumentation response time.
| |
| Verify the ECCS RESPONSE TIME for each ECCS 24 -meRths injection/spray subsystem is within limits.
| |
| FINSERT1 NMP2 3.5.1-6 Amendment 91
| |
| | |
| ECCS - Shutdown 3.5.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action C.2 D.1 Initiate action to Immediately and associated restore secondary Completion Time not containment to met. OPERABLE status.
| |
| AND D.2 Initiate action to Immediately restore one standby gas treatment subsystem to OPERABLE status.
| |
| AND D.3 Initiate action to Immediately restore isolation capability in each required secondary containment penetration flow path not isolated.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for each required low pressure ECCS S injection/spray subsystem, the suppression pool water level is >_195 ft.
| |
| (continued)
| |
| FINSERT 1 1 NMP2 3.5.2-2 Amendment 91
| |
| | |
| ECCS - Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.2.2 Verify, for the required High Pressure Core !2 hOUrsF-Spray (HPCS) System, the:
| |
| : a. Suppression pool water level is
| |
| >_195 ft; or
| |
| : b. Condensate storage tank B water level is > 26.9 ft.
| |
| SR 3.5.2.3 Verify, for each required ECCS injection/ 31 days spray subsystem, the piping is filled with water from the pump discharge valve to the injection valve. [INSERT 1 SR 3.5.2.4 ------------------- NOTE ----------------------------------
| |
| One low pressure coolant injection (LPCI) subsystem may be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned and not otherwise inoperable.
| |
| Verify each required ECCS injection/spray subsystem manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
| |
| (continued)
| |
| NMP2 3.5.2-3 Amendment 91
| |
| | |
| ECCS - Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY i
| |
| SR 3.5.2.5 Verify each required ECCS pump develops the In accordance specified flow rate with the specified with the developed head. Inservice Testing TOTAL Program SYSTEM FLOW RATE DEVELOPED HEAD LPCS > 6350 gpm _ 284 psid LPCI A, B > 7450 gpm _ 127 psid LPCI C _ 7450 gpm _ 140 psid HPCS _ 6350 gpm > 327 psid i
| |
| SR 3.5.2.6 -NOTE Vessel injection/spray may be excluded.
| |
| Verify each required ECCS injection/spray 24-menths subsystem actuates on an actual or A simulated automatic initiation signal.
| |
| SR 3.5.2.7 -NOTE Instrumentation response time may be assumed to be the design instrumentation response time.
| |
| Verify the ECCS RESPONSE TIME for each ECCS injection/spray subsystem is within limits.
| |
| NMP2 3.5.2-4 Amendment 91
| |
| | |
| RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System piping is filled 3! days<
| |
| with water from the pump discharge valve to the injection valve.
| |
| SR 3.5.3.2 Verify each RCIC System manual, power ,1 day s..
| |
| operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
| |
| SR 3.5.3.3 --------------------- NOTE---------------
| |
| Not required to be performed until 12 hours JIET1 after reactor steam pressure and flow are adequate to perform the test.
| |
| Verify, with reactor pressure < 1035 psig 92 days<
| |
| and > 935 psig, the RCIC pump can develop a flow rate > 600 gpm against a system head corresponding to reactor pressure.
| |
| SR 3.5.3.4 --------------------- NOTE ---------------
| |
| Not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test.
| |
| Verify, with reactor pressure 5 165 psig, 24-RneRth 4:z the RCIC pump can develop a flow rate
| |
| > 600 gpm against a system head corresponding to reactor pressure.
| |
| (continued)
| |
| NMP2 3.5.3-2 Amendment 91
| |
| | |
| RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.3.5 ---------------------- NOTE ---------------
| |
| Vessel injection may be excluded.
| |
| Verify the RCIC System actuates on an 2".. .....
| |
| actual or simulated automatic initiation signal.
| |
| JINSERT 1 NMIP2 3.5.3-3 Amendment 91
| |
| | |
| Primary Containment 3.6.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.1.1 Perform required visual examinations and In accordance leakage rate testing except for primary with 10 CFR 50 containment air lock testing, in Appendix J accordance with 10 CFR 50 Appendix J Testing Program Testing Program Plan. Plan SR 3.6.1.1.2 Verify the drywell-to-suppression chamber In accordance bypass leakage rate is less than or equal with the to the equivalent leakage rate through a Type A testing orifice 0.0054 ft2 at an initial frequency of differential pressure of > 3 psid. the 10 CFR 50 Appendix J Testing Program Plan AND
| |
| -------- NOTE--------
| |
| Only required after two consecutive tests fail and continues until two consecutive tests pass 24 -. enths (continued)
| |
| JINSERT 1 NMP2 3.6.1.1-2 Amendment 91
| |
| | |
| Primary Containment 3.6.1.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.1.3 -------------------- NOTE ----------------
| |
| SR 3.6.1.1.2 may be performed in lieu of SR 3.6.1.1.3.
| |
| Verify, at an initial differential pressure of > 3 psid:
| |
| : a. The leakage rate through each drywell-to-suppression chamber bypass leak path containing suppression chamber-to-drywell vacuum breakers is I R less than or equal to the equivalent through an orifice 0.000648 ft2 ; and
| |
| : b. The combined leakage rate through all four drywell-to-suppression chamber bypass leak paths containing suppression chamber-to-drywell vacuum breakers is less than or equal to the equivalent through an orifice 0.001296 ft 2.
| |
| NMP2 3.6.1.1-3 Amendment 91
| |
| | |
| Primary Containment Air Locks 3.6.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.2.1 ---------------
| |
| 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 3.6.1.1.1.
| |
| Perform required primary containment air In accordance lock leakage rate testing in accordance with 10 CFR 50 with 10 CFR 50 Appendix J Testing Program Appendix J Plan. Testing Program Plan
| |
| +
| |
| SR 3.6.1.2.2 Verify only one door in the primary FRGRthS containment air lock can be opened at a N time.
| |
| FINSERT 1 NMP2 3.6.1.2-5 Amendment 91
| |
| | |
| PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i
| |
| SR 3.6.1.3.1 --------------------- NOTE---------------
| |
| Not required to be met when the 12 inch and 14 inch primary containment purge valves are open for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open, provided:
| |
| a) the Standby Gas Treatment (SGT)
| |
| System is OPERABLE; or b) the primary containment full flow line to the SGT System is isolated and one SGT subsystem is OPERABLE.
| |
| Verify each 12 inch and 14 inch primary containment purge valve is closed.
| |
| SR 3.6.1.3.2 --------------------- NOTES ---------------
| |
| : 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| : 2. Not required to be met for PCIVs that are open under administrative controls.
| |
| Verify each primary containment isolation 31-days.<-
| |
| manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed.
| |
| (continued)
| |
| NMP2 3.6.1.3-10 Amendment 91
| |
| | |
| PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.3.3 -------------------- NOTES ---------------
| |
| : 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| : 2. Not required to be met for PCIVs that are open under administrative controls.
| |
| Verify each primary containment isolation Prior to manual valve and blind flange that is entering MODE 2 located inside primary containment and or 3 from not locked, sealed, or otherwise secured MODE 4, if and is required to be closed during primary accident conditions is closed, containment was de-inerted while in MODE 4, if not performed within the previous 92 days SR 3.6.1.3.4 Verify continuity of the traversing incore probe (TIP) shear isolation valve INSERT 1 explosive charge.
| |
| SR 3.6.1.3.5 Verify the isolation time of each power In accordance operated, automatic PCIV, except MSIVs, with the is within limits. Inservice Testing Program (continued)
| |
| NMP2 3.6.1.3-11 Amendment 91
| |
| | |
| PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY i
| |
| SR 3.6.1.3.6 Perform leakage rate testing for each 144 days <-
| |
| primary containment purge valve with resilient seals. AND Once within 92 days after opening the valve
| |
| .4 SR 3.6.1.3.7 Verify the isolation time of each MSIV is In accordance
| |
| _>3 seconds and < 5 seconds. with the JINSERT 1 Inservice Testing Program SR 3.6.1.3.8 Verify each automatic PCIV actuates to the isolation position on an actual or simulated isolation signal.
| |
| SR 3.6.1.3.9 Verify a representative sample of reactor 24 meRth- <
| |
| instrumentation line EFCVs actuates to the isolation position on an actual or simulated instrument line break signal.
| |
| SR 3.6.1.3.10 Remove and test the explosive squib from 22 m4nthc on a each shear isolation valve of the TIP STAGGERED TEST System. BASIS SR 3.6.1.3.11 Verify the leakage rate for the secondary In accordance containment bypass leakage paths is with 10 CFR 50 within the limits of Table 3.6.1.3-1 when Appendix J pressurized to >_40 psig. Testing Program Plan (continued)
| |
| NMP2 3.6.1.3-12 Amendment Q-4, 96
| |
| | |
| Drywell and Suppression Chamber Pressure 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drywell and Suppression Chamber Pressure LCO 3.6.1.4 Drywell and suppression chamber pressure shall be
| |
| >_14.2 psia and < 15.45 psia.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell or suppression A.1 Restore drywell and 1 hour chamber pressure not suppression chamber within limits, pressure to within limits.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.4.1 Verify drywell and suppression chamber !2 h9-US pressure is within limits. I
| |
| [INSERT 1 NMP2 3.6.1.4-1 Amendment 91
| |
| | |
| Drywell Air Temperature 3.6.1.5 3.6 CONTAINMENT SYSTEMS 3.6.1.5 Drywell Air Temperature LCO 3.6.1.5 Drywell average air temperature shall be < 1500 F.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell average air A.1 Restore drywell 8 hours temperature not within average air limit. temperature to within limit.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.5.1 Verify drywell average air temperature is 2'4 h'-s within limit.
| |
| [INSERT 1 NMP2 3.6.1.5-1 Amendment 91
| |
| | |
| RHR Drywell Spray 3.6.1.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i
| |
| SR 3.6.1.6.1 Verify each RHR drywell spray subsystem manual and power operated valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.
| |
| SR 3.6.1.6.2 Verify, by administrative means, that each required RHR pump is OPERABLE.
| |
| SR 3.6.1.6.3 Verify each drywell spray nozzle is Following unobstructed, maintenance that could result in nozzle blockage NMP2 3.6.1.6-2 Amendment 94, 122
| |
| | |
| Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.7.1 ---------------
| |
| NOTES---------------------------
| |
| : 1. Not required to be met for vacuum breakers that are open during Surveillances.
| |
| : 2. Not required to be met for vacuum breakers open when performing their intended function.
| |
| Verify each vacuum breaker is closed.
| |
| 14 days',
| |
| SR 3.6.1.7.2 ---------------
| |
| -- NOTE --------------
| |
| Not required to be met for vacuum breaker 21SC*RV36B for the remainder of Cycle 9.
| |
| INSERT 1 Perform a functional test of each vacuum 31 dwgiqE breaker.
| |
| AND Within 12 hours after any discharge of steam to the suppression chamber from the safety/relief valves I.
| |
| SR 3.6.1.7.3 Verify the opening setpoint of each vacuum breaker is < 0.25 psid.
| |
| NMP2 3.6.1.7-3 Amendment 9-1-, 98, 108
| |
| | |
| Suppression Pool Average Temperature 3.6.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1.1 Verify suppression pool average temperature is within the applicable limits. AND 5 minutes when performing testing that adds heat to the suppression pool NMP2 3.6.2.1-3 Amendment 91
| |
| | |
| Suppression Pool Water Level 3.6.2.2 3.6 CONTAINMENT SYSTEMS 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression pool water level shall be _>199 ft 6 inches and
| |
| < 201 ft APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Suppression pool water A.1 Restore suppression 2 hours level not within pool water level to limits, within limits.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.2.1 Verify suppression pool water level is 24 rcur within limits.
| |
| JINSERT 1 NMP2 3.6.2.2-1 Amendment 91
| |
| | |
| RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling 31 days subsystem manual and power operated valve in the flow path that is not locked, sealed, or otherwise secured in position, INSER is in the correct position or can be aligned to the correct position.
| |
| SR 3.6.2.3.2 Verify each required RHR pump develops a In accordance flow rate > 7450 gpm through the with the associated heat exchanger while operating Inservice in the suppression pool cooling mode. Testing Program NMP2 3.6.2.3-2 Amendment 91
| |
| | |
| RHR Suppression Pool Spray 3.6.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.4.1 Verify each RHR suppression pool spray 1-days subsystem manual and power operated valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position or can be IINSERT 1 aligned to the correct position.
| |
| SR 3.6.2.4.2 Verify each required RHR pump develops a In accordance flow rate > 450 gpm while operating in with the the suppression pool spray mode. Inservice Testing Program NMP2 3.6.2.4-2 Amendment 91
| |
| | |
| Primary Containment Oxygen Concentration 3.6.3.2 3.6 CONTAINMENT SYSTEMS 3.6.3.2 Primary Containment Oxygen Concentration LCO 3.6.3.2 The primary containment oxygen concentration shall be
| |
| < 4.0 volume percent.
| |
| APPLICABILITY: MODE 1 during the time period:
| |
| : a. From 24 hours after THERMAL POWER is > 15% RTP following startup, to
| |
| : b. 24 hours prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment A.1 Restore oxygen 24 hours oxygen concentration concentration to not within limit, within limit.
| |
| B. Required Action and B.1 Reduce THERMAL POWER 8 hours associated Completion to < 15% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.2.1 Verify primary containment oxygen days concentration is within limits.
| |
| FINSERT 1 NMP2 3.6.3.2-1 Amendment 91
| |
| | |
| Secondary Containment 3.6.4.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. Secondary containment C.1 -------- NOTE------
| |
| inoperable during LCO 3.0.3 is not movement of recently irradiated applicable.
| |
| fuel assemblies in the secondary containment or during OPDRVs. Suspend movement of Immediately recently irradiated fuel assemblies in the secondary containment.
| |
| AND C.2 Initiate action to Immediately suspend OPDRVs.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is 2A4 h.-hrG
| |
| ..... INSER
| |
| _>0.25 inch of vacuum water gauge.
| |
| SR 3.6.4.1.2 Verify all secondary containment ,1 days equipment hatches are closed and sealed.
| |
| (continued)
| |
| NMP2 3.6.4.1-2 Amendment Q-4, 101
| |
| | |
| Secondary Containment 3.6.4.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY
| |
| +
| |
| SR 3.6.4.1.3 Verify one secondary containment access door in each access opening is closed.
| |
| +
| |
| SR 3.6.4.1.4 Verify the secondary containment can be 24 monOthe on a drawn down to > 0.25 inch of vacuum STA^GGERED TEST water gauge in < 66.7 seconds using one BASIS forOach <
| |
| standby gas treatment (SGT) subsystem. -G-T su*bsystefa SR 3.6.4.1.5 Verify the secondary containment can be maintained > 0.25 inch of vacuum water STAGGERED TEST gauge for 1 hour using one SGT subsystem A SIS for o*r at a flow rate < 2670 cfm. SG- sUI§ysteF JINrT1 NMP2 3.6.4.1-3 Amendment 91
| |
| | |
| SCIVs 3.6.4.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.2.1 ---------------
| |
| NOTES------------
| |
| : 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| : 2. Not required to be met for SCIVs that are open under administrative controls.
| |
| Verify each secondary containment 3 1 days isolation manual valve and blind flange that is not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed.
| |
| SR 3.6.4.2.2 Verify the isolation time of each power 2-days operated, automatic SCIV is within limits.
| |
| SR 3.6.4.2.3 Verify each automatic SCIV actuates to the isolation position on an actual or simulated automatic isolation signal.
| |
| NMP2 3.6.4.2-4 Amendment 91
| |
| | |
| SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. (continued) AND E.3 Initiate action to Immediately suspend OPDRVs.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SIR 3.6.4.3.1 Operate each SGT subsystem for 34-day's <-
| |
| >_10 continuous hours with heaters operating.
| |
| 1*
| |
| SR 3.6.4.3.2 Perform required SGT filter testing in In accordance accordance with the Ventilation Filter with the VFTP INSERT 1 Testing Program (VFTP).
| |
| +
| |
| SR 3.6.4.3.3 Verify each SGT subsystem actuates on an 24 FAeR <
| |
| actual or simulated initiation signal.
| |
| +
| |
| SR 3.6.4.3.4 Verify each SGT decay heat removal air 24-me th inlet valve can be opened.
| |
| NMP2 3.6.4.3-3 Amendment 9-4, 101
| |
| | |
| SW System and UHS 3.7.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i
| |
| SR 3.7.1.1 --------------------- NOTE Not required to be met if SR 3.7.1.5 and SR 3.7.1.8 satisfied.
| |
| Verify the water temperature of the intake I P heIr*;
| |
| tunnels is > 38°F.
| |
| SR 3.7.1.2 Verify the water level in the SW pump 24- h','S intake bay is >_233.1 ft.
| |
| SR 3.7.1.3 Verify the water temperature of each SW subsystem supply header is < 84°F.
| |
| AND 4 hours when supply header water temperature is
| |
| > 78 0 F SR 3.7.1.4 Verify each required SW pump is in 24-h9UFG<
| |
| operation.
| |
| (continued)
| |
| NMP2 3.7.1-3 Amendment 9.1-, 119
| |
| | |
| SW System and UHS 3.7.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.7.1.5 ---------------------- NOTE ---------------
| |
| Not required to be met if SR 3.7.1.1 satisfied.
| |
| Verify, for each intake deicer heater 7-eye dy division, the current of each required heater feeder cable is within the limit.
| |
| SR 3.7.1.6 -- ------------------- NOTE ---------------
| |
| Isolation of flow to individual components does not render SW System inoperable.
| |
| Verify each SW subsystem manual, power 31 days <
| |
| operated, and automatic valve in the flow path servicing safety related systems or components, that is not locked, sealed, or F otherwise secured in position, is in the correct position.
| |
| SR 3.7.1.7 Verify each SW subsystem actuates on an 24- M.eithe actual or simulated initiation signal.
| |
| SR 3.7.1.8 ---------------------- NOTE ---------------
| |
| Not required to be met if SR 3.7.1.1 satisfied.
| |
| Verify, for each intake deicer heater 24 meRth& <
| |
| division, the resistance of each required heater feeder cable and associated heater elements is within the limit.
| |
| NMIP2 3.7.1-4 Amendment 91
| |
| | |
| CREF System 3.7.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME F. Two CREF subsystems --------------- NOTE ---------
| |
| inoperable with safety LCO 3.0.3 is not applicable.
| |
| function not maintained during movement of recently irradiated fuel assemblies in F.1 Suspend movement of Immediately the secondary containment recently irradiated or during OPDRVs. fuel assemblies in the secondary OR containment.
| |
| One or more CREF AND subsystems inoperable due to inoperable CRE boundary F.2 Initiate action to suspend Immediately during movement of recently OPDRVs.
| |
| irradiated fuel assemblies in the secondary containment or during OPDRVs.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Operate each CREF subsystem for _>1 continuous hour.
| |
| SR 3.7.2.2 Perform required CREF System filter testing in accordance with the Ventilation Filter Testing Program (VFTP).
| |
| SR 3.7.2.3 Verify each CREF subsystem actuates on an actual or simulated initiation signal.
| |
| (continued)
| |
| NMP2 3.7.2-3 Amendment 91, 95, 97, 12*, 126
| |
| | |
| Control Room Envelope AC System 3.7.3 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME H. Required Action and ------------ NOTE ----------
| |
| associated Completion LCO 3.0.3 is not applicable.
| |
| Time of Condition C or D not met during movement of recently irradiated fuel H.1 Suspend movement of Immediately assemblies in the recently irradiated fuel secondary containment assemblies in the or during OPDRVs. secondary containment.
| |
| AND H.2 Initiate action to suspend OPDRVs. Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Verify each control room envelope AC subsystem has the capability to remove the assumed heat load for the Main Control Room area and the Relay Room area.
| |
| [INSERT 1 NMP2 3.7.3-4 Amendment 128
| |
| | |
| Main Condenser Offgas 3.7.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 --------------------- NOTE --------------
| |
| Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.
| |
| Verify the gross gamma activity rate of the noble gases is < 350,000 pCi/second after decay of 30 minutes. AND Once within 4 hours after a
| |
| > 50% increase in the nominal steady state fission gas release after factoring out increases due to changes in THERMAL POWER level NMP2 3.7.4-2 Amendment 91
| |
| | |
| Main Turbine Bypass System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Main Turbine Bypass System LCO 3.7.5 The Main Turbine Bypass System shall be OPERABLE.
| |
| OR LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR, are made applicable.
| |
| APPLICABILITY: THERMAL POWER > 23% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A. 1 Satisfy the 2 hours LCO not met. requirements of the LCO.
| |
| B. Required Action and B.1 Reduce THERMAL POWER 4 hours associated Completion to < 23% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Perform a system functional test. 2I mtNSERT SR 3.7.5.2 Verify the TURBINE BYPASS SYSTEM RESPONSE 24"-e.ths TIME is within limits.
| |
| NMP2 3.7.5-1 Amendment 9-1-, 140
| |
| | |
| Spent Fuel Storage Pool Water Level 3.7.6 3.7 PLANT SYSTEMS 3.7.6 Spent Fuel Storage Pool Water Level LCO 3.7.6 The spent fuel storage pool water level shall be _>22 ft 3 inches over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks.
| |
| APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel storage pool, During movement of new fuel assemblies in the spent fuel storage pool with irradiated fuel assemblies seated in the spent fuel storage pool.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel storage A.1 - --------- NOTE------
| |
| pool water level not LCO 3.0.3 is not within limit, applicable.
| |
| Suspend movement of Immediately fuel assemblies in the spent fuel storage pool.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify the spent fuel storage pool water level is >_22 ft 3 inches over the top of irradiated fuel assemblies seated in the S spent fuel storage pool racks.
| |
| NMP2 3.7.6-1 Amendment 91
| |
| | |
| AC Sources - Operating 3.8.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME E. Two required DGs E.1 Restore one required 2 hours inoperable. DG to OPERABLE status. OR 24 hours if Division 3 DG is inoperable F. Required Action and F.1 Be in MODE 3. 12 hours Associated Completion Time of Condition A, AND B, C, D, or E not met.
| |
| F.2 Be in MODE 4. 36 hours G. Three or more required G.1 Enter LCO 3.0.3. Immediately AC sources inoperable.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power availability for each required offsite circuit.
| |
| kcontinued)
| |
| IINSERT I NMP2 3.8.1-5 Amendment 91
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.2 NOTE ---------------
| |
| All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.
| |
| Verify each required DG starts from standby conditions and achieves:
| |
| : a. In < 10 seconds, voltage _>3950 V for Division 1 and 2 DGs and _>3820 V for Division 3 DG, and frequency > 58.8 Hz for Division 1 and 2 DGs and > 58.0 Hz for Division 3 DG; and
| |
| : b. Steady state voltage >_3950 V and
| |
| < 4370 V and frequency >_58.8 Hz and
| |
| < 61.2 Hz.
| |
| (continued)
| |
| NMP2 3.8.1-6 Amendment 91
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.3 NOTES ---------------
| |
| : 1. DG loadings may include gradual loading as recommended by the manufacturer.
| |
| : 2. Momentary transients outside the load range do not invalidate this test.
| |
| : 3. This Surveillance shall be conducted on only one DG at a time.
| |
| : 4. This SR shall be preceded by, and immediately follow, without shutdown, a successful performance of SR 3.8.1.2.
| |
| Verify each required DG is synchronized and 31 days loaded and operates for _>60 minutes at a load __3960 kW and _<4400 kW for Division 1 and 2 DGs, and _>2340 kW and *< 2600 kW for Division 3 DG.
| |
| SR 3.8.1.4 Verify each required day tank contains
| |
| >_403 gal of fuel oil for Division 1 and 2 DGs and Ž_282 gal for Division 3 DG.
| |
| SR 3.8.1.5 Check for and remove accumulated water from each required day tank.
| |
| SR 3.8.1.6 Verify each required fuel oil transfer subsystem operates to automatically transfer fuel oil from the storage tank to the day tank.
| |
| (continued)
| |
| NMP2 3.8.1-7 Amendment 91
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.7 NOTES ---------------
| |
| : 1. This Surveillance shall not be performed in MODE 1 or 2 (not applicable to Division 3 DG). However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 2. If performed with DG synchronized with offsite power, it shall be performed within the power factor limit. However if grid conditions do not permit, the power factor limit is not required to be met. Under this condition the power factor shall be maintained as close to the limit as practicable.
| |
| Verify each required DG rejects a load greater than or equal to its associated single largest post-accident load, and following load rejection, the frequency is
| |
| < 64.5 Hz for Division 1 and 2 DGs and
| |
| < 66.75 Hz for Division 3 DG.
| |
| (continued)
| |
| NMP2 3.8.1-8 Amendment @-1-, 133
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.8 NOTES ---------------
| |
| : 1. This Surveillance shall not be performed in MODE 1 or 2 (not applicable to Division 3 DG). However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 2. If grid conditions do not permit, the power factor limit is not required to be met. Under this condition the power factor shall be maintained as close to the limit as practicable.
| |
| Verify each required DG operating within 24 R~the the power factor limit does not trip and voltage is maintained:
| |
| : a. <_4576 V during and following a load rejection of a load Ž_4400 kW for Division 1 and 2 DGs; and
| |
| : b. < 5824 V during and following a load rejection of a load >_2600 kW for Division 3 DG.
| |
| (continued)
| |
| NMIP2 3.8.1-9 Amendment 9-1-, 133
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.9 ---------------------- NOTES ---------------
| |
| : 1. All DG starts may be preceded by an engine prelube period.
| |
| : 2. This Surveillance shall not be performed in MODE 1, 2, or 3 (not INSERT1 applicable to Division 3 DG).
| |
| However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify on an actual or simulated loss of 24 ent offsite power signal:
| |
| : a. De-energization of emergency buses;
| |
| : b. Load shedding from emergency buses for Divisions 1 and 2 only; and
| |
| : c. DG auto-starts from standby condition and:
| |
| : 1. energizes permanently connected loads in < 13.20 seconds,
| |
| : 2. energizes auto-connected shutdown loads for Division 1 and 2 DGs only, through the associated automatic load sequence time delay relays,
| |
| : 3. maintains steady state voltage
| |
| > 3950 V and:< 4370 V,
| |
| : 4. maintains steady state frequency
| |
| > 58.8 Hz and < 61.2 Hz, and
| |
| : 5. supplies permanently connected and auto-connected shutdown loads for
| |
| > 5 minutes for Division 1 and 2 DGs and supplies permanently connected shutdown loads for
| |
| > 5 minutes for Division 3 DG.
| |
| (continued)
| |
| NMP2 3.8.1-10 Amendment 94, 133
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.10 --------------------- NOTES ---------------
| |
| : 1. All DG starts may be preceded by an engine prelube period.
| |
| INSERT 1
| |
| : 2. This Surveillance shall not be performed in MODE 1 or 2 (not applicable to Division 3 DG). However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify on an actual or simulated Emergency 24"!,/Rh.
| |
| Core Cooling System (ECCS) initiation signal each required DG auto-starts from standby condition and:
| |
| : a. In < 10 seconds after auto-start, achieves voltage > 3950 V for Division 1 and 2 DGs and _>3820 V for Division 3 DG, and frequency >_58.8 Hz for Division 1 and 2 DGs and >_58.0 Hz for Division 3 DG;
| |
| : b. Achieves steady state voltage >_3950 V and < 4370 V and frequency > 58.8 Hz and < 61.2 Hz;
| |
| : c. Operates for _>5 minutes;
| |
| : d. Permanently connected loads remain energized from the offsite power system for Divisions 1 and 2 only; and
| |
| : e. Emergency loads are auto-connected through the associated automatic load sequence time delay relays to the offsite power system for Divisions 1 and 2 only.
| |
| (continued)
| |
| NMP2 3.8.1-11 Amendment -4,133
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY i
| |
| SR 3.8.1.11 NOTE ---------------
| |
| This Surveillance shall not be performed in MODE 1, 2, or 3 (not applicable to Division 3 DG). However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify each required DG's automatic trips are bypassed on actual or simulated loss of voltage signal on the emergency bus concurrent with an actual or simulated ECCS initiation signal except:
| |
| : a. Engine overspeed; and
| |
| : b. Generator differential current.
| |
| (continued)
| |
| NMP2 3.8.1-12 Amendment 4-, 133
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.12 --------------------- NOTES ---------------
| |
| : 1. Momentary transients outside the load and power factor ranges do not invalidate this test.
| |
| : 2. This Surveillance shall not be performed in MODE 1 or 2 unless the other two DGs are OPERABLE. If either of the other two DGs become inoperable, this Surveillance shall be suspended. However, credit may be taken for unplanned events that INSERT 1I satisfy this SR.
| |
| : 3. If grid conditions do not permit, the power factor limit is not required to be met. Under this condition the power factor shall be maintained as close to the limit as practicable.
| |
| Verify each required DG operating within -4 FX-Rth, the power factor limit operates for
| |
| _>24 hours:
| |
| : a. For __2 hours loaded __4620 kW and
| |
| < 4840 kW for Division 1 and 2 DGs, and _>2730 kW and < 2860 kW for Division 3 DG; and
| |
| : b. For the remaining hours of the test loaded __3960 kW and __4400 kW for Division 1 and 2 DGs, and > 2340 kW and < 2600 kW for Division 3 DG.
| |
| (continued)
| |
| NMP2 3.8.1-13 Amendment 91
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.13 NOTES ---------------
| |
| : 1. This Surveillance shall be performed within 5 minutes of shutting down the DG after the DG has operated _>2 hours loaded > 3960 kW for Division 1 and 2 DGs, and > 2340 kW for Division 3 DG.
| |
| -~T1 Momentary transients below the load limit do not invalidate this test.
| |
| : 2. All DG starts may be preceded by an engine prelube period.
| |
| Verify each required DG starts and 24) 4ARth-q achieves:
| |
| : a. In _<10 seconds, voltage >_3950 V for Division 1 and 2 DGs and >_3820 V for Division 3 DG, and frequency _ 58.8 Hz for Division 1 and 2 DGs and > 58.0 Hz for Division 3 DG; and
| |
| : b. Steady state voltage >_3950 V and
| |
| < 4370 V and frequency _>58.8 Hz and < 61.2 Hz.
| |
| (continued)
| |
| NMP2 3.8.1-14 Amendment 91
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY i
| |
| SR 3.8.1.14 NOTE ---------------
| |
| This Surveillance shall not be performed in MODE 1, 2, or 3 (not applicable to Division 3 DG). However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify each required DG: 24 mReR~th
| |
| : a. Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite power;
| |
| : b. Transfers loads to offsite power source; and
| |
| : c. Returns to ready-to-load operation.
| |
| SR 3.8.1.15 NOTE ---------------
| |
| This Surveillance shall not be performed in MODE 1, 2, or 3 (not applicable to Division 3 DG). However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify, with a DG operating in test mode 24-menth and connected to its bus, an actual or simulated ECCS initiation signal overrides the test mode by:
| |
| : a. Returning DG to ready-to-load operation; and
| |
| : b. Automatically energizing the emergency load from offsite power.
| |
| (continued)
| |
| NMIP2 3.8.1-15 Amendment 94, 133
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY
| |
| +
| |
| SR 3.8.1.16 NOTE ---------------
| |
| This Surveillance shall not be performed in MODE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify interval between each sequenced load block, for the Division 1 and 2 DGs only, is > 90% of the design interval for each automatic load sequence time delay relay.
| |
| (continued)
| |
| NMP2 3.8.1-16 Amendment 91
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.17 --------------------- NOTES ---------------
| |
| : 1. All DG starts may be preceded by an engine prelube period.
| |
| : 2. This Surveillance shall not be performed in MODE 1, 2, or 3 (not INSERT 1 applicable to Division 3 DG).
| |
| However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify, on an actual or simulated loss of OA nth-offsite power signal in conjunction with an actual or simulated ECCS initiation signal:
| |
| : a. De-energization of emergency buses;
| |
| : b. Load shedding from emergency buses for Divisions 1 and 2 only; and
| |
| : c. DG auto-starts from standby condition and:
| |
| : 1. energizes permanently connected loads in < 10 seconds,
| |
| : 2. for Divisions 1 and 2, energizes auto-connected emergency loads through the associated automatic load sequence time delay relays and for Division 3, energizes auto-connected emergency loads,
| |
| : 3. maintains steady state voltage
| |
| >_3950 V and:< 4370 V,
| |
| : 4. maintains steady state frequency
| |
| > 58.8 Hz and < 61.2 Hz, and
| |
| : 5. supplies permanently connected and auto-connected emergency loads for
| |
| > 5 minutes.
| |
| (continued)
| |
| NMP2 3.8.1-17 Amendment 9-1-, 133
| |
| | |
| AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.18 T-NO E -----------------------------
| |
| All DG starts may be preceded by an engine INSERT 1 prelube period.
| |
| Verify, when started simultaneously from ',-aF~s standby condition, each Division 1, 2, and 3 DG achieves, in < 10 seconds, voltage
| |
| > 3950 V for Division 1 and 2 DGs and
| |
| _ 3820 V for Division 3 DG, and frequency
| |
| _ 58.8 Hz for Division 1 and 2 DGs and
| |
| _ 58.0 Hz for Division 3 DG.
| |
| NMP2 3.8.1-18 Amendment 91
| |
| | |
| Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify each fuel oil storage tank contains: days
| |
| ,1
| |
| : a. > 50,000 gal of fuel for Division 1 DG and Division 2 DG; and
| |
| : b. > 35,342 gal of fuel for Division 3 DG.
| |
| SR 3.8.3.2 Verify lube oil inventory is:- 1*days <
| |
| : a. > 99 gal for Division 1 DG and Division 2 DG; and
| |
| : b. _ 168 gal for Division 3 DG.
| |
| SR 3.8.3.3 Verify fuel oil properties of new and In accordance stored fuel oil are tested in accordance with the Diesel with, and maintained within the limits of, Fuel Oil the Diesel Fuel Oil Testing Program. Testing Program SR 3.8.3.4 Verify each DG air start receiver pressure 31 days<
| |
| is:
| |
| : a. _ 225 psig for Division 1 DG and Division 2 DG; and
| |
| : b. _>190 psig for Division 3 DG.
| |
| SR 3.8.3.5 Check for and remove accumulated water from 31 days<
| |
| each fuel oil storage tank.
| |
| NMP2 3.8.3-3 Amendment 91
| |
| | |
| DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is _>130 V on float charge.
| |
| battery Verify no visible corrosion at SR 3.8.4.2 Verify no visible corrosion at battery terminals and connectors.
| |
| OR Verify battery connection resistance is
| |
| < 20% above the resistance as measured during installation for intercell and terminal connections.
| |
| i SR 3.8.4.3 Verify battery cells, cell plates, and racks show no visual indication of physical damage or abnormal deterioration that could degrade battery performance.
| |
| SR 3.8.4.4 Remove visible corrosion and verify battery 2"4me. then cell to cell and terminal connections are coated with anti-corrosion material.
| |
| SR 3.8.4.5 Verify battery connection resistance is 24, mRthe
| |
| < 20% above the resistance as measured during installation for intercell and terminal connections.
| |
| (continued)
| |
| NMP2 3.8.4-2 Amendment 91
| |
| | |
| DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.6 Verify each required Division 1 and 2 24 mnths<
| |
| battery charger supplies >_300 amps and the required Division 3 battery charger supplies _> 40 amps at >_130 V for
| |
| _Ž4 hours.
| |
| SR 3.8.4.7 ---------------------- NOTES ---------------
| |
| : 1. The modified performance discharge test in SR 3.8.4.8 may be performed in'INSE lieu of the service test in SR 3.8.4.7 provided the modified performance discharge test completely envelops the service test.
| |
| : 2. This Surveillance shall not be performed in MODE 1, 2, or 3 (not applicable to Division 3). However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify battery capacity is adequate to 24 monthc<
| |
| supply, and maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test.
| |
| (continued)
| |
| NMP2 3.8.4-3 Amendment -4, 136
| |
| | |
| DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.8 ---------------------- NOTE ---------------
| |
| This Surveillance shall not be performed in MODE 1, 2, or 3 (not applicable to Division 3).
| |
| However, credit may be taken for INSERT1 unplanned events that satisfy this SR.
| |
| Verify battery capacity is > 80% of the 60,MeRthS manufacturer's rating when subjected to a performance discharge test or a modified AND performance discharge test.
| |
| 12 months when battery shows degradation or has reached 85%
| |
| of 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 NMP2 3.8.4-4 Amendment 9-1-, 136
| |
| | |
| Battery Cell Parameters 3.8.6 ACTIONS (continued)
| |
| 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
| |
| < 650 F.
| |
| OR One or more batteries with one or more battery cell parameters not within Table 3.8.6-1 Category C limits.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters meet Table 3.8.6-1 Category A limits.
| |
| (continued)
| |
| INSERT 1 NMP2 3.8.6-2 Amendment 91
| |
| | |
| Battery Cell Parameters 3.8.6 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.6.2 Verify battery cell parameters meet Table 3.8.6-1 Category B limits.
| |
| AND Once within 7 days after battery discharge
| |
| < 107 V
| |
| [INSERT 1I AND Once within 7 days after battery overcharge
| |
| > 142 V 4
| |
| SR 3.8.6.3 Verify average electrolyte temperature of representative cells is - 65°F.
| |
| NMP2 3.8.6-3 Amendment 91
| |
| | |
| Inverters - Operating 3.8.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct emergency UPS inverter voltage, frequency, and alignment to 120 VAC uninterruptible panels.
| |
| -- INSERT I1 NMP2 3.8.7-2 Amendment 91
| |
| | |
| Distribution Systems - Operating 3.8.8 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME F. Two or more electrical F.1 Enter LCO 3.0.3. Immediately power distribution subsystems inoperable that result in a loss of function.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignment and power 7-A availability to required AC, DC, and 120 VAC uninterruptible electrical power distribution subsystems.
| |
| 1INSERT 1 NMP2 3.8.8-3 Amendment 91
| |
| | |
| Distribution Systems - Shutdown 3.8.9 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.3 Initiate action to Immediately suspend operations with a potential for draining the reactor vessel.
| |
| AND A.2.4 Initiate actions to Immediately restore required AC, DC, and 120 VAC uninterruptible electrical power distribution subsystems to OPERABLE status.
| |
| AND A.2.5 Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and power and availability to required AC, DC, 120 VAC uninterruptible electrical power distribution subsystems.
| |
| NMP2 3.8.9-2 Amendment 91
| |
| | |
| Refueling Equipment Interlocks 3.9.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Perform CHANNEL FUNCTIONAL TEST on each of the following required refueling equipment interlock inputs:
| |
| : a. All-rods-in,
| |
| : b. Refueling platform position,
| |
| : c. Refueling platform fuel grapple, fuel-loaded,
| |
| : d. Refueling platform monorail hoist, fuel-loaded,
| |
| : e. Refueling platform frame-mounted hoist, fuel-loaded, and
| |
| : f. Service platform hoist, fuel-loaded.
| |
| NMP2 3.9.1-2 Amendment 91
| |
| | |
| Refuel Position One-Rod-Out Interlock 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Refuel Position One-Rod-Out Interlock LCO 3.9.2 The refuel position one-rod-out interlock shall be OPERABLE.
| |
| APPLICABILITY: MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refuel position one- A.1 Suspend control rod Immediately rod-out interlock withdrawal.
| |
| inoperable.
| |
| AND A.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Verify reactor mode switch locked in refuel 2 '-
| |
| position.
| |
| (continued)
| |
| --INSERT 1I NMP2 3.9.2-1 Amendment 91
| |
| | |
| Refuel Position One-Rod-Out Interlock 3.9.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.9.2.2 ---------------------- NOTE ---------------
| |
| Not required to be performed until 1 hour after any control rod is withdrawn.
| |
| Perform CHANNEL FUNCTIONAL TEST. 7-deye
| |
| --INSERT 1 NMP2 3.9.2-2 Amendment 91
| |
| | |
| Control Rod Position 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Control Rod Position LCO 3.9.3 All control rods shall be fully inserted.
| |
| APPLICABILITY: When loading fuel assemblies into the core.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more control A.1 Suspend loading fuel Immediately rods not fully assemblies into the inserted, core.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify all control rods are fully inserted.
| |
| NMP2 3.9.3-1 Amendment 91
| |
| | |
| Control Rod OPERABILITY - Refueling 3.9.5 3.9 REFUELING OPERATIONS 3.9.5 Control Rod OPERABILITY - Refueling LCO 3.9.5 Each withdrawn control rod shall be OPERABLE.
| |
| APPLICABILITY: MODE 5.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more withdrawn A.1 Initiate action to Immediately control rods fully insert inoperable, inoperable withdrawn control rods.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1 ------------------- NOTE ---------------------------- --- INSERT 1 Not required to be performed until 7 days after the control rod is withdrawn.
| |
| Insert each withdrawn control rod at least one notch.
| |
| 7 dey SR 3.9.5.2 Verify each withdrawn control rod scram days accumulator pressure is > 940 psig.
| |
| NMP2 3.9.5-1 Amendment 91
| |
| | |
| RPV Water Level - Irradiated Fuel 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Reactor Pressure Vessel (RPV) Water Level - Irradiated Fuel LCO 3.9.6 RPV water level shall be >_22 ft 3 inches above the top of the RPV flange.
| |
| APPLICABILITY: During movement of irradiated fuel assemblies within the RPV.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RPV water level not A.1 Suspend movement of Immediately within limit, irradiated fuel assemblies within the RPV.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify RPV water level is > 22 ft 3 inches 2 h-rS above the top of the RPV flange.
| |
| -FINSERT 1 NMP2 3.9.6-1 Amendment 91
| |
| | |
| RPV Water Level - New Fuel or Control Rods 3.9.7 3.9 REFUELING OPERATIONS 3.9.7 Reactor Pressure Vessel (RPV) Water Level - New Fuel or Control Rods LCO 3.9.7 RPV water level shall be > 22 ft 3 inches above the top of irradiated fuel assemblies seated within the RPV.
| |
| APPLICABILITY: During movement of new fuel assemblies or handling of control rods within the RPV when irradiated fuel assemblies are seated within the RPV.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RPV water level not A.1 Suspend movement of Immediately within limit, new fuel assemblies and handling of control rods within the RPV.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify RPV water level is > 22 ft 3 inches -$
| |
| above the top of irradiated fuel assemblies seated within the RPV.
| |
| --[INSERT 1 NMP2 3.9.7-1 Amendment 91
| |
| | |
| RHR - High Water Level 3.9.8 NMP2 3.9.8-3 Amendment 91
| |
| | |
| RHR - Low Water Level 3.9.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.9.1 Verify one RHR shutdown cooling subsystem 7 eWhc-is operating. G f Lýýý NMP2 3.9.9-3 Amendment 91
| |
| | |
| Reactor Mode Switch Interlock Testing 3.10.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3.1 Place the reactor 1 hour mode switch in the shutdown position.
| |
| OR A.3.2-------- NOTE-------
| |
| Only applicable in MODE 5.
| |
| Place the reactor 1 hour mode switch in the refuel position.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.2.1 Verify all control rods are fully inserted 12 heE in core cells containing one or more fuel assemblies. IN-SET R1 I
| |
| I SR 3.10.2.2 Verify no CORE ALTERATIONS are in progress. 24 heu~s<
| |
| NMP2 3.10.2-2 Amendment 91
| |
| | |
| Single Control Rod Withdrawal - Hot Shutdown 3.10.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.3.1 Perform the applicable SRs for the required According to LCOs. the applicable SRs SR 3.10.3.2 -------------------- NOTE -----------------
| |
| Not required to be met if SR 3.10.3.1 is INSERT 1 satisfied for LCO 3.10.3.d. 1 requirements.
| |
| Verify all control rods, other than the 24-heu~s control rod being withdrawn, in a five by five array centered on the control rod being withdrawn, are disarmed.
| |
| SR 3.10.3.3 Verify all control rods, other than the control rod being withdrawn, are fully inserted.
| |
| NMP2 3.10.3-3 Amendment 91
| |
| | |
| Single Control Rod Withdrawal - Cold Shutdown 3.10.4 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. One or more of the B.1 Suspend withdrawal of Immediately above requirements not the control rod and met with the affected removal of associated control rod not CRD.
| |
| insertable.
| |
| AND B.2.1 Initiate action to Immediately fully insert all control rods.
| |
| OR B.2.2 Initiate action to Immediately satisfy the requirements of this LCO.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.4.1 Perform the applicable SRs for the required According to LCOs. applicable SRs SR 3.10.4.2 --------------------- NOTE---------------------- INSERT1 Not required to be met if SR 3.10.4.1 is satisfied for LCO 3.10.4.c.1 requirements.
| |
| Verify all control rods, other than the 24 heU'-
| |
| control rod being withdrawn, in a five by five array centered on the control rod being withdrawn, are disarmed.
| |
| (continued)
| |
| NMP2 3.10.4-3 Amendment 91
| |
| | |
| Single Control Rod Withdrawal - Cold Shutdown 3.10.4 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.10.4.3 Verify all control rods, other than the 24-heu <
| |
| control rod being withdrawn, are fully inserted.
| |
| SR 3.10.4.4 ---------------
| |
| NOTE--------------- INSERT 1 Not required to be met if SR 3.10.4.1 is satisfied for LCO 3.10.4.b. 1 requirements.
| |
| Verify a control rod withdrawal block is 21 hAou-r,<
| |
| inserted.
| |
| NMP2 3.10.4-4 Amendment 91
| |
| | |
| Single CRD Removal - Refueling 3.10.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.1 Initiate action to Immediately fully insert all control rods.
| |
| OR A.2.2 Initiate action to Immediately satisfy the requirements of this LCO.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.5.1 Verify all controls rods, other than the 24 hnu <
| |
| control rod withdrawn for the removal of the associated CRD, are fully inserted.
| |
| SR 3.10.5.2 Verify all control rods, other than the 2,4 hoe.....
| |
| control rod withdrawn for the removal of IINSE the associated CRD, in a five by five array centered on the control rod withdrawn for the removal of the associated CRD, are disarmed.
| |
| SR 3.10.5.3 Verify a control rod withdrawal block is .24h*o.r..
| |
| inserted.
| |
| (continued)
| |
| NMP2 3.10.5-2 Amendment 91
| |
| | |
| Single CRD Removal - Refueling 3.10.5 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.10.5.4 Perform SR 3.1.1.1. According to SR 3.1.1.1 SR 3.10.5.5 Verify no other CORE ALTERATIONS are in ,, ...
| |
| progress.
| |
| -- INSERT 1 I NMP2 3.10.5-3 Amendment 91
| |
| | |
| Multiple Control Rod Withdrawal - Refueling 3.10.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3.1 Initiate action to Immediately fully insert all control rods in core cells containing one or more fuel assemblies.
| |
| OR A.3.2 Initiate action to Immediately satisfy the requirements of this LCO.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.6.1 Verify the four fuel assemblies are removed 24 heues<
| |
| from core cells associated with each control rod or CRD removed.
| |
| SR 3.10.6.2 Verify all other control rods in core cells containing one or more fuel assemblies are fully inserted.
| |
| INSERT 1 SR 3.10.6.3 ---------------------- NOTE -----------------------------
| |
| Only required to be met during fuel loading.
| |
| Verify fuel assemblies being loaded are in 24 h*U, r< 1 compliance with an approved spiral reload sequence.
| |
| NMP2 3.10.6-2 Amendment 91
| |
| | |
| SDM Test - Refueling 3.10.8 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.10.8.2 --------------------- NOTE ---------------
| |
| Not required to be met if SR 3.10.8.3 satisfied.
| |
| According to Perform the MODE 2 applicable SRs for the applicable LCO 3.3.2.1, Function 2 of Table 3.3.2.1-1. SRs SR 3.10.8.3 --------------------- NOTE ---------------
| |
| Not required to be met if SR 3.10.8.2 satisfied.
| |
| Verify movement of control rods is in During control compliance with the approved control rod rod movement sequence for the SDM test by a second licensed operator or other qualified member of the technical staff.
| |
| SR 3.10.8.4 Verify no other CORE ALTERATIONS are in progress.
| |
| (continued)
| |
| INSERT 1 NMP2 3.10.8-3 Amendment 91
| |
| | |
| SDM Test - Refueling 3.10.8 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.10.8.5 Verify each withdrawn control rod does not Each time the go to the withdrawn overtravel position. control rod is withdrawn to "full out" position AND Prior to satisfying LCO 3.10.8.c requirement after work on control rod or CRD System that could affect coupling SR 3.10.8.6 Verify CRD charging water header pressure
| |
| _>940 psig.
| |
| INSERT1 I NMP2 3.10.8-4 Amendment 91
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Control Room Envelope Habitability Program (continued)
| |
| : 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 IINSERT 2 CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.
| |
| NMP2 5.5-13 Amendment 126
| |
| | |
| ATTACHMENT 4 License Amendment Request Nine Mile Point Nuclear Station, Unit 2 Docket No. 50-410 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
| |
| Proposed Technical Specification Bases Page Changes - Unit 2 (NOTE: TS Bases pages below marked with an asterisk (*) do not contain any mark-ups.
| |
| These pages are provided for completeness and for information purposes only.)
| |
| B 3.1.3-7 B 3.3.2.1-12 B 3.3.7.2-6 B 3.5.2-6 B 3.6.4.2-5 B 3.8.3-8 B 3.1.3-8 B 3.3.2.2-5 B 3.3.7.2-7 B 3.5.3-4 B 3.6.4.2-6 B 3.8.3-9 B 3.1.4-4* B 3.3.2.2-6 B 3.3.8.1-6* B 3.5.3-5 B 3.6.4.2-7 B 3.8.4-5 B 3.1.4-5 B 3.3.2.2-7 B 3.3.8.1-7 B 3.5.3-6 B 3.6.4.3-5 B 3.8.4-6 B 3.1.5-5 B 3.3.3.1-10* B 3.3.8.2-7 B 3.6.1.1-4* B 3.6.4.3-6 B 3.8.4-7 B 3.1.6-4 B 3.3.3.1-11 B 3.3.8.2-8 B 3.6.1.1-5 B 3.7.1-9 B 3.8.4-9 B 3.1.7-4 B 3.3.3.1-12 B 3.3.8.3-6 B 3.6.1.2-7* B 3.7.1-10 B 3.8.6-3 B 3.1.7-5 B 3.3.3.2-4 B 3.3.8.3-7 B 3.6.1.2-8 B 3.7.1-11 B 3.8.6-4 B 3.1.7-6 B 3.3.3.2-5 B 3.4.1-6 B 3.6.1.3-12* B 3.7.1-12 B 3.8.7-4 B 3.1.7-7 B 3.3.4.1-8 B 3.4.2-3* B 3.6.1.3-13 B 3.7.2-8 B 3.8.8-11 B 3.1.8-4 B 3.3.4.1-9 B 3.4.2-4 B 3.6.1.3-14 B 3.7.2-9 B 3.8.9-4 B 3.1.8-5 B 3.3.4.1-10 B 3.4.3-3* B 3.6.1.3-15 B 3.7.3-7 B 3.9.1-4 B 3.2.1-3 B 3.3.4.2-7* B 3.4.3-4 B 3.6.1.3-16 B 3.7.4-3 B 3.9.2-3 B 3.2.2-3 B 3.3.4.2-8 B 3.4.5-5 B 3.6.1.3-17 B 3.7.5-3 B 3.9.3-2 B 3.2.3-3 B 3.3.4.2-9 B 3.4.7-6 B 3.6.1.3-18 B 3.7.6-2 B 3.9.5-3 B 3.3.1.1-28 B 3.3.4.2-10 B 3.4.7-7 B 3.6.1.3-19 B 3.8.1-18 B 3.9.6-3 B 3.3.1.1-29 B 3.3.5.1-37 B 3.4.8-4 B 3.6.1.4-2 B 3.8.1-19 B 3.9.7-3 B 3.3.1.1-30 B 3.3.5.1-38 B 3.4.9-5 B 3.6.1.5-2* B 3.8.1-20 B 3.9.8-4 B 3.3.1.1-31 B 3.3.5.1-39 B 3.4.10-4 B 3.6.1.5-3 B 3.8.1-21 B 3.9.9-4 B 3.3.1.1-32 B 3.3.5.2-11 B 3.4.11-6 B 3.6.1.6-3* B 3.8.1-22 B 3.10.2-4*
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| B 3.3.1.1-33 B 3.3.5.2-12 B 3.4.11-7 B 3.6.1.6-4 B 3.8.1-23* B 3.10.2-5 B 3.3.1.1-34 B 3.3.5.2-13 B 3.4.11-9 B 3.6.1.7-5 B 3.8.1-24 B 3.10.3-4 B 3.3.1.1-35 B 3.3.6.1-34 B 3.4.11-10* B 3.6.1.7-6 B 3.8.1-25 B 3.10.4-5 B 3.3.1.1-36 B 3.3.6.1-35 B 3.4.12-2 B 3.6.2.1-5 B 3.8.1-26 B 3.10.5-4*
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| B 3.3.1.2-5* B 3.3.6.1-36 B 3.5.1-9 B 3.6.2.2-3 B 3.8.1-27 6 3.10.5-5 B 3.3.1.2-6 B 3.3.6.1-37 B 3.5.1-10 B 3.6.2.3-3* B 3.8.1-28 B 3.10.6-3 B 3.3.1.2-7 B 3.3.6.2-9 B 3.5.1-11" B 3.6.2.3-4 B 3.8.1-29 B 3.10.8-5 B 3.3.1.2-8 B 3.3.6.2-10 B 3.5.1-12 B 3.6.2.4-3* B 3.8.1-30 B 3.10.8-6 B 3.3.1.2-9 B 3.3.6.2-11 B 3.5.1-13* B 3.6.2.4-4 B 3.8.1-31 B 3.3.2.1-8 B 3.3.7.1-8 B 3.5.1-14 B 3.6.3.2-3 B 3.8.1-32 B 3.3.2.1-9 B 3.3.7.1-9 B 3.5.1-15 B 3.6.4.1-4 B 3.8.1-33 B 3.3.2.1-10 B 3.3.7.1-10 B 3.5.2-4 B 3.6.4.1-5 B 3.8.3-5*
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| B 3.3.2.1-11 B 3.3.7.2-5 B 3.5.2-5 B 3.6.4.1-6 B 3.8.3-6
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| | |
| INSERT 3 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| Control Rod OPERABILITY B 3.1.3 BASES ACTIONS E.1 (continued)
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| If any Required Action and associated Completion Time of Condition A, C, or D are not met or nine or more inoperable control rods exist, 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 12 hours. This ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active function (i.e., scram) of the control rods. The number of control rods permitted to be inoperable when operating above 10% RTP (i.e., no CRDA considerations) could be more than the value specified, but the occurrence of a large number of inoperable control rods could be indicative of a generic problem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach 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 position of each control rod must be determined, to ensure adequate information on control rod position is available to the operator for determining control rod OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, (full-in, full-out, or numeric indicator), by verifying the indicators one notch "out" and one notch "in" are OPERABLE, or by the use of other appropriate methods. Thc 24 hour Froquency of this SR is based on oporating oxporionco rolated to cxpoctod changcsi con-trol rod pocition and the availability Of control rod position indications an the.......
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| hnsert
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| ....... ,. < 3 (Note: SR 3.1.3.2 has been deleted)
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| SR 3.1.3.3 Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves.
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| The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal. This Surveillance is not required when THERMAL POWER is less than or equal to the (continued)
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| NMP2 B 3.1.3-7 Revision 0, 29 (A130)
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| | |
| Control Rod OPERABILITY B 3.1.3 BASES SURVEILLANCE SR 3.1.3.3 (continued)
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| REQUIREMENTS actual LPSP of the RWM since the notch insertions may not be compatible with the requirements of the banked position withdrawal sequence (BPWS) (LCO 3.1.6) and the RWM ilnsert 3 'LCO . _ 4 - T , d y . .. . . - ,,-;, ,
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| account epcratfing cxpcricncc rclatcd to changcz in ORD
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| ,Pre ee-.At
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| , any time, if a control rod is immovable, a determination of that control rod's trippability (OPERABILITY) must be made and appropriate action taken.
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| This SR is modified by a Note that allows 31 days after withdrawal of the control rod and increasing power to above the LPSP, to perform the Surveillance. This acknowledges that the control rod must be first withdrawn and THERMAL POWER must be increased to above the LPSP before performance of the Surveillance, and therefore, the Note avoids potential conflicts with SR 3.0.3 and SR 3.0.4.
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| SR 3.1.3.4 Verifying the scram time for each control rod to notch position 05 is < 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function.
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| This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1, "Reactor Protection System (RPS)
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| Instrumentation," and the functional testing of SDV vent and drain valves in LCO 3.1.8, "Scram Discharge Volume (SDV)
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| Vent and Drain Valves," overlap this Surveillance to provide complete testing of the assumed safety function. The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.
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| (continued)
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| NMP2 B 3.1.3-8 Revision 0, 29 (A130)
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| | |
| Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE core scram, the CRD pump head would be seen by all control REQUIREMENTS rods and would have a negligible effect on the scram (continued) insertion times.
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| SR 3.1.4.1 The scram reactivity used in DBA and transient analyses is based on assumed control rod scram time. Measurement of the scram times with reactor steam dome pressure > 800 psig demonstrates acceptable scram times for the transients analyzed in References 5 and 6.
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| Maximum scram insertion times occur at a reactor pressure of approximately 800 psig because of the competing effects of reactor steam dome pressure and stored accumulator energy.
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| Therefore, demonstration of adequate scram times at reactor steam dome pressure >_800 psig ensures that the scram times will be within the specified limits at higher pressures.
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| Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be performed. To ensure scram time testing is performed within a reasonable time following a shutdown _>120 days, control rods are required to be tested before exceeding 40% RTP. This Frequency is acceptable, considering the additional Surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by fuel movement within the associated core cell and by work on control rods or the CRD System.
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| SR 3.1.4.2 Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle. A representative sample contains at least 10% of the control rods. The sample remains representative if no more than 7.5% of the control rods in the sample tested are determined to be "slow." If more than 7.5 % of the sample is declared to be "slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this 7.5% criterion (i.e., 7.5% of the entire sample size) is satisfied, or until the total number of "slow" control rods (throughout the core, from all Surveillances) exceeds the LCO limit. For (continued)
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| NMP2 B 3.1.4-4 Revision 9, 30 (Al 32)
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| Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE SR 3.1.4.2 (continued)
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| REQUIREMENTS planned testing, the control rods selected for the sample should be different for each test. Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data were previously tested in a sample. The 200 day F=roqunc. ic basod on oporating oxporionco that has sh(Own con-.trol ro-d 6cram times do not .igniflcantly change evc aA epcrating cyclo. This FrcequencY us a169 Foasonablo, based on tho additional Survoillancoc dono oni the GRDs at moRcF frcqucnt intcryals in AccordAnco wi'Fth A 1LCQ 32. *I'2and1LCo 3.1.5, "Control Rod Scram Acc9umulatorc.-"
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| SR 3.1.4.3 When work that could affect the scram insertion time is performed on a control rod or the CRD System, testing must be done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure. The scram testing must be performed once before declaring the control rod OPERABLE. The required scram time testing must demonstrate that the affected control rod is still within acceptable limits. The scram time limits for reactor pressures < 800 psig are found in the Technical Requirements Manual (Ref. 9) and are established based on a high probability of meeting the acceptance criteria at reactor pressures Ž! 800 psig. Limits for reactor pressures Ž 800 psig are found in Table 3.1.4-1.
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| If testing demonstrates the affected control rod does not meet these limits, but is within the 7-second limit of Table 3.1.4-1, Note 2, the control rod can be declared OPERABLE and "slow."
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| Specific examples of work that could affect the scram times include (but are not limited to) the following: removal of any CRD for maintenance or modification; replacement of a control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator isolation valve, or check valves in the piping required for scram.
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| (continued)
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| NMP2 B 3.1.4-5 Revision 0, 30 (Al 32)
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| | |
| Control Rod Scram Accumulators B 3.1.5 BASES (continued)
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| SURVEILLANCE SR 3.1.5.1 periodically REQUIREMENTS SR 3.1.5.1 requires that the accumulator pressure be hecked ey,.Fy-:days to ensure adequate accumulator pressure exists to provide sufficient scram force. The primary indicator of accumulator OPERABILITY is the accumulator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of 940 psig is well below the expected pressure of 1050 psig to 1100 psig (Ref. 2). Declaring the accumulator inoperable when the minimum pressure is not maintained ensures that significant degradation in scram times does not occur. The- 7 4ay Froquoncy has bc..,hown to be accoptablo through .p..at.ng aind takes fint, errrnr aGmiint indic~iti9nG i'iiiaihlc On tho conQtrol roo.
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| MOM.- 3 REFERENCES 1. USAR, Section 4.3.2.5.1.
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| : 2. USAR, Section 4.6.1.1.2.
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| : 3. USAR, Section 5.2.2.2.3.
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| : 4. USAR, Section 15.4.1.
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| : 5. 10 CFR 50.36(c)(2)(ii).
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| NMP2 B 3.1.5-5 Revision 0
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| Rod Pattern Control B 3.1.6 BASES ACTIONS B.1 and B.2 (continued)
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| If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed sequence. Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has less impact on control rod worth than withdrawals have. Required Action B.1 is modified by a Note that allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO 3.3.2.1 requires verification of control rod movement by a second licensed operator (Reactor Operator or Senior Reactor Operator) or by a qualified member of the technical staff (e.g.,
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| a qualified shift technical advisor or reactor engineer).
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| With nine or more OPERABLE control rods not in compliance with BPWS, the reactor mode switch must be placed in the shutdown position within 1 hour. With the reactor mode switch in shutdown, the reactor is shut down, and therefore does not meet the applicability requirements of this LCO. The allowed Completion Time of 1 hour is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence.
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| SURVEILLANCE SR 3.1.6.1 periodically REQUIREMENTS The control rod pattern is verified to be in compliance with the BPWS at a 24 hour Fr..qucn.y, ensuring the assumptions of the CRDA Ilnset 3 I analyses a Tho 21 hour F=.oquon.y of this ur..,.llanco
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| .. was-pattFRn complianco with. tho BPWAIS is po~formod by the RWVM (LCOQ 3.3.2.1). The RWM provides control rod blocks to enforce the required control rod sequence and is required to be OPERABLE when operating at < 10% RTP.
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| REFERENCES 1. Supplemental Reload Licensing Report for Nine Mile Point Nuclear Station Unit 2, (revision specified in the COLR).
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| : 2. Letter from T.A. Pickens (BWROG) to G.C. Laines (NRC),
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| "Amendment 17 to General Electric Licensing Topical Report NEDE-2401 1-P-A," BWROG-8644, August 15, 1988.
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| (continued)
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| NMP2 B 3.1.6-4 Revision 0, 18 (A120)
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| SLC System B 3.1.7 BASES ACTIONS C.1 and C.2 (continued)
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| If any Required Action and associated Completion Time is 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 12 hours and to MODE 4 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 3.1.7.1, SR 3.1.7.2, and SR 3.1.7.3 REQUIREMENTS SR 3.1.7.1 through SR 3.1.7.3 aro 24 hour Suro-coc, verifyi.eq certain characteristics of the SLC System (e.g.,
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| the volume and temperature of the borated solution in the storage tank), thereby ensuring the SLC System OPERABILITY without disturbing normal plant operation. These Surveillances ensure the proper borated solution and temperature, including the temperature of the pump suction piping, are maintained. Maintaining a minimum specified borated solution temperature is important in ensuring that the boron remains in solution and does not precipitate out in the storage tank or in the pump suction piping. T-he 24 hour Frequ*n*y of thoso is
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| ,Rebased oAn .p*at*ng varmitmne ianthomm mom d paralmmtors of vyoeluWo and f
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| SR 3.1.7.4 and SR 3.1.7.6 lInsert 3 SR 3.1.7.4 verifies the continuity of the explosive charges in the injection valves to ensure proper operation will occur if required. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The 31 day Frog ony "6based on operating cxperiencc that has dcmonctrated thoe reliability of the oxplocive charge continuityý.<
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| SR 3.1.7.6 verifies each valve in the system is in its correct position, but does not apply to the squib (i.e.,
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| explosive) valves. Verifying the correct alignment for (continued)
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| NMP2 B 3.1.7-4 Revision 0, 26 (A125)
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| SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.4 and SR 3.1.7.6 (continued)
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| REQUIREMENTS manual, power operated, and automatic valves in the SLC System flow path ensures that the proper flow paths will exist for system operation. A valve is also allowed to be in the nonaccident position, provided it can be aligned to the accident position from the control room, or locally by a dedicated operator at the valve control. This is acceptable since the SLC System is a manually initiated system. This Surveillance does not apply to valves that are locked, sealed, or otherwise secured in position, since they were verified to be in the correct position prior to locking, sealing, or securing. This verification of valve alignment does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation; rather, it involves ,, 3 verification that those valves capable of bei mispositioned are in the correct positions. I e,
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| ,ie-'-^,
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| FrFequoncy is based onegnern judgmont -andi onitn with the procoduraI controic governling valve operation that onzUre correct valve 139Giti9nG.
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| SR 3.1.7.5 This Surveillance requires an examination of the sodium pentaborate solution by using chemical analysis to ensure the proper concentration of boron (measured in weight %
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| sodium pentaborate decahydrate) exists in the storage tank.
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| SR 3.1.7.5 must be performed anytime boron or water is added to the storage tank solution to establish that the boron solution concentration is within the specified limits. This Surveillance must be performed anytime the temperature is restored to within the limit (i.e., > 701F), to ensure no significant boron precipitation occurred. The 31 day Frequency of thiG Su:.eilla. e is appropriate becau.e of th.e, relatively slew vatatr;+iato oAboron Aoncent-sRratioA-n between SR 3.1.7.7 Demonstrating each SLC System pump develops a flow rate
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| Ž 41.2 gpm at a discharge pressure _>1327 psig ensures that pump performance has not degraded during the fuel cycle.
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| This minimum pump flow rate requirement ensures that, when (continued)
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| NMP2 B 3.1.7-5 Revision 0, 17 (A!-17), 24 (A!23), 26 (A!125),
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| 37 (140)
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| SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.7 (continued)
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| REQUIREMENTS combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve, and is indicative of overall performance. Such inservice tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance. The Frequency of this Surveillance is in accordance with the Inservice Testing Program.
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| SR 3.1.7.8 and SR 3.1.7.9 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired. The pump anid oplosiv.,,o t..,.,,d should be altcrl-ated such that both complot- flow pathG a,-
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| tested eycy 48 months, at altcrfating 21 month .,,- :,als.
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| The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV. Wholc thcsc Sur'cillanccs can ha ^^.nFarrnaA OfK+ fK- rarM~r p4- n^ ^~r annprpfun gs~rp has shown th*,se mponcnts usually pass the Su mmeillancc who, porform.d at the 21..monh Froquoncy, which is based on hroucling eyclc. Thcrcforc, th Froqucncy . ,,as,onc...
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| to be a..optablc.......fro.m a rcliability standpoint.
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| *"..........nse~rt - 3 Demonstrating that all heat traced piping between the boron solution storage tank and the suction valve to the injection pumps is unblocked ensures that there is a functioning flow path for injecting the sodium pentaborate solution. An acceptable method for verifying that the suction piping up to the suction valve is unblocked is to pump from the storage tank to the test tank. Upon completion of this verification, the pump suction piping between the pump (continued)
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| NMP2 B 3.1.7-6 Revision ,12- (A1!-!), 26 (A125)
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| SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.8 and SR 3.1.7.9 (continued)
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| REQUIREMENTS and it suction valve and pump suction m: drained and flushed with demineralized water, since th'4l iping is not heat
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| - 3-r tracedýTho 24 moRth Froqurncy is acceptable since there is t 3a lowprobability that the subject piping will be blocked due to precipitation of the boron from solution in the heat
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| [eriodic1-traced piping. This is especially true in light of the
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| ' ytemperature verification of this piping required by SR 3.1.7.3. However, if, in performing SR 3.1.7.3, it is determined that the temperature of this piping has fallen below the specified minimum, SR 3.1.7.9 must be performed once within 24 hours after the piping temperature is restored within the limits of SR 3.1.7.3.
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| SR 3.1.7.10 Enriched sodium pentaborate solution is made by mixing granular, enriched sodium pentaborate with water. Isotopic tests on the granular sodium pentaborate to verify the actual B-10 enrichment must be performed prior to addition to the SLC tank in order to ensure that the proper B-10 atom percentage is being used.
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| REFERENCES 1. 10 CFR 50.62.
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| : 2. USAR, Section 9.3.5.3.
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| : 3. 10 CFR 50.36(c)(2)(ii).
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| : 4. NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants," USNRC, February 1995.
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| : 5. 10 CFR 50.67, "Accident Source Term."
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| NMP2 B 3.1.7-7 Revision 26 (A125)
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| SDV Vent and Drain Valves B 3.1.8 BASES (continued)
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| SURVEILLANCE SR 3.1.8.1 REQUIREMENTS During normal operation, the SDV vent and drain valves should be in the open position (except when performing SR 3.1.8.2) to allow for drainage of the SDV piping.
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| Verifying that each valve is in the open position ensures that the SDV vent and drain valves will perform their intended function during normal operation. This SR does not require any testing or valve manipulation; rather, it involves verification that the valves are in the correct position. The 31 day Fr..qucn.y is based on cnginccrlng judgmnR t and-i n cicton t with the procodural ceontroe
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| : p. ,,t"e, . -_prnp.r valve position (closed) would not affect the isolation function.
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| SR 3.1.8.2 During a scram, the SDV vent and drain valves should close to contain the reactor water discharged to the SDV piping.
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| Cycling each valve through its complete range of motion (closed and open) ensures that the valve will function properly during a scram. Thc 92 day Fc..qu.ncy is based on operating oxperienco and takes into c o n the !eye! et rod undancy in the syst m dc g . <
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| Insert 3 sR 3.1.8.3 SR 3.1.8.3 is an integrated test of the SDV vent and drain valves to verify total system performance. After receipt of a simulated or actual scram signal, the closure of the SDV vent and drain valves is verified. The closure time of 30 seconds after a receipt of a scram signal is based on the bounding leakage case evaluated in the accident analysis.
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| Similarly, after receipt of a simulated or actual scram reset signal, the opening of the SDV vent and drain valves is verified. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1 and the scram time testing of control rods in LCO 3.1.3, "Control Rod OPERABILITY," overlap this Surveillance to provide complete testing of the assumed safety function. -ip, 2, I,,,,h Frc"'-c,, is basd on th, aply dur in a l ant Iou ca _ dt ..c pot ntialfo r anu uplannod transient if thecSu rvcllanccw rcpo reefrmod wiithc (continued)
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| NMP2 B 3.1.8-4 Revision 0
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| SDV Vent and Drain Valves B 3.1.8 BASES SURVEILLANCE SR 3.1.8.3 (continued)
| |
| REQUIREMENTS the roactor at power. Oporating oxpeOrionc hac GhoWn thoc componcnts usually pass the Surwcillancc when pc~fermcd at the 24 month FrcguoncGY; thorforo, the FrogqucncY waG concludcd to be accoptabic fromn a roliability standpoint.
| |
| REFERENCES 1. USAR, Section 4.6.1.1.2.
| |
| : 2. 10 CFR 50.67, "Accident Source Term."
| |
| : 3. NUREG-0803, "Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping,"
| |
| August 1981.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.1.8-5 Revision 0, 26 (A125)
| |
| | |
| APLHGR B 3.2.1 BASES ACTIONS B. 1 (continued)
| |
| If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 23% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 23% RTP in an orderly manner and without challenging plant systems.
| |
| SR 3.2.1.1 periodically SURVEILLANCE REQUIREMENTS APLHGRs are required to be initially calculated within 12 hours after THERMAL POWER is Ž 23% RTP and then 24-heU-s thereafter. They are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The-24 heu Frcqucncy is based on both cngineG-,n' judgment and roconiton f tho low nos of chang npwrdgrbto undcr nrmEal conditions. The 12 hour allowance after THERMAL POWER > 23% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.
| |
| * lnsert 3 REFERENCES 1. NEDE-2401 1-P-A, "GE Standard Application for Reactor Fuel," (revision specified in the COLR).
| |
| : 2. USAR, Chapter 15B.
| |
| : 3. USAR, Chapter 15G.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.2.1-3 Revision 0, 37 (A140)
| |
| | |
| MCPR B 3.2.2 BASES APPLICABILITY occurs. When in MODE 2, the intermediate range monitor (continued) (IRM) provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels
| |
| < 23% RTP, the reactor is operating with substantial margin to the MCPR limits and this LCO is not required.
| |
| ACTIONS A.1 If any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met. Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within analyzed conditions. The 2 hour Completion Time is normally sufficient to restore the MCPR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of specification.
| |
| B.1 If the MCPR cannot be restored to within the required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 23% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 23% RTP in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE SR 3.2.2.1 periodicallY REQUIREMENTS The MCPR is required to be initially calculated within 12 hours after THERMAL POWER is Ž_23% RTP and the,-ne'lei 24-heWr-e thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24. h'-
| |
| Frcequcncy is based on both cnginccring judgment and rccognition of the slewnczs Of changes in power dictributin during norm.al op"atio The 12 hour allowance after THERMAL POWER reaches > 23% RTP is acceptable given the large inherent margin to operating limits at low power levels. ( i (continued)
| |
| NMP2 B 3.2.2-3 Revision 9, 37 (A140)
| |
| | |
| LHGR B 3.2.3 BASES ACTIONS B.1 (continued)
| |
| If the LHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 23% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 23% RTP in an orderly manner and without challenging plant systems.
| |
| SR 3.2.3.1 periodically REQUIREMENTS)
| |
| SURVEILLANCE The LHGRs are required to be initially calculated within 12 hours after THERMAL POWER is Ž 23% RTP and 24-her-&-thereafter. They are compared with the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The-24 hew Frcqucncy is based on both cngincc"i"g judgmont aFd undcr normal conditioc. The 12 hour allowance after THERMAL POWER > 23% RTP is achieved is acceptable given the large inherent margin to operating limits at lower power levels. N Insert 3 REFERENCES 1. NEDE-2401 1-P-A, "GE Standard Application for Reactor Fuel," (revision specified in the COLR).
| |
| : 2. Supplemental Reload Licensing Report for Nine Mile Point Nuclear Station Unit 2, (revision specified in the COLR).
| |
| : 3. NUREG-0800, Section II A.2(g), Revision 2, July 1981.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| : 5. NEDC-33286P, "Nine Mile Point Nuclear Station Unit 2 -
| |
| APRM/RBM/Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)," March 2007.
| |
| NMP2 B 3.2.3-3 Revision ,2,4 ^(A!23), 37 (A140)
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.1 and SR 3.3.1.1.2 REQUIREMENTS (continued) Performance of the CHANNEL CHECK Rncc ovry 12 hourscr ovor'; 21 hou,,, as applicablo, 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 instrument channels could be an indication of excessive instrument drift on one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
| |
| 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 instrument has drifted outside its limit. Insert 3 The PFrouonc'; is basod upon oporatln poRIARnceeha d*monStratoc channcl-fAiurc ns rare-. he CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
| |
| SR 3.3.1.1.3 To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. The Froquncy of enco pnr 7 days imc bak-d Aon mrir c-hagc:in LPRM sensitivity, whih*
| |
| couldd a;ffoc-t the APRM roading betwoon performAnccR of S
| |
| .... I.nsert 3 An allowance is provided that requires the SR to be performed only at > 23% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER (continued)
| |
| NMP2 B 3.3.1.1-28 Revision , 1, o24 ^(A1-34, 37 (A140)
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.3 (continued)
| |
| REQUIREMENTS consistent with a heat balance when < 23% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large inherent margin to thermal limits (MCPR, APLHGR, and LHGR). At >_23% RTP, the Surveillance is required to have been satisfactorily performed within the last 7 days in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 23% if the 7-dey Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after reaching or exceeding 23% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
| |
| SR 3.3.1.1.4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| As noted, for Functions 1.a and 1.b, SR 3.3.1.1.4 is not required to be performed when entering MODE 2 from MODE 1 since testing of the MODE 2 required IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This allows entry into MODE 2 if the 7-4ay Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
| |
| A Froqunccy f:7 days p vid.s .. an a .,.ptablo level of system avcragc unavailabilit' ovcr the Frcequcncy intcrval and is basoPd- onR oliability analysis (Ref. 10). (The Manual';; Scram; Functieon HANNEL FUNCTIONAl TEST Frcquoncy Was c..ditcd in the analysis to cxtcd m automatic
| |
| .any .cram. Functi*ns F=Feq*I9AGie&4 SR 3.3.1.1.5 and SR 3.3.1.1.6 These Surveillances are established to ensure that no gaps in neutron flux indication exist from subcritical to power (continued)
| |
| NMP2 B 3.3.1.1-29 Revision G-,4, 37 (Al 40)
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.5 and SR 3.3.1.1.6 (continued)
| |
| REQUIREMENTS operation for monitoring core reactivity status.
| |
| The overlap between SRMs and IRMs is required to be demonstrated to ensure that reactor power will not be increased into a region without adequate neutron flux indication. This is required prior to fully withdrawing SRMs since indication is being transitioned from the SRMs to the IRMs.
| |
| The overlap between IRMs and APRMs is of concern when reducing power into the IRM range. On power increases, the system design will prevent further increases (initiate a rod block) if adequate overlap is not maintained. Overlap between IRMs and APRMs exists when sufficient IRMs and APRMs concurrently have onscale readings such that the transition between MODE 1 and MODE 2 can be made without either APRM downscale rod block, or IRM upscale rod block. Overlap between SRMs and IRMs similarly exists when, prior to withdrawing the SRMs from the fully inserted position, IRMs are above mid-scale on range 1 before SRMs have reached the upscale rod block. The IRM/APRM and SRM/IRM overlaps are also acceptable if a 1/2 decade overlap exists.
| |
| As noted, SR 3.3.1.1.6 is only required to be met during entry into MODE 2 from MODE 1. That is, after the overlap requirement has been met and indication has transitioned to the IRMs, maintaining overlap is not required (APRMs may be reading downscale once in MODE 2).
| |
| If overlap for a group of channels is not demonstrated (e.g., IRM/APRM overlap), the reason for the failure of the Surveillance should be determined and the appropriate channel(s) declared inoperable. Only those appropriate channel(s) that are required in the current MODE or condition should be declared inoperable.
| |
| A Frcqucncy of 7 days i.. based on cnginccFing
| |
| .as.nablc judgment and the rcliability of the IRMs and APRMs.
| |
| SR 3.3.1.1.7 Insert 3 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP)
| |
| System. This establishes the relative local flux profile (continued)
| |
| NMP2 B 3.3.1.1-30 Revision@0,1
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.7 (continued)
| |
| REQUIREMENTS for appropriate representative input to the APRM System.
| |
| >-Th 1000 o.8,tivo full p3.3.1.1 hoU, (EFPH) 3 Frquncy i2 based on operating cxpcrioncc with LPRM sensitiity chan ges.-
| |
| SR 3.3.1.1.8, SR 3.3.1.1.10. and SR 3.3.1.1.12 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. For Function 2.b, the CHANNEL FUNCTIONAL TEST also includes the flow input function, excluding the flow transmitters. Note 1 is provided for SR 3.3.1.1.10 that requires the APRM SR for Function 2.a to be performed within lInsert 3 1 12 hours of entering MODE 2 from MODE 1. Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. Note 2 is provided for SR 3.3.1.1.10 that allows the Function 2.e CHANNEL FUNCTIONAL TEST to consist of toggling the appropriate outputs of the APRM. This is acceptable since this will test all of the hardware required to produce the trip signal, but not directly re-test software-controlled logic. Also, the automatic self-test logic will automatically detect a hardware fault that results in a change to the software.
| |
| II LI ~ 1t4, I I------. .. . . 15Lr1,J IJL4~J .JI I s, IL, roliability analysis of Roforonco 10. TheI I181 day Froequency of SR 3.3 1. isih baovd on tho roliability analy siso Rcfcrcnco 11. The 24 month Frcqucncy of SR 3.3.1.1.12 is conditions that apply during a plant eutagc and h potontial for an uRpIanRn tAR-n- ct if the Sur~oIanee,,
| |
| woro perfermcdd with thc rcacter at pewcr. Opcralting 44-fl VII 4
| |
| I I..I .... -
| |
| .r. ,n.i In nnn saul~%flfl nfl. flrrnflrl n~ +kn ')A m*nTn *r*Ni B* *
| |
| '1A (continued)
| |
| (continued)
| |
| NMP2 B 3.3.1.1-31 Revision0, 1
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.9 REQUIREMENTS (continued) The calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.1.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
| |
| J.1~
| |
| -rL-- r----. . -Z nn - Z-- e1rn --I It 4 4 n !- L-- Ck &6ýý
| |
| .cliability a" aly.
| |
| . .. of f... nc. 10..
| |
| LI!~II SR 3.3.1.1.11 and SR 3.3.1.1.13 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies.the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. For Function 2.b, the CHANNEL CALIBRATION also includes the flow input function.
| |
| Note 1 to SR 3.3.1.1.13 states that neutron detectors are excluded from CHANNEL CALIBRATION because of the difficulty of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performing the 7-day calorimetric calibration (SR 3.3.1.1.3) and the 1000 EFPH LPRM calibration against the TIPs (SR 3.3.1.1.7).
| |
| Note 2 to SR 3.3.1.1.13 requires the APRM and IRM SRs for Functions 1.a and 2.a to be performed within 12 hours of entering MODE 2 from MODE 1. Testing of the MODE 2 APRM and IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. Note 3 is provided for SR 3.3.1.1.13 that allows the Function 2.e CHANNEL CALIBRATION to consist of a verification of OPRM - Upscale setpoints in the APRM by the review of the (continued)
| |
| NMP2 B 3.3.1.1-32 RevisionO , 1
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.11 and SR 3.3.1.1.13 (continued)
| |
| REQUIREMENTS "Show Parameters" display. This is acceptable because, other than the flow and LPRM input processing, all OPRM functional processing is performed digitally involving equipment or components that cannot be calibrated. Pie Frequency of SR 3.3.1.1.11 c based upon the asccumptieon of 18 monqAth calibrFation intcrval inthe doteFrmination of tho Fmagnitud of oequipmont drift in thoe otpoint analysis. The Frcequcncy of SR 3.3.1.1.13 is based on the assumption of a 24 monFth c-alibrationR intorval intho dctormin~ationR of the magnitude of oquipmcnt drift inthe setpoint analysi. <
| |
| SR 3.3.1.1.14 The LOGIC SYSTEM FUNCTIONAL TEST (LSFT) demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods, in LCO 3.1.3, "Control Rod OPERABILITY," and SDV vent and drain valves, in LCO 3.1.8, "Scram Discharge Volume (SDV) Vent and Drain Valves," overlaps this Surveillance to provide In sert 3 complete testing of the assumed safety function. In addition, for Function 2.f, the LSFT includes simulating APRM trip conditions at the APRM channel inputs to the 2-Out-Of-4 Voter channel to check all combinations of two tripped APRM channel inputs to the 2-Out-Of-4 Voter logic in the 2-Out-Of-4 Voter channels.
| |
| Tho 24 month .Fequ.nc.yi basod on the Reod to porform thi.
| |
| Survillnccundcr tho conditionc that apply during a plant outage and tho p.tc.t.al for.-R an...p.anno.d transi.nt if the Surillanc porformod Icc With tho-rc-actor at poWor.
| |
| Opcrating cxpcricncc has shown that these componcnts usually mo;FCc tho Sur'..'ollancoe whon Gorfofrmd at the 241 month I=.equeney. -.1 SR 3.3.1.1.15 This SR ensures that scrams initiated from the Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions will not be inadvertently bypassed when THERMAL POWER is _>26% RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodology are incorporated into the Allowable Value and the actual setpoint. Because main (continued)
| |
| NMP2 B 3.3.1.1-33 Revision 0r--4, 37 (A140)
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 (continued)
| |
| REQUIREMENTS turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine bypass valves must remain closed during an in-service calibration at THERMAL POWER _> 26%
| |
| RTP to ensure that the calibration is valid.
| |
| If any bypass channel setpoint is nonconservative (i.e., the Functions are bypassed at Ž_26% RTP, either due to open main turbine bypass valve(s) or other reasons), then the affected Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition, this SR is met and the channel is considered OPERABLE.
| |
| The Froequoncy of 21 months is based on enqinccrinq judgmont And r- liabilit-l of.the
| |
| ,.,* ., - -ets -4 SR 3.3.1.1.16 This SR ensures that scrams initiated from the APRM OPRM - Upscale Function will not be inadvertently bypassed when THERMAL POWER is _>26% RTP and recirculation drive flow is < 60% rated recirculation drive flow. elr nsert 3 If any bypass channel setpoint is nonconservative (i.e., the Function is bypassed at _>26% RTP and < 60% rated recirculation drive flow), then the affected channel is considered inoperable.
| |
| R- roquo-rivy E; Mori 0 0 MOO Wrl V .
| |
| SR 3.3.1.1.17 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. The RPS RESPONSE TIME acceptance criteria are included in Reference 12.
| |
| (continued)
| |
| NMP2 B 3.3.1.1-34 Revision 0,1, -29, 37 (A140)
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.17 (continued)
| |
| REQUIREMENTS As noted (Note 1), the Function 2.f digital electronics are excluded. This is allowed since self-testing and calibration checks the time base of the digital electronics (confirmation of the time base is adequate to assure required response times are met). In addition, Note 2 states the response time of the sensors for Functions 3 and 4 may be assumed to be the design sensor response time and therefore, are excluded from the RPS RESPONSE TIME testing.
| |
| This is allowed since the sensor response time is a small part of the overall RPS RESPONSE TIME (Ref. 13). Note 4 modifies the starting point of the RPS RESPONSE TIME test for Function 9, since this starting point (start of turbine control valve fast closure) corresponds with the safety analysis assumptions.
| |
| RPS RESPONSE TIME tests are conducted on a 24 merth STAGGERED TEST BASIS. Note 3 requires STAGGERED TEST BASIS Frequency to be determined based on 4 channels per trip in accordance system, in lieu of the 8 channels specified in with the Table 3.3.1.1-1 for the MSIV Closure Function. This Surveillance Frequency is based on the logic interrelationships o , u various channels required to produce an RP ram signal. Frequency Therefore, staggered testing result i ponse time Control Program verification of these devices A."p -4cteth. The-24-.t Frcequcncy us eensistent with the typical indUctr, rcfucling cy.lo and is based upon plant operating experience, which that random failUrcsEof nhow3 mt mpencnts causing co-rio-uc ti~mo degradation, but not channel failuro, a~i~qet ý_Iset REFERENCES 1. USAR, Section 7.2.
| |
| : 2. USAR, Section 5.2.2 and Appendix A Section A.5.2.2.
| |
| : 3. USAR, Section 6.3.3.
| |
| : 4. USAR, Chapter 15 and Appendix A.
| |
| : 5. 10 CFR 50.36(c)(2)(ii).
| |
| : 6. USAR, Section 15.4.1.
| |
| : 7. NEDO-23842, "Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.
| |
| (continued)
| |
| NMP2 B 3.3.1.1-35 RevisionO , 1
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES REFERENCES 8. USAR, Section 15.4.9.
| |
| (continued)
| |
| : 9. Letter, P. Check (NRC) to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.
| |
| : 10. NEDO-30851-P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System,"
| |
| March 1988.
| |
| : 11. NEDC-3241 0-P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC-PRNM)
| |
| Retrofit Plus Option III Stability Trip Function,"
| |
| October 1995.
| |
| : 12. Technical Requirements Manual.
| |
| : 13. NEDO-32291-A, "System Analyses for the Elimination of Selected Response Time Testing Requirements,"
| |
| October 1995.
| |
| 41ed
| |
| : 14. USAR, Section 7.6.1.4.3.
| |
| : 15. NEDC 32P10 P A, "NUMAC PRNI..Rotrofit Plus Option 111 Stability Trip-Pun Spmcnt 1,"
| |
| Novcmbcr 1997.
| |
| : 16. NEDC-31336P-A, GE Nuclear Energy, "General Electric Instrument Setpoint Methodology," Class III (Proprietary),
| |
| September 1996.
| |
| NMP2 B 3.3.1.1-36 Revision 0_-4, 37 (A140)
| |
| | |
| SRM Instrumentation B 3.3.1.2 BASES ACTIONS E.1 and E.2 (continued)
| |
| With one or more required SRMs inoperable in MODE 5, the capability to detect local reactivity changes in the core during refueling is degraded. CORE ALTERATIONS must be immediately suspended, and action must be immediately initiated to fully insert all insertable control rods in core cells containing one or more fuel assemblies.
| |
| Suspending CORE ALTERATIONS prevents the two most probable causes of reactivity changes, fuel loading and control rod withdrawal, from occurring. Inserting all insertable control rods ensures that the reactor will be at its minimum reactivity, given that fuel is present in the core.
| |
| Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe, conservative position.
| |
| Action (once required to be initiated) to insert control rods must continue until all insertable rods in core cells containing one or more fuel assemblies are inserted.
| |
| SURVEILLANCE As noted at the beginning of the SRs, the SRs for each SRM REQUIREMENTS Applicable MODE or other specified condition are found in the SRs column of Table 3.3.1.2-1.
| |
| SR 3.3.1.2.1 and SR 3.3.1.2.3 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 the same parameter indicated on other similar channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
| |
| Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, (continued)
| |
| NMP2 B 3.3.1.2-5 Revision 0
| |
| | |
| SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.1 and SR 3.3.1.2.3 (continued)
| |
| REQUIREMENTS including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
| |
| The Frnquncey of onc cvcr;n 12 horz frf SR 3.3.1.2.1hin based on oporating oxporicncc that domonc~tratoc channol failuro *6 raro. W~hvieoi OE and 4, Fersoaiit,' changcc rclaxod to24 ho'-rs for SR 3.3.1.2.3.4eCHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
| |
| SR 3.3.1.2.2 To provide adequate coverage of potential reactivity changes in the core when the fueled region encompasses more than one SRM, one SRM is required to be OPERABLE in the quadrant where CORE ALTERATIONS are being performed, and the other OPERABLE SRM must be in an adjacent quadrant containing fuel. Note 1 states that this SR is required to be met only during CORE ALTERATIONS. It is not required to be met at other times in MODE 5 since core reactivity changes are not occurring. This Surveillance consists of a review of plant logs to ensure that SRMs required to be OPERABLE for given CORE ALTERATIONS are, in fact, OPERABLE. In the event that only one SRM is required to be OPERABLE (when the fueled region encompasses only one SRM), per Table 3.3.1.2-1, footnote (b), only the a. portion of this SR is required.
| |
| Note 2 clarifies that more than one of the three requirements can be met by the same OPERABLE SRM. The 12 hour Froequoncy as based upon opcrating experienee and supplomcnts opcrational conRolc6 ever rofuo-ling activitioc, which includc stcpsc to on.Ur that the SRM roquired . by the LCOG arc inthe propor quadrant.<
| |
| SR 3.3.1.2.4 This Surveillance consists of a verification of the SRM instrument readout to ensure that the SRM reading is greater than a specified minimum count rate. This ensures that the detectors are indicating count rates indicative of neutron flux levels within the core. With few fuel assemblies (continued)
| |
| NMP2 B 3.3.1.2-6 Revision 0
| |
| | |
| SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.4 (continued)
| |
| REQUIREMENTS loaded, the SRMs will not have a high enough count rate to satisfy the SR. Therefore, allowances are made for loading sufficient "source" material, in the form of irradiated fuel assemblies, to establish the minimum count rate.
| |
| To accomplish this, the SR is modified by a Note that states that the count rate is not required to be met on an SRM that has less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies are in the associated core quadrant. With four or less fuel assemblies loaded around each SRM and no other fuel assemblies in the associated quadrant, even with a control rod withdrawn the configuration will not be critical.
| |
| The Froqucncy is based upon channcl rcdwndancy and ethc informfation ayailable in the control roomR, and enSUrcc that the rcquirod channcls arc frcqucntly mRonitocd whilo coro rcactivity changcs arc occurring. When no rcactivity
| |
| ;hangcs ar i*n*pO*gss, OR tho ,r y IS F.oqu.n. lacd from.
| |
| 12 hourF te 24 hoUS.
| |
| SR 3.3.1.2.5 and SR 3.3.1.2.6 Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated channel will function properly. SR 3.3.1.2.5 is required in MODE 5, and the 7 day FcqU,*..y ensures that the channels are OPERABLE while core reactivity changes could be in progress. This 7 day *F.quc..y is.oasonablo, baod on epcrating and on othcl SurlIiIlanc's (such as a CH4ANNEL CHECGK) that onsuro propor functioning betweon CHANNEL FUNCTIONAL.TESTS. Insert 3 SR 3.3.1.2.6 is required in MODE 2 with IRMs on Range 2 or below and in MODES 3 and 4. S-ncc co.. r.activity chan@gs do not nrmFFally tako placo in MOIDES 3 and 4 and coro Fcactivity changcs arc due only toVcotrol rod mvomlt l in MODE 2, the F..qucn.y has been extended from 7 days to 31 days. The 31 day F-oquenc*y is based on oepeating cxpor-cne and on othc- SuFveillan-,s (such as-CHANNEL CHEK) hat onsro~ propcr functioning between CHANNEL FUNICICTIONAL TESTS-(continued)
| |
| NMP2 B 3.3.1.2-7 Revision 0
| |
| | |
| SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.5 and SR 3.3.1.2.6 (continued)
| |
| REQUIREMENTS Verification of the signal to noise ratio also ensures that the detectors are inserted to a normal operating level. In a fully withdrawn condition, the detectors are sufficiently removed from the fueled region of the core to essentially eliminate neutrons from reaching the detector. Any count rate obtained while fully withdrawn is assumed to be "noise" only. With few fuel assemblies loaded, the SRMs will not have a high enough count rate to determine the signal to noise ratio. Therefore, allowances are made for loading sufficient "source" material, in the form of irradiated fuel assemblies, to establish the conditions necessary to determine the signal to noise ratio. To accomplish this, SR 3.3.1.2.5 is modified by a Note that states that the determination of signal to noise ratio is not required to be met on an SRM that has less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies are in the associated core quadrant. With four or less fuel assemblies loaded around each SRM and no other fuel assemblies in the associated quadrant, even with a control rod withdrawn the configuration will not be critical.
| |
| The Note to SR 3.3.1.2.6 allows the Surveillance to be delayed until entry into the specified condition of the Applicability. The SR must be performed in MODE 2 within 12 hours of entering MODE 2 with IRMs on Range 2 or below.
| |
| The allowance to enter the Applicability with the 31-day Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels. Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances.
| |
| SR 3.3.1.2.7 Performance of a CHANNEL CALIBRATION verifies the performance of the SRM detectors and associated circuitry.
| |
| The Frequency considers the plant conditions required to perform the test, the ease of performing the test, and the likelihood of a change in the system or component status.
| |
| (continued)
| |
| NMP2 B 3.3.1.2-8 Revision 0
| |
| | |
| SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.7 (continued)
| |
| REQUIREMENTS The neutron detectors are excluded from the CHANNEL CALIBRATION (Note 1) because they cannot readily be adjusted. The detectors are fission chambers that are designed to have a relatively constant sensitivity over the range, and with an accuracy specified for a fixed useful life.
| |
| Note 2 to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the Applicability. The SR must be performed in MODE 2 within 12 hours of entering MODE 2 with IRMs on Range 2 or below.
| |
| The allowance to enter the Applicability with the 24-FeRth Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels. Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances.
| |
| REFERENCES 1. USAR, Section 7.7.1.7.2.
| |
| NMP2 B 3.3.1.2-9 Revision 0
| |
| | |
| Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS E.1 and E.2 (continued)
| |
| LCO 3.1.1. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are therefore not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.
| |
| SURVEILLANCE As noted at the beginning of the SRs, the SRs for each Control Rod REQUIREMENTS Block instrumentation Function are found in the SRs column of Table 3.3.2.1-1.
| |
| The Surveillances are modified by a second Note to indicate that when an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability.
| |
| Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 7) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary.
| |
| SR 3.3.2.1.1 and SR 3.3.2.1.2 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function. The CHANNEL FUNCTIONAL TEST for the RWM is performed by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs and by attempting to select a control rod not in compliance with the prescribed sequence and verifying a selection error occurs. As noted in the SRs, SR 3.3.2.1.1 is not required to be performed until 1 hour after any control rod is withdrawn at < 10% RTP in MODE 2, and SR 3.3.2.1.2 is not required to be performed until 1 hour after THERMAL POWER is < 10% RTP in MODE 1. This allows entry into MODE 2 (and if entering during a shutdown, concurrent power reduction to < 10%
| |
| RTP) for SR 3.3.2.1.1, and THERMAL POWER reduction to < 10%
| |
| RTP in MODE 1 for SR 3.3.2.1.2, to perform the required Surveillances if the 92 day Frequency is not met per SR 3.0.2. The 1 hour (continued)
| |
| NMP2 B 3.3.2.1-8 Revision 0, 24 (A123)
| |
| | |
| Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.1 and SR 3.3.2.1.2 (continued)
| |
| REQUIREMENTS allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. pcratti.g has shown that these
| |
| .xpc.i.ncc compenonts usbually pass the S-uroiellanco Whon porformced at the 02 day Froqucncy. T-hcrcforc, the Frcequcncy was concIudod to be aceoptabll from a roliability Gtan dp*Int.
| |
| SR 3.3.2.1.3 Insert 3 A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended function. It includes the Reactor Manual Control Multiplexing System input.
| |
| Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. Th*c Fcqucncy of 181 days is based on the analyses in Rof.ronco8..
| |
| SR 3.3.2.1.4 The RBM setpoints are automatically varied as a function of power.
| |
| Three Allowable Values are specified in Table 3.3.2.1-1, each within a specific power range. The power at which the control rod block Allowable Values automatically change are based on the APRM signal's input to each RBM channel. Below the minimum power setpoint or if a peripheral control rod is selected, the RBM is automatically bypassed. These power Allowable Values must be verified periodically to be less than or equal to the specified values. If any power range setpoint is nonconservative, then the affected RBM channel is considered inoperable. Alternatively, the power range channel can be placed in the conservative condition (i.e., enabling the proper RBM setpoint). If placed in this condition, the SR is met and the RBM channel is not considered inoperable. As noted, neutron detectors are excluded from the Surveillance because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.3 and SR 3.3.1.1.7. Thc 24 month Fr..quny.-i.
| |
| basedr 3 the .(continued)
| |
| NMP2 B 3.3.2.1-9 Revision G, 24 (A123)
| |
| | |
| Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.5 REQUIREMENTS (continued) The RWM is automatically bypassed when power is above a specified value. The power level is determined from steam flow signals. The automatic bypass setpoint must be verified periodically to be > 10% RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodology are incorporated into the Allowable Value and the actual setpoint. If the RWM low power setpoint is nonconservative, then the RWM is considered inoperable. Alternately, the low power setpoint channel can be placed in the conservative condition (nonbypass). If placed in the nonbypassed condition, the SR is met and the RWM is not considered inoperable. The FrcquRncy is based on the trip sctp"int methodology utiliz fo.r tho loV powor G[tpnint channel.
| |
| SR 3.3.2.1.6 A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch - Shutdown Position Function to ensure that the entire channel will perform the intended function. The CHANNEL FUNCTIONAL TEST for the Reactor Mode Switch - Shutdown Position Function is performed by attempting to withdraw any control rod with the reactor mode switch in the shutdown position and verifying a control rod block occurs.
| |
| As noted in the SR, the Surveillance is not required to be performed until 1 hour after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links. This allows entry into MODES 3 and 4 if the 24". mneth Frequency is not met per SR 3.0.2. The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs.
| |
| linsert 3 I-- >Thc 24 month Frcqucncy is based on the need to porformA thi
| |
| ~ur:oiiianco unuor mu conc itncthat apply duFrig a plant outagc and the potcntial for an unplanned transllnt i8f. the Surveillancc wcrc pcrforffc G I.M.P.E.1 1 18 FrBaGEEK a1z p19 ur.
| |
| Uperating experioneo hacs shoWNR these compononts usually pasc thP crrma+~i Rncn WARRciiac r
| |
| FAF tne ;44 montn 1--rcouccV.
| |
| .................... *1¸ (continued)
| |
| (continued)
| |
| NMP2 B 3.3.2.1-10 Revision 0
| |
| | |
| Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SIR 3.3.2.1.7 REQUIREMENTS (continued) A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy.
| |
| As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal.
| |
| Neutron detectors are adequately tested in SR 3.3.1.1.3 and SR Ilnsert 3 3.3.1.1.7.
| |
| ii
| |
| ~The I=ro~iuoncY IGiDacd on theo analysis in KotoFrAno A.
| |
| SR 3.3.2.1.7 for SL-LSSS functions is modified by a Note as identified in Table 3.3.2.1-1. The Note requires declaring the channel inoperable if the as-found setpoint does not match the NTSP. Due to the digital nature of the RBM, there are no as-found or as-left tolerances. Identification of a channel setpoint different than the NTSP is indicative of a channel that is not functioning correctly.
| |
| The Note also requires that the NTSP be specified in the COLR and the methodology used to determine the NTSP be specified in the Bases (see Ref. 11).
| |
| SR 3.3.2.1.8 The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer. This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. The Surveillance is performed once prior to declaring RWM OPERABLE following loading of sequence into RWM, since this is when rod sequence input errors are possible.
| |
| REFERENCES 1. USAR, Section 7.7.1.7.
| |
| : 2. USAR, Section 7.7.1.3.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| : 4. USAR, Section 15.4.2.3.
| |
| : 5. USAR, Section 15.4.9.
| |
| (continued)
| |
| NMP2 B 3.3.2.1-11 Revision 0, 24 (A123)
| |
| | |
| Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES 6. NRC SER, "Acceptance of Referencing of Licensing (continued) Topical Report NEDE-2401 1-P-A," "General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17," December 27, 1987.
| |
| : 7. GENE-770-06-1-A, "Addendum To Bases For Changes To Surveillance Test Intervals And Allowed Out-of-Service Times For Selected Instrumentation Technical
| |
| -*-- Specifications," December 1992.
| |
| Stcblir tMycur-mot Analycic and q, M',
| |
| RCtroflt Pocu GO Stability' Trip. Function,;..,
| |
| QOctoboer 10959&
| |
| : 9. NEDO-33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," July 2004.
| |
| : 10. NEDC-33286P, "Nine Mile Point Nuclear Station Unit 2 -
| |
| APRM/RBM/Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)," March 2007.
| |
| : 11. NEDC-31336P-A, "General Electric Setpoint Methodology,"
| |
| September 1996.
| |
| NMP2 B 3.3.2.1-12 Revision ,-1-8-(A! 20), 24 (A123)
| |
| | |
| Feedwater System and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES ACTIONS C.1 and C.2 (continued)
| |
| With a channel not restored to OPERABLE status or placed in trip, THERMAL POWER must be reduced to < 23% RTP within 4 hours. As discussed in the Applicability section of the Bases, operation below 23% RTP results in sufficient margin to the required limits, and the Feedwater System and Main Turbine High Water Level Trip Instrumentation is not required to protect fuel integrity during the feedwater controller failure, maximum demand event. Alternately, if a channel is inoperable solely due to an inoperable feedwater pump breaker, the affected feedwater pump breaker may be removed from service since this performs the intended function of the instrumentation. The allowed Completion Time of 4 hours is based on operating experience to reduce THERMAL POWER to < 23% RTP from full power conditions in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the Function maintains feedwater system and main turbine high water level trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 4) assumption that 6 hours is the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the feedwater pumps and main turbine will trip when necessary.
| |
| SR 3.3.2.2.1 Performance of the CHANNEL CHECK onco evory 24 hours 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 something even more serious. A CHANNEL CHECK will detect (continued)
| |
| NMP2 B 3.3.2.2-5 Revision 0, 37 (A140)
| |
| | |
| Feedwater System and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES SURVEILLANCE SR 3.3.2.2.1 (continued)
| |
| REQUIREMENTS gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
| |
| 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 instrument has drifted outside its limits.
| |
| -- --31 FnSeR The F=FeA Jnnc~' V hi"nd nn nnnmtinn .~-I--..-J.
| |
| n~rir~ncr~ thit
| |
| -... o---..........
| |
| domonctratoc chnno f"',,.ro .... r..
| |
| . The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with the channels required by the LCO.
| |
| SR 3.3.2.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| LIIIO1*I I. ,J ** r I*c-.,** n-* ,,:0**, ,-;- h-t,,- -,*,,.;'k lh -,*-l,,¢ The Froeqwoncy of 92 days iGbased on roliability analysis (Ref.4).
| |
| SR 3.3.2.2.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| linsert 3 T--ý 1419 Frcequcncy is based upon the assumption of an 24 month ea"Wbatien intorval intho dotormnination of tho magnitudo of cciuimcmnt drift inthe sctrpoint analysis.
| |
| (continued)
| |
| NMP2 B 3.3.2.2-6 Revision 0
| |
| | |
| Feedwater System and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES SURVEILLANCE SIR 3.3.2.2.4 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the feedwater pump breakers and main turbine stop valves is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function. Therefore, if a main turbine stop valve or Insert feedwater pump breaker is incapable of operating, the In 3 associated instrumentation would also be inoperable.
| |
| 21 month Fqie,*,y is base d en the need to pc.Form thik SuOR-6oilAlane un-dor the con-ditions that apply duFrin a plant outago and the potential for an unplanncd transiont if the Surweillancc wcro pcrfeffmed with the rcactor at powcr.
| |
| Oporating oxpoFFrionc h-as -;hq.oRn that thoso compencntc Wuually pasG the Su~ilnowhen pe&Frmcd at the 24 month F=eq~eRey-REFERENCES 1. USAR, Sections 15.1.2 and A.15.1.2.
| |
| : 2. USAR, Sections 15.1 and 15.3.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| : 4. GENE-770-06-1-A, "Bases For Changes To Surveillance Test Intervals And Allowed Out-Of-Service Times For Selected Instrumentation Technical Specifications,"
| |
| December 1992.
| |
| NMP2 B 3.3.2.2-7 Revision 0
| |
| | |
| PAM Instrumentation B 3.3.3.1 BASES ACTIONS D..1 (continued)
| |
| This Required Action directs entry into the appropriate Condition referenced in Table 3.3.3.1-1. The applicable Condition referenced in the Table is Function dependent.
| |
| Each time an inoperable channel has not met the Required Action of Condition C and the associated Completion Time has expired, Condition D is entered for that channel and provides for transfer to the appropriate subsequent Condition.
| |
| E.1 For the majority of Functions in Table 3.3.3.1-1, ifthe Required Action and associated Completion Time of Condition C is not met, the plant must be placed in a MODE in which the LCO does not apply. This is done by placing the plant in at least MODE 3 within 12 hours.
| |
| The allowed Completion Times are reasonable, based on operating experience, to reach the required plant condition from full power conditions in an orderly manner and without challenging plant systems.
| |
| F. 1 Since alternate means of monitoring drywell radiation have been developed and tested, the Required Action is not to shut down the plant but rather to follow the directions of Specification 5.6.6. These alternate means may be temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. 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.
| |
| SURVEILLANCE As noted at the beginning of the SRs, the following SRs REQUIREMENTS apply to each PAM instrumentation Function in Table 3.3.3.1-1, except where identified in the SR.
| |
| The Surveillances are modified by a second Note to indicate that when a channel is placed in an inoperable status solely (continued)
| |
| NMP2 B 3.3.3.1 -10 Revision 0
| |
| | |
| PAM Instrumentation B 3.3.3.1 BASES SURVEILLANCE for performance of required Surveillances, entry into REQUIREMENTS associated Conditions and Required Actions may be delayed (continued) for up to 6 hours, provided the other required channel in the associated Function is OPERABLE. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
| |
| The 6 hour testing allowance is acceptable since it does not significantly reduce the probability of properly monitoring post-accident parameters, when necessary.
| |
| SR 3.3.3.1.1 Performance of the CHANNEL CHECK onc. ovor'y 31 days ensures that a gross instrumentation failure 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. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The high radiation instrumentation should be compared to similar plant instruments located in the drywell. For Function 8, the CHANNEL CHECK shall consist of verifying the valve is at its correct position by checking as appropriate its red, green, or analog indication in the control room.
| |
| 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 Insert 3 outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.
| |
| The Frcequcncy of 31 days is based upon plant opcrating With Fcgard to channcl OPERA~BILITY and drift I excicc Which domROnstratos that failuro of monr than ono channol of a ~jVfW ie lUeclo in any 31 dauy intcr~ai IS Farc.- I he~
| |
| (continued)
| |
| NMP2 B 3.3.3.1-11 Revision 0
| |
| | |
| PAM Instrumentation B 3.3.3.1 BASES SURVEILLANCE SR 3.3.3.1.1 (continued)
| |
| REQUIREMENTS CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of those displays associated with the channels required by the LCO.
| |
| SR 3.3.3.1.2 A CH,ANNEL CA LIBP,,A, TION iOpcorfrmod oe.'o' 24 monthc.
| |
| CHANNEL CALIBRATION is a complete check of the instrument loop including the sensor. The test verifies that the channel responds to the measured parameter with the necessary range and accuracy. For Function 5, the CHANNEL CALIBRATION shall consist of an electronic calibration of the channel, excluding the detector, for range decades
| |
| Ž_10 R/hour and a one point calibration check of the detector with an installed or portable gamma source for the range decade < 10 R/hour. For Function 8, the CHANNEL I*CALIBRATION shall consist of a position verification using ongioorig judgement.
| |
| REFERENCES 1. Regulatory Guide 1.97, "Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 3, May 1983.
| |
| : 2. USAR, Section 7.5.2.1.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.3.3.1-12 Revision 0, 25 (A124)
| |
| | |
| Remote Shutdown System B 3.3.3.2 BASES ACTIONS B.1 (continued) the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE The Surveillances are modified by a Note to indicate REQUIREMENTS that when an instrument channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
| |
| The 6 hour testing allowance is acceptable since it does not significantly reduce the probability of properly operating the associated equipment, when necessary.
| |
| SR 3.3.3.2.1 Performance of the CHANNEL CHECK on.. ovo.; 31 days 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 instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
| |
| 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. As specified in the Surveillance, a CHANNEL CHECK is only required for those channels that are normally energized.
| |
| (continued)
| |
| NMP2 B 3.3.3.2-4 Revision 0
| |
| | |
| Remote Shutdown System B 3.3.3.2 BASES SURVEILLANCE SR 3.3.3.2.1 (continued)
| |
| REQUIREMENTS
| |
| ~~tratcschanobase upon
| |
| , epoatfig xperi-*en; that
| |
| ;lhru;R re-r SR 3.3.3.2.2 SR 3.3.3.2.2 verifies each required Remote Shutdown System transfer switch and control circuit performs the intended function. This verification is performed from the remote shutdown panel. Operation of the equipment from the remote shutdown panel is not necessary. The Surveillance can be Ilnserl satisfied by performance of a continuity check. This will ensure that if the control room becomes inaccessible, the plant can be placed and maintained in MODE 3 from the remote shutdown pael >Ocating cxpcricncc dcmons~tratcs that 148FRE)te 11-8-11 -,-tE)FA GE)RIFel GHaRRe!G UGwally SaGG tHe SuF,,illancc when p*r.f*rmd at the 24 month Frcquony.
| |
| SR 3.3.3.2.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies the channel responds to measured parameter values with the necessary range and accuracy.
| |
| ->The 20 month Frcqucncy is based upon opcrating .. p..ic...
| |
| and F,, f,, ev,.,, judgmot, a,,* onginoo.in.
| |
| t F j G y:l.. v rv ,,e, and ic concictont v , . ~
| |
| With tho typical
| |
| , L . . ., * . ) i.
| |
| REFERENCES 1. USAR, Section 7.4.2.4.
| |
| : 2. 10 CFR 50, Appendix A, GDC 19.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| : 4. Technical Requirements Manual.
| |
| NMP2 B 3.3.3.2-5 Revision 0
| |
| | |
| EOC-RPT Instrumentation B 3.3.4.1 BASES SURVEILLANCE entered and Required Actions taken. This Note is based on REQUIREMENTS the reliability analysis (Ref. 4) assumption of the average (continued) time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the recirculation pumps will trip when necessary.
| |
| SR 3.3.4.1.1 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| The Frcq ucncy of 92 days as based on rcliability anialysis (Ref. 4).
| |
| SR 3.3.4.1.2 Insert 3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| Tho PFrqucncy is based upon the assumptien of an 24 mon~th Gallurji. A t..
| |
| ..... , iR the...
| |
| RIn..ie.. e+ t.. . . .i.u. .
| |
| of equipment drift in the stpint analyi*. aRd o.gin,,,ing judgment SR 3.3.4.1.3 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the pump breakers is included as a part of this test, overlapping the LOGIC SYSTEM FUNCTIONAL TEST, to provide complete testing of the associated safety function. Therefore, if a breaker is incapable of operating, the associated instrument channel would also be inoperable.
| |
| (continued)
| |
| NMP2 B 3.3.4.1-8 Revision 0
| |
| | |
| EOC-RPT Instrumentation B 3.3.4.1 BASES SURVEILLANCE SR 3.3.4.1.3 (continued)
| |
| REQUIREMENTS Thoh!- 2 month F,.roquocY ic.tbasod on the poody to apofr t r84e RSurvoillanco undorjP_ tho con~di9tionzR that apply duFrig a plant outagc and the potential for an unpla~nnd tranciont of the C. cirrqu
| |
| .NnIn c3rit7o nn~r -F-wro poll-1111-8 -4,r.
| |
| ....;A+k 1-11 +kn v r8clo +n~rn or a +.uApw Or.
| |
| Insert 3
| |
| .... ,L....... II ......
| |
| uPoraiiin oxporioncoI I Har- qtnAOw i f these comigenonw usually pass I III
| |
| * A th 6urvoiI"ancc t~rot ;An porrormoa at tho 24i montn S3e.4e.ey.
| |
| SR 3.3.4.1.4 This SR ensures that an EOC-RPT initiated from the TSV - Closure and TCV Fast Closure, Trip Oil Pressure - Low Functions will not be inadvertently bypassed when THERMAL POWER is Ž_26% RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint.
| |
| Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from first stage pressure), the main turbine bypass valves must remain closed during an in-service calibration at THERMAL POWER > 26% RTP to ensure that the calibration remains valid. If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at >_26% RTP either due to open main turbine bypass valves or other reasons), the affected TSV - Closure and TCV Fast Closure, Trip Oil Pressure - Low Functions are considered inoperable.
| |
| Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition, this SR is met and the channel Ilnsert 3 considered OPERABLE.
| |
| The F*rqu*n* y of 21-on*thc i- bascd on onginccring judgmcnt and rcliability of thce copencntS.
| |
| (continued)
| |
| NMP2 B 3.3.4.1-9 Revision 92-9, 37 (A140)
| |
| | |
| EOC-RPT Instrumentation B 3.3.4.1 BASES SURVEILLANCE SR 3.3.4.1.5 REQUIREMENTS (continued) This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. The EOC-RPT SYSTEM RESPONSE TIME acceptance criteria are included in Reference 5.
| |
| A Note to the Surveillance states that breaker arc suppression time may be assumed from the most recent performance of SR 3.3.4.1.6. This is allowed since the arc suppression time is short and does not appreciably change, due to the design of the breaker opening device and the fact that the breaker is not routinely cycled.
| |
| EGG RPT- SYSTEM RESPONSE TIME tests arc conductcd on an 24 mornth STAGGERED TEST BASIS. Response times cannot be determined at power because operation of final actuated devices is re dThorofor, tho 24 mneth Froquoncy d*yl and is based upon plant opcrating cxpcricncc, which shows that riandovim fai*usic of ieitRitumiintation compo*not* that iauco serleiouc re pencc tome dogradation, but not channol failur!e, aro in~froquont occurroncoc.
| |
| SR 3.3.4.1.6
| |
| 'Ilnser 31 This SR ensures that the EOC-RPT breaker arc suppression time is provided to the EOC-RPT SYSTEM RESPONSE TIME test.
| |
| dificultY of pm*fommRmRi v tho tost and the Frliabilit*
| |
| *f the circu-fit brcakcrci.
| |
| REFERENCES 1. USAR, Section 7.6.1.5.
| |
| : 2. USAR, Sections 15.2.2, 15.2.3, and 15.2.5.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| : 4. GENE-770-06-1 -A, "Bases For Changes To Surveillance Test Intervals And Allowed Out-Of-Service Times For Selected Instrumentation Technical Specifications,"
| |
| December 1992.
| |
| : 5. USAR, Table 7.6-7.
| |
| NMP2 B 3.3.4.1-10 Revision 0
| |
| | |
| ATWS-RPT Instrumentation B 3.3.4.2 BASES ACTIONS C.1 (continued)
| |
| Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within both Functions result in both Functions not maintaining ATWS-RPT trip capability. The description of a Function maintaining ATWS-RPT trip capability is discussed in the Bases for Required Action B.1, above.
| |
| The 1 hour Completion Time is sufficient for the operator to take corrective action and takes into account the likelihood of an event requiring actuation of the ATWS-RPT instrumentation during this period.
| |
| D.1 and D.2 With any Required Action and associated Completion Time not met, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours (Required Action D.2). Alternately, the associated recirculation pump breaker(s) may be removed from service since this performs the intended Function of the instrumentation (Required Action D.1). The allowed Completion Time of 6 hours is reasonable, based on operating experience, both to reach MODE 2 from full power conditions and to remove recirculation pump breaker(s) from service in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains ATWS-RPT trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 4) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the recirculation pumps will trip when necessary.
| |
| (continued)
| |
| NMP2 B 3.3.4.2-7 Revision 0
| |
| | |
| ATWS-RPT Instrumentation B 3.3.4.2 BASES SURVEILLANCE SR 3.3.4.2.1 REQUIREMENTS (continued) Performance of the CHANNEL CHECK nco, ovc."; 12 hourc 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 instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. 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.
| |
| 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 instrument has drifted outside its limit.
| |
| ~L. I~.. L IJ
| |
| ,,-+He r-ruguuney is Udiituu up19- -#purillng L-A[pc~Ric that d*.emAt.at .aF. ro1..*
| |
| csi*"nnol faiu The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of this LCO.
| |
| [Insert 3J SR 3.3.4.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| ý "Ifl,.= r-I---.. -Irrn A- - !- I-
| |
| ;0'~aJiI~!
| |
| r!I.JyLmý_ = =
| |
| U: U..
| |
| we :L, Aji; +R8U IUIIUI analysis 01 KefecRcrc co'i SR 3.3.4.2.3 Calibration of analog trip modules provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less (continued)
| |
| NMP2 B 3.3.4.2-8 Revision 0
| |
| | |
| ATWS-RPT Instrumentation B 3.3.4.2 BASES SURVEILLANCE SR 3.3.4.2.3 (continued)
| |
| REQUIREMENTS conservative than the Allowable Value specified in SR 3.3.4.2.5. If the trip setting is discovered to be less conservative than the setting accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the ATWS analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
| |
| analysis
| |
| . .. Refc,*,,2., ,
| |
| SR 3.3.4.2.4 This SR ensures that the LFMG breaker trip portion of the ATWS-RPT initiated from the Reactor Vessel Steam Dome Pressure - High Function will not be inadvertently bypassed for > 29 seconds when THERMAL POWER is > 5% RTP. This involves verification of the time delay and calibration of the APRM Downscale trip channel. Adequate margins for the r~ii instrument setpoint methodologies are incorporated into the actual APRM setpoint. If any time delay or APRM Downscale setpoint is nonconservative (i.e., the Reactor Vessel Steam Dome Pressure - High Function is bypassed for > 29 seconds when THERMAL POWER is > 5% RTP), the affected Reactor Vessel Steam Dome Pressure - High Function is considered inoperable.
| |
| Alternately, if only the APRM Downscale setpoint is nonconservative, the APRM channel can be placed in the conservative condition (e.g., placed in the inop trip condition). If placed in the conservative condition, this SR is met and the associated Reactor Vessel Steam Dome Pressure - High Function is considered OPERABLE.
| |
| >The F..qu...Y iG basod upon tho ac..umption of a 24 month eaIiR~atien intervaii in the deteurminati en oT flthe mayREIit of cquipmcnt drift in the APRM sopitaayi and is also based upo cgcrIg udcct and the rcliability of the tome dclay comonnb (continued)
| |
| NMP2 B 3.3.4.2-9 Revision 0
| |
| | |
| ATWS-RPT Instrumentation B 3.3.4.2 BASES SURVEILLANCE SR 3.3.4.2.5 REQUIREMENTS (continued) A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| no i-rcucnV is basoo unn the aGssumm~ion or an ;-1 Fmenm
| |
| .I i. f
| |
| * A calfibration intcr:alI Ain the dotoFrminationi of the m~agnitudc of eu.pmant drift in tho setpeint analysis andogiern judgment.
| |
| SR 3.3.4.2.6 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the pump breakers, included as part of this Surveillance, overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function. Therefore, if a breaker is incapable of operating, the associated instrument channel(s) would be inoperable.
| |
| ....... III*IILIIAM Ih rý- I--
| |
| 11 S iý--
| |
| J Mý nýý ^n Tm^ ýn T^ T Survoillancc undcr the coenditions that apply duFrin a plant oUtago and the potential for;i an uplanned transicnt if the Su Fvoillanco wo-ropo- rc With thc rcacto at power.
| |
| ORPrapting V-exooeriencae has shown that these GeRmoononP-tS uc6ua414
| |
| . .*lzo.. ---
| |
| Bass the uvi~ac when Be~frmed at the 24 moenth REFERENCES 1. USAR, Section 7.6.1.8.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| : 3. USAR, Section 15.8.
| |
| : 4. GENE-770-06-1 -A, "Bases For Changes To Surveillance Test Intervals And Allowed Out-of-Service Times For Selected Instrumentation Technical Specifications,"
| |
| December 1992.
| |
| NMP2 B 3.3.4.2-10 Revision 0
| |
| | |
| ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for (continued) performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours as follows: (a) for Functions 3.e, 3.g, 3.h, and 3.i; and (b) for Functions other than 3.e, 3.g, 3.h, and 3.i provided the associated Function or redundant Function maintains ECCS initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on reliability analyses (Refs. 5 and 6) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the ECCS will initiate when necessary.
| |
| SR 3.3.5.1.1 Performance of the CHANNEL CHECK ocoR cv,"r; 12 hours 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 instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
| |
| 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
| |
| ! l instrument has drifted outside its limit.
| |
| demc,"trates .hann.l failurc iz Fa. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
| |
| (continued)
| |
| NMP2 B 3.3.5.1-37 Revision 0
| |
| | |
| ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE SR 3.3.5.1.2 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| P AA I I I l II rio rPen 01 U dave lc based on the roliaiity SR 3.3.5.1.3 The calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be not within its required Allowable Value specified in Table 3.3.5.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analyses. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than the setting accounted for in the appropriate setpoint methodology.
| |
| .T-hc Frauncf 02- d-avzs iqba~cd;- onR the reliabilt P
| |
| aRalyScc v1 m omvomronco and v.
| |
| SR 3.3.5.1.4 and SR 3.3.5.1.5 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| ->Thc Fro.u.....of.S 3D*.3.. i azcd uoo t.hc....... +on of a02 ay... alibration
| |
| . . . . .... J...................
| |
| intcrwal in the .......
| |
| dctcrmlinationi r
| |
| of the r magnitd of cquipmon~t drft inthe sctpeint analysis. The I i Vi
| |
| * 214 ffmonthI coal 0 fibraion- intoPrgal inthe determfination of the magnitude of equipment drioft inthe zcpon anay.
| |
| (continued)
| |
| NMP2 B 3.3.5.1-38 Revision 0
| |
| | |
| ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE SR 3.3.5.1.6 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.5.1, LCO 3.5.2, LCO 3.8.1, and LCO 3.8.2 overlaps this Surveillance to provide complete testing of the assumed safety function.
| |
| Surveiltancoe un'dor tho cAROndiio that apply duFrin a plant outagc (cxccpt for the Divisien 3 DG, which can be tested in any oporational condition) and the potential for unplanned trancionts if the nc N1iuRvo9P ..RAWQr p--r-ormod-
| |
| -- ith the reactor at poWer.
| |
| __J.;__
| |
| oweatino cxocrience e na~ ~nown inc~c com~~oncnt~ u~uaii'.~ ua~
| |
| thc u,*.*eillane wfhen pwomod*4at the mnMtn ireruenRc.
| |
| REFERENCES 1. USAR, Section 5.2.
| |
| : 2. USAR, Section 6.3.
| |
| : 3. USAR, Chapter 15.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| : 5. NEDC-30936-P-A, "BWR Owners' Group Technical Specification Improvement Analyses for ECCS Actuation Instrumentation, Part 2," December 1988.
| |
| : 6. NEDC-30851-P-A, Supplement 2, "Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.
| |
| NMP2 B 3.3.5.1-39 Revision Q, 32 (A133)
| |
| | |
| RCIC System Instrumentation B 3.3.5.2 BASES SURVEILLANCE taken. This Note is based on the reliability analysis REQUIREMENTS (Ref. 2) assumption of the average time required to perform (continued) channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the RCIC will initiate when necessary.
| |
| SR 3.3.5.2.1 Performance of the CHANNEL CHECK nco, ovory 12 ho'rc 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 instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. 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.
| |
| 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 instrument has drifted outside its limit.
| |
| >The Froguency is based upon oporati thAt domonctratoc channol failurc i5 Frar. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
| |
| Insert 3 SIR 3.3.5.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| The Frequency of 92 days is based on the roliabRAiy analyses of Rcfernnnf,2.
| |
| NMP2 B 3.3.5.2-11 Revision 0
| |
| | |
| RCIC System Instrumentation B 3.3.5.2 BASES SURVEILLANCE SR 3.3.5.2.3 REQUIREMENTS (continued) The calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.5.2-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be re-adjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
| |
| hoFrqucy of 02 days as based on the rcliabty analycico9f RPofPoPRonc2.
| |
| SR 3.3.5.2.4 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter with the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| Insert 3 ,TheF...qu ..y is based on thc a..umptin of a 24 mo-nth calfibrationitrali
| |
| '. ' 9 LI'" *'..' *""'0
| |
| '"J ... .,v the .. dctorm~inati9A
| |
| : 1. A*
| |
| of tho magnitude
| |
| 'I-" .....
| |
| 11 V -1,--,"-*" -,, 11 1 SR 3.3.5.2.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.5.3 overlaps this Surveillance to provide complete testing of the safety function.
| |
| m II I
| |
| -oThc 24 mont'h FrOpUCwncv Is base ont he necd to pc*Or'" tfr" I
| |
| . - - - --- -- 'e.P outagc and the potcntial for an unplanncd tranzicnt if the Swrvcillanco wcro pcrformcdi with tho roactor at poWor.
| |
| (continued)
| |
| NMP2 B 3.3.5.2-12 Revision 0
| |
| | |
| RCIC System Instrumentation B 3.3.5.2 BASES SURVEILLANCE SR 3.3.5.2.5 (continued)
| |
| REQUIREMENTS Oprating oxprinc, has .hown that there ,..*-*,,*.nt- u, ,lhY pass tho Su-,oillan," whon porFo*nmod at the 24 moneth FFequency.
| |
| REFERENCES 1. 10 CFR 50.36(c)(2)(ii).
| |
| : 2. GENE-770-06-2-A, "Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," December 1992.
| |
| NMP2 B 3.3.5.2-13 Revision 0
| |
| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1 BASES (continued)
| |
| SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Primary Containment Isolation Instrumentation Function are found in the SRs column of Table 3.3.6.1-1.
| |
| The Surveillances are also modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains isolation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analyses (Refs. 11 and 12) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the PCIVs will isolate the penetration flow path(s) when necessary.
| |
| SR 3.3.6.1.1 Performance of the CHANNEL CHECK oc.o .v..y 12 ho.. 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 instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
| |
| 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 la___..*. ,-i I instrument has drifted outside its limit.
| |
| LlEnsýýerto ý.
| |
| ~The FrcequcncY as based on oporating experienco that d....t*ratcc .hann.lfai.,.. ic ..... The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
| |
| (continued)
| |
| NMP2 B 3.3.6.1-34 Revision 0
| |
| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.2 REQUIREMENTS (continued) This Surveillance verifies that, when the Allowable Value for the Main Steam Line Tunnel Lead Enclosure Temperature - High Function is adjusted based on the formula in footnote (b) to Table 3.3.6.1-1, the actual ambient temperature, as measured by the Main Steam Line Tunnel Lead Enclosure Temperature - High Function channels, is greater than or equal to the ambient temperature (Tam.b) used to adjust the Allowable Value. Only the OPERABLE Main Steam Line Tunnel Lead Enclosure Temperature - High Function channels are required to be verified. As stated in the Note to the SR, the SR is only required to be met when the Allowable Value is adjusted in accordance with Table 3.3.6.1-1, footnote (b), since the normal Allowable Value is based on a sufficiently low ambient temperature that the verification is not necessary.
| |
| j, .- no hrcnLJcnfl'.~ 0? 1~ floum
| |
| ~i -..-J-............ flfl~Od on mc noon to pe*i*divally monitor the ambient tcmpcvatuWr to enGnrn the Allowablc Value Femains valid, and wac chocon to coincido with thc CHANNEL CHECK Frcqucqcy. As required by.SR 3.0.1, the SR must also be performed prior to adjusting the Allowable Value, since the Surveillance must be met at all times when the Allowable Value has been adjusted (thus to meet the SR when the Allowable Value has been adjusted, it
| |
| ~iI1 must actually be performed prior to the adjustment).
| |
| SR 3.3.6.1.3 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| The, IF r c y e 92- dayc is ua,
| |
| * u U,* ruI,14bl',y i'-,ayjs, dczcribcd in Refercnees 11 and 1 2.
| |
| SR 3.3.6.1.4 The calibration of trip units consists of a test to provide a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in (continued)
| |
| NMP2 B 3.3.6.1-35 Revision 0
| |
| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.4 (continued)
| |
| REQUIREMENTS Table 3.3.6.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
| |
| hrounev ot 92z daysis nased An dI The1-e f A* I I i I J - - - - __ - - -
| |
| th rciianiim11 SR 3.3.6.1.5 CHANNEL CALIBRATION is a complete check of the instrument
| |
| ~i~J loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument -drifts between successive calibrations, consistent with the plant specific setpoint methodology.
| |
| cThobroquncty is based on thc accumptin of a 24 month calibhrrAtio intoval in the detcrmfinatien of the maaRNWItud oteaowimont drltt in thec zcteeint analvcis.
| |
| SR 3.3.6.1.6 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel. The system functional testing performed on PCIVs in LCO 3.6.1.3 overlaps this Surveillance to provide complete testing of the assumed safety function.
| |
| 24 mon~th iFrquoncy is based on the Reed to prform this Swuyecillancc undcr the eonditiens that apply during a plant outage and the potontial for An Iunplannod trancicnt if the 44 k 4-k
| |
| -..---..-......ii ,
| |
| mc ~urciI~ ....enn pcrrormcci at the 24 mentn I-rcqucnay.
| |
| (continued)
| |
| (continued)
| |
| NMP2 B 3.3.6.1-36 Revision 0
| |
| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.7 REQUIREMENTS (continued) This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. Testing is performed only on channels where the assumed response time does not correspond to the diesel generator (DG) start time. For channels assumed to respond within the DG start time, sufficient margin exists in the 10 second start time when compared to the typical channel response time (milliseconds) so as to assure adequate response without a specific measurement test. The instrument response times must be added to the PCIV closure times to obtain the ISOLATION SYSTEM RESPONSE TIME.
| |
| However, failure to meet the ISOLATION SYSTEM RESPONSE TIME due to a PCIV closure time not within limits does not require the associated instrumentation to be declared inoperable; only the PCIV is required to be declared inoperable. ISOLATION SYSTEM RESPONSE TIME acceptance criteria are included in Reference 13.
| |
| A Note to the Surveillance states that the response time of the sensors may be assumed to be the design sensor response time and therefore, are excluded from the ISOLATION SYSTEM RESPONSE TIME testing. This is allowed since the sensor response time for the affected Functions (Functions 1 .a, 1 .b, and 1 .c) is a small part of the overall ISOLATION SYSTEM RESPONSE TIME (Ref. 14).
| |
| Ilnsert 3 ISOLATION SYSTEM RESPONSE TIME tcsts arc condu-ctd on a 24 month S.TAGGRFE_,D,- TE-ST B-,,-,-RASnIS. The 214 mon.th test F-qU........
| |
| isconsistcnt with the typical industry rcfueling cyclo and is based upon plant opcrating cxpcricncc that shows that random failurcc of instrumon~tation comRp9nontc caUucig sorious rccpense timo dogradation, but not channcl failurc, a~e iRfequeRt.
| |
| REFERENCES 1. USAR, Table 6.2-56.
| |
| : 2. USAR, Section 6.2.
| |
| : 3. USAR, Chapter 15 and Appendix A.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| : 5. USAR, Section 15.1.3.
| |
| (continued)
| |
| NMP2 B 3.3.6.1-37 Revision 0
| |
| | |
| Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ACTIONS C.1.1, C.1.2, C.2.1. and C.2.2 (continued)
| |
| One hour is sufficient for plant operations personnel to establish required plant conditions or to declare the associated components inoperable without challenging plant systems.
| |
| SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Secondary Containment Isolation instrumentation Function are located in the SRs column of Table 3.3.6.2-1.
| |
| The Surveillances are also modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains isolation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Action(s) taken.
| |
| This Note is based on the reliability analysis (Refs. 4 and 5) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the SCIVs will isolate the associated penetration flow paths and the SGT System will initiate when necessary.
| |
| SR 3.3.6.2.1 Performance of the CHANNEL CHECK nco, cvcry 12 hourc ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the indicated parameter for one instrument 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 instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
| |
| (continued)
| |
| NMP2 B 3.3.6.2-9 Revision 0
| |
| | |
| Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE SR 3.3.6.2.1 (continued)
| |
| REQUIREMENTS 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 instrument has drifted outside its limit.
| |
| - -rL-- r-__
| |
| dcmnstlRtcs channcl fail'-rc i Farc. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
| |
| SR 3.3.6.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| >Thcrcquncy of 92 days is ba .. d upon the r-liability; SR 3.3.6.2.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.6.2-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
| |
| >T ho 1-Fquonc'; ot 92 days, is b9ascd on the rcl a13 1ty anaii~is ef I-Pfecronc4 and 55.
| |
| ..... ,Z .....
| |
| (continued)
| |
| NMP2 B 3.3.6.2-10 Revision 0
| |
| | |
| Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE SR 3.3.6.2.4 REQUIREMENTS (continued) CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| am F A *
| |
| - i no -.......... Yi ,-a- upon the a.-umpt,,n A? 3* 24 ,month calibhratioi the dotorminRatin*
| |
| ,f the mRait- 'PIP
| |
| .. equ-pm t drit m in - -the-*tpnt analycic.
| |
| SR 3.3.6.2.5 Insert 3 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel. The system functional testing, performed on SCIVs and the SGT System in LCO 3.6.4.2 and LCO 3.6.4.3, respectively, overlaps this Surveillance to provide complete testing of the assumed safety function.
| |
| - I he 24 Month hrcaucncv is based on the need to pei~eFm th 6 I L, :11 "PAR RAGe bindeF tHe GE)RE3111ORS tHat apply GURR9 a plaR mutag and the petnntianm ior an unplanned trannient ni then 56urOPoilanco woro pedformo-d With the roactor at pewer. *lh,
| |
| * Ii A .,, Ir . . . . . . . . . .
| |
| mc~ ~ ~~. pueneanowhnDoomecl at the 21 month brpuoncY.-
| |
| g g REFERENCES 1. USAR, Section 15.6.5.
| |
| : 2. USAR, Section 15.7.4.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| : 4. NEDC-31677-P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation,"
| |
| July 1990.
| |
| : 5. NEDC-30851-P-A, Supplement 2, "Technical Specifications Improvement Analysis for BWR Isolation Instrumentations Common to RPS and ECCS Instrumentation," March 1989.
| |
| NMP2 B 3.3.6.2-11 Revision 0
| |
| | |
| CREF System Instrumentation B 3.3.7.1 BASES ACTIONS D.1 and D.2 (continued)
| |
| With any Required Action and associated Completion Time not met, the associated CREF subsystem must be placed in the emergency pressurization mode of operation (Required Action D. 1) to ensure that control room personnel will be protected in the event of a Design Basis Accident. The method used to place the CREF subsystem in operation must provide for automatically reinitiating the subsystem upon restoration of power following a loss of power to the CREF subsystem. Alternately, if it is not desired to start the subsystem, the CREF subsystem associated with inoperable, untripped channels must be declared inoperable within 1 hour.
| |
| The 1 hour Completion Time is intended to allow the operator time to place the CREF subsystem in operation. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels, or for placing the associated CREF subsystem in operation.
| |
| SURVEILLANCE As noted at the beginning of the SRs, the SRs for each CREF REQUIREMENTS System Instrumentation Function are located in the SRs column of Table 3.3.7.1-1.
| |
| The Surveillances are also modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains CREF System initiation capability. Upon completion of the surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Refs. 5, 6, and 7) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the CREF System will initiate when necessary.
| |
| SR 3.3.7.1.1 Performance of the CHANNEL CHECK ocRc evcr;y 12 hourc ensures that a gross failure of instrumentation has not occurred. A (continued)
| |
| NMP2 B 3.3.7.1-8 Revision 0
| |
| | |
| CREF System Instrumentation B 3.3.7.1 BASES SURVEILLANCE SR 3.3.7.1.1 (continued)
| |
| REQUIREMENTS CHANNEL CHECK is normally a comparison of the indicated parameter for one instrument 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 instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
| |
| 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 instrument has drifted outside its limit.
| |
| >The PFrcgcncy is based upon oporating cxpericncc that d,..on.tato. ,hann. l fail.. .. c Fa. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with channels required by the LCO.
| |
| Insert 31 SR 3.3.7.1.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| >The PFrquoncY of 92 days is based on the roliability analyseG of Rcfcrcnccz 5, 66,and 7.
| |
| SR 3.3.7.1.3 The calibration of trip units provides a check of the actual trip setpoints. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value. If the trip setting is discovered to be (continued)
| |
| NMP2 B 3.3.7.1-9 Revision 0
| |
| | |
| CREF System Instrumentation B 3.3.7.1 BASES SURVEILLANCE SR 3.3.7.1.3 (continued)
| |
| REQUIREMENTS less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
| |
| XThe Frociucncy of 92 days is basod onA the reliabilit' anIV e =Trwcccc .~ n SR 3.3.7.1.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| lizsr I J
| |
| *m The P-rcciuncy is based on the assumetion ot a 24 montm AA1*hFAt8AR thp dPtPr.R;4RAt AR Af the rRý mob We SR 3.3.7.1.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.7.2, "Control Room Envelope Filtration (CREF) System,"
| |
| overlaps this Surveillance to provide complete testing of the assumed safety function.
| |
| ->1 hr-'O A1 mnth4 [Frreiiirnre hisbase enf the' needr te, me":-Frm thi" Survcillanco undcr the conditions that apply during a plant outage and the potential for an unplannced transient if the Surveillanee were pcr-formcd with the reactor at power.
| |
| i-- intratina r ........
| |
| * tevnerie~nce
| |
| ---- r nas* snowAn.............
| |
| tntss comoone*ntts r
| |
| uluanlyU~ eas m
| |
| thinR'-inrriin w~hion nr'rnamcmac mi thin 2A mnnth4 Frrriiirnn"a-(continued)
| |
| NMP2 B 3.3.7.1-10 Revision 0
| |
| | |
| Mechanical Vacuum Pump Isolation Instrumentation B 3.3.7.2 BASES ACTIONS C.1, C.2. C.3, and C.4 (continued)
| |
| With any Required Action and associated Completion Time not met, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours (Required Action C.4). Alternately, the associated mechanical vacuum pump may be removed from service since this performs the intended function of the instrumentation (Required Actions C.1 and C.2). An additional option is provided to isolate the main steam lines (Required Action C.3), which may allow operation to continue. Isolating the main steam lines effectively provides an equivalent level of protection by precluding fission product transport to the condenser.
| |
| The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions, or to remove the mechanical vacuum pump from service, or to isolate the main steam lines, in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours provided mechanical vacuum pump isolation trip capability is maintained. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
| |
| This Note is based on the reliability analysis (Ref. 5) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the mechanical vacuum pumps will trip and isolate when necessary.
| |
| SR 3.3.7.2.1 Performance of the CHANNEL CHECK ncc cvcry 12 hou'rc 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 (continued)
| |
| NMP2 B 3.3.7.2-5 Revision 0
| |
| | |
| Mechanical Vacuum Pump Isolation Instrumentation B 3.3.7.2 BASES SURVEILLANCE SR 3.3.7.2.1 (continued)
| |
| REQUIREMENTS channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
| |
| 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 instrument has drifted outside its limit.
| |
| >The Frcgquency is based upon 9pcrating cxperiencc that d.m.nstats*
| |
| , ,a... The CHANNEL CHECK hann, l fail-,, is supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of this LCO.
| |
| Insert 3 SR 3.3.7.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| >Tho Froquoncy "*of 02 d... is ba.sod. on. tho ro.,iabilit, analysis of Rofc~rcnco 5 SR 3.3.7.2.3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| (continued)
| |
| NMP2 B 3.3.7.2-6 Revision 0
| |
| | |
| Mechanical Vacuum Pump Isolation Instrumentation B 3.3.7.2 BASES SURVEILLANCE SR 3.3.7.2.3 (continued)
| |
| REQUIREMENTS In X Th F=Feawerocloy 1G D3aced upon theo acumqPtion 01 A :21 mentv; d g
| |
| *l i aiior-atmo lnrire a, in nte e;meFrrm-ia"R-A n Gn- iR0A-
| |
| -T ;n' mR'--R4r-A ot cpUimonFP-t d-reft in the sctEpofint analyrcis.
| |
| SR 3.3.7.2.4 lnsert 3 ] The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the mechanical vacuum pump breakers and isolation valve is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function. Therefore, if a breaker or the isolation valve is incapable of operating, the associated instrument channel(s) would be inoperable.
| |
| k 9i A J li II I I f LI a I- - +~Ttr~lLAf
| |
| &_v- 11 1 MI 1t tI
| |
| +~ ^ m- - ^ ^ lii l Sur-ve9ianc under the.o that apply during a plant
| |
| ,dition, outage and the potontial for an unplanned t.ansi.nt if the ar IKM
| |
| *I lanrmý tpirp na^r.f larm~l j.,.fk +k rganntr n+ ^ ^atjr r,
| |
| mE Uneratinci expcriencc has sho.. nthese coemponcnts usually pass g
| |
| i dl
| |
| * i II
| |
| ~u...
| |
| .n .ncoiac
| |
| ......... ..... .......hco ofc
| |
| * . w--I ..... .r ° REFERENCES 1. USAR, Section 7.3.1.1.2.
| |
| : 2. USAR, Section 10.4.2.
| |
| : 3. USAR, Section 15.4.9.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| : 5. NEDC-30851 -P-A, "Supplement 2, "Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.
| |
| NMP2 B 3.3.7.2-7 Revision 0
| |
| | |
| LOP Instrumentation B 3.3.8.1 BASES ACTIONS A.1 (continued) allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.
| |
| Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the channel in trip would result in a DG initiation), Condition B must be entered and its Required Action taken.
| |
| The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
| |
| B.1 If any Required Action and associated Completion Time is not met, the associated Function may not be capable of performing the intended function. Therefore, the associated DG(s) are declared inoperable immediately. This requires entry into applicable Conditions and Required Actions of LCO 3.8.1 and LCO 3.8.2, which provide appropriate actions for the inoperable DG(s).
| |
| SURVEILLANCE As noted at the beginning of the SRs, the SRs for each LOP REQUIREMENTS Instrumentation Function are located in the SRs column of Table 3.3.8.1-1.
| |
| The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 2 hours provided the associated Function maintains LOP initiation capability. LOP initiation capability is maintained provided bus load shedding control scheme can be initiated by the Loss of Voltage or Degraded Voltage Functions for two of the three 4.16 kV emergency buses.
| |
| Upon completion of the Surveillance, or expiration of the 2 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
| |
| (continued)
| |
| NMP2 B 3.3.8.1-6 Revision 0
| |
| | |
| LOP Instrumentation B 3.3.8.1 BASES SURVEILLANCE SR 3.3.8.1.1 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| m d w m mI* m m*UUUN mug UI U B LA.y.JICLIUCFLLA
| |
| -- y- - - i V-1 " wv-. LAti*
| |
| J*LA* it
| |
| . . -1 . :.1..
| |
| -4 4. 6 i 01=MADII 1-rV .4 A :4 UrdU1UILU-M9~FA8G LfIJl RA Iiu-e A-+ itiuri. ui...i ARA- GR.A;..-
| |
| SR 3.3.8.1.2 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| I lInsr m I 71hc N"
| |
| Ircaucnc 4-:
| |
| Icbased on thec assumpt on ot a 21 montf A
| |
| P X of equipment drift inthe sctpeint analysis.
| |
| SR 3.3.8.1.3 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required actuation logic for a specific channel. The system functional testing performed in LCO 3.8.1 and LCO 3.8.2 overlaps this Surveillance to provide complete testing of the assumed safety functions.
| |
| 4
| |
| .Th,-, ')A ý^n +k r.,, ,,n,,, . k
| |
| *o.., p,- ^m +k^ m l +^ ^.rFr, +h.L mmr#*mmui*Fn unur tnc eennitienn that apply amUrmng a plant eutage and theopotntial for An u- ,planncdtrancic.*,
| |
| . t fifthe SuR'c0Ianc r pcfoRmcd with the rcactor at peWcr.
| |
| A I II Upcrating cxpcricncc has di own tncscc cmpon nts usually pass
| |
| * "1l l the Survciiianc when pmreormoa at mo 24 monmn t equcncy (continued)
| |
| (continued)
| |
| NMP2 B 3.3.8.1-7 Revision 0
| |
| | |
| RPS Electric Power Monitoring-Logic B 3.3.8.2 BASES SURVEILLANCE SR 3.3.8.2.1 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each overvoltage, undervoltage, and underfrequency channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| The 484-day Frequency is based on guidance provided in Generic Letter 91-09 (Ref. 3).
| |
| SR 3 2 R3.3I8.2n.2 3 I CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| calibration ntAP.a1 i thoe doomnto f the magnitude of 8quipmont drift i the sctpoint analysis.
| |
| S R ... 2
| |
| * S3.3.8.2.3.3 Performance of a system functional test demonstrates that, with a required system actuation (simulated or actual) signal, the logic of the system will automatically trip open the associated power monitoring assembly circuit breaker.
| |
| Only one signal per power monitoring assembly is required to be tested. This Surveillance overlaps with the CHANNEL CALIBRATION to provide complete testing of the safety function. The system functional test of the Class 1E circuit breakers is included as part of this test to provide complete testing of the safety function. If the breakers lInsert 3 1 *'" are incapable
| |
| ,monitoring of operating, assembly would the be associated inoperable. electric power Survcillancc undcr the conditions that apply dUFInlg a plant eutage and the petcntial for an unplanned transient if the Surye*Iancc wcrc pcrformced with the rcactor at pewcr.
| |
| (continued)
| |
| NMP2 B 3.3.8.2-7 Revision0, 4 (A99)
| |
| | |
| RPS Electric Power Monitoring - Logic B 3.3.8.2 BASES SURVEILLANCE SR 3.3.8.2.3 (continued)
| |
| REQUIREMENTS Oporating expcrienco hAc; Ahown 9 that those compenontG ucually pass tho 9uRgoillanco whon porformo;Aed at tho- 2-4 month Frcequc,",y.-
| |
| REFERENCES 1. USAR, Section 8.3.1.1.3.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| : 3. NRC Generic Letter 91-09, "Modification of Surveillance Interval for the Electric Protective Assemblies in Power Supplies for the Reactor Protection System."
| |
| NMP2 B 3.3.8.2-8 Revision 0
| |
| | |
| RPS Electric Power Monitoring - Scram Solenoids B 3.3.8.3 BASES SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when an RPS electric power monitoring assembly is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the other RPS electric power monitoring assembly for the associated RPS scram solenoid bus maintains trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the assembly must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This 6 hour allowance is acceptable since it does not significantly reduce the probability that the RPS electric power monitoring assembly function will initiate when necessary.
| |
| SR 3.3.8.3.1 A CHANNEL FUNCTIONAL TEST is performed on each overvoltage, undervoltage, and underfrequency channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| >Thc 181 day FrFouoncy is based on .uidancc .. ,vidcdin I -4,-~.- fl4 AA ~ ~
| |
| WN "If, " "I Wv r i k V - i.
| |
| SR 3.3.8.3.2 CHANNEL CALIBRATION is a complete check of the instrument linsert 3 1 loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| "IPL p !-- I. ... I - -- -- I ot cpuafflmcnt dr-Iff intheo sct~oent analys s.
| |
| SR 3.3.8.3.3 Performance of a system functional test demonstrates that, with a required system actuation (simulated or actual)
| |
| (continued)
| |
| NMP2 B 3.3.8.3-6 RevisionO , 4 (A99)
| |
| | |
| RPS Electric Power Monitoring - Scram Solenoids B 3.3.8.3 BASES SURVEILLANCE SR 3.3.8.3.3 (continued)
| |
| REQUIREMENTS signal, the logic of the system will automatically trip open the associated power monitoring assembly circuit breaker.
| |
| Only one signal per power monitoring assembly is required to be tested. This Surveillance overlaps with the CHANNEL CALIBRATION to provide complete testing of the safety function. The system functional test of the Class 1E circuit breakers is included as part of this test to provide complete testing of the safety function. If the breakers are incapable of operating, the associated electric power monitoring assembly would be inoperable.
| |
| Surycillancce undcr the conditions that apply during a plant outago and tho potential for an unplAnnodj tAraniont if the Survoillanco worc perfoFrmod with the rcact9r at powcr.
| |
| -! ..... L---- -- I ....... JLL __4. L u~oratina ox~orionco nac snown mat mnoco comoononic ucuaiiv pass the u~oIac whcn pcrtormea at the 24 montr
| |
| - egue.ey-REFERENCES 1. USAR, Section 8.3.1.1.3.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| * NRC Gon;ric. ,.o.or 91 00, ,Mod.ific*.ation o Su..eillanc1 Itcr,-RIA.l fo-,r the Elcctric Protc;tivc A,,1cmbVic, in Powcr Supplies for tho Roacteo Protcction Syctem."
| |
| NMP2 B 3.3.8.3-7 Revision 0
| |
| | |
| Recirculation Loops Operating B 3.4.1 BASES (continued)
| |
| SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loop flows are within the allowable limits for mismatch. At low core flow (i.e.,
| |
| effective core flow < 70% of rated core flow), the APLHGR and MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when effective core flow is < 70% of rated core flow. The jet pump loop flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop. The effective core flow shall be calculated by assuming both loops are at the smaller value of the two jet pump loop flows.
| |
| The mismatch is measured in terms of percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered not in operation.
| |
| This SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation.
| |
| Ilnsert3 *'i',.*
| |
| I,. '*.. II.. .. .. .- .~... ... .... :IL L ,ILL I. . .... Jr jet pump OPERABILIT,,* Yvificatin. and has been ,hown by oporating oproc ob dqaot oetofnra o PUMP loop flo'S in tIm II I II I nII -
| |
| REFERENCES 1. USAR, Section 6.3 and Appendix A Section 6.
| |
| : 2. USAR, Section 6.3.3.7.
| |
| : 3. USAR, Section 5.4.1.3.
| |
| : 4. USAR Chapter 15B.
| |
| : 5. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.4.1-6 Revision0, 1
| |
| | |
| FCVs B 3.4.2 BASES ACTIONS A.1 (continued) violation of the Safety Limit MCPR. The FCVs are designed to lockup (high drywell pressure interlock) under LOCA conditions. When the FCVs "lock-up," the recirculation flow coastdown is adequate and the resulting calculated clad temperatures are acceptable. In addition, it has been calculated with the FCVs closing at the specified limit, the resulting calculated clad temperatures will also be acceptable. Closing an FCV faster than the limit assumed in the LOCA analysis (Ref. 3) could affect the recirculation flow coastdown, resulting in higher peak clad temperatures.
| |
| Therefore, if an FCV is inoperable, deactivating the valve will essentially lock the valve in position, which will prohibit the FCV from adversely affecting the DBA and transient analyses. Continued operation is allowed in this Condition.
| |
| The 4 hour Completion Time is a reasonable time period to complete the Required Action, while limiting the time of operation with an inoperable FCV.
| |
| B.1 If the FCVs are not deactivated ("locked up") within the associated Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 12 hours. This brings the unit to a condition where the flow coastdown characteristics of the recirculation loop are not important. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems.
| |
| SURVEILLANCE SR 3.4.2.1 REQUIREMENTS Hydraulic power unit pilot operated isolation valves located between the servo valves and the common "open" and "close" lines are required to close in the event of a loss of hydraulic pressure. When closed, these valves inhibit FCV motion by blocking hydraulic pressure from the servo valve to the common open and close lines as well as to the alternate subloop. This Surveillance verifies FCV lockup on a loss of hydraulic pressure.
| |
| (continued)
| |
| NMP2 B 3.4.2-3 Revision 0
| |
| | |
| FCVs B 3.4.2 BASES SURVEILLANCE SR 3.4.2.1 (continued)
| |
| REQUIREMENTS The 24 month FrFequency is based on tho Recd to perfoFrm thi Survillaco udor-thoeodiin that apply during a plant TutagS anrd the potertial for an unplanned tFansient of tho Suraollncoiwe pFsimainainedwith the realtyr at power.
| |
| Oporating experienco has shown the so componcntG usually pass the SR when performed at the 24 mon~th FrFegucncY. Thferefoe-,
| |
| the Froguoncy was concluded to be acceptable from a reliability Standpoint.
| |
| SR 3.4.2.2 This SR ensures the overall average rate of FCV movement at all positions is maintained within the analyzed limits.
| |
| The 21 mont; FrFequency is jbaG9o onR the neePd- to pel49rmf thi Survillnceunder the conditions that apply durFing a plant outage and the potential for an unplanned transient if the Sureilanc wee prrme wih te ractr t power.
| |
| Operating experiencee has shown these componients usually pass 4.R e 2r-P %
| |
| wnun BcriolRmc At- *HP a-44 mon irogquencY.+ re*ret-*r.
| |
| the Frequency was eoncluded to be ace. eptable #eRma Fcliability standpoint.
| |
| REFERENCES 1. USAR, Section 15.3.2.
| |
| : 2. USAR, Section 15.4.5.
| |
| : 3. USAR, Section 6.3.3.4.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.4.2-4 Revision 0
| |
| | |
| Jet Pumps B 3.4.3 BASES (continued)
| |
| SURVEILLANCE SR 3.4.3.1 REQUIREMENTS This SR is designed to detect significant degradation in jet pump performance that precedes jet pump failure (Ref. 3).
| |
| This SR is required to be performed only when the loop has forced recirculation flow since surveillance checks and measurements can only be performed during jet pump operation. The jet pump failure of concern is a complete mixer displacement due to jet pump beam failure. Jet pump plugging is also of concern since it adds flow resistance to the recirculation loop. Significant degradation is indicated if any two of the three specified criteria confirm unacceptable deviations from established patterns or relationships. The allowable deviations from the established patterns have been developed based on the variations experienced at plants during normal operation and with jet pump assembly failures (Refs. 3 and 4). The baseline patterns were established during the startup test program. Since refueling activities (fuel assembly replacement or shuffle, as well as any modifications to fuel support orifice size or core plate bypass flow) can affect the relationship between core flow, jet pump flow, and recirculation loop flow, these relationships may need to be re-established each cycle. Similarly, initial entry into extended single loop operation may also require establishment of these relationships. During the initial weeks of operation under such conditions, while baselining new "established patterns", engineering judgement of the daily Surveillance results is used to detect significant abnormalities which could indicate a jet pump failure.
| |
| The recirculation flow control valve (FCV) operating characteristics (jet pump loop flow versus FCV position) are determined by the flow resistance from the loop suction through the jet pump nozzles. A change in the relationship may indicate a flow restriction, loss in pump hydraulic performance, leak, or new flow path between the recirculation pump discharge and jet pump nozzle. For this criterion, the jet pump loop flow versus FCV position relationship must be verified.
| |
| Jet pump loop flow can be determined from measurement of the recirculation loop drive flow. Once this relationship has been established, increased or reduced jet pump loop flow for the same recirculation loop drive flow may be an indication of failures in one or several jet pumps.
| |
| (continued)
| |
| NMP2 B 3.4.3-3 Revision 0
| |
| | |
| Jet Pumps B 3.4.3 BASES SURVEILLANCE SR 3.4.3.1 (continued)
| |
| REQUIREMENTS Individual jet pumps in a recirculation loop typically do not have the same flow. The unequal flow is due to the drive flow manifold, which does not distribute flow equally to all risers. The jet pump diffuser to lower plenum differential pressure pattern or relationship of one jet pump to the loop average is repeatable. An appreciable change in this relationship is an indication that increased (or reduced) resistance has occurred in one of the jet pumps.
| |
| The deviations from normal are considered indicative of a potential problem in the recirculation drive flow or jet pump system (Ref. 3). Normal flow ranges and established jet pump differential pressure patterns are established by plotting historical data as discussed in Reference 3.
| |
| to be adequate to Yerify jet pump OPERABILITY anid consistent with the FrFequency for recirculation loop OPERA.BILIP(Yverification.
| |
| This SIR is modified by two Notes. Note 1 allows this Surveillance not to be performed until 4 hours after the associated recirculation loop is in operation, since these checks can only be performed during jet pump operation. The 4 hours is an acceptable time to establish conditions appropriate for data collection and evaluation.
| |
| Note 2 allows this SIR not to be performed until 24 hours after THERMAL POWER exceeds 23% RTP. During low flow conditions, jet pump noise approaches the threshold response of the associated flow instrumentation and precludes the collection of repeatable and meaningful data. The 24 hours is an acceptable time to establish conditions appropriate to perform this SR.
| |
| REFERENCES 1. USAR, Section 6.3.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| : 3. GE Service Information Letter No. 330 including Supplement 1, "Jet Pump Beam Cracks," June 9, 1980.
| |
| (continued)
| |
| NMP2 B 3.4.3-4 Revision 0, 37 (A140)
| |
| | |
| RCS Operational LEAKAGE B 3.4.5 BASES ACTIONS C.1 and C.2 (continued) based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE SR 3.4.5.1 REQUIREMENTS The RCS LEAKAGE is monitored by a variety of instruments designed to provide alarms when LEAKAGE is indicated and to quantify the various types of LEAKAGE. Leakage detection instrumentation is discussed in more detail in the Bases for LCO 3.4.7, "RCS Leakage Detection Instrumentation." Drain tank level and flow rate are typically monitored to determine actual LEAKAGE rates. However, any method may be used to quantify LEAKAGE within the guidelines of
| |
| {Insert 3 _. _ .,4..,,..,.o .,,.,,."
| |
| adm~ink~trativc controlz, a 12 hour FrcquoncY for thic Survillaco i appopRito for idcntifying GhangeG in LEA~KAGE anid for_trFacking rcquircd trcnds (Rcfr 0).
| |
| REFERENCES 1. 10 CER 50.2.
| |
| : 2. 10 CFR 50.55a(c).
| |
| : 3. 10 CER 50, Appendix A, GDC 55.
| |
| : 4. GEAP-5620, "Failure Behavior in ASTM A106B Pipes Containing Axial Through-Wall Flaws," April 1968.
| |
| : 5. NUREG-751067, "Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactors," October 1975.
| |
| : 6. USAR, Section 5.2.5.5.3.
| |
| : 7. 10 CFR 50.36(c)(2)(ii).
| |
| : 8. Regulatory Guide 1.45, May 1973.
| |
| 0- GnricP~~_ LottorF 88 01, Supplomont 4, eFcbruary 1992 NMP2 B 3.4.5-5 Revision 0
| |
| | |
| RCS Leakage Detection Instrumentation B 3.4.7 BASES ACTIONS E. 1 (continued)
| |
| With all required monitors inoperable, no required automatic means of monitoring LEAKAGE are available, and immediate plant shutdown in accordance with LCO 3.0.3 is required.
| |
| SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the other required instrumentation (either the drywell floor drain tank fill rate monitoring system or the drywell atmospheric monitoring channel, as applicable) is OPERABLE. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. The 6 hour testing allowance is acceptable since it does not significantly reduce the probability of properly monitoring RCS leakage.
| |
| SR 3.4.7.1 This SR requires the performance of a CHANNEL CHECK of the Ilns ]-----------
| |
| rt required drywell atmospheric s reasonable monitoring confidence that system.
| |
| the channel The check is operating
| |
| ~~prope* ThG FrqJ..yOf ! 2 hc'Jr6 is based G..!"cr'm,.
| |
| roliabillty and IS ro,-onblo fo'r *-ertctinR of norml SR 3.4.7.2 and SR 3.4.7.4 These SRs require the performance of a CHANNEL FUNCTIONAL TEST of the required RCS leakage detection instrumentation.
| |
| The test ensures that the monitors can perform their Insert 3 function in the desired manner. The test also verifies the rm setpoint and relative accuracy of the instrument string. Thc Frcqucncy of 31 days and 184 days, as applicablo, conidorS instFruot roliability, and o poating oxoioc has shoenF i t propor for dotocsting dogradation.
| |
| In.addition, the 184 day frcgucncvy of SR 3.4.7.14i i II A
| |
| * J B *
| |
| * i f suppertcp by the 31 day GOU~rcc cnock rociuicment or R-P 2-4.7.3-(continued)
| |
| NMP2 B 3.4.7-6 Revision 0, 36 (A139)
| |
| | |
| RCS Leakage Detection Instrumentation B 3.4.7 BASES SURVEILLANCE SR 3.4.7.3 REQUIREMENTS (continued) This SR requires performance of a source check of the required drywell atmospheric monitoring system. The source check is a qualitative assessment of the radiation detector response when the radiation detector is exposed to a source of increased radioactivity. This will ensure that the radiation detector can detect a significant increase in Insert 3
| |
| * radioactivity levels in the drywell atmosphere. The Frcequcncy of 31 days conzideorc dotoctor Fcliability, and oporating oxporionco hafc Ish(oOn it propcr for dotocting deffadateff- 7 SR 3.4.7.5 This SR requires the performance of a CHANNEL CALIBRATION of the required RCS leakage detection instrumentation channels.
| |
| The calibration verifies the accuracy of the instrument Ilnert3 -- ,strin.g, including the instruments located inside the cyc~le and GensideFc channel roliability. Oporatg expcricncc has pFEovcn this Frequnc icacptablc.
| |
| REFERENCES 1. 10 CFR 50, Appendix A, GDC 30.
| |
| : 2. Regulatory Guide 1.45, Revision 0, "Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
| |
| : 3. GEAP-5620, "Failure Behavior in ASTM A106B Pipes Containing Axial Through-Wall Flaws," April 1968.
| |
| : 4. NUREG-75/067, "Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactors," October 1975.
| |
| : 5. USAR, Section 5.2.5.5.3.
| |
| : 6. 10 CFR 50.36(c)(2)(ii).
| |
| : 7. USAR, Section 5.2.5.1.1.
| |
| NMP2 B 3.4.7-7 Revision 36 (A139)
| |
| | |
| RCS Specific Activity B 3.4.8 BASES (continued)
| |
| SURVEILLANCE SR 3.4.8.1 Insert 3 REQUIREMENTS This Surveillance is performed to e re iodine remains within limit during normal operation. "-dey Fr-y s*adequate to t*rnd chango* in thed ci level.
| |
| This SR is modified by a Note that requires this Surveillance to be performed only in MODE 1 because the level of fission products generated in other MODES is much less.
| |
| REFERENCES 1. 10 CFR 50.67, "Accident Source Term."
| |
| : 2. USAR, Section 15.6.4.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.4.8-4 Revision 0, 26 (A125)
| |
| | |
| RHR Shutdown Cooling System - Hot Shutdown B 3.4.9 BASES (continued)
| |
| SURVEILLANCE SR 3.4.9.1 REQUIREMENTS This Surveillance verifies that one RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay hea remova cap ThcFrcqucncyofl2ho'rcic ufIcinti view of othor Yicual andAuboidcaon available to the epcrater for m9onitorig tho RHR GuboYctom1 i n the control room.
| |
| This Surveillance is modified by a Note allowing sufficient time to align the RHR System for shutdown cooling operation after clearing the pressure interlock that isolates the system, or for placing a recirculation pump in operation.
| |
| The Note takes exception to the requirements of the Surveillance being met (i.e., forced coolant circulation is not required for this initial 2 hour period), which also allows entry into the Applicability of this Specification in accordance with SR 3.0.4 since the Surveillance will not be "not met" at the time of entry into the Applicability.
| |
| REFERENCES 1. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.4.9-5 Revision 0
| |
| | |
| RHR Shutdown Cooling System - Cold Shutdown B 3.4.10 BASES ACTIONS B.1 and B.2 (continued)
| |
| With no RHR shutdown cooling subsystem and no recirculation pump in operation, except as is permitted by LCO Note 1, and until RHR or recirculation pump operation is re-established, an alternate method of reactor coolant circulation must be placed into service. This will provide the necessary circulation for monitoring coolant temperature. The 1 hour Completion Time is based on the coolant circulation function and is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation. Furthermore, verification of the functioning of the alternate method must be reconfirmed every 12 hours thereafter. This will provide assurance of continued temperature monitoring capability.
| |
| During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling system or recirculation pump),
| |
| the reactor coolant temperature and pressure must be periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate.
| |
| SURVEILLANCE SR 3.4.10.1 REQUIREMENTS This Surveillance verifies that one RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The F,,,quo' ;.of 12 hoUr, i.
| |
| Gufficicnt in Yicw of othor visual and audible indicationG availablo to tho oporator for mon48itrig the RHR Gubcy~teom in the eontrol roeo.
| |
| REFERENCES 1. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.4.10-4 Revision 0
| |
| | |
| RCS P/T Limits B 3.4.11 BASES ACTIONS B.1 and B.2 (continued)
| |
| Pressure and temperature are reduced by bringing the plant to at least MODE 3 within 12 hours and to MODE 4 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.
| |
| C.1 and C.2 Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored.
| |
| Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 2000 F. Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components.
| |
| ASME Section Xl, Appendix E (Ref. 6), may be used to support the evaluation; however, its use is restricted to evaluation of the beltline.
| |
| Condition C is modified by a Note requiring Required Action C.2 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 RCPB integrity.
| |
| SURVEILLANCE SR 3.4.11.1 REQUIREMENTS Verification that operation is within limits is required vc,"; 30 ,m,,-tc when RCS pressure and temp= re conditions are undergoing planned changes. is-F.eqen-Gy r, concidcred rcasonablc in view of the control room indication available to mon~itor ROS status. Alo, Gince temperature rate of c-hange limits arc specified in hourlY (continued)
| |
| NMP2 B 3.4.11-6 Revision 0
| |
| | |
| RCS P/T Limits B 3.4.11 BASES SURVEILLANCE SR 3.4.11.1 (continued)
| |
| REQUIREMENTS increments, 30 minutes permits assessment and correction ot minor deviations. The limits in the PTLR are met when operation is to the right of the applicable limit curve.
| |
| Surveillance for heatup, cooldown, or system leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.
| |
| This SR has been modified by a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and system leakage and hydrostatic testing.
| |
| SR 3.4.11.2 A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical. The limits in the PTLR are met when operation is to the right of the applicable limit curve.
| |
| Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.
| |
| SR 3.4.11.3 and SR 3.4.11.4 Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of (continued)
| |
| NMP2 B 3.4.11-7 Revision 0, 40 (A145)
| |
| | |
| RCS P/T Limits B 3.4.11 BASES SURVEILLANCE SR 3.4.11.5 and SR 3.4.11.6 (continued)
| |
| REQUIREMENTS rate > 50% of rated jet pump loop flow. Therefore, SR 3.4.11.5 and SR 3.4.11.6 have been modified by a Note that requires the Surveillance to be met only under these conditions. The Note for SR 3.4.11.6 further limits the requirement for this Surveillance to exclude comparison of the idle loop temperature if the idle loop is isolated from the RPV since the water in the loop cannot be introduced into the remainder of the Reactor Coolant System.
| |
| SR 3.4.11.7. SR 3.4.11.8. and SR 3.4.11.9 Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO as specified in the Surveillance limits.
| |
| Frequency Control Program The ange temperatures must be verified to be above the limits within 30 minutcs bcforc and e.e.; 30 mFnm to thcerefte while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. When in MODE 4 with RCS temperature < 801F, 30 m..'-te checks of the flange temperatures are required because of the reduced margin to the limits. When in MODE 4 with RCS temperature < 90 0 F, monitoring of the flange temperature is required evrey 12 he-,i to ensure the temperatures are within the specified limits.
| |
| IiIne3h 30 m..inut F=r.quoncY rofle.t. tho.u.g... o m ,.aintai*ning the tcnpcFratu--rcz wit-hin lim~itc, and alco; imi~ltheth time that the temporatur *imi*t. could b" o.c.ode.. Th. 12 hour rFrcqucnc';GiF rasnable based on the Fate of tcmpcraturo change p..;ibl, at these tomporaturo.
| |
| SR 3.4.11.7 is modified by a Note that requires the Surveillance to be performed only when tensioning the reactor vessel head bolting studs. SR 3.4.11.8 is modified by a Note that requires the Surveillance to be initiated 30 minutes after RCS temperature < 80°F in MODE 4. SR 3.4.11.9 is modified by a Note that requires the Surveillance to be initiated 12 hours after RCS temperature
| |
| < 90°F in MODE 4.
| |
| (continued)
| |
| NMP2 B 3.4.11-9 Revision 0
| |
| | |
| RCS P/T Limits B 3.4.11 BASES SURVEILLANCE SR 3.4.11.7, SR 3.4.11.8, and SR 3.4.11.9 (continued)
| |
| REQUIREMENTS The Notes contained in these SRs are necessary to specify when the reactor vessel flange and head flange temperatures are required to be verified to be within the specified limits.
| |
| REFERENCES 1. 10 CFR 50, Appendix G.
| |
| : 2. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.
| |
| : 3. (Deleted)
| |
| : 4. 10 CFR 50, Appendix H.
| |
| : 5. Regulatory Guide 1.99, Revision 2, May 1988.
| |
| : 6. ASME, Boiler and Pressure Vessel Code, Section Xl, Appendix E.
| |
| : 7. GEH Report, NEDC-33178P-A, Revision 1, "General Electric Methodology for Development of Reactor Pressure Vessel Pressure- Temperature Curves," July 2009.
| |
| : 8. GEH Report, NEDC-33414P, Revision 1, "Pressure-Temperature Curves for Constellation Generation Group Nine Mile Point Nuclear Station Unit 2," October 2012.
| |
| : 9. 10 CFR 50.36(c)(2)(ii).
| |
| : 10. USAR, Section 15.4.4.
| |
| : 11. BWRVIP-86, Rev 1, "BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan," September 2008.
| |
| NMP2 B 3.4.11 -10 Revision 0n,-1, 1-4, 40 (A145)
| |
| | |
| Reactor Steam Dome Pressure B 3.4.12 BASES (continued)
| |
| APPLICABILITY In MODES 1 and 2, the reactor steam dome pressure is required to be less than or equal to the limit. In these MODES, the reactor may be generating significant steam, and events that may challenge the overpressure limits are possible.
| |
| In MODES 3, 4, and 5, the limit is not applicable because the reactor is shut down. In these MODES, the reactor pressure is well below the required limit, and no anticipated events will challenge the overpressure limits.
| |
| ACTIONS A.1 With the reactor steam dome pressure greater than the limit, prompt action should be taken to reduce pressure to below the limit and return the reactor to operation within the bounds of the analyses. The 15 minute Completion Time is reasonable considering the importance of maintaining the pressure within limits. This Completion Time also ensures that the probability of an accident while pressure is greater than the limit is minimal.
| |
| B._1 If the reactor steam dome pressure cannot be restored to within the limit 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 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE SR 3.4.12.1 REQUIREMENTS Verification that reactor steam dome pressure is < 1035 psig ensures that the initial conditions of the vessel lnsert 3 overpressure protection analysis is met.
| |
| cxpecrincc has shwn the 12 hou.r Frcqu, ,yto be ,uffie*;fint for idnti*yAg t,,ndG and vcif,'i,, oporatin within safoty analyses ascumptintnn.
| |
| (continued)
| |
| NMP2 B 3.4.12-2 Revision 0
| |
| | |
| ECCS - Operating B 3.5.1 BASES (continued)
| |
| SURVEILLANCE SR 3.5.1.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the HPCS System, LPCS System, and LPCI subsystems full of water ensures that the systems will perform properly, injecting their full capacity into the RCS upon demand. This will also prevent a water hammer following an ECCS initiation signal. One acceptable method of ensuring Insert 3 the lines are full is to vent at the high points. iM 31 day Frogucncy is based On operating experience, on the pr*Bd.*.al GontrolS goevcrning system opcFation, and en the gradual nature o-f,o-id, buildupin the EGGS pimp*in.
| |
| SR 3.5.1.2 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 initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves potentially capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
| |
| [Insert 3 \
| |
| vj nt thr HF-? w~v' dr~ri'~'r~d trnm thr~
| |
| Thc 31. -da. F.rc.q*ucncyl,*..
| |
| ,'.nsvic Terting Progr a. rc ,urmcntc for pelo*rming Yalve testing at learst oncc- ever; 02 days. The Frequency oe 31 days is furthcr justified bccausc the valvcS arc opeffatcd under procedural control and bccsause imRproper valve alignment would only affcct a single subsystem. This FrFequency has been shWown to be acceptable thrOUgh operating (continued)
| |
| NMP2 B 3.5.1-9 Revision 0
| |
| | |
| ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.2 (continued)
| |
| REQUIREMENTS In MODE 3 with reactor steam dome pressure less than the actual RHR cut-in permissive pressure, the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Therefore, this SR is modified by a Note that allows LPCI subsystems to be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. Alignment and operation for decay heat removal includes when the required RHR pump is not operating or when the system is realigned from or to the RHR shutdown cooling mode. At the low pressures and decay heat loads associated with operation in MODE 3 with reactor steam dome pressure less than the RHR cut-in permissive pressure, a reduced complement of low pressure ECCS subsystems should provide the required core cooling, thereby allowing operation of RHR shutdown cooling, when necessary.
| |
| SR 3.5.1.3 Verification every ,3..dys that ADS nitrogen receiver discharge header pressure is _>160 psig and ADS nitrogen receiver tank pressure is >_334 psig assures adequate nitrogen pressure for reliable ADS operation. The accumulator on each ADS valve provides nitrogen pressure for valve actuation. The designed nitrogen supply pressure requirements for the accumulator are such that, following a failure of the nitrogen supply to the accumulator, at least one valve actuation can occur with the drywell at 100% of design pressure (Ref. 15). The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required nitrogen receiver discharge header pressure of 160 psig is provided by two ADS nitrogen receiver tanks.
| |
| The minimum ADS nitrogen receiver tank pressure of 334 psig nsertnsures a 5 day supply of nitrogen is available to recharge
| |
| ~~~the ADS
| |
| * To3 ~ rqcc .zkc!t considorationR dntAtocontrol or epcatn of tho nitrogcn rcccivcr tanks and alaFRms for lWA nitrogcn PFeGGU~e.
| |
| (continued)
| |
| NMP2 B 3.5.1-10 Revision 0
| |
| | |
| ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.4 REQUIREMENTS (continued) The performance requirements of the ECCS pumps are determined through application of the 10 CFR 50, Appendix K, criteria (Ref. 8). This periodic Surveillance is performed (in accordance with the ASME OM Code requirements for the ECCS pumps) to verify that the ECCS pumps will develop the flow rates required by the respective analyses.
| |
| The ECCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of 10 CFR 50.46 (Ref. 10).
| |
| The pump flow rates are verified against a system head that is equivalent to the RPV pressure expected during a LOCA.
| |
| The total developed head is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure present during LOCAs. These values may be established during pre-operational testing. A 92 day Frequency for this Surveillance is in accordance with the Inservice Testing Program requirements.
| |
| SR 3.5.1.5 The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCS, LPCS, and LPCI will cause the systems or subsystems to operate as designed, i.e., actuation of the system throughout its emergency operating sequence, which includes automatic pump startup and actuation of all automatic valves (including the LPCI flow diversion valves closed on a Reactor Vessel Water Level - Low, Level 3 or a Drywell Pressure - High (Boundary Isolation) signal) to their required positions. This Surveillance also ensures that the HPCS System will automatically restart (i.e., injection valve re-open) on an RPV low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) signal and that the suction is automatically transferred from the CST to the suppression pool. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlaps this Surveillance to provide complete testing of the assumed safety function.
| |
| (continued)
| |
| NMP2 B 3.5.1 -11 Revision 0, 28 (A129)
| |
| | |
| ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.5 (continued)
| |
| REQUIREMENTS
| |
| - Th o 24FA nth F...quonc.; scb3oed on ttio n..d to p*e"Rorm t+
| |
| iuoiiveIancc undca the conditions that apply during a plat*
| |
| outage (cXccpt for the Division 3 diesel genReator, which canb tested in any oporational condition) and the potcntial for an unplanncd transicnt if the Survcillancc wcrc perfofrmed with h rcactVI at power. Oporating oxpooIncoV has shown that these compoenets usually pass the SR when porfo.moPd- at the 214 month Fr.equency, which i. baeP-d on the rFu' ...c In.g . Therefore, the F-o"u o'n'c ,wAch,-oncl-udod to bo ;Ao,.ntabl.
| |
| -r- ............
| |
| from.. a li -1
| |
| .r.l*2.b stamJpein-This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
| |
| SR 3.5.1.6 The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals.
| |
| A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e.,
| |
| solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
| |
| >The 21 month Frequency is based en the need to perfoFrm thi outage and the potential for an unplanned transient if the SUF~eillanco were perfoFrmed with the reactor at power.
| |
| Operating experience has shown that theSe components Wusualy pass the SR when perfoFrmed at the 24 moneth Frequency, which
| |
| - ^mfnnr;^ "^1 ý lnA^
| |
| y
| |
| ", ,- ~- fI-,.---.-
| |
| k . . nl-f ,J l
| |
| , ..11 v Jv
| |
| + I EvvVIG larit;~
| |
| av -to vuUl Ex olumarIanrll ilGll standpeint This SR is modified by a Note that excludes valve actuation.
| |
| This prevents an RPV pressure blowdown.
| |
| (continued)
| |
| NMIP2 B 3.5.1-12 Revision 0, 32 (A133)
| |
| | |
| ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.7 REQUIREMENTS (continued) A manual actuation of each ADS actuator is performed to verify that the valve and solenoids are functioning properly. This can be demonstrated by one of two methods.
| |
| Each ADS actuator can be tested using either method. The first method is a manual actuation of the ADS valve with verification of the response of the turbine control or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow (e.g.,
| |
| tailpipe temperature or acoustic monitor). Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the ADS valves divert steam flow upon opening. Sufficient time is therefore allowed, after the required pressure and flow are achieved, to perform this test. Adequate pressure at which this test is to be performed is 950 psig (the pressure recommended by the valve manufacturer). Adequate steam flow is represented by at least 2 turbine bypass valves open.
| |
| Reactor startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test. The 12 hours allowed for manual actuation after the required pressure and flow are reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR.
| |
| The second method involves cycling the actuator of the ADS valves prior to startup of the unit. A sample population of S/RVs are normally removed and bench tested as required by SR 3.4.4.1. Since the ADS valves are also S/RVs, bench testing the S/RVs also cycles the ADS valves. The successful performance of the bench test for this sample provides a reasonable assurance that the remaining ADS valves that were not bench tested in this sample will perform (i.e., open when required) in a similar fashion.
| |
| After each ADS valve is reinstalled following a bench test, the actuator of the ADS valve shall be uncoupled from the ADS valve stem and cycled using one solenoid to ensure that no damage has occurred to the ADS valve actuator during transportation and installation and to ensure one solenoid is functioning properly. In addition, the ADS valves not (continued)
| |
| NMP2 B 3.5.1-13 Revision 0
| |
| | |
| ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.7 (continued)
| |
| REQUIREMENTS removed and bench tested shall be tested in a similar manner to verify one solenoid and the actuator. Uncoupling from the ADS valve stem allows operation of the actuator without disturbing the valve disc seat interface and thereby minimizes the likelihood of steam leaking past the ADS valve seat when it is closed. If performed using this method, ADS valve OPERABILITY has been demonstrated while shutdown for each ADS valve based upon the successful bench testing of a sample population of S/RVs and the successful operation of the actuator and a solenoid for the remaining ADS valves.
| |
| Thus the Note that modifies this SR is not needed since reactor steam pressure is not used to lift the valve.
| |
| SR 3.5.1.6 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
| |
| lInsert 3 -. .
| |
| The PFe(3 ucncy of 21 menths on a STAGGFEREID TEST BASIS enzurcs fk-f k-fk 8,4ý f-, -k A nQ -
| |
| alye arc altcrnatcly tcstcd. I hc I-rcgucncy is 13ascl on the need to po~foRm tn~
| |
| S.,lurvllanc und.. tho conditionc that apply just prior to orduring a Startup fromA a plant outagc. Opcrating cxpcriencc has shown that these compencnts usually pass the SR whcni pcrformced at thc- 21 month Frcequoncy, which; is bao on the rcfucling cyclc. Thercforc, tho- Frq c was concl-udcd to hp acccotabloP from aroliabilitv CtARNdROit.
| |
| SR 3.5.1.8 This SR ensures that the ECCS RESPONSE TIME for each ECCS injection/spray subsystem is less than or equal to the maximum value assumed in the accident analysis. Response time testing acceptance criteria are included in Reference 16. This SR is modified by a Note that allows the instrumentation portion of the response time to be assumed to be the design instrumentation response time and therefore, is excluded from the ECCS RESPONSE TIME testing.
| |
| This is allowed since the instrumentation response time is a small part of the ECCS RESPONSE TIME (e.g., sufficient margin exists in the diesel generator start time when compared to the instrumentation response time) (Ref. 17).
| |
| (continued)
| |
| NMP2 B 3.5.1-14 Revision 0
| |
| | |
| ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.8 (continued)
| |
| REQUIREMENTS ECCS RESPONSE TIME tests arc coductod ever' 24 months. Th-24 month Froqcc icnstntWith the typical in~dustr-y
| |
| [Insert 3 -- rofuel!ng cyclc and is bascd upon plant oporat',.
| |
| e)v reF* R G8- ' .l*
| |
| REFERENCES 1. USAR, Section 6.3.2.2.3.
| |
| : 2. USAR, Section 6.3.2.2.4.
| |
| : 3. USAR, Section 6.3.2.2.1.
| |
| : 4. USAR, Section 6.3.2.2.2.
| |
| : 5. USAR, Section 15.6.6.
| |
| : 6. USAR, Section 15.6.4.
| |
| : 7. USAR, Section 15.6.5.
| |
| : 8. 10 CFR 50, Appendix K.
| |
| : 9. USAR, Section 6.3.3.
| |
| : 10. 10 CFR 50.46.
| |
| : 11. USAR, Section 6.3.3.3.
| |
| : 12. 10 CFR 50.36(c)(2)(ii).
| |
| : 13. Memorandum from R.L. Baer (NRC) to V. Stello, Jr.
| |
| (NRC), "Recommended Interim Revisions to LCO's for ECCS Components," December 1, 1975.
| |
| : 14. USAR, Chapter 15C.
| |
| : 15. USAR, Section 5.2.2.4.1.
| |
| : 16. Technical Requirements Manual.
| |
| : 17. NEDO-32291-A, "System Analyses for the Elimination of Selected Response Time Testing Requirements,"
| |
| October 1995.
| |
| NMP2 B 3.5.1-15 Revision 0
| |
| | |
| ECCS - Shutdown B 3.5.2 BASES ACTIONS C.1, C.2, D.1, D.2, and D.3 (continued) information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, the Surveillances may need to be performed to restore the component to OPERABLE status.
| |
| Actions must continue until all required components are OPERABLE.
| |
| SURVEILLANCE SR 3.5.2.1 and SR 3.5.2.2 REQUIREMENTS The minimum water level of 195 ft required for the suppression pool is periodically verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for the ECCS pumps, recirculation volume, and vortex prevention. With the suppression pool water level less than the required limit, all ECCS injection/spray subsystems are inoperable unless they are aligned to an OPERABLE CST.
| |
| When the suppression pool level is < 195 ft, the HPCS System is considered OPERABLE only if it can take suction from CST B and CST B water level is sufficient to provide the required NPSH for the HPCS pump. Therefore, a verification that either the suppression pool water level is >_195 ft or the HPCS System is aligned to take suction from the CST and the CST contains > 253,000 gallons of water, equivalent to 26.9 ft, ensures that the HPCS System can supply 135,000 gallons of makeup water to the RPV. In addition, to ensure the 135,000 gallons of makeup water is available, the HPCS suction source auto-swap from the CST to the suppression pool must be disabled (e.g., by closing the suppression pool suction valve and deenergizing the breaker for the valve motor operator). This is necessary since the actual trip setpoint of the HPCS Pump Suction Pressure - Low Function is at a pressure sufficiently high such that 135,000 gallons would not be available before the auto-swap occurred.
| |
| r nf thc"r~ ~P'~ wi" dc"cInnr~rI r~nn"idr~rinri c*-rating cepartcecc ratcdtp*-rs.ST c pool and waterF cticI yariati Ans; andd ionstru .mi."ent drift-du-rinRg th (continued)
| |
| NMP2 B 3.5.2-4 Revision 0
| |
| | |
| ECCS - Shutdown B 3.5.2 BASES SURVEILLANCE SR 3.5.2.1 and SR 3.5.2.2 (continued)
| |
| REQUIREMENTS applicablo MODS.Futhormoro, tho* 1:2 hour Frequency iS con...idored adequate in vew foto indicatioeR in the The Bases provided for SR 3.5.1.1, SR 3.5.1.4, SR 3.5.1.5, and SR 3.5.1.8 are applicable to SR 3.5.2.3, SR 3.5.2.5, SR 3.5.2.6, and SR 3.5.2.7, respectively.
| |
| SR 3.5.2.4 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 initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The 31 day Fr.equency is appropriate becau. e the valves are oporatod unider procedural cenRe!1 and the probability of their being mnispesitiened duFrig this time pe~ed is !ww In MODES 4 and 5, the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Therefore, this SR is modified by a Note that allows one LPCI subsystem to be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable.
| |
| Alignment and operation for decay heat removal includes when the required RHR pump is not operating or when the system is being realigned from or to the RHR shutdown cooling mode.
| |
| (continued)
| |
| NMP2 B 3.5.2-5 Revision 0
| |
| | |
| ECCS - Shutdown B 3.5.2 BASES SURVEILLANCE SR 3.5.2.4 (continued)
| |
| REQUIREMENTS Because of the low pressure and low temperature conditions in MODES 4 and 5, sufficient time will be available to manually align and initiate LPCI subsystem operation to provide core cooling prior to postulated fuel uncovery.
| |
| lnsert 3 _ This will ensure adequate core cooling if an inadvertent vessel draindown should occur.
| |
| REFERENCES 1. USAR, Section 6.3.3.3.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.5.2-6 Revision 0
| |
| | |
| RCIC System B 3.5.3 BASES (continued)
| |
| SURVEILLANCE SR 3.5.3.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge line of the RCIC System full of water ensures that the system will perform properly, injecting its full capacity into the Reactor Coolant System upon demand. This will also prevent a water hammer following an initiation signal. One acceptable method of ensuring the line is full is to vent at lInsert 3 ghd
| |
| . .,,, hinu, , ,... of ,,., ,,, ,, ;., . CiC oi3d4 bida4upntyh, ,, , piping,,
| |
| .,-;.. ,the procd-ral coAnt8rl governing ... tem . p..atien, and operating experience.
| |
| SR 3.5.3.2 Verifying the correct alignment for manual, power operated, and automatic valves (including the RCIC pump flow controller) in the RCIC flow path provides assurance that the proper flow path will exist for RCIC operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time.
| |
| This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. For the RCIC System, this SR also includes the steam flow path for the turbine and the flow controller position.
| |
| . ~~- ........
| |
| h 31 da 'R-R ..... ......thi e .}f'd ef .... ... .......
| |
| IInse rt 3 'RnPPAric Tesin Prga eurmnts for per-forming valve totig t ist ovor,' 02 daseh rqec f 31 days is further justified because tevlearopaedunder procedural control and bocause imrpe alve position would affert only the R-IC System. This Fr-equency has been shown tu be ueue uble throu, uu h e,ltn
| |
| ,,i crii, u.
| |
| (continijed'~
| |
| (continued)
| |
| NMVP2 B 3.5.3-4 Revision 0
| |
| | |
| RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.3 and SR 3.5.3.4 REQUIREMENTS (continued) The RCIC pump flow rates ensure that the system can maintain reactor coolant inventory during pressurized conditions with the RPV isolated. The flow tests for the RCIC System are performed at two different pressure ranges such that system capability to provide rated flow against a system head corresponding to reactor pressure is tested both at the higher and lower operating ranges of the system. The required system head should overcome the RPV pressure and associated discharge line losses. Adequate reactor steam pressure must be available to perform these tests.
| |
| Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the RCIC System diverts steam flow. Therefore, sufficient time is allowed after adequate pressure and flow are achieved to perform these SRs.
| |
| Adequate reactor steam pressure to perform SR 3.5.3.3 is 935 psig and to perform SR 3.5.3.4 is 150 psig. Adequate steam flow is represented by at least one turbine bypass valve open. Reactor startup is allowed prior to performing the low pressure Surveillance because the reactor pressure is low and the time to satisfactorily perform the Surveillance is short. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure test has been satisfactorily completed and there is no indication or reason to believe that RCIC is inoperable. Therefore, these SRs are modified by Notes that state the Surveillances are not required to be performed until 12 hours after the reactor steam pressure and flow are adequate to perform the test. The 12 hours allowed for the flow tests after the required pressure and flow are reached are sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SRs.
| |
| linsert A 92 day FrcequcncY for SR 3.6.3.3 is -- nsis-tnt with the IncorVico Tocting Progra roqu Rmntq. The 24 month Frcequcncy for SR 3.6.3.4 sbsdo the need to pcrfoFRm this Survcillancc undcr thcconditions that apply during startup from a plant eutagc. Opcrating cxpcricncc has shoWn that these compencnts usually pass the SR when pcrfermdat the 21 month Froequoncy, Wh ih is basod on tho rofuoling cyclo. T-hcrcforc, the Frcequcncy was concludcd to be acccptablc from a rci~abii~ty ~staROee+Rt (continued)
| |
| NMP2 B 3.5.3-5 Revision 0
| |
| | |
| RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.5 REQUIREMENTS (continued) The RCIC System is required to actuate automatically to perform its design function. This Surveillance verifies that with a required system initiation signal (actual or simulated) the automatic initiation logic of RCIC will cause the system to operate as designed, i.e., actuation of the system throughout its emergency operating sequence, which includes automatic pump startup and actuation of all automatic valves to their required positions. This Surveillance also ensures that the RCIC System will automatically restart on an RPV low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool on a CST low water level signal. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.2 overlaps this Surveillance to
| |
| [Insert 3 provide complete testing of the assumed design function.
| |
| u.,ally PaRl; thI S;R Whn pc-formod at the 24 moenth PFrqucncY, Which is baccd on the rofuoling cYclc.
| |
| Thcroforc, the Frcqucncy was concluidcd to be acccptablc ro a rcliability standpoint.
| |
| This SR is modified by a Note that excludes vessel injection during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
| |
| REFERENCES 1. 10 CFR 50, Appendix A, GDC 33.
| |
| : 2. USAR, Section 5.4.6.1.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| : 4. Memorandum from R.L. Baer (NRC) to V. Stello, Jr.
| |
| (NRC), "Recommended Interim Revisions to LCO's for ECCS Components," December 1, 1975.
| |
| NMP2 B 3.5.3-6 Revision 0
| |
| | |
| Primary Containment B 3.6.1.1 BASES SURVEILLANCE SR 3.6.1.1.1 (continued)
| |
| REQUIREMENTS (SR 3.6.1.3.11), resilient seal primary containment purge valve leakage limit (SR 3.6.1.3.6), or main steam isolation valve leakage limit (SR 3.6.1.3.12) does not necessarily result in a failure of this SR. The impact of the failure to meet these SRs must be evaluated against the Type A, B, and C acceptance criteria of 10 CFR 50 Appendix J Testing Program Plan.
| |
| As left leakage prior to the first startup after performing a required 10 CFR 50 Appendix J Testing Program Plan leakage test is required to be < 0.6 La for combined Type B and C leakage, and < 0.75 La for overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of < 1.0 La. At _ 1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis. The Frequency is required by the 10 CFR 50 Appendix J Testing Program Plan.
| |
| SR 3.6.1.1.2 and SR 3.6.1.1.3 Maintaining the pressure suppression function of the primary containment requires limiting the leakage from the drywell to the suppression chamber. Thus, if an event were to occur that pressurized the drywell, the steam would be directed through the downcomers into the suppression pool.
| |
| SR 3.6.1.1.2 measures drywell-to-suppression chamber differential pressure to ensure that the leakage paths that would bypass the suppression pool are within allowable limits. The suppression chamber-to-drywell vacuum breakers are the most likely source of potential bypass leakage, therefore, these valves are normally tested on a more frequent basis.
| |
| Satisfactory performance of SR 3.6.1.1.2 can be achieved by establishing a known differential pressure (_>3.0 psid) between the drywell and the suppression chamber and verifying that the A/v'K calculated from the measured bypass leakage is equivalent to that through an area < 0.0054 ft2 .
| |
| The leakage test is performed at the same Frequency as the Type A testing requirements of the 10 CFR 50 Appendix J Testing Program Plan. This Frequency was developed since historically the leakage is much less than the design value and that the most credible source of potential bypass (continued)
| |
| NMP2 B 3.6.1.1-4 Revision 0
| |
| | |
| Primary Containment B 3.6.1.1 BASES SURVEILLANCE SR 3.6.1.1.2 and SR 3.6.1.1.3 (continued)
| |
| REQUIREMENTS leakage, the suppression chamber-to-drywell vacuum breakers will normally be tested more frequently in accordance with SR 3.6.1.1.3. Two consecutive as-found test failures of SR 3.6.1.1.2, however, would indicate unexpected primary containment degradation; in this event, as the Note indicates, increasing the Frequency to nco, ovoy 24 mo,,thg
| |
| .. .e.4iFed until the situation is remediated as evidenced by passing two consecutive tests.
| |
| Conservative test criteria was chosen for SR 3.6.1.1.3 based Insert 3 on the assumed bypass leakage in the LOCA analysis.
| |
| :24 month Froequoncy spocifiod for SR 3.6.1.1.3 was dovoloped considcring iisprudent that this Survcillancc be porf*o..m.d during a unit "utag.* A Note has been added to SR 3.6.1.1.3 which provides an allowance not to perform SR 3.6.1.1.3 when SR 3.6.1.1.2 is required to be performed since SR 3.6.1.1.2 will provide adequate information on the capacity of the pressure suppression function of the primary containment.
| |
| REFERENCES 1. USAR, Section 6.2.
| |
| : 2. USAR, Section 15.6.5.
| |
| : 3. 10 CFR 50, Appendix J, Option B.
| |
| : 4. USAR, Section 6.2.6.1.
| |
| : 5. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.6.1.1-5 Revision 0
| |
| | |
| Primary Containment Air Locks B 3.6.1.2 BASES ACTIONS D.1 and D.2 (continued)
| |
| If the inoperable primary containment air lock 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 12 hours and to MODE 4 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.
| |
| SURVEILLANCE SR 3.6.1.2.1 REQUIREMENTS Maintaining the primary containment air locks OPERABLE requires compliance with the leakage rate test requirements of the 10 CFR 50 Appendix J Testing Program Plan. This SR reflects the leakage rate testing requirements with regard to air lock leakage (Type B leakage tests). The acceptance criteria were established as a small fraction of the total allowable primary containment leakage. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall primary containment leakage rate. The Frequency is required by the 10 CFR 50 Appendix J Testing Program Plan.
| |
| The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.
| |
| This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR, requiring the results to be evaluated against the acceptance criteria which is applicable to SR 3.6.1.1.1. This ensures that air lock leakage is properly accounted for in determining the combined Types B and C primary containment leakage rate.
| |
| SR 3.6.1.2.2 The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident primary (continued)
| |
| NMP2 B 3.6.1.2-7 Revision 0
| |
| | |
| Primary Containment Air Locks B 3.6.1.2 BASES SURVEILLANCE SR 3.6.1.2.2 (continued)
| |
| REQUIREMENTS containment pressure (Ref. 2), closure of either door will support primary containment OPERABILITY. Thus, the interlock feature supports primary containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not inadvertently occur. te1 8 3
| |
| .. othi'
| |
| .Ature , interlock, aRd -oventhat1hoint rl k Insert 3 SmochaniGsl
| |
| * i not normnally Ihallc I hc II thcimP111a1 con-tafinmcntR ;air locr-k door is used- for entry and exist (procedures Foquire GBFict adhoFronc to single door opening), this test Is onlY required to be pe~rmcff~d cver;
| |
| : 24. mon-t-hs. Tho24. Imonth FrFequency is based on the nood to pecf9Fm this Surveillancc uinder the conditions that apply during a plant outag, , and the potonti-al for' o66; Of p..m*a;....
| |
| containment URfL-AW41b Ilif~I the burweiiianc were penormo with the reaet*o at power. Operatin e has shown these compents uIually pa" ,the ..rw,,lla,-, whon peofermed at tho 24 mRonth Frequency. The FrFequenc~y isbased on ooinoRAn judgment and 06 Geoiderod adecquate gfiyo V
| |
| * V
| |
| ! l I * [ l ll is not mlmmmii ged during the use of the aiF i96k ainmcmmli REFERENCES 1. USAR, Section 3.8.1.1.2.
| |
| : 2. USAR, Section 6.2.6.1.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.6.1.2-8 Revision 0
| |
| | |
| PCIVs B 3.6.1.3 BASES ACTIONS E.1, E.2. and E.3 (continued) containment purge exhaust valve does not increase during the time the penetration is isolated. The normal Frequency for SR 3.6.1.3.6 is 184 days. Since more reliance is placed on a single valve while in this Condition, it is prudent to perform the SR more often. Therefore, a Frequency of once per 92 days was chosen and has been shown acceptable based on operating experience.
| |
| F. 1 and F.2 If any Required Action and associated Completion Time cannot be met in MODE 1, 2, or 3, 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 12 hours and to MODE 4 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.
| |
| G.1 and G.2 If any Required Action and associated Completion Time cannot be met for PCIV(s) required OPERABLE in MODE 4 or 5, the plant must be placed in a condition in which the LCO does not apply. Action must be immediately initiated to suspend operations with a potential for draining the reactor vessel (OPDRVs) to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. If suspending the OPDRVs would result in closing the residual heat removal (RHR) shutdown cooling isolation valves, an alternative Required Action is provided to immediately initiate action to restore the valves to OPERABLE status. This allows RHR shutdown cooling to remain in service while actions are being taken to restore the valve.
| |
| SURVEILLANCE SR 3.6.1.3.1 REQUIREMENTS This SR verifies that the 12 inch and 14 inch primary containment purge valves are closed as required or, if open, opened for an allowable reason.
| |
| (continued)
| |
| NMP2 B 3.6.1.3-12 Revision 0;-38, 39
| |
| | |
| PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.1 (continued)
| |
| REQUIREMENTS The SR is modified by a Note stating that the SR is not required to be met when the purge valves are open for the stated reasons. The Note states that these valves may be opened for inerting, de-inerting, pressure control, ALARA, or air quality considerations for personnel entry, or for Surveillances that require the valves to be open, provided that either: a) the SGT System is OPERABLE (i.e., both subsystems); or b) the primary containment full flow line to the SGT System is isolated and one SGT subsystem is OPERABLE. These primary containment purge valves are capable of closing in the environment following a LOCA.
| |
| Therefore, these valves are allowed to be open for limited periods of time. The allowance is intended to balance the operational needs of the unit with the requirement to preclude a radiological release through the purge exhaust lines. With the primary containment atmosphere being exhausted through the containment full flow line to the SGT System, a pressure transient could damage the operating SGT subsystem. Thus both subsystems are required to be OPERABLE when the full flow line is in service. This ensures that, if an accident occurs that damages the operating SGT subsystem, the remaining SGT subsystem is still available to perform the intended SGT System safety function. When the full flow line is not in service (i.e., the two inch bypass valve is open), then only one SGT subsystem is required to Ilnsert 3 Ibe OPERABLE since a pressure transient cannot damage the operating . .Th 31 day
| |
| , Frqucny is c.n.i'tnt with othr prim~ary containmont isolation valvo roquiromentc di.us..d inR SR 3.6841.3.2.
| |
| SR 3.6.1.3.2 This SR verifies that each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions, is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the primary containment boundary is within design limits.
| |
| This SR does not require any testing or valve manipulation.
| |
| Rather, it involves verification that those PCIVs outside primary containment, and capable of being mispositioned, are in the correct position. Sin.. Y...ifi.ati.n of p..iti.n for PCIV* outsidc primary c ;'latively easy, thc ic
| |
| .ntainmcnt (continued)
| |
| NMP2 B 3.6.1.3-13 Revision 0
| |
| | |
| PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.2 (continued)
| |
| REQUIREMENTS 31 day FroguoncY war, chocon to provido addod accranco that tho PCr-,4,.. ro in tho"
| |
| ...... t po"itin*r". This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing. < lInsert 3 1 Two Notes are added to this SR. The first Note applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these PCIVs, once they have been verified to be in the proper position, is low. A second Note is included to clarify that PCIVs open under administrative controls are not required to meet the SR during the time the PCIVs are open. These controls consist of stationing a dedicated operator at the controls.of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated.
| |
| SR 3.6.1.3.3 This SR verifies that each primary containment manual isolation valve and blind flange located inside primary containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions, is closed.
| |
| The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits. For PCIVs inside primary containment, the Frequency of "prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days," is appropriate since these PCIVs are operated under administrative controls and the probability of their misalignment is low. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.
| |
| (continued)
| |
| NMP2 B 3.6.1.3-14 Revision 0
| |
| | |
| PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.3 (continued)
| |
| REQUIREMENTS Two Notes are added to this SR. The first Note allows valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable since the primary containment is inerted and access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA and personnel safety.
| |
| Therefore, the probability of misalignment of these PCIVs, once they have been verified to be in their proper position, is low. A second Note is included to clarify that PCIVs that are open under administrative controls are not required to meet the SR during the time that the PCIVs are open.
| |
| These controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated.
| |
| SR 3.6.1.3.4 The traversing incore probe (TIP) shear isolation valves are actuated by explosive charges. Surveillance of explosive charge continuity provides assurance that TIP valves will actuate when required. Other administrative controls, suchInsert 3 as those that limit the shelf life and operating life, as applicable, of the explosive charges, must be followed.
| |
| 31 day Froequoncey is based on operating e~perienco that ha dcmROnctratcd the rcliability of tho expleswye cahargc SR 3.6.1.3.5 Verifying the isolation time of each power operated, automatic PCIV is within limits is required to demonstrate OPERABILITY. MSIVs may be excluded from this SR since MSIV full closure isolation time is demonstrated by SR 3.6.1.3.7.
| |
| The isolation time test ensures that each valve will isolate in a time period less than or equal to that assumed in the safety analysis. The Frequency of this SR is in accordance with the Inservice Testing Program.
| |
| (continued)
| |
| NMP2 B 3.6.1.3-15 Revision 0
| |
| | |
| PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.6 REQUIREMENTS (cont inued) For primary containment purge valves with resilient seals, additional leakage rate testing beyond the test requirements of 10 CFR 50, Appendix J Option B (Ref. 7), is required to ensure OPERABILITY. The primary containment purge supply valves, which are secondary containment bypass leakage pathway valves, are tested at a pressure of 40.0 psig and the primary containment purge exhaust valves, which are not secondary containment bypass leakage pathway valves, are tested at Pa, 39.75 psig. The leakage limit for the 12 inch supply and exhaust valves are 3.75 scfh while the 14 inch supply and exhaust valve leakage limit is 4.38 scfh.
| |
| Operating experience has demonstrated that this type of seal Ilnsert 3 I has the potential to degrade in a shorter time period than do ot er sea . Bacod on thic obco."-rationA, and the importance of maintaining these pcnctratiens leak tight (due to the diroct path beten prima eR'cntainmont and the cnv.ronmAt in some cases), a Frequency of 184 days wa-s es*ablie*hd. Additionally, this SR must be performed within 92 days after opening the valve. The 92 day Frequency was chosen recognizing that cycling the valve could introduce additional seal degradation (beyond that which occurs to a valve that has not been opened). Thus, decreasing the interval (49fti-.184- dys) is a prudent measure after a valve has been opened.
| |
| SR 3.6.1.3.7 Verifying that the full closure isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY. The full closure isolation time test ensures that the MSIV will isolate in a time period that does not exceed the times assumed in the DBA and transient analyses.
| |
| The Frequency of this SR is in accordance with the Inservice Testing Program.
| |
| SR 3.6.1.3.8 Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a DBA. This SR ensures that each automatic PCIV will actuate to its isolation position on a primary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.1, "Primary Containment (continued)
| |
| NMP2 B 3.6.1.3-16 Revision 0
| |
| | |
| PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.8 (continued) Insert 3 REQUIREMENTS Isolation Instrumentation," overlaps thi R to provide complete testing of the safety function. 4he-A-me FrequcncY ic bazcd on the need to pcrfoFRm this Survcillanc undcr tho condi0tione- that apply during a plant outago and the potcntial for an unplannod trancionPt if thoP Surwillanco woro perfoFRmcd with the roactor at powor. Oporating oxoDcc h-As Sheown th-at thocoe comFponontG Wuually pacc this; Survcillanc when pcrformcnd at the 24 month PFrcqUecY. I I
| |
| * m I F
| |
| , n;teorc, the -F;eroUec WSa, concuaeoa to bo accop.tab, ,rofn
| |
| ---- I!--L!I!L -- { .. J .. ! I -- -- --I J !1! -- -- LL ) -- f*
| |
| aI rciiaoiirn:
| |
| II I
| |
| . j cianoDoint. in apomon. inic ~ur:ciiiancc
| |
| ... . i-- -- . ........ F .... .
| |
| cIa p Ffol in IIV1VDE I e*artormd oer .
| |
| SR 3.6.1.3.9 This SR requires a demonstration that a representative sample of reactor instrumentation line EFCVs is OPERABLE by verifying that the valves actuate to the isolation position on an actual or simulated instrument line break condition.
| |
| The representative sample consists of an approximately equal number of reactor instrumentation line EFCVs, such that each EFCV is tested at least once every 5 refueling cycles. In addition, the reactor instrumentation line EFCVs in the sample are representative of the various plant configurations, models, sizes, and operating environments.
| |
| This ensures that any potentially common problem with a specific type or application of reactor instrumentation line EFCV is detected at the earliest possible time. This SR provides assurance that the reactor instrumentation line EFCVs will perform as designed.
| |
| isod oll ntho Inod to porforlml th iI Smupvellcane unamr tih enditml*n tmat apply m uming a pmant outago and the potontial for an u mplannod ta*mi*ont if tho Sur~illncoWoro porformod- with tho rcactor at powor. Thoe nomin~al 10 year intew~al is based on porform~anco testing as di,.Ucs.d in NEDO 32977 A, FREAWo;, PChopck Valvo To-'ting Rclamation" (Ref. 8). Fwutheffmor, any rcactOr accoraancc thn Mn vaintenanec rwice 1rouramf to ensburc
| |
| -v.rall roeliabil;,ity is na*inta.i-nd. Appropr-at' .or-c-tov, actiono Will bo takon if fa*ilu-ro o Iood mthc established mn rrrnrmin nriR*RR t inrririn 10)nriffilRAO . R=0 nRffnn~rin..........-.f-.
| |
| V*
| |
| that thpcc componcntc aoehighly roliab'Po ad that fai lurc to isolate arc Ycry infrcqucnt. Thcrcforc, testing of a A
| |
| rcn~iceataivei. sample: wa Puuiludedi in He: ae-e-etableui i+uri A rcllability standpeint.
| |
| (continued)
| |
| NMP2 B 3.6.1.3-17 Revision@, 3 (A96)
| |
| | |
| PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.10 REQUIREMENTS (continued) The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired, and shall be installed in accordance with the manufacturer's recommendations. Other administrative controls, such as those that limit the shelf Ilnsert 3 life and operating life, as applicable, of the explosive charges, musT De Tollowe dTho FroquoncY of 24 months on a STAGGERED TEST BASIS iz... ,,de.,d adequate giv.. thc admnitrtio onrolc.4F-I on rplaccmcnt charge6 and the frcqucnt chccks of circuit continufity (SR 3.6.1.34).
| |
| SR 3.6.1.3.11 This SR ensures that the leakage rate of secondary containment bypass leakage paths (with the exception of the MSIVs, which are tested per SR 3.6.1.3.12) is less than or equal to the specified leakage rate. While the MSIVs are also classified as secondary containment bypass leakage pathway valves, they are evaluated according to SR 3.6.1.3.12, and if not within limits, actions are required to be taken in accordance with ACTION D. This provides assurance that the assumptions in the radiological evaluations that form the basis of the USAR (Ref. 2) are met. The leakage rate of each bypass leakage path is assumed to be the maximum pathway leakage (leakage through the worse of the two isolation valves) unless the penetration is isolated by use of one closed and de-activated automatic valve, closed manual valve, or blind flange. In this case, the leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation device. If both isolation valves in the penetration are closed, the actual leakage rate is the lesser leakage rate of the two valves. The Frequency is required by the 10 CFR 50 Appendix J Testing Program Plan.
| |
| Bypass leakage is considered part of La.
| |
| (continued)
| |
| NMP2 B 3.6.1.3-18 Revision 0
| |
| | |
| PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.12 REQUIREMENTS (continued) The analyses in Reference 1 are based on leakage that is less than the specified leakage rate. Leakage through each MSIV must be < 24 scfh when tested at 40 psig. This ensures that MSIV leakage is properly accounted for in determining the overall primary containment leakage rate. The Frequency is required by the 10 CFR 50 Appendix J Testing Program Plan.
| |
| MSIV leakage is considered part of La.
| |
| SR 3.6.1.3.13 Surveillance of hydrostatically tested lines provides assurance that the calculation assumptions of Reference 1 are met. The acceptance criteria for the combined leakage of all hydrostatically tested lines is 1 gpm times the total number of hydrostatically tested PCIVs when tested at
| |
| _>1.10 Pa (43.73 psig). The combined leakage rates must be demonstrated in accordance with the leakage test Frequency required by the 10 CFR 50 Appendix J Testing Program Plan.
| |
| REFERENCES 1. Technical Requirements Manual.
| |
| : 2. USAR, Section 15.6.5.
| |
| : 3. USAR, Section 15.6.4.
| |
| : 4. USAR, Section 15.2.4.
| |
| : 5. 10 CFR 50.36(c)(2)(ii).
| |
| : 6. USAR, Section 6.2.4.3.2.
| |
| : 7. 10 CFR 50, Appendix J Option B.
| |
| 8- NEDU 3297( A, 'Exeess Flew' V Uflk V.alye Iczting Relaxatien," junc 2000.
| |
| NMP2 B 3.6.1.3-19 Revision 0, 3 (A96)
| |
| | |
| Drywell and Suppression Chamber Pressure B 3.6.1.4 BASES (continued)
| |
| APPLICABILITY In MODES 1, 2, and 3, a DBA could result in a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining drywell and suppression chamber internal pressure within limits is not required in MODE 4 or 5.
| |
| ACTIONS A._1 When drywell or suppression chamber internal pressure is not within the limits of the LCO, drywell and suppression chamber internal pressure must be restored to within limits within 1 hour. The Required Action is necessary to return operation to within the bounds of the primary containment analysis. The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1.1, "Primary Containment," which requires that primary containment be restored to OPERABLE status within 1 hour.
| |
| B.1 and B.2 If drywell and suppression chamber internal pressure cannot be restored to within limits 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 12 hours and to MODE 4 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.
| |
| SURVEILLANCE SR 3.6.1.4.1 REQUIREMENTS Verifying that drywell and suppression chamber internal pressure is within limits ensures that operation remains within the limits assumed in the primary containment IinseaIýi as , Tkh 12 ho-ur F=..quc..Y.f thIS SR wac dovolopcd bsased en*c.rating.. ." latd to tnding prim.a.y cnntiInmr~nt nrr~tirr~ '~'irIntIorn' cliirinn thr~ nnnlimblo MODES view of othc. indieatio.. available inhcotro.l r.oo, ,m, including alarmzG, te alcr h oprto to an abnormnal BFFavetmmn cntanmcn orzzur orimr: codition.
| |
| m--*p-'s..-i,.-,,..1÷: A.
| |
| (continued)
| |
| NMP2 B 3.6.1.4-2 Revision 0
| |
| | |
| Drywell Air Temperature B 3.6.1.5 BASES LCO across the primary containment boundary is not exceeded. As (continued) a result, the ability of primary containment to perform its design function is ensured.
| |
| APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining drywell average air temperature within the limit is not required in MODE 4 or 5.
| |
| ACTIONS A.1 When drywell average air temperature is not within the limit of the LCO, it must be restored within 8 hours. This Required Action is necessary to return operation to within the bounds of the primary containment analysis. The 8 hour Completion Time is acceptable, considering the sensitivity of the analysis to variations in this parameter, and provides sufficient time to correct minor problems.
| |
| B.1 and B.2 If the drywell average air temperature cannot be restored to within the limit 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 12 hours and to MODE 4 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.
| |
| SURVEILLANCE SR 3.6.1.5.1 REQUIREMENTS Verifying that the drywell average air temperature is within the LCO limit ensures that operation remains within the limits assumed for the primary containment analyses. In order to determine the drywell average air temperature, an arithmetic average is calculated, using measurements taken at locations within the drywell selected to provide a representative sample of the overall drywell atmosphere.
| |
| (continued)
| |
| NMP2 B 3.6.1.5-2 Revision 0
| |
| | |
| Drywell Air Temperature B 3.6.1.5 BASES SURVEILLANCE SR 3.6.1.5.1 (continued)
| |
| REQUIREMENTS The required locations are approximately as follows, in elevation and azimuth: 244 ft 0 inches, 1100; 244 ft 0 inches, 2840; 253 ft 11 inches, 169'; 255 ft 6 inches, 3260; 262 ft 3 inches, 280; 268 ft 0 inches, 2030; 282 ft 6 inches, 2430; 283 ft 0 inches, 580; 294 ft 5 inches, 1170; 296 ft 4 inches, 3230; 306 ft 9 inches, 1890; and 306 ft 9 inches, 3540.
| |
| ý The 24 heUF Frcequcncy of this SR was developed based on linset ýýýeReFat@AR e 'n r -n~r rnlited te d r' wel aveI n.rr r ni tcmpcraturce yariatioens and tempora r dep nd nt drift of a n~tFUmcntatien located inthe dr,'wol' duringthc applicabi MOD-PES -;ndtho loW prob9ability of a DBA. occurringbcF.:oo Survoillanccs. uFuthcrmorc~, the 24 hour Frcequc c I concidorod adoquate in Aioof otho' r ORIn iA~tiO n ayafilablc in the conRtrol room-R, incl'uding alarMG, to alort the operator tc an abnrm~fal BFrimarv containmont air tmvcraturc REFERENCES 1 USAR, Section 6.2.1.1.3.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.6.1.5-3 Revision 0
| |
| | |
| RHR Drywell Spray B 3.6.1.6 BASES ACTIONS A.1 (continued) primary containment bypass leakage mitigation, mixing, and fission product removal functions. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in reduced primary containment bypass leakage mitigation, mixing capability, and fission product removal capability. The 7 day Completion Time was chosen in light of the redundant RHR drywell spray capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.
| |
| B.1 With two RHR drywell spray subsystems inoperable, one subsystem must be restored to OPERABLE status within 8 hours. In this condition, there is a substantial loss of the primary containment bypass leakage mitigation and fission product removal functions. The 8 hour Completion Time is based on this loss of function and is considered acceptable due to the low probability of a DBA and because alternative methods to reduce primary containment pressure and ensure adequate mixing in the primary containment are available.
| |
| C.1 and C.2 If any Required Action and associated Completion Time cannot be 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 12 hours and to MODE 4 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.
| |
| SURVEILLANCE SR 3.6.1.6.1 REQUIREMENTS Verifying the correct alignment for manual and power operated valves in the RHR containment spray mode flow path provides assurance that the proper flow paths will exist for system operation. This SR does not apply to valves that are (continued)
| |
| NMP2 B 3.6.1.6-3 Revision 0, 26 (A125)
| |
| | |
| RHR Drywell Spray B 3.6.1.6 BASES SURVEILLANCE SR 3.6.1.6.1 (continued)
| |
| REQUIREMENTS locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable, since the RHR drywell spray mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
| |
| Insele 3 ... .p..atdu.dor procodural
| |
| .alyecntrl, impr"pcr position would affecat only a single subsystom,th probability of*c*qu*iig an cvcnt -w., the system is initi.tionof and-the system is a manually intto 1ytm. This FrFequencY has bo shown to be acceeptable based on.A operating experience.
| |
| SR 3.6.1.6.2 Verifying, by administrative means, that each required RHR pump is OPERABLE ensures that the RHR pump is capable of performing its intended function (i.e., capable of developing the assumed drywell spray flow rate) when in the drywell spray mode. This Surveillance is met by verifying that another required Surveillance, which demonstrated the RHR pump OPERABILITY, was performed within the required Frequency. The verification can be performed by examining logs or other information, to determine if a required RHR pump is out of service for maintenance or other reasons. It is not necessary to perform an additional Surveillance needed to ,/-IIn t demonstrate the OPERABILITY of the required RHR pumps.
| |
| FrFequency of 92 days Is con~'frtent with the nrmFFal RHR pump flow rate Survelllaneo FrFequency ("in accordance with the lnser~ice Testin*g ProFeam") in ether S:rveillances.
| |
| (continued)
| |
| NMP2 B 3.6.1.6-4 Revision 0,--22-(A!22), 26 (A125)
| |
| | |
| Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.7 BASES ACTIONS D.1 and D.2 (continued) brought to at least MODE 3 within 12 hours and to MODE 4 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.
| |
| SURVEILLANCE SR 3.6.1.7.1 REQUIREMENTS Each vacuum breaker is verified closed to ensure that this potential large bypass leakage path is not present. This Surveillance is performed by observing the vacuum breaker position indication or by verifying that a differential pressure of _> 0.25 psid between the suppression chamber and drywell is maintained for 1 hour without makeup. 1 InseAny Fr*.quocy is based on cmmimccmmng judgment, is considcrmd adequate in view of othcr indications of Yacwuum brakor status aya*iablo to 9peratienc porconnol18, and h-Ac1 boonP- GhoWn to be accceptable through opcrating cxpcricncc.
| |
| Two Notes are added to this SR. This first Note allows suppression chamber-to-drywell vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers. The second Note is included to clarify that vacuum breakers open due to an actual differential pressure are not considered as failing this SR.
| |
| Insert 3 SR 3.6.1.7.2 Each vacuum breaker must be c cled to ensure that it opens adequately to perform its design7unction and returns to the fully closed position. This ensur s that the safety analysis assumptions are valid. he -1 ,a ....
| |
| this SR was dvolopod, ba..od, on .I""ico Toeting Program.
| |
| roquiromon~ts to pcrform )valve testing at lcast oncc cvcr; 02 days. A 31 day Froquoncy waG chocon to provido additional assurancc that the YacuumA brcakcrs arc OPEmRABL-E-,
| |
| (continued)
| |
| NMP2 B 3.6.1.7-5 Revision 0
| |
| | |
| Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.7 BASES SURVEILLANCE SR 3.6.1.7.2 (continued)
| |
| REQUIREMENTS SiRnc they aFe Iocatod OR a harsh en.ironment (the suppressic,* GhambeF*airspace). In addition, this functional test is required within 12 hours after a discharge of steam to the suppression chamber from the safety/relief valves.
| |
| SIR 3.6.1.7.3 Verification of the vacuum breaker opening setpoint is necessary to ensure that the safety analysis assumption regarding vacuum breaker opening (i.e., starts to open from the fully closed position) differential pressure of jlnsert 3 n.25.io 2m<
| |
| : i. vli~l5.The 24 ,,nth Fr.equency is based on the need to perfoFrm thiG Surycillancc under the condfitins that apply during a plant outage and the potential for an unplannedd transienP-t if theP* nureiae were performned with~
| |
| the reactor at power. The 214 m*nth Frequeny has been shown to be acceptable, based oR operating eX*FieR*e, aRndis further justified because of other sur.'eillanccs performfed at shorter FrFequencies that eenvey the pro~per functioning status of each VacHuum breaker.
| |
| REFERENCES 1. USAR, Section 6.2.1.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.6.1.7-6 Revision 0
| |
| | |
| Suppression Pool Average Temperature B 3.6.2.1 BASES ACTIONS E.1 and E.2 (continued) temperature or pressure. Furthermore, if a blowdown were to occur when temperature was > 120 0 F, the maximum allowable bulk and local temperatures could be exceeded very quickly.
| |
| SURVEILLANCE SR 3.6.2.1.1 REQUIREMENTS The suppression pool average temperature is regularly monitored to ensure that the required limits are satisfied.
| |
| Average temperature is determined by taking an arithmetic average of at least one OPERABLE post accident monitoring instrumentation channel in each suppression pool quadrant.
| |
| Alternatively, average temperature can be determined by taking an arithmetic average of 10 OPERABLE suppression pool Insert 3 water temperature channels, which are distributed in different suppression pool sectors. There is no divisional fqfuirement with respect to the instrument channels for this Sl. The 24 hour Frcquccy hasc boon shown to bo accoptablo based on epcrating cxpcrfiencc. When heat isbeing addcdt the GUPPrcccien pool by testing, howcvcr, it isnccoccar; to monitor SUPP., ssien pool tmperat... more frcqucntly. The Ifrequency is minute Frequency during testing is justified by the rates Sat ch testing will heat up the suppression pool, has been shown I e acceptable based on operating experience, and provides as eunce that allowable pool temperatures are not exceeded. Th u further justified in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool average temperature condition.
| |
| REFERENCES 1. USAR, Section 6.2.1.1.3.
| |
| : 2. USAR, Appendix 6A.10.1.
| |
| : 3. NUREG-0783.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.6.2.1-5 Revision 0
| |
| | |
| Suppression Pool Water Level B 3.6.2.2 BASES ACTIONS B.1 and B.2 (continued) achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 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.
| |
| SR 3.6.2.2.1 Ilnsert 3-1 SURVEILLANCE REQUIREMENTS Verification of the suppression pool water evel is to ensure that the required limits are satisfie . The-24-heu Frequency has bccn shown to be a..eptablc based on operating ep~Erinc. FurthcFRmOre, the 24 hour FrFequency i
| |
| %ýcnidered adequate in view *fotho indic tion.ylailable in the control roomf, including alaFrms, to alort the operator to annUPProssion abnormalncvel pool water condition.
| |
| REFERENCES 1. USAR, Section 6.2.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.6.2.2-3 Revision 0
| |
| | |
| RHR Suppression Pool Cooling B 3.6.2.3 BASES ACTIONS B.1 (continued)
| |
| With two RHR suppression pool cooling subsystems inoperable, one subsystem must be restored to OPERABLE status within 8 hours. In this condition, there is a substantial loss of the primary containment pressure and temperature mitigation function. The 8 hour Completion Time is based on this loss of function and is considered acceptable due to the low probability of a DBA and the potential avoidance of a plant shutdown transient that could result in the need for the RHR suppression pool cooling subsystems to operate.
| |
| C.1 and C.2 If any Required Action and associated Completion Time cannot be 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 12 hours and to MODE 4 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.
| |
| SURVEILLANCE SR 3.6.2.3.1 REQUIREMENTS Verifying the correct alignment for manual and power operated valves in the RHR suppression pool cooling mode flow path provides assurance that the proper flow path exists for system 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 being locked, sealed, or secured.
| |
| A valve is also allowed to be in the nonaccident position, provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable, since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
| |
| (continued)
| |
| NMP2 B 3.6.2.3-3 Revision 0
| |
| | |
| RHR Suppression Pool Cooling B 3.6.2.3 BASES SURVEILLANCE SR 3.6.2.3.1 (continued)
| |
| REQUIREMENTS loosero3 u......dpr
| |
| . con.trol, imprepr v.p.iti.
| |
| .Oc.dural woul-id affcct only a singlo subcystem, tho probab Hit,' of anf ovont roquiring initiation of the system is low, and the system ic a manually :nitiatod
| |
| * ystem. This F*c.u...y has beenuhown to be accoptablo, based on oporatkig SR 3.6.2.3.2 Verifying each required RHR pump develops a flow rate
| |
| _>7450 gpm, while operating in the suppression pool cooling mode with flow through the associated heat exchanger, ensures that the primary containment peak pressure and temperature can be maintained below the design limits during a DBA (Ref. 1). The flow is also a normal test of centrifugal pump performance required by the ASME OM Code (Ref. 3). This test confirms one point on the pump design curve, and the results are indicative of overall performance. Such inservice tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program.
| |
| REFERENCES 1. USAR, Section 6.2.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| : 3. ASME Code for Operation and Maintenance of Nuclear Power Plants.
| |
| NMP2 B 3.6.2.3-4 Revision 0, 28 (A129)
| |
| | |
| RHR Suppression Pool Spray B 3.6.2.4 BASES ACTIONS A.1 (continued)
| |
| However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in reduced primary containment bypass mitigation capability. The 7 day Completion Time was chosen in light of the redundant RHR suppression pool spray capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.
| |
| B.1 With both RHR suppression pool spray subsystems inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours. In this condition, there is a substantial loss of the primary containment bypass leakage mitigation function. The 8 hour Completion Time is based on this loss of function and is considered acceptable due to the low probability of a DBA and because alternative methods to reduce pressure in the primary containment are available.
| |
| C.1 and C.2 If any Required Action and associated Completion Time cannot be 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 12 hours and MODE 4 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.
| |
| SURVEILLANCE SR 3.6.2.4.1 REQUIREMENTS Verifying the correct alignment for manual and power operated valves in the RHR suppression pool spray mode flow path provides assurance that the proper flow paths will exist for system 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 is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is (continued)
| |
| NMP2 B 3.6.2.4-3 Revision 0
| |
| | |
| RHR Suppression Pool Spray B 3.6.2.4 BASES SURVEILLANCE SR 3.6.2.4.1 (continued)
| |
| REQUIREMENTS acceptable since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
| |
| linsert 3 >The .rcquencyf 31 days is justified bccausc th.e alvc. arc oprao u ndor pfrocodural con~trol, imfffprop valve pesitien would affcct only a singlo subsystem, the probability of an event requiring initiation of the system or, low, and the subsystem i- a manualy initiatod system. This Froequoncy has been shown to be acceptable based on oer~eatingE SR 3.6.2.4.2 Verifying each required RHR pump develops a flow rate
| |
| > 450 gpm while operating in the suppression pool spray mode helps ensure that the primary containment pressure can be maintained below the design limits during a DBA (Ref. 1).
| |
| The normal test of centrifugal pump performance required by the ASME OM Code (Ref. 3) is covered by the requirements of LCO 3.6.2.3, "RHR Suppression Pool Cooling."
| |
| The Frequency of this SR is in accordance with the Inservice Testing Program.
| |
| REFERENCES 1. USAR, Section 6.2.2.2.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| : 3. ASME Code for Operation and Maintenance of Nuclear Power Plants.
| |
| NMP2 B 3.6.2.4-4 Revision 0, 28 (A129)
| |
| | |
| Primary Containment Oxygen Concentration B 3.6.3.2 BASES ACTIONS B.1 (continued)
| |
| If oxygen concentration cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, power must be reduced to < 15% RTP within 8 hours. The 8 hour Completion Time is reasonable, based on operating experience, to reduce reactor power from full power conditions in an orderly manner and without challenging plant systems.
| |
| Insert 3 SURVEILLANCE SR 3.6.3.2.1 REQUIREMENTS The primary containment must be determined to be inerted by verifying that oxygen concentration is < 4.0 v/o. T-.rv a FrFequency i basod on the slew rate at Which oxygen conccntration can chang and on othcr indications Et abnormnal conditioe%(which could lead to moRo~ froquont checking by operators in accordance with plant pocuo)
| |
| Also, this FroequcncY haG boon shown to be accoeptablc threoug operating cxcicc.
| |
| REFERENCES 1. USAR, Section 6.2.5.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.6.3.2-3 Revision 0
| |
| | |
| Secondary Containment B 3.6.4.1 BASES ACTIONS C.1 and C.2 (continued) reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
| |
| SURVEILLANCE SR 3.6.4.1.1 REQUIREMENTS This SR ensures that the secondary containment boundary is sufficiently leak tight to preclude exfiltration. e 24 hGHr Fr"nniinR.' Gf thi" RR was deyeriond basednp Insert 3I oporating oxporionco rol-atcd to cocon~dar,' cntainment vacuumR Yariations duFrig the applicablc MODES and the low probability of a DBA occ-Gurring botwoon IurvcillaAG*G.
| |
| Furthormoro, the 24 hour Frouoncy ic conidercd ad.quatciin Yecw or etncr indcateons avail aIcI in Inh L eln roI F8Im.
| |
| incl~uding alarmG, to alert thoerjpeuwwr1 4- Rn A14nRmi
| |
| - L. . . . . .
| |
| cocondary cont-ainmont Yacui UFA 59RitiAR SR 3.6.4.1.2 and SR 3.6.4.1.3 Verifying that secondary containment equipment hatches and one access door in each access opening are closed ensures that the infiltration of outside air of such magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from the secondary containment will not occur. In this application, the term "sealed" has no connotation of leak tightness.
| |
| Maintaining secondary containment OPERABILITY requires verifying one door in the access opening is closed. An access opening contains one Inner and one outer door. In some cases, a secondary containment barrier contains multiple inner or multiple outer doors. For these cases, the access openings share the inner door or the outer door, i.e., the access openings have a common inner door or outer door. The intent is not to breach the secondary containment at any time when secondary containment is required.
| |
| This is achieved by maintaining the inner or outer portion of the barrier closed at all times; i.e., all inner doors closed or all outer doors closed. Thus, each access opening has one door closed. However all secondary containment access doors are normally kept closed, except when the access opening is being used for entry and exit or when maintenance is being (continued)
| |
| NMP2 B 3.6.4.1-4 Revision 0, 5
| |
| | |
| Secondary Containment B 3.6.4.1 BASES SURVEILLANCE SR 3.6.4.1.2 and SR 3.6.4.1.3 (continued) Insert 3 REQUIREMENTS performed on an access opening. *dey-e F Ire- fer thcco~ RSRs has becn &hownA..R to boeout based on oporating cxpcicnc, and is considcrcd adequate in vie;A. of tho othor indicAation-s of door and hatch status taaoavailablo to the epeFater-SR 3.6.4.1.4 and SR 3.6.4.1.5 The SGT System exhausts the secondary containment atmosphere to the environment through appropriate treatment equipment.
| |
| Each SGT subsystem is designed to draw down pressure in the secondary containment to >_0.25 inches of vacuum water gauge in <_66.7 seconds and maintain pressure in the secondary containment at _>0.25 inches of vacuum water gauge for 1 hour at a flow rate of < 2670 cfm. To ensure that all fission products released to the secondary containment are treated, SR 3.6.4.1.4 and SR 3.6.4.1.5 verify that a pressure in the secondary containment that is less than the lowest postulated pressure external to the secondary containment boundary can rapidly be established and maintained. When the SGT System is operating as designed, the establishment and maintenance of secondary containment pressure cannot be accomplished if the secondary containment boundary is not intact. Establishment of this pressure is confirmed by SR 3.6.4.1.4, which demonstrates that the secondary containment can be drawn down to >_0.25 inches of vacuum water gauge in < 66.7 seconds with the initial secondary containment pressure >_0 psig, using one SGT subsystem. SR 3.6.4.1.5 demonstrates that the pressure in the secondary containment can be maintained > 0.25 inches of vacuum water gauge for 1 hour using one SGT subsystem at a flow rate < 2670 cfm. This flow rate is the assumed secondary containment leak rate during the drawdown period.
| |
| The 1 hour test period allows secondary containment to be in thermal equilibrium at steady state conditions. The drawdown test conditions must be adjusted based on the methodology in Reference 5 to compensate for actual inleakage flow and initial conditions during the test. The primary purpose of these SRs is to ensure secondary containment boundary integrity. The secondary purpose of these SRs is to ensure that the SGT subsystem being tested functions as designed. There is a separate LCO with Surveillance Requirements that serves the primary purpose of (continued)
| |
| NMP2 B 3.6.4.1-5 Revision 0
| |
| | |
| Secondary Containment B 3.6.4.1 BASES SURVEILLANCE SR 3.6.4.1.4 and SR 3.6.4.1.5 (continued)
| |
| REQUIREMENTS ensuring OPERABILITY of the SGT System. Thcs SRq, need not be *poformm d with oAch ST*vm Fubsymto..m. The SGTiuby* ommm uscd foer thcsc Sur'.'illanccs is staggcrcd to ensUrc that i
| |
| -addit-ionRto the rogUiromonP-tc. of LCOG 3.6.4.3, oithor SGT-subystom will p+....R. this test. The inoperability of the SGT System does not necessarily constitute a failure of these Surveillances relative to the secondary containment OPERABILITY. cxpi*,,nc. has shown the s+. .ndary
| |
| +pc-rting cOntainmn+t bouindary usually pasoc theos surcillancRs whRn I *m
| |
| * J nortormod ~t tflr~ ~i mnntn i-rc~riucnr~v I ncrrnnrn +h^
| |
| Frogqunc,,y was con.hludod to bo acc.ptablo from a ro.iabimty standpeint-REFERENCES 1. USAR, Section 3.6A.2.1.5.
| |
| : 2. USAR, Section 15.6.5.
| |
| : 3. USAR, Section 15.7.4.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| : 5. USAR, Section 6.2.3.4.
| |
| NMP2 B 3.6.4.1-6 Revision 0
| |
| | |
| SCIVs B 3.6.4.2 BASES ACTIONS C.1 and C.2 (continued) reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
| |
| D.1 and D.2 If any Required Action and associated Completion Time cannot be met, the plant must be placed in a condition in which the LCO does not apply. If applicable, the movement of recently irradiated fuel assemblies in the secondary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be immediately initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.
| |
| Required Action D.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
| |
| SURVEILLANCE SR 3.6.4.2.1 REQUIREMENTS This SR verifies each secondary containment isolation manual valve and blind flange that is not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the secondary containment boundary is within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification that those SCIVs in secondary containment that are capable of being mispositioned are in the correct position.
| |
| 5Sinco- these SCIVz, arc rcadily accoccsiblo to porconnol duFrin nrAmal unt peration and Yerification of their pocition is rolativcly easy, the 31 day Frcequency war, chozcn to providc (continued)
| |
| NMP2 B 3.6.4.2-5 Revision 0, 5
| |
| | |
| SCIVs B 3.6.4.2 BASES SURVEILLANCE SR 3.6.4.2.1 (continued)
| |
| REQUIREMENTS ad-ded- accuranc that the 8CIVs arc i the corroct pocitionc.
| |
| ,*This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were lInsert 3 verified to be in the correct position upon locking, sealing, or securing.
| |
| Two Notes have been added to this SR. The first Note applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these SCIVs, once they have been verified to be in the proper position, is low.
| |
| A second Note has been included to clarify that SCIVs that are open under administrative controls are not required to meet the SR during the time the SCIVs are open. These controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated. is in accordance with the I-Jnservice Testing Program SR 3.6.4.2.2 and Verifying the isolation time of each powe operated, automatic SCIV is within limits is requir d to demonstrate OPERABILITY. The isolation time test ensures that the SCIV will isolate in a time period less than or equal to that assumed in the safety analyses. The- Fq ency of this SR SR 3.6.4.2.3 Verifying that each automatic SCIV closes on a secondary containment isolation signal is required to prevent leakage of radioactive material from secondary containment following a DBA or other accidents. This SR ensures that each automatic SCIV will actuate to the isolation position on a secondary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.2, "Secondary Containment (continued)
| |
| NMP2 B 3.6.4.2-6 Revision 0
| |
| | |
| SCIVs B 3.6.4.2 BASES SURVEILLANCE SR 3.6.4.2.3 (continued)
| |
| REQUIREMENTS Isolation Instrumentation," overlaps this SR to provide complete testing of the safety function. While this Surefcillancc can be pecformcd with the rcactGr at pewcr, oporating oXporionco has shoWn thoco coempenents usually pass tho Sumollancoe whcn pcrfoFrmcd at the 24 month
| |
| - WOe-Frcgucncy, Which us on tho rbased.efulig cylo. Thoroforo, tho FroguoncywaG co ncludoed- to bo accoptabic fromn a roliability stasndpein-REFERENCES 1. USAR, Section 15.6.5.
| |
| : 2. USAR, Section 15.7.4.
| |
| : 3. Technical Requirements Manual.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.6.4.2-7 Revision 0
| |
| | |
| SGT System B 3.6.4.3 BASES (continued)
| |
| SURVEILLANCE SR 3.6.4.3.1 REQUIREMENTS Operating (from the control room using the manual initiation switch) each SGT subsystem for >_10 continuous hours ensures that both subsystems are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on (automatic heater cycling to maintain period temperature) for Ž_10 continuous hours e,.y days eliminates moisture on the adsorbers and HEPA filters.
| |
| a'-allable 'Athe system.l S R 3.6.4.3.2 --Insert3 This SR verifies that the required SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The SGT System filter tests are in accordance with Regulatory Guide 1.52 (Ref. 6). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specified test frequencies and additional information are discussed in detail in the VFTP.
| |
| SR 3.6.4.3.3 This SR requires verification that each SGT subsystem starts upon receipt of an actual or simulated initiation signal.
| |
| Insert 3 --Tho, GIC SYSTEM FUNCTION,.L TEST- in LCO 3.3.6.2, "Se..dary Conta.min.t lsolation n.etrum..tation," ovo..ap. this SR to proYido complctc testing of the safcy function. Whole thin Sur~illacoGan bo pe&Knrmd with the rcactor at po.. c, epcrating cxpcricnco has chown thoco components usually pass the Surve*Iancc when pcrformcd at the 24 moneth Frcequcncy,-
| |
| which OGbac-d-onR tho rcfucling cycalc. Thcroforc, the Frcequcncy waG concludcd to be acooptablo frm a rcliability Gtandpein-(continued)
| |
| NMP2 B 3.6.4.3-5 Revision 0
| |
| | |
| SGT System B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.4 REQUIREMENTS (continued) This SR requires verification that the SGT decay heat removal air inlet valves can be opened. This ensures that the decay heat removal mode of SGT System operation is avaI While
| |
| .. ... this
| |
| ...-.. Su...illancc
| |
| ..... can bc pcf.R.ed CL'3 ........ with Inset - the Feafi9Fu at pewer, epcruung expeRicnuc FiaG SriuwRi iric
| |
| .A compencnts usually pass the Surwcillancc when pecfeRned at the 24 moenth rFrcquency, which is based on the rcfuclig cycic. TheTrcfor, the Frcquency was nrGcnuded to be a*,cptabl, from a r.,- ability standoJit.
| |
| REFERENCES 1. 10 CFR 50, Appendix A, GDC 41.
| |
| : 2. USAR, Section 6.5.1.2.2.
| |
| : 3. USAR, Section 15.6.5.
| |
| : 4. USAR, Section 15.7.4.
| |
| : 5. 10 CFR 50.36(c)(2)(ii).
| |
| : 6. Regulatory Guide 1.52, Rev. 2.
| |
| NMP2 B 3.6.4.3-6 Revision 0
| |
| | |
| SW System and UHS B 3.7.1 BASES ACTIONS G.1 and G.2 (continued)
| |
| The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
| |
| The Required Actions are modified by a Note indicating that the applicable Conditions of LCO 3.4.9, "Residual Heat Removal (RHR)
| |
| Shutdown Cooling System - Hot Shutdown," be entered and the Required Actions taken if the inoperable SW System or UHS results in an inoperable RHR shutdown cooling subsystem. This is in accordance with LCO 3.0.6 and ensures the proper actions are taken for the RHR Shutdown Cooling System.
| |
| SURVEILLANCE SR 3.7.1.1 REQUIREMENTS Verification that the water temperature of the intake tunnels is > 38°F ensures that frazil ice, which can block the intake tunnels, cannot form.
| |
| This ensures that the intake tunnels can perform their intended function.
| |
| This Surveillance is only required to be met when SR 3.7.1.5 and SR 3.7.1.8 are not satisfied. With the Intake Deicer Heater System OPERABLE (and SR 3.7.1.5 and SR 3.7.1.8 met), frazil ice cannot form even with the intake tunnels water temperature < 38 F. ,he,2c'-
| |
| 12 FroequcncY iGbased on operating cxp8rience rol-ated1to 4Ronig of the parFamctcr variatins during thc applicable MODES.
| |
| Ilnsert 3 SR 3.7.1.2 This SR verifies the water level in the SW pump intake bay to be sufficient for the proper operation of the SW pumps (net positive suction head and pump vortexing are considered in determining this limit). The water level limit, 233.1 ft, is referenced to mean sea level. The. 24 hour Fr.equenc i isr3 S R 3.7.1.3 Verification of each SW subsystem supply header temperature ensures that the heat removal capability of the SW System is within the assumptions of the DBA analysis. The 24 hour F.requency i6 ba.od 9A trnntinm ipri)
| |
| NMP2 B 3.7.1-9 Revision , 13*,(A!*1,3), 19 (Al 19)
| |
| | |
| SW System and UHS B 3.7.1 BASES SURVEILLANCE SR 3.7.1.3 (continued)
| |
| REQUIREMENTS oporating oeoioc olatod to trending of the paramotor YariatineR
| |
| )"Gatble ,MG owever, if a SW subsystem supply header water temperature is > °F, the Surveillance must be performed more frequently (every 4 hours), ince the condition is closer to the maximum water temperature limif I-Lnsert" 3 SR 3.7.1.4 Verification that each required SW pump is in operation ensures that an adequate number of SW pumps are operating to perform the long term containment cooling function during a LOCA. Tho 2R hour Froquocny ic based on operating experiense And the opora R'iherent knowledge oe SR 3.7.1.5 nsert 3 The current for each required heater feeder cable is required to be checked to ensure the proper number of heaters are OPERABLE for each intake deicer heater division. The Surveillance is performed by verifying, at the motor control centers, that the current is > 20 amps (total for all three phases when adjusted to degraded voltage conditions, i.e., 518 volts) in each intake structure for each division. The current limit is based upon ensuring 14 heaters are OPERABLE (which includes in operation) in an intake structure. This Surveillance is only required to be met when SR 3.7.1.1 is not satisfied, since with the intake tunnels water temperature > 38°F (i.e., SR 3.7.1.1 met), frazil ice cannot form even with the intake deicer heaters inoperable The 7. day Frequency is ba..P.d on ep-k at-, e..x.e*II,.n-,-,- tha thas - Ie..lr-* a.* .,ass,se-1-,;nt usually, i"
| |
| S R 3.7.1.6 Verifying the correct alignment for each manual, power operated, and automatic valve in each SW subsystem flow path provides assurance that the proper flow paths will exist for SW 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 is also allowed to be in the nonaccident (continued)
| |
| NMP2 B 3.7.1 -10 Revision, O ^1-3(A!1-3), 19 (Al 19)
| |
| | |
| SW System and UHS B 3.7.1 BASES SURVEILLANCE SR 3.7.1.6 (continued)
| |
| REQUIREMENTS position and yet considered in the correct position, provided it can be automatically realigned to its accident position within the required time.
| |
| This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
| |
| This SR is modified by a Note indicating that isolation of the associated SW subsystem to components or systems may render those components or systems inoperable, but does not affect the OPERABILITY of the SW subsystem. As such, when all SW pumps, valves, and piping are OPERABLE, but a branch connection off the main header is isolated, the SW subsystem is still OPERABLE.
| |
| >_Thc 31 day Frcequcncy I based en nanria ~udement.I meoc~t and oncurcz wircth thev pozitionz.e~t els aovorninGH.aAKP__oFoatln.
| |
| Ao SR 3.7.1.7 lnsert 3 This SR verifies that the automatic isolation valves (i.e., SW isolation valves servicing non-safety related equipment, SW supply header cross connect valves, and SW pump discharge valves of non-operating SW pumps) of the SW System will automatically switch to the safety or emergency position to provide cooling water exclusively to the safety related equipment during a transient event (i.e., LOOP). This is demonstrated by use of an actual or simulated initiation signal. This SR also verifies the automatic start capability of the SW pump (and associated pump discharge valve opening capability) in each subsystem.
| |
| pcricnco hac chown that thccc eomp enontc usually paszth Sk-wheR-peý ..rmd on the 24 month Fro-quoncy. TFhcrcforo, this FFequieRGY 0-concludcd to be accceptablc from a Fi eliability standpoint.
| |
| SR 3.7.1.8 The resistance of each required heater feeder cable and associated heater elements is required to be checked to ensure the required heaters are OPERABLE for each intake deicer heater division. The Surveillance is performed by verifying that the resistance is _>28 ohms for each required heater feeder cable and associated heater element.
| |
| (continued)
| |
| NMP2 B 3.7.1-11 Revision 0, 13 (Al 13)
| |
| | |
| SW System and UHS B 3.7.1 BASES SURVEILLANCE SR 3.7.1.8 (continued)
| |
| REQUIREMENTS The minimum resistance is based on ensuring the intake structure bar racks are heated sufficiently such that the SW flow assumed to safely shutdown the unit can be achieved through the intake structures. This Surveillance is only required to be met when SR 3.7.1.1 is not satisfied, since with the intake tunnels water temperature _> 38 0 F (i.e., SR 3.7.1.1 met), frazil ice cannot form even with the intake deicer heaters Insert 3 1*._nper)->Thc 21 month Fr--quc*'; us i. barod on ating
| |
| .p .. n that has shown that thcco compencntc ucually pascc thicsurcla when pcrfermcd at this Frcequcncy.
| |
| REFERENCES 1. Regulatory Guide 1.27, Revision 2, January 1976.
| |
| : 2. USAR, Section 9.2.1.
| |
| : 3. USAR, Section 9.2.5.
| |
| : 4. USAR, Tables 9.2-1 and 9.2-1A.
| |
| : 5. USAR, Section 6.2.
| |
| : 6. USAR, Section 6.3.
| |
| : 7. USAR, Chapter 15.
| |
| : 8. USAR, Appendix A.
| |
| : 9. USAR, Section 6.2.2.
| |
| : 10. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.7.1-12 Revision 0, 13 (Al 13)
| |
| | |
| CREF System B 3.7.2 BASES ACTIONS If applicable, movement of recently irradiated fuel (continued) assemblies in the secondary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. If applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.
| |
| SURVEILLANCE SR 3.7.2.1 REQUIREMENTS Operating (from the control room) each CREF subsystem for >_1 continuous hour ensures that both subsystems are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, filter booster or air conditioning unit fan or motor failure, or excessive vibration can be detected for corrective action. In addition, it is not necessary to operate all components of a single subsystem simultaneously for the 1 hour period. It is acceptable to operate the fan portion of the air conditioning unit(s) of one subsystem with the CROASFT of the other subsystem, such that the CROASFTs and fan portion of the air conditioning units are each operated for 1 continuous llnsert Iw Fel!abilib*o ef f-AR moAtor=S OR..---'~ a d Ge~trc!= a =d the redu~dan cy available in thecytm SR 3.7.2.2 This SR verifies that the required CROASFT testing is performed in accordance with Specification 5.5.7, "Ventilation Filter Testing Program (VFTP)." The CROASFT filter tests are in accordance with Regulatory Guide 1.52 (Ref. 8). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test Frequencies and additional information are discussed in detail in the VFTP.
| |
| SR 3.7.2.3 This SR verifies that each CREF subsystem starts and operates on an actual or simulated initiation signal. This SR also includes ensuring the air conditioning units (fan portion only) start on a low flow signal after the appropriate time delay. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.7.1, "Control Room Envelope Filtration (CREF)
| |
| System Instrumentation," overlaps this SR to provide complete testing of the safety function. Oporating 'xpr*"inc has- .hown that these (continued)
| |
| NMP2 B 3.7.2-8 Revision 0, 1, 2, 27 (Al 26)
| |
| | |
| CREF System B 3.7.2 BASES SURVEILLANCE c...pono-..tc. normally pass the SR when p...orm..d at the 24 mon-th REQUIREMENTS Fr.quoRny. Thor.foro, the Fr. *uncy Wa, .. f9oun.d, to. b...acc.ptabl (continued) from a roliability standpoint.
| |
| SR 3.7.2.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE.
| |
| The details of the testing are specified in the Control Room Envelope Habitability Program.
| |
| The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of DBA consequences is no more than 5 rem TEDE and the CRE occupants are protected from hazardous chemicals and smoke. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences. When unfiltered air inleakage is greater than the assumed flow rate, Condition B must be entered. Required Action B.3 allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within the licensing basis habitability limits for the occupants following an accident. Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref. 11)which endorses, with exceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 12). These compensatory measures may also be used as mitigating actions as required by Required Action B.2. Temporary analytical methods may also be used as compensatory measures to restore OPERABILITY (Ref. 13). Options for restoring the CRE boundary to OPERABLE status include changing the licensing basis DBA consequence analysis, repairing the CRE boundary, or a combination of these actions. Depending upon the nature of the problem and the corrective action, a full scope inleakage test may not be necessary to establish that the CRE boundary has been restored to OPERABLE status.
| |
| REFERENCES 1. USAR, Section 6.4.2.1.
| |
| : 2. Regulatory Guide 1.183, July 2000.
| |
| : 3. USAR, Section 6.4.1.
| |
| : 4. USAR, Section 9.4.1.
| |
| : 5. USAR, Chapter 6.
| |
| (continued)
| |
| NMP2 B 3.7.2-9 Revision 0,-*2, 26'(A*l2-5), 27 (A126)
| |
| | |
| Control Room Envelope AC System B 3.7.3 BASES ACTIONS H.1 and H.2 (continued)
| |
| During movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs if Required Actions of Condition C or Condition D cannot be met within the required Completion Times, action must be taken to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room envelope. This places the unit in a condition that minimizes risk.
| |
| If applicable, handling of recently irradiated fuel in the secondary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until the OPDRVs are suspended.
| |
| SURVEILLANCE SR 3.7.3.1 REQUIREMENTS This SR verifies that the- heat removal capability of the system is sufficient to remove the control room envelope heat load assumed in the safety analyses for the Main Control Room area and the Relay Room area. The SR consists of a combination of testing and calculation. Thp, 24 month Froquc.c; i ...... p*t sinc" signif*"ant dogradatiOn of th Ct*ontrol R*oo Envolopc AC e,,sm is not cxpceetd ovc this time. pcr.... ner-**ne3 REFERENCES 1. USAR, Section 6.4.
| |
| : 2. USAR, Section 9.4.1.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.7.3-7 Revision 28 (Al 28)
| |
| | |
| Main Condenser Offgas B 3.7.4 BASES (continued)
| |
| SURVEILLANCE SR 3.7.4.1 REQUIREMENTS This SR, on a 31 day Fr.... uoy, requires an isotopic analysis of an offgas sample (taken before holdup and discharge downstream of the recombiner) to ensure that the required limits are satisfied. The noble gases to be sampled are Xe-1 33, Xe-1 35, Xe-1 38, Kr-85, Kr-87, and Kr-88.
| |
| If the measured rate of radioactivity increases significantly as indicated by either offgas pretreatment radiation monitor (by > 50% after correcting for expected increases due to changes in THERMAL POWER), an isotopic analysis is also performed within 4 hours after the increase is noted, to ensure that the increase is not indicative of a sustained increase in the radioactivity rate. The** -,*y
| |
| -Frguecncy as adequate in vicw of otho, that continuouY mon..itor thoeoga, i9...accoptablo
| |
| ., bao*d on op,.ating cxpcr',cncc. < -lnset This SR is modified by a Note indicating that the SR is not required to be performed until 31 days after any main steam line is not isolated and the SJAE is in operation. Only in this condition can radioactive fission gases be in the Main Condenser Offgas System at significant rates.
| |
| REFERENCES 1. USAR, Section 15.7.1.
| |
| : 2. NUREG-1047.
| |
| : 3. 10 CFR 100.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.7.4-3 Revision 0
| |
| | |
| Main Turbine Bypass System B 3.7.5 BASES ACTIONS B._1 (continued)
| |
| If the Main Turbine Bypass System cannot be restored to OPERABLE status and the MCPR limits for an inoperable Main Turbine Bypass System are not applied, THERMAL POWER must be reduced to < 23% RTP. As discussed in the Applicability section, operation at < 23% RTP results in sufficient margin to the required limits, and the Main Turbine Bypass System is not required to protect fuel integrity during the feedwater controller failure, maximum demand event. The 4 hour Completion Time is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
| |
| SURVEILLANCE SR 3.7.5.1 REQUIREMENTS The Main Turbine Bypass System is required to actuate automatically to perform its design function. This SR demonstrates that, with the required system initiation signals, the valves will actuate to their required position.
| |
| pA/vvwc', eperating ex1 hrincn haR horWn that thes2cmpnonth
| |
| + '* usually pass the SR' whon porformoPd at the 2-4 month I HOU F-- Frcequcncy, which is based on thP rfucling cy.
| |
| T-hercforc, tho Froequoncy was conclu dod to he acccptable fromR a rcliability standpoint.
| |
| SR 3.7.5.2 This SR ensures that the TURBINE BYPASS SYSTEM RESPONSE TIME is in compliance with the assumptions of the appropriate safety analysis. The response time limits are specified in the Technical Requirements Manual (Ref. 4).
| |
| * Ie 3 6urveoIancce can be pcrformced with tho rcacto at7powoL eperating cxpcricncc has shown that these compencnts usuall4y pass the SR when pcrformod at the 24 month Froqucncy, Which is based 9R the rcfucli~ng cyclo. T-hcrcforc, the Froequency was concludod to be accoptablo fromA a rcliabikiy sta*dpieVV (continued)
| |
| NMP2 B 3.7.5-3 Revision 0, 37 (A140)
| |
| | |
| Spent Fuel Storage Pool Water Level B 3.7.6 BASES (continued)
| |
| APPLICABILITY This LCO applies whenever movement of irradiated fuel assemblies occurs in the spent fuel storage pool or whenever movement of new fuel assemblies occurs in the spent fuel storage pool with irradiated fuel assemblies in the spent fuel storage pool, since the potential for a release of fission products exists.
| |
| ACTIONS A.1 LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since fuel assembly movement can occur in MODE 1, 2, or 3, Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. If moving fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Entering LCO 3.0.3 while in MODE 1, 2, or 3 would require the unit to be shutdown, but would not require immediate suspension of movement of fuel assemblies.
| |
| The Note to the ACTIONS, "LCO 3.0.3 is not applicable,"
| |
| ensures that the actions for immediate suspension of fuel assembly movement are not postponed due to entry into LCO 3.0.3.
| |
| When the initial conditions for an accident cannot be met, steps should be taken to preclude the accident from occurring. With the spent fuel storage pool level less than required, the movement of fuel assemblies in the spent fuel storage pool is suspended immediately. Suspension of this activity shall not preclude completion of movement of a fuel assembly to a safe position. This effectively precludes a spent fuel handling accident from occurring.
| |
| SURVEILLANCE SR 3.7.6.1 Insert 3 REQUIREMENTS This SR verifies that sufficient water is available in the event of a fuel handling accident. The water level in th spent fuel storage pool must be checked periodically.
| |
| 7 day Fqe..ncy c acptablo, bacod on peFr*a*ng eoXP eF AGnc,concGiR
| |
| -.do_1In that tho wator esVcnL,,).lumo in the pool i.c nrmFFally stable and oatcer level c-hang ae eeA#91ed by WRit PFeeedWFes.
| |
| REFERENCES 1. USAR, Section 9.1.2.
| |
| : 2. USAR, Section 15.7.4.
| |
| (continued)
| |
| NMP2 B 3.7.6-2 Revision 0
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE Where the SRs discussed herein specify voltage and frequency REQUIREMENTS tolerances, the following summary is applicable. The (continued) minimum steady state output voltage of 3950 V is approximately 95% of the nominal 4160 V output voltage.
| |
| This value, which is specified in ANSI C84.1 (Ref. 14),
| |
| allows for voltage drop to the terminals of 4000 V motors whose minimum operating voltage is specified as 90%, or 3600 V. It also allows for voltage drops to motors and other equipment down through the 120 V level where minimum operating voltage is also usually specified as 90% of name plate rating. The specified maximum steady state output voltage of 4370 V is equal to the maximum operating voltage specified for 4000 V motors. It ensures that for a lightly loaded distribution system, the voltage at the terminals of 4000 V motors is no more than the maximum rated operating voltages. The specified minimum and maximum frequencies of the DG are 58.8 Hz and 61.2 Hz, respectively. These values are equal to +/- 2% of the 60 Hz nominal frequency and are derived from the recommendations given in Regulatory Guide 1.9 (Ref. 11).
| |
| SR 3.8.1.1 This SR ensures proper circuit continuity for the offsite AC electrical power supply to the onsite distribution network and availability of offsite AC electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure that distribution buses and loads are connected to their preferred power source and that appropriate independence of offsite circuits is maintained.
| |
| ---r~Thc- ý7day Frcqucncy is adequate sinco brcakcr poSition i linsert 3 k--" notlikly o hango without the oporator being awarc of it and bccauso Its statuis is disolaycd in the control roomA.
| |
| SR 3.8.1.2 This SR helps to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and maintain the unit in a safe shutdown condition.
| |
| To minimize the wear on moving parts that do not get lubricated when the engine is not running, this SR has been modified by a Note to indicate that all DG starts for this Surveillance may be preceded by an engine prelube period. In addition, to minimize wear and tear on the DG, the Note (continued)
| |
| NMP2 B 3.8.1-18 Revision 0, 35 (Al 38)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.2 (continued)
| |
| REQUIREMENTS also allows all DG starts to be followed by a warmup period prior to loading.
| |
| For the purposes of this testing, the DGs are started from standby conditions. Standby conditions for a DG mean that the diesel engine coolant (Division 1 and 2 DGs only) and lube oil are being continuously circulated and temperature is being maintained consistent with manufacturer recommendations.
| |
| SR 3.8.1.2 requires that the DG starts from standby conditions and achieves required voltage and frequency within 10 seconds. The 10 second start requirement supports the assumptions in the design basis LOCA analysis (Ref. 15).
| |
| In addition, the DG is required to maintain proper voltage and frequency limits after steady state is achieved. The voltage and frequency limits are normally achieved within 13 seconds for the Division 1 and 2 DGs and within 15 seconds for the Division 3 DG. The time for the DG to reach steady state operation is periodically monitored and the trend evaluated to identify degradation of governor and voltage regulator performance.
| |
| LI HbtlL 0 r -',,*.Th e 31 day.,, F=e'
| |
| ... .. e .. G 5;,,. 2DA 1 " 2 r,. r ,,,,.,....,,, w,:ith R.gulate-y Guide 1.9 (Ref. 11). This Froquoncy providc
| |
| . . * . . .. . . r-* r¢- -.
| |
| ir* r-* i~mITX"J .... :L1 : :* _
| |
| ardeEluate aSSUFaRGG 94. d6%B11=1-F-T-, Hile FAIRIFRIZIRE1 ueemF Fesulting from mation iesin,.
| |
| SR 3.8.1.3 This Surveillance demonstrates that the DGs are capable of synchronizing and accepting a load approximately equivalent to that corresponding to the continuous rating. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source.
| |
| Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0 when running synchronized with the grid. The 0.8 power factor value is the design rating of the machine at a particular KVA. The 1.0 power factor value is an operational condition where the reactive power component is zero, which minimizes the reactive heating of (continued)
| |
| NMP2 B 3.8.1-19 Revision 0, 29 (A!29), 35 (A138)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.3 (continued)
| |
| REQUIREMENTS the generator. Operating the generator at a power factor between 0.8 lagging and 1.0 avoids adverse conditions associated with underexciting the generator and more closely represents the generator operating requirements when performing its safety function (running isolated on its associated 4.16 kV emergency bus). The load band is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY.
| |
| lInsert 3 with wReguiatory Guide 1.9 (Ref. 11)
| |
| Note 1 modifies this Surveillance to indicate that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized.
| |
| Note 2 modifies this Surveillance by stating that momentary transients because of changing bus loads do not invalidate this test.
| |
| Note 3 indicates that this Surveillance must be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations.
| |
| Note 4 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance.
| |
| SR 3.8.1.4 This SR provides verification that the level of fuel oil in the day tank is at or above the level at which the low-low level alarm is annunciated. The level is expressed as an equivalent volume in gallons, and is selected to ensure adequate fuel oil for a minimum of 1 hour of DG operation at full load plus 10%.
| |
| The 31 day Frcqucncy as adequate to aG.U.. that a suffi...n, supply of fuel oil is -available,sine low level alarmcs arc (continued)
| |
| NMP2 B 3.8.1-20 Revision 0, 32 (A! 3), 35 (Al 38)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.4 (continued)
| |
| REQUIREMENTS DryddAnd facilityi eporators r Would bo AWAro of any lARao IIA: PT 43'PI~ Ail 641irir ig tn; pnor1n.
| |
| ... t Insert 3 i SR 3.8.1.5 Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a periodically environment in order to survive a of water from the fuel oil day tanks cG-ey3 days eliminates the necessary environment for bacterial survival. This is most effective means in controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system.
| |
| rha Q, ;H C:rrn an,: f~ef.Wn ki 14 k 1 Da m,l+nJ 0 clf-rK7 c3riviv rv C1 - V J VC10 CK V. I d
| |
| ii - n r rýtjýmrn#-
| |
| mnaintcnancc. The prISPnoo of 4Atc docs not ncccssarily roprosont a failuroe of thiG SR proYidod that accrumulated watc iSrcmoycd duFrin porfofRmancc of this Surycillancc.
| |
| SR 3.8.1.6 Insert 3 This Surveillance demonstrates that each fuel oil transfer pump (two per DG) operates and automatically transfers fuel oil from its associated storage tank to its associated day tank. It is required to support the continuous operation of standby power sources. This Surveillance provides assurance that each fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for automatic fuel transfer systems are OPERABLE. Two fuel oil transfer pumps per DG are required since each pump only has a simplex strainer.
| |
| ;ý' IcReI Feun t---
| |
| TorA-
| |
| *L:--
| |
| ;-nR.i-i'*t'* :_
| |
| HS&A-conccrvauvc wit rcs~ccl to the tcstnmg roqu:romonts .pum.ps r as cnta.. a in the AS.ME GM o*d, (Ref. 16).
| |
| (continued)
| |
| NMP2 B 3.8.1-21 Revision 0, 35 (Al 38)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.7 REQUIREMENTS (continued) Each DG is provided with an engine overspeed trip to prevent damage to the engine. Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed, which, if excessive, might result in a trip of the engine. This Surveillance demonstrates the DG load response characteristics and capability to reject the largest single load without exceeding predetermined frequency and while maintaining a specified margin to the overspeed trip. The load referenced for Division 1 DG is the 1125 kW low pressure core spray pump; for Division 2 DG, the 750 kW residual heat removal (RHR) pump; and for Division 3 DG the 2435 kW HPCS pump. The specified load values conservatively bound the expected kW rating of the single largest loads under accident conditions. This Surveillance may be accomplished by:
| |
| : a. Tripping the DG output breaker with the DG carrying greater than or equal to its associated single largest post-accident load while paralleled to offsite power, or while solely supplying the bus; or
| |
| : b. Tripping its associated single largest post-accident load with the DG solely supplying the bus.
| |
| Consistent with Regulatory Guide 1.9 (Ref. 11), the load rejection test is acceptable if the diesel speed does not exceed the nominal (synchronous) speed plus 75% of the difference between nominal speed and the overspeed trip setpoint, or 115% of nominal speed, whichever is lower.
| |
| This corresponds to < 64.5 Hz for the Division 1 and 2 DGs and < 66.75 Hz for the Division 3 DG, which is the nominal speed plus 75% of the difference between nominal speed and the overspeed trip setpoint.
| |
| I I I I II E I The 24 mointri i-rcglucncy takes into consiacration the piant linsert 3 to perform thn roquirod.enditionc and I I . . . .. n F I n intcndcdl to bcoe GE)G'tcnt w ith expecteci fuel cycle iongtfc.-
| |
| This SR has been modified by two Notes. The reason for Note 1 is that during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, plant safety systems (this portion of Note 1 is not applicable to the Division 3 DG). Credit may be taken for unplanned events that satisfy this SR. In order to ensure that the DG is (continued)
| |
| NMP2 B 3.8.1-22 Revision ,O -2(A! 33), 35 (A138)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.7 (continued)
| |
| REQUIREMENTS tested under load conditions that are as close to design basis conditions as possible, Note 2 requires that, if synchronized to offsite power, testing must be performed at a power factor as close to the power factor of the single largest post-accident load as practicable. The power factor limit is < 0.92 for Division 1 and 2 DGs and < 0.93 for Division 3 DG. This power factor is representative of the actual design basis inductive loading that the DG could experience. However, since the offsite electrical power transmission network is not balanced, it may not be possible to raise DG voltage sufficiently to meet the power factor limit without one phase of the DG exceeding the current limit. Therefore, to ensure the DG is not placed in an unsafe condition during this test, the power factor limit does not have to be met ifthe offsite grid phase imbalance does not permit the power factor limit to be met when the DG is tied to the grid. When this occurs, the power factor should be maintained as close to the limit as practicable.
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| SR 3.8.1.8 Consistent with Regulatory Guide 1.9 (Ref. 11), paragraph C.2.2.8, this Surveillance demonstrates the DG capability to reject a full load without overspeed tripping or exceeding the predetermined voltage limits. The DG full load rejection may occur because of a system fault or inadvertent breaker tripping. This Surveillance ensures proper engine generator load response under the simulated test conditions.
| |
| This test simulates the loss of the total connected load that the DG experiences following a full load rejection and verifies that the DG does not trip upon loss of the load.
| |
| These acceptance criteria provide DG damage protection.
| |
| While the DG is not expected to experience this transient during an event, and continues to be available, this response ensures that the DG is not degraded for future application, including reconnection to the bus if the trip initiator can be corrected or isolated.
| |
| In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, testing must be performed at a power factor as close to the accident load power factor as practicable. The power factor limit is < 0.91 for Division 1 and 2 DGs and
| |
| < 0.93 for Division 3 DG. This power factor is (continued)
| |
| NMP2 B 3.8.1-23 Revision 0, 32 (A133), 35 (A138)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.8 (continued)
| |
| REQUIREMENTS representative of the actual design basis inductive loading that the DG would experience.
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| ; "Tho24 mon.th Froquen.* y takes into considration the plant lnsert 3 - cnditions requ'red to po.focrm the Sur'veiance, anrd s intended o bo coniLtent with oxpected fuel cyclec engths.
| |
| This SR has been modified by two Notes. The reason for Note 1 is that during operation with the reactor critical, performance of this SR could cause perturbation to the electrical distribution systems that could challenge continued steady state operation and, as a result, plant safety systems (this portion of Note 1 is not applicable to the Division 3 DG).
| |
| Credit may be taken for unplanned events that satisfy this SR.
| |
| Note 2 is provided in recognition that since the offsite electrical power transmission network is not balanced, it may not be possible to raise DG voltage sufficiently to meet the power factor limit without one phase of the DG exceeding the current limit. Therefore, to ensure the DG is not placed in an unsafe condition during this test, the power factor limit does not have to be met if the offsite grid phase imbalance does not permit the power factor limit to be met when the DG is tied to the grid.
| |
| When this occurs, the power factor should be maintained as close to the limit as practicable.
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| SR 3.8.1.9 Consistent with Regulatory Guide 1.9 (Ref. 11),
| |
| paragraph C.2.2.4, this Surveillance demonstrates the as designed operation of the standby power sources during loss of the offsite source. This test verifies all actions encountered from the loss of offsite power, including shedding of the nonessential loads (Divisions 1 and 2 only) and energization of the emergency buses and respective loads from the DG. It further demonstrates the capability of the DG to automatically achieve the required voltage and frequency within the specified time.
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| The DG auto-start and energization of permanently connected loads time of 13.20 seconds is derived from the 3.20 second Loss of Voltage - Time Delay Function Allowable Value (LCO 3.3.8.1) and the requirements of the accident analysis for responding to a design basis large break LOCA (Ref. 14).
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| The Surveillance should be continued for a minimum of (continued)
| |
| NMIP2 B 3.8.1-24 Revision 0, 35 (A138)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.9 (continued)
| |
| REQUIREMENTS 5 minutes in order to demonstrate that all starting transients have decayed and stability has been achieved.
| |
| The requirement to verify the connection and power supply of permanently connected loads and auto-connected loads (Division 1 and 2 only) is intended to satisfactorily show the relationship of these loads to the DG loading logic. In certain circumstances, many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation. For instance, ECCS injection valves are not desired to be stroked open, systems are not capable of being operated at full flow, or RHR systems performing a decay heat removal function are not desired to be realigned to the ECCS mode of operation. In lieu of actual demonstration of the connection and loading of these loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.
| |
| lnsert 3 t A...
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| 43....nd .... t
| |
| " ....... the intondodG to bohe ctn with cxpoctod fuel cyclo longths.
| |
| This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant (Division 1 and 2 DGs only) and lube oil being continuously circulated and temperature maintained consistent with manufacturer recommendations. The reason for Note 2 is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge plant safety systems (this portion of Note 2 is not applicable to the Division 3 DG).
| |
| Credit may be taken for unplanned events that satisfy this SR.
| |
| SR 3.8.1.10 Consistent with Regulatory Guide 1.9 (Ref. 11), paragraph C.2.2.5, this Surveillance demonstrates that the DG automatically starts and achieves the required voltage and frequency within the specified time (10 seconds) from the design basis actuation signal (LOCA signal). In addition, (continued)
| |
| NMP2 B 3.8.1-25 Revision 0* 2-33-.),
| |
| A! 35 (A138)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.10 (continued)
| |
| REQUIREMENTS the DG is required to maintain proper voltage and frequency limits after steady state is achieved. The voltage and frequency limits are normally achieved within 13 seconds for the Division 1 and 2 DGs and within 15 seconds for the Division 3 DG. The DG is required to operate for
| |
| _>5 minutes. The 5 minute period provides sufficient time to demonstrate stability. SR 3.8.1.10.d and SR 3.8.1.10.e ensure that permanently connected loads and emergency loads are energized from the offsite electrical power system on an ECCS signal without loss of offsite power (for Divisions 1 and 2 only).
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| The requirement to verify the connection and power supply of permanent and autoconnected loads is intended to satisfactorily show the relationship of these loads to the loading logic for loading onto offsite power. This is only required for Divisions 1 and 2 because the loading logic is different based on the power source. In certain circumstances, many of these loads cannot actually be
| |
| *connected or loaded without undue hardship or potential for undesired operation. For instance, ECCS injection valves are not desired to be stroked open, systems are not capable of being operated at full flow, or RHR systems performing a decay heat removal function are not desired to be realigned to the ECCS mode of operation. In lieu of actual demonstration of the connection and loading of these loads, testing that adequately shows the capability of the AC electrical power system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.
| |
| 33 The PFrqucncy cf 21 m onths takes into ooncidcration plant Insertlinset conditions required to pefform the Surveillance, and is inRtended to be consistent with the expected fuel cycle IegthG.
| |
| This SR is modified by two Notes. The reason for the Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant (Division 1 and 2 DGs only) and lube oil being continuously circulated and temperature maintained consistent with manufacturer recommendations. The reason for Note 2 is that during operation with the reactor critical, performance of (continued)
| |
| NMP2 B 3.8.1-26 Revision 9, 35 (A138)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.10 (continued)
| |
| REQUIREMENTS this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, plant safety systems (this portion of Note 2 is not applicable to the Division 3 DG). Credit may be taken for unplanned events that satisfy this SR.
| |
| SR 3.8.1.11 Consistent with Regulatory Guide 1.9 (Ref. 11), paragraph C.2.2.12, this Surveillance demonstrates that DG non-critical protective functions (e.g., high jacket water temperature) are bypassed on a loss of voltage signal concurrent with an ECCS initiation test signal and critical protective functions (engine overspeed and generator differential current) trip the DG to avert substantial damage to the DG unit. The non-critical trips are bypassed during DBAs and provide an alarm on an abnormal engine condition. This alarm provides the operator with sufficient time to react appropriately. The DG availability to mitigate the DBA is more critical than protecting the engine against minor problems that are not immediately detrimental to emergency operation of the DG.
| |
| ,/The 24 month Frcgucncy as based on cngincoring judgment, taking inocnioainplant co~nditionc roquirod to Insert 3 p---~peFoM the Su ,.illanc.,and is intended to be.. n.*tnt with oxpoctod fuel GYclIc cgh..
| |
| The SR is modified by a Note. The reason for the Note is that performing the Surveillance removes a required DG from service (this portion of the Note is not applicable to the Division 3 DG).
| |
| Credit may be taken for unplanned events that satisfy this SR.
| |
| SR 3.8.1.12 Consistent with Regulatory Guide 1.9 (Ref. 11), paragraph C.2.2.9, this Surveillance requires demonstration that the DGs can start and run continuously at full load capability for an interval of not less than 24 hours, 22 hours of which is at a load equivalent to 90% to 100% of the continuous rating of the DG and 2 hours of which is at a load equivalent to 105% to 110% of the continuous rating of the DG. The DG starts for this Surveillance can be performed (continued)
| |
| NMP2 B 3.8.1-27 Revision G, 35 (A138)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.12 (continued)
| |
| REQUIREMENTS either from standby or hot conditions. The provisions for prelube and warmup, discussed in SR 3.8.1.2, and for gradual loading, discussed in SR 3.8.1.3, are applicable to this SR.
| |
| In order to ensure that the DG is tested under load conditions that are as close to design conditions as possible, testing must be performed at a power factor as close to the accident load power factor as practicable. The power factor limit is < 0.91 for Division 1 and 2 DGs and
| |
| _ 0.93 for Division 3 DG. This power factor is representative of the actual design basis inductive loading that the DG could experience.
| |
| intended to be Gen~istent with oxpcctcd fuel cyclc IcngthG.
| |
| This Surveillance is modified by three Notes. Note 1 states that momentary transients due to changing bus loads do not invalidate this test. The load band is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY. Similarly, momentary power factor transients above the limit do not invalidate the test. The reason for Note 2 is that during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that would challenge continued steady state operation and, as a result, plant safety systems. However, it is acceptable to perform this SR in MODES 1 and 2 provided the other two DGs are OPERABLE, since a perturbation can only affect one divisional DG. If during performance of this Surveillance one of the other DGs becomes inoperable, this Surveillance is to be suspended.
| |
| Credit may be taken for unplanned events that satisfy this SR. Note 3 is provided in recognition that since the offsite electrical power transmission network is not balanced, it may not be possible to raise DG voltage sufficiently to meet the power factor limit without one phase of the DG exceeding the current limit. Therefore, to ensure the DG is not placed in an unsafe condition during this test, the power factor limit does not have to be met if the offsite grid phase imbalance does not permit the power factor limit to be met when the DG is tied to the grid.
| |
| (continued)
| |
| NMP2 B 3.8.1-28 Revision O,32-((A!33), 35 (Al 38)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.12 (continued)
| |
| REQUIREMENTS When this occurs, the power factor should be maintained as close to the limit as practicable.
| |
| SR 3.8.1.13 This Surveillance demonstrates that the diesel engine can restart from a hot condition, such as subsequent to shutdown from normal Surveillances, and achieve the required voltage and frequency within 10 seconds. The 10 second time is derived from the requirements of the accident analysis for responding to a design basis large break LOCA (Ref. 15). In addition, the DG is required to maintain proper voltage and frequency limits after steady state is achieved. The voltage and frequency limits are normally achieved within 13 seconds for the Division 1 and 2 DGs and within 15 seconds for the Division 3 DG.
| |
| The 24 monRth FroeqUoncY taker, into concsidor-ation the plant condi00tiAonzI rcquircd to pcrformA the Survcillanco, and is Insert 3 antn*ded to be con si.stont with cxpoct. d fuol
| |
| *cy, o l,ngthS.
| |
| This SR has been modified by two Notes. Note 1 ensures that the test is performed with the diesel sufficiently hot. The requirement that the diesel has operated for at least 2 hours at approximately full load conditions prior to performance of this Surveillance is based on manufacturer recommendations for achieving hot conditions. Momentary transients due to changing bus loads do not invalidate this test. Note 2 allows all DG starts to be preceded by an engine prelube period to minimize wear and tear on the diesel during testing.
| |
| SR 3.8.1.14 Consistent with Regulatory Guide 1.9 (Ref. 11),
| |
| paragraph C.2.2. 11, this Surveillance ensures that the manual synchronization and load transfer from the DG to the offsite source can be made and that the DG can be returned to ready-to-load status when offsite power is restored. It also ensures that the auto-start logic is reset to allow the DG to reload if a subsequent loss of offsite power occurs.
| |
| The DG is considered to be in ready-to-load status when the DG is at rated speed and voltage, the output breaker is open (continued)
| |
| NMP2 B 3.8.1-29 Revision 9- 2-(A!3-3), 35 (Al 38)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.14 (continued)
| |
| REQUIREMENTS and can receive an auto-close signal on bus undervoltage, and the individual load timers are reset.
| |
| -> 4uonc; of _24 monRths takes into consideration plant Froge conditieoc roquircd to pcr*FoRm the Su.'eilla,,.. and i inRtendcd to be consistont With oXpoctod fue IyI 4&.
| |
| This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems (this portion of the Note is not applicable to the Division 3 DG). Credit may be taken for unplanned events that satisfy this SR.
| |
| SR 3.8.1.15 Consistent with Regulatory Guide 1.9 (Ref. 11), paragraph C.2.2.13, demonstration of the parallel test mode override
| |
| [Insert 3 ensures that the DG availability under accident conditions is not compromised as the result of testing. Interlocks to the LOCA sensing circuits cause the DG to automatically reset to ready-to-load operation if an ECCS initiation signal is received during operation in the test mode.
| |
| Ready-to-load operation is defined as the DG running at rated speed and voltage with the DG output breaker open.
| |
| These provisions for automatic switchover are required by IEEE-308 (Ref. 17), paragraph 6.2.6(2).
| |
| The requirement to automatically energize the emergency loads with offsite power is essentially identical to that of SR 3.8.1.10. The intent in the requirement associated with SR 3.8.1.15.b is to show that the emergency loading is not affected by the DG operation in parallel test mode. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the emergency loads to perform these functions is acceptable.
| |
| This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.
| |
| The 24 month FrFequency takes into consideration plant RnnR1iffifiR" F8qbiiid te nrerFn the' Suiir"r'Ianeer and is i ntended to be consistent with e, p ected fuel cycale lengths.
| |
| (continued)
| |
| (continued)
| |
| NMP2 B 3.8.1-30 Revision 0, 32(A!.33), 35 (Al 38)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.15 (continued)
| |
| REQUIREMENTS This SR has been modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems (this portion of the Note is not applicable to the Division 3 DG).
| |
| Credit may be taken for unplanned events that satisfy this SR.
| |
| SR 3.8.1.16 Under accident conditions loads are sequentially connected to the bus by the automatic load sequence time delay relays.
| |
| The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading of the DGs due to high motor starting currents. The -10% load sequence time interval limit ensures that a sufficient time interval exists for the DG to restore frequency and voltage prior to applying the next load. There is no upper limit for the load sequence time interval since, for a single load interval (i.e., .the time between two load blocks), the capability of the DG to restore frequency and voltage prior to applying the second load is not negatively affected by a longer than designed load interval, and if there are additional load blocks (i.e., the design includes multiple load intervals), then the lower limit requirements (-10%)
| |
| will ensure that sufficient time exists for the DG to restore frequency and voltage prior to applying the remaining load blocks (i.e., all load intervals must be
| |
| > 90% of the design interval). Reference 2 provides a summary of the automatic loading of emergency buses. Since only the Division 1 and 2 DGs have more than one load block, this SR is only applicable to these DGs.
| |
| EInsertEEý This SR is modified by a Note. The reason for the Note is that performing the Surveillance during these MODES would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge plant safety systems. Credit may be taken for unplanned events that satisfy this SR.
| |
| (continued)
| |
| NMP2 B 3.8.1-31 Revision 0,-28-(A!-29), 35 (A138)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.17 REQUIREMENTS (continued) In the event of a DBA coincident with a loss of offsite power, the DGs are required to supply the necessary power to ESF systems so that the fuel, RCS, and containment design limits are not exceeded.
| |
| This Surveillance demonstrates the DG operation, as discussed in the Bases for SR 3.8.1.9, during a loss of offsite power actuation test signal in conjunction with an ECCS initiation signal. Since the Loss of Voltage - Time Delay Functions are bypassed during an ECCS initiation signal, a 10 second DG start time applies. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.
| |
| The FroquoncY of 24 moneths takes into considcration plant cOndRition rcquircd to pcrforFm the Survcillancce, and is Insert 3 - in*Wn*dd to bo*....t-.t with an oxpoctod fuol cy.l.
| |
| This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant (Division 1 and 2 DGs only) and lube oil being continuously circulated and temperature maintained consistent with manufacturer recommendations. The reason for Note 2 is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge plant safety systems (this portion of Note 2 is not applicable to the Division 3 DG). Credit may be taken for unplanned events that satisfy this SR.
| |
| SR 3.8.1.18 This Surveillance demonstrates that the DG starting independence has not been compromised. Also, this Surveillance demonstrates that each engine can achieve proper frequency and voltage within the specified time when the DGs are started simultaneously.
| |
| (continued)
| |
| NMP2 B 3.8.1-32 Revision 35 (Al 38)
| |
| | |
| AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.18 (continued)
| |
| REQUIREMENTS jInsert 3 - ý i ogulate.-y Guid,"- o 4 9(Re- )D,*p...a .... h G-,.2-.
| |
| * This SR is modified by a Note. The reason for the Note is to minimize wear on the DG during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
| |
| REFERENCES 1. 10 CFR 50, Appendix A, GDC 17.
| |
| : 2. USAR, Chapter 8.
| |
| : 3. USAR, Tables 8.3-1, 8.3-2, and 8.3-3.
| |
| : 4. Regulatory Guide 1.9, Revision 2, December 1979.
| |
| : 5. USAR, Chapter 6.
| |
| : 6. USAR, Chapter 15.
| |
| : 7. 10 CFR 50.36(c)(2)(ii).
| |
| : 8. Regulatory Guide 1.93, Revision 0, December 1974.
| |
| : 9. Generic Letter 84-15, July 2, 1984.
| |
| : 10. 10 CFR 50, Appendix A, GDC 18.
| |
| : 11. Regulatory Guide 1.9, Revision 3, July 1993.
| |
| : 12. Regulatory Guide 1.108, Revision 1, August 1977.
| |
| : 13. Regulatory Guide 1.137, Revision 1, October 1979.
| |
| : 14. ANSI C84.1, 1982.
| |
| [Deleted 15. USAR, Section 15.6.5.
| |
| : 17. IEEE Standard 308-1980.
| |
| NMP2 B 3.8.1-33 Revision 35 (Al 38)
| |
| | |
| Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES ACTIONS D.1 (continued) or to restore the stored fuel oil properties. This restoration may involve feed and bleed procedures, filtering, or a combination of these procedures. Even if a DG start and load was required during this time interval and the fuel oil properties were outside limits, there is high likelihood that the DG would still be capable of performing its intended function.
| |
| E.1 With any starting air receiver pressure < 225 psig for a Division 1 or 2 DG or < 190 psig for the Division 3 DG, sufficient capacity for five successive DG starts does not exist. However, as long as the receiver pressure in both receivers is >_175 psig for a Division 1 or 2 DG and
| |
| > 110 psig for the Division 3 DG, there is adequate capacity for at least one start, and the DG can be considered OPERABLE while the air receiver pressure is restored to the required limit. A period of 48 hours is considered sufficient to complete restoration to the required pressure prior to declaring the DG inoperable. This period is acceptable based on the remaining air start capacity, the fact that most DG starts are accomplished on the first attempt, and the low probability of an event during this brief period.
| |
| F. 1 With a Required Action and associated Completion Time of Condition A, B, C, D, or E not met, or the stored diesel fuel oil, lube oil, or starting air subsystem not within limits for reasons other than addressed by Conditions A through E, the associated DG may be incapable of performing its intended function and must be immediately declared inoperable.
| |
| SURVEILLANCE SR 3.8.3.1 REQUIREMENTS This SR provides verification that there is an adequate inventory of fuel oil in the storage tanks to support each DG's operation for 7 days at full load. The 7 day period is (continued)
| |
| NMP2 B 3.8.3-5 Revision 0
| |
| | |
| Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES SURVEILLANCE SR 3.8.3.1 (continued)
| |
| REQUIREMENTS sufficient time to place the unit in a safe shutdown condition and to bring in replenishment fuel from an offsite location.
| |
| The 31 day Fc.qucn.y is adcquate to cn.... that a uff,*eicnt supply of fueloIl s available, ,inc,Iow leovl alarms are providod and unit epcraterc would hc awar-o of any largo uco of fuel oil dur~ing thiG period.
| |
| SR 3.8.3.2 This Surveillance ensures that sufficient lube oil inventory (above the manufacturers minimum recommended level) is Insert 3 available to support at least 7 days of full load operation for each DG. The 99 gallon requirement for the Division 1 and 2 DGs and the 168 gallon requirement for the Division 3 DG are based on the DG manufacturer's consumption values for the run time of the DG. The 7 day inventory can be in the engine oil sump or a combination of the engine oil sump and remote storage location. Implicit in this SR is the requirement to verify the capability to transfer the lube oil from its storage location to the DG when the DG lube oil sumps do not hold adequate inventory for 7 days of full load operation without the level reaching the manufacturer's recommended minimum level.
| |
| ....31yFqucy is adequato to- ncRsure that a suA'"Ont lub. o1 supply .n.it.,
| |
| is smog DG starts and FRu times arc closely onWItorcd by the plant staff, SR 3.8.3.3 The tests of new fuel oil prior to addition to the storage tanks are a means of determining whether new fuel oil is of the appropriate grade and has not been contaminated with substances that would have an immediate detrimental impact on diesel engine combustion and operation. If results from these tests are within acceptable limits, the fuel oil may be added to the storage tanks without concern for contaminating the entire volume of fuel oil in the storage tanks. These tests are to be conducted prior to adding the new fuel to the storage tank(s).
| |
| (continued)
| |
| NMP2 B 3.8.3-6 Revision 0
| |
| | |
| Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES SURVEILLANCE SR 3.8.3.3 (continued)
| |
| REQUIREMENTS not burn properly in a diesel engine. However, the particulate can cause fouling of filters and fuel oil injection equipment, which can cause engine failure.
| |
| Particulate concentrations should be determined in accordance with ASTM D2276-78, Method A (Ref. 7). This method involves a gravimetric determination of total particulate concentration in the fuel oil and has a limit of 10 mg/l. It is acceptable to obtain a field sample for subsequent laboratory testing in lieu of field testing.
| |
| The Frequency of this Surveillance takes into consideration fuel oil degradation trends indicating that particulate concentration is unlikely to change between Frequency intervals.
| |
| SR 3.8.3.4 This Surveillance ensures that, without the aid of the refill compressor, sufficient air start capacity for each DG is available. The system design requirements provide for a minimum of five engine starts without recharging. The pressure specified in this SR is intended to support the lowest value in both receivers at which the five starts can be accomplished.
| |
| The 31 day FroeqUoncy takes into account tho capacity, capailit, v , :
| |
| rdundancy,
| |
| .:*. Hp p " I and dmyeorciy 4,,,.
| |
| 41 of* the ,
| |
| AC :vv soUrccsi,, :~.
| |
| and Insert 3 t2t Wr " U0 ""-" LAf.W% R MR R...FE FROM... , *%fflAV.I
| |
| ........ t ..e . ... ... e ih ...... .. . ..... ......... ... .....
| |
| PFessHFe.
| |
| SR 3.8.3.5 Periodic removal Microbiological fouling is a majo cause of fuel oil degradation. There are numero s bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. of water from the storage tanks oncc ovory 31 days eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment (continued)
| |
| NMP2 B 3.8.3-8 Revision 0
| |
| | |
| Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES SURVEILLANCE SR 3.8.3.5 (continued)
| |
| REQUIREMENTS in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and from breakdown of the fuel oil by bacteria. Frequent checking for and
| |
| [Insert ý3 }-----.-. removal of accumulated ta regarding waterintegrity the watertight minimizes fouling of the and provides fuel oil sy-h SuTk-cwillancc Frcequency is established by Regulator; Guide 1.137 (Ref. 2). This SR is for preyentiye m~aintnAncoR~. The proconce of wator dooc not nococcarily rcprcsent a failUrc of this SR provided that accumulated watcr is removed during pecforma%,e o~f the Suryoillanco.
| |
| REFERENCES 1. USAR, Section 9.5.4.
| |
| : 2. Regulatory Guide 1.137, Revision 1, October 1979.
| |
| : 3. ANSI N195, Appendix B, 1976.
| |
| : 4. USAR, Chapter 6.
| |
| : 5. USAR, Chapter 15 and Appendix A.
| |
| : 6. 10 CFR 50.36(c)(2)(ii).
| |
| : 7. ASTM Standards: D4057-81; D975-81; D4176-82; D1552-79; D2622-82; D2276-78.
| |
| NMIP2 B 3.8.3-9 Revision 0
| |
| | |
| DC Sources - Operating B 3.8.4 BASES ACTIONS C.1 and C.2 (continued) conditions from full power conditions in an orderly manner and without challenging plant systems. The Completion Time to bring the unit to MODE 4 is consistent with the time specified in Regulatory Guide 1.93 (Ref. 8).
| |
| SURVEILLANCE SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a fully charged state. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the Insert 3 battery Si7innp crc.*Ir~i tinnQ Tho 7 day Froquoncy is GeRse ~R fl.J* et .%~eR 11md tR- RH EAU rR FS Feeemmenuatiens nad IFF 460u tRef. to).
| |
| SR 3.8.4.2 Visual inspection to detect corrosion of the battery cells and connections, or measurement of the resistance of each intercell and terminal connection, provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance.
| |
| The connection resistance limits are < 20% above the resistance as measured during installation (Ref. 9).
| |
| linsert 3 Fc onyf rt.h i + h cn A 4- J: ;
| |
| , - hai
| |
| ,itancc h n,.i,.
| |
| _1I I_ I _
| |
| 92 days. ......This Fr-oEiu. . , i-I.... J
| |
| ---" ..... I J concioorool aecoplaoio Bacco; on oporalfng oXpoRionc roiel to- doPtocGting corrocion tProde1.
| |
| SR 3.8.4.3 Visual inspection of the battery cells, cell plates, and battery racks provides an indication of physical damage or abnormal deterioration that could potentially degrade (continued)
| |
| NMP2 B 3.8.4-5 Revision 0
| |
| | |
| DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.3 (continued)
| |
| REQUIREMENTS battery performance. The presence of physical damage or deterioration does not necessarily represent a failure of this SR, provided an evaluation determines that the physical damage or deterioration does not affect the OPERABILITY of the battery (its ability to perform its design function).
| |
| ýým i- ii^m^ ^r 4- ^ aimi^mým- in n-" ^-
| |
| i engieerig judgoment. Opcralting cx(pcricncc has shown tha th Anomoet usuially pass the SR when porfo9FRmd at h 24 mon~th Frcequcncy. Thferforc, the Frcequcncy was concludcd to be .;Accoptablo fromn a reliability 6tandpoint.
| |
| SR 3.8.4.4 and SR 3.8.4.5 Visual inspection and resistance measurements of intercell and terminal connections provides an indication of physical lIiset damage or abnormal deterioration that could indicate degraded battery condition. The anti-corrosion material is used to ensure good electrical connections and to reduce terminal deterioration. The visual inspection for corrosion is not intended to require removal of and inspection under each terminal connection.
| |
| The removal of visible corrosion is a preventive maintenance SR. The presence of visible corrosion does not necessarily represent a failure of this SR, provided visible corrosion is removed during performance of this Surveillance.
| |
| The connection resistance limits are < 20% above the resistance as measured during installation (Ref. 9).
| |
| -The 24 month Frcequcncy for the Survcillanco is bascc eR ongieeoring judgment. Operating expcrionce has shown at these componcnts usually pass the SR whcni pcrformcd at h 24 mon~th Frcquoncy. Therefore, the PFequenc'; waS concludc to be acccptablc fromR a reliability standpoint.
| |
| SR 3.8.4.6 Battery charger capability requirements are based on the design capacity of the chargers (Ref. 4). According to Regulatory Guide 1.32 (Ref. 10), the battery charger supply (continued)
| |
| NMP2 B 3.8.4-6 Revision 0
| |
| | |
| DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.6 (continued)
| |
| REQUIREMENTS is required to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and duration ensure that these requirements can be satisfied.
| |
| r~*To Su~oilanc Froucny icaccptablo gk'on the adminiztrativc cOntroks oxicting to cnSUrc adequate charger porformnanco duFrin thtqco 24 month itra.InAddition, thic Froauon i inodod to be concictent with ecxooctod fuel cyclc lengths.
| |
| Insert 3 SR 3.8.4.7 A battery service test is a special test of the battery's capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length correspond to the design duty cycle requirements as specified in Reference 4.
| |
| - - Thc Su...oillanco. Froqucnc;y of 24 mRe*th- is a.ccptablo, g uit condition, r to p.rfor. the test
| |
| .quir.d . and the.thc.
| |
| roquiRontc oxfisting to cncUro adoquato battory porformancc.
| |
| durin thse 24 moneth entervals. In addition, thi Froquoncy i-" into-dod to be e"oni.t.. ""t with poco*rd fuol This SR is modified by two Notes. Note 1 allows the performance of a modified performance discharge test in lieu of a service test provided the modified performance discharge test completely envelops the service test. This substitution is acceptable because a modified performance discharge test represents a more severe test of battery capacity than SR 3.8.4.7. The reason for Note 2 is that performing the Surveillance would remove a required DC electrical power subsystem from service, perturb the electrical distribution system, and challenge safety systems. The Division 3 test may be performed in MODE 1, 2, or 3 in conjunction with HPCS system outages.
| |
| Credit may be taken for unplanned events that satisfy the Surveillance.
| |
| (continued)
| |
| NMP2 B 3.8.4-7 Revision 0, 34 (Al 36)
| |
| | |
| DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.8 (continued) in the battery size calculation. A capacity of 80% shows that the battery is getting old and capacity will decrease more rapidly, even if there is ample capacity to meet the load requirements.
| |
| Ilne.crt o, I L,,
| |
| A AR--
| |
| ::1--P9P ... ý- "- X^gr g m--11 60 teFths-,If the battery shows degradation, or if the battery has reached 85% of its expected life and capacity is
| |
| < 100% of the manufacturer's rating, the Surveillance Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity > 100% of the manufacturer's rating. Degradation is indicated, consistent with IEEE-450 (Ref. 9), when the battery capacity drops by
| |
| , more than 10% of rated capacity in the previous 72 months or Ifrequency is it is below 90% of the manufacturer's rating. The 12mon ,t , consistent with the recommendations in IEEE-450 (Ref. 9). The 24 month Frequency is derived from the recommendations of IEEE-450 (Ref. 9).
| |
| This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DC electrical power subsystem from service, perturb the electrical distribution system, and challenge safety systems. The Division 3 test may be performed in MODE 1, 2, or 3 in conjunction with HPCS system outages.
| |
| Credit may be taken for unplanned events that satisfy the Surveillance.
| |
| REFERENCES 1. 10 CFR 50, Appendix A, GDC 17.
| |
| : 2. Regulatory Guide 1.6, Revision 0, March 10, 1971.
| |
| : 3. IEEE Standard 308, 1974.
| |
| : 4. USAR, Section 8.3.2.
| |
| : 5. USAR, Chapter 6.
| |
| : 6. USAR, Chapter 15 and Appendix A.
| |
| : 7. 10 CFR 50.36(c)(2)(ii).
| |
| (continued)
| |
| NMP2 B 3.8.4-9 Revision 0, 34 (Al 36)
| |
| | |
| Battery Cell Parameters B 3.8.6 BASES ACTIONS A.1, A.2, and A.3 (continued) 7 day intervals until the parameters are restored to Category A and B limits. This periodic verification is consistent with the normal Frequency of pilot cell Surveillances.
| |
| Continued operation is only permitted for 31 days before battery cell parameters must be restored to within Category A and B limits. Taking into consideration that while battery capacity is degraded, sufficient capacity exists to perform the intended function and to allow time to fully restore the battery cell parameters to normal limits, this time is acceptable for operation prior to declaring the associated DC batteries inoperable.
| |
| B.1 When any battery parameter is outside the Table 3.8.6-1 Category C limit for any connected cell, sufficient capacity to supply the maximum expected load requirement is not assured and the corresponding DC electrical power subsystem must be declared inoperable. Additionally, other potentially extreme conditions, such as any Required Action of Condition A and associated Completion Time not met or average electrolyte temperature of representative cells
| |
| < 65 0 F, also are cause for immediately declaring the associated DC electrical power subsystem inoperable.
| |
| SURVEILLANCE SR 3.8.6.1 REQUIREMENTS The SR verifies that Table 3.8.6-1 Category A battery cell parameters are consistent with IEEE-450 (Ref. 4), which recommends regular battery inspections (atleast ,ne.peF FeRth-) including voltage, specific gravity, and electrolyte level of pilot cells.
| |
| SR 3.8.6.2 lilnsert 3 ** *h.. +-"...,*.. ... f ... ; .. ,...,
| |
| o1~locroet level for eaeh eepnncctd Gel! 66 cozistont with IFE=E 1*5"0- (Ref. )., In addition, within 7 days of a battery discharge < 107 V or a battery overcharge > 142 V, the battery must be demonstrated to meet Table 3.8.6-1 (continued)
| |
| NMP2 B 3.8.6-3 Revision 0
| |
| | |
| Battery Cell Parameters B 3.8.6 BASES SURVEILLANCE SR 3.8.6.2 (continued)
| |
| REQUIREMENTS Category B limits. Transients, such as motor starting transients, which may momentarily cause battery voltage to drop to < 107 V, do not constitute a battery discharge provided the battery terminal voltage and float current return to pre-transient values. This inspection is also consistent with IEEE-450 (Ref. 4), which recommends special inspections following a severe discharge or overcharge, to ensure that no significant degradation of the battery occurs as a consequence of such discharge or overcharge. The 7 day requirement is based on engineering judgement.
| |
| SR 3.8.6.3 This Surveillance verification that the average temperature of representative cells is >_651F is consistent with a recommendation of IEEE-450 (Ref. 4), which statc, that the tmp r-at*r, of ,-Iet l,4-, . in rp-,r,,ntatio clis heould bo doetFormnod on a baci;. -uro, For this SR, a check of 20% of the connected cells is considered representative.
| |
| Lower than normal temperatures act to inhibit or reduce battery capacity. This SR ensures that the operating temperatures remain within an acceptable operating range.
| |
| This limit is based on manufacturers recommendations and the battery sizing calculations.
| |
| Table 3.8.6-1 This Table delineates the limits on electrolyte level, float voltage, and specific gravity for three different categories. The meaning of each category is discussed below.
| |
| Category A defines the normal parameter limit for each designated pilot cell in each battery. The cells selected as pilot cells are those whose temperature, voltage, and electrolyte specific gravity approximate the state of charge of the entire battery.
| |
| The Category A limits specified for electrolyte level are based on manufacturers recommendations and are consistent with the guidance in IEEE-450 (Ref. 4), with the extra 1/4 inch allowance above the high water level indication for (continued)
| |
| NMP2 B 3.8.6-4 Revision 0
| |
| | |
| Inverters - Operating B 3.8.7 BASES ACTIONS B.1 and B.2 (continued)
| |
| If a required emergency UPS inverter cannot be restored to OPERABLE status within the associated Completion Time (by either restoring the inoperable emergency UPS inverter to OPERABLE status or, alternatively, by placing the standby divisional emergency UPS inverter into service), 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 12 hours and to MODE 4 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.
| |
| SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This Surveillance verifies that the required emergency UPS inverters are functioning properly with all required circuit breakers closed and 120 VAC uninterruptible panels energized from the emergency UPS inverter. The verification of proper voltage and frequency output ensures that the required power is readily available for the instrumentation connected to the 120 VAC uninterruptible panels. ý-h,, Edt- ...... I Insert 3 i Qthior Onclon-Atonn'- 'v'ii4ihbr' n thim anntrnl rniqm thit Alp;*t I g a gf dl the opoFator to iinvorter il alt'ulVnDI .
| |
| REFERENCES 1. USAR, Section 8.3.1.1.2.
| |
| : 2. USAR, Chapter 6.
| |
| : 3. USAR, Chapter 15 and Appendix A.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.8.7-4 Revision 0, 4-5, 16
| |
| | |
| Distribution Systems - Operating B 3.8.8 BASES SURVEILLANCE SR 3.8.8.1 (continued)
| |
| REQUIREMENTS appropriate separation and independence of the electrical divisions is maintained, and power is available to each required bus. The verification of energization of the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. This is normally performed by verifying correct voltage for the AC and DC switchgear and by verifying that no inoperability status indicator lights (that indicate a loss of power to one or more of the required load centers, motor control centers (MCCs), or distribution panels) are lit in the control room.
| |
| Alternately, when the normal method is not available, verification that a load powered from the associated bus is Inser 3 energized is also acceptable. t' 7 .... F....-,ae3 into account the rcdundant capability of the AC, DG, and12 VAC unintrruptiblc clc*triGal pew or ditributien subsy-*tems, and othchr indiatiens available in the control roomF that alcrt the opcrator to subsystem malfunctionz.
| |
| REFERENCES 1. USAR, Chapter 6.
| |
| : 2. USAR, Chapter 15 and Appendix A.
| |
| : 3. Regulatory Guide 1.93, Revision 0, December 1974.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| NMIP2 B 3.8.8-11 Revision 0
| |
| | |
| Distribution Systems - Shutdown B 3.8.9 BASES ACTIONS A.1, A.2.1, A.2.2, A.2.3, A.2.4, and A.2.5 (continued)
| |
| Notwithstanding performance of the above conservative Required Actions, a required residual heat removal - shutdown cooling (RHR-SDC) subsystem may be inoperable. In this case, Required Actions A.2.1 through A.2.4 do not adequately address the concerns relating to coolant circulation and heat removal. Pursuant to LCO 3.0.6, the RHR-SDC ACTIONS would not be entered. Therefore, Required Action A.2.5 is provided to direct declaring RHR-SDC inoperable, which results in taking the appropriate RHR-SDC ACTIONS.
| |
| The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required distribution subsystems should be completed as quickly as possible in order to minimize the time the plant safety systems may be without power.
| |
| SURVEILLANCE SR 3.8.9.1 REQUIREMENTS This Surveillance verifies that the AC, DC, and 120 VAC uninterruptible electrical power distribution subsystems are functioning properly, with the correct breaker alignment.
| |
| The correct breaker alignment ensures power is available to each required bus. The verification of energization of the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. This is normally performed by verifying correct voltage for the AC and DC switchgear and by verifying that no inoperability status indicator lights (that indicate a loss of power to one or more of the required load centers, MCCs, or distribution panels) are lit in the control room. Alternately, when the normal method is not available, verification that a load powered from the associated bus is energized is also acceptable.1, ,n'--
| |
| Frogquncy takes into account the rcdunda. . c i of tc.......
| |
| clcctrical poword6646b1949A c Ubcystom , as well aG othcr indicati-nc a.ailablo iFnthe control rOOm that aIcrt the oporator to ,ubyGtc,*n mnaIfURntion.-
| |
| REFERENCES 1. USAR, Chapter 6.
| |
| : 2. USAR, Chapter 15 and Appendix A.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.8.9-4 Revision 0
| |
| | |
| Refueling Equipment Interlocks B 3.9.1 BASES ACTIONS A.1, A.2.1, and A.2.2 (continued) that would potentially not be blocked from unacceptable operations (e.g., loading fuel into a cell with a control rod withdrawn). Suspension of in-vessel fuel movement shall not preclude completion of movement of a component to a safe position. Alternately, Required Actions A.2.1 and A.2.2 require that a control rod withdrawal block be inserted and that all control rods are subsequently verified to be fully inserted in core cells containing one or more fuel assemblies. Required Action A.2.1 ensures that no control rods can be withdrawn. This action ensures that control rods cannot be inappropriately withdrawn since an electrical or hydraulic block to control rod withdrawal is in place.
| |
| Required Action A.2.2 is normally performed after placing the rod withdrawal block in effect and provides a verification that all control rods in core cells containing one or more fuel assemblies are fully inserted. Like Required Action A.1, Required Actions A.2.1 and A.2.2 ensure that unacceptable operations are blocked (e.g., loading fuel into a core cell with the control rod withdrawn).
| |
| SURVEILLANCE SR 3.9.1.1 REQUIREMENTS Performance of a CHANNEL FUNCTIONAL TEST demonstrates each required refueling equipment interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested.
| |
| Insert3 >The day ,7 *,* G ,*eu ba e en, e,,g,,,e .... ju ,,,m ,,t a,, ,
| |
| consiidcrc-pd adequate Yin iw of othcr indications of rcfucling intcrlock and thcir aGsociatcd input statuG that aro available to unit oporations porconnol.
| |
| REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.
| |
| : 2. USAR, Section 7.1.4.
| |
| : 3. USAR, Section 15.4.1.1.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.9.1-4 Revision 0
| |
| | |
| Refuel Position One-Rod-Out Interlock B 3.9.2 BASES ACTIONS A.1 and A.2 (continued) control rods are fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and, therefore, do not have to be inserted.
| |
| SURVEILLANCE SR 3.9.2.1 REQUIREMENTS Proper functioning of the refueling position one-rod-out interlock requires the reactor mode switch to be in Refuel.
| |
| During control rod withdrawal in MODE 5, improper positioning of the reactor mode switch could, in some instances, allow improper bypassing of required interlocks.
| |
| Therefore, this Surveillance imposes an additional level of assurance that the refueling position one-rod-out interlock will be OPERABLE when required. By "locking" the reactor mode switch in the proper position (i.e., removing the reactor mode switch key from the console while the reactor mode switch is positioned in refuel), an additional administrative control is in place to preclude operator errors from resulting in unanalyzed operation.
| |
| linsert D3---_ý_ adminictrFatlvc controlz utilizod during rofuoeing .*, .. ati9 to cncuro cafc oporatin.
| |
| SR 3.9.2.2 Performance of a CHANNEL FUNCTIONAL TEST on each channel demonstrates the associated refuel position one-rod-out interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested. _L Insert 3 Froqucncy is. .nider.dadequate b. of d,... t,,t I circuit reliability, preodu*rl GRt'ol-s* control rod withdvawalu, a.nd nzual and audible *ndfiations ayailabli the control room to alcrt the epcFatOF of coAnFto rFod nt*
| |
| f*U',i,", ^,e" To perform the required testing, the applicable condition must be entered (i.e., a control rod must be withdrawn from its full-in position). Therefore, SR 3.9.2.2 has been modified by a Note that states the CHANNEL FUNCTIONAL TEST is not required to be performed until 1 hour after any control rod is withdrawn.
| |
| (continued)
| |
| NMP2 B 3.9.2-3 Revision 0
| |
| | |
| Control Rod Position B 3.9.3 BASES (continued)
| |
| LCO All control rods must be fully inserted during applicable refueling conditions to minimize the probability of an inadvertent criticality during refueling.
| |
| APPLICABILITY During MODE 5, loading fuel into core cells with control rods withdrawn may result in inadvertent criticality.
| |
| Therefore, the control rods must be inserted before loading fuel into a core cell. All control rods must be inserted before loading fuel to ensure that a fuel loading error does not result in loading fuel into a core cell with the control rod withdrawn.
| |
| In MODES 1, 2, 3, and 4, the reactor pressure vessel head is on, and no fuel loading activities are possible. Therefore, this Specification is not applicable in these MODES.
| |
| ACTIONS A.1 With all control rods not fully inserted during the applicable conditions, an inadvertent criticality could occur that is not analyzed in the USAR. All fuel loading operations must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position.
| |
| SURVEILLANCE SR 3.9.3.1 REQUIREMENTS During refueling, to ensure that the reactor remains subcritical, all control rods must be fully inserted prior to and during fuel loading. Periodic checks of the control rod position ensure this condition is maintained.
| |
| -Thc 12 hourF F,..uen.y takes into idc.ation the jlnsertý ýý procedural controlsG on control rod MovomonRIPt during rofueling as well as the redundant functions of the rcfucling interloks REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.
| |
| : 2. USAR, Section 15.4.1.1.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.9.3-2 Revision 0
| |
| | |
| Control Rod OPERABILITY - Refueling B 3.9.5 BASES SURVEILLANCE SR 3.9.5.1 and SR 3.9.5.2 (continued)
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| REQUIREMENTS concidoration eguipmonet
| |
| [Insert 3 .t-/tatcumulatorlo,acd cumtuol tro alaFRmc and indicating lights iargo proccuroc.
| |
| SR 3.9.5.1 is modified by a Note that allows 7 days after withdrawal of the control rod to perform the Surveillance.
| |
| This acknowledges that the control rod must first be withdrawn before performance of the Surveillance, and therefore avoids potential conflicts with SR 3.0.1.
| |
| REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.
| |
| : 2. USAR, Section 15.4.1.1.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.9.5-3 Revision 0
| |
| | |
| RPV Water Level - Irradiated Fuel B 3.9.6 BASES (continued)
| |
| SURVEILLANCE SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level of 22 ft 3 inches above the top of the RPV flange ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to result from a fuel handling accident in containment (Ref. 2).
| |
| no i-Fre-1uonrv or OA koure ls se eno n onoinoReeno jiuagment f*AI
| |
| * IIn s e rt and is considercd adequate in viow of the large volume ot w inu-iR tnc nRE)mai 19rocuuurui ARuTruIR An 49 We puiiuris, whoeh-mal - -' I" IIIl*Ul I"!IL unplanned Icvl -hange, unlikely.
| |
| REFERENCES 1. Regulatory Guide 1.183, July 2000.
| |
| : 2. USAR, Section 15.7.4.
| |
| : 3. NUREG-0800, Section 15.0.1.
| |
| : 4. USAR, Table 15.7-9.
| |
| : 5. 10 CFR 50.67, "Accident Source Term."
| |
| : 6. 10 CFR 50.36(c)(2)(ii).
| |
| : 7. Regulatory Issue Summary (RIS) 2006-04, March 7, 2006.
| |
| NMIP2 B 3.9.6-3 Revision 0, 26 (A125)
| |
| | |
| RPV Water Level - New Fuel or Control Rods B 3.9.7 BASES SURVE ILLANCE SR 3.9.7.1 (continued)
| |
| REQUIFREMENTS operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to result from a fuel handling accident in containment (Ref. 2).
| |
| Tlný mr*,,I *nl ,,n "O , h*,nj c* o6AI I Mnni n* : Mrinn i,,l ~emn water and onormdal th prccdural 688491w ctR onryvlucpitinc,
| |
| ,.,hic.h mako cQianific4At .. n.lanned !.. el hane.. unlikely.
| |
| REFERENCES 1. Regulatory Guide 1.183, July 2000.
| |
| : 2. USAR, Section 15.7.4.
| |
| : 3. NUREG-0800, Section 15.0.1.
| |
| : 4. USAR, Table 15.7-9.
| |
| : 5. 10 CFR 50.67, "Accident Source Term."
| |
| : 6. 10 CFR 50.36(c)(2)(ii).
| |
| : 7. Regulatory Issue Summary (RIS) 2006-04, March 7, 2006.
| |
| NMIP2 B 3.9.7-3 Revision 0, 26 (A125)
| |
| | |
| RHR - High Water Level B 3.9.8 BASES ACTIONS B.1, B.2, B.3, and B.4 (continued) must be restored to OPERABLE status. In this case, a surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.
| |
| C.1 and C.2 If no RHR shutdown cooling subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour. The Completion Time is modified such that 1 hour is applicable separately for each occurrence involving a loss of coolant circulation.
| |
| During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.
| |
| SURVEILLANCE SR 3.9.8.1 REQUIREMENTS This Surveillance demonstrates that the required RHR shutdown cooling subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat LInsert 3 remo Thc F-..qu.n.y of 12 hours is SUffi.ic'"
| |
| inR view Of othor visu al and audiblo inRdications available toe the opcrato-r frmntrn the RHR shutdown cooling subsystem in the coAntrol roo-m.
| |
| REFERENCES 1. 10 CFR 50, Appendix A, GDC 34.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.9.8-4 Revision 0
| |
| | |
| RHR - Low Water Level B 3.9.9 BASES ACTIONS B.1, B.2, and B.3 (continued) need for secondary containment isolation is indicated).
| |
| This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components.
| |
| If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, a surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.
| |
| C.1 and C.2 If no RHR shutdown cooling subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour. The Completion Time is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation.
| |
| During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem), the reactor coolant temperature must be periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate.
| |
| SURVEILLANCE SR 3.9.9.1 REQUIREMENTS This Surveillance demonstrates that one RHR shutdown cooling subsystem is in operation and circulating reactor coolant.
| |
| The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal 3 caucnc; Fqnsert of 1' hourc'i L~efofothcrVku~al and audible indicationc availablo to the epcratOr for mRonitorig the RHR shutdewn cooling subsystc in tho control roomR.
| |
| REFERENCES 1. 10 CFR 50, Appendix A, GOC 34.
| |
| : 2. 10 CER 50.36(c)(2)(ii).
| |
| NMP2 B 3.9.9-4 Revision 0
| |
| | |
| Reactor Mode Switch Interlock Testing B 3.10.2 BASES ACTIONS A.1, A.2, A.3.1, and A.3.2 (continued)
| |
| All CORE ALTERATIONS, except control rod insertion, if in progress, are immediately suspended in accordance with Required Action A.1, and all insertable control rods in core cells that contain one or more fuel assemblies are fully inserted within 1 hour, in accordance with Required Action A.2. This will preclude potential mechanisms that could lead to criticality. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and, therefore, do not have to be inserted.
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| Suspension of CORE ALTERATIONS shall not preclude the completion of movement of a component to a safe condition.
| |
| Placing the reactor mode switch in the shutdown position will ensure that all inserted control rods remain inserted and result in operation in accordance with Table 1.1-1.
| |
| Alternatively, if in MODE 5, the reactor mode switch may be placed in the refuel position, which will also result in operating in accordance with Table 1.1-1. A Note is added to Required Action A.3.2 to indicate that this Required Action is not applicable in MODES 3 and 4, since only the shutdown position is allowed in these MODES. The allowed Completion Time of 1 hour for Required Actions A.2, A.3.1, and A.3.2 provides sufficient time to normally insert the control rods and place the reactor mode switch in the required position, based on operating experience, and is acceptable given that all operations that could increase core reactivity have been suspended.
| |
| SURVEILLANCE SR 3.10.2.1 and SR 3.10.2.2 REQUIREMENTS Meeting the requirements of this Special Operations LCO maintains operation consistent with or conservative to operating with the reactor mode switch in the shutdown position (or the refuel position for MODE 5). The functions of the reactor mode switch interlocks that are not in effect, due to the testing in progress, are adequately compensated for by the Special Operations LCO requirements.
| |
| The administrative controls are to be periodically verified to ensure that the operational requirements continue to be met. In addition, the all rods fully inserted Surveillance (SR 3.10.2.1) must be verified by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff (e.g., a qualified (continued)
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| NMP2 B 3.10.2-4 Revision 0
| |
| | |
| Reactor Mode Switch Interlock Testing B 3.10.2 BASES SURVEILLANCE SR 3.10.2.1 and SR 3.10.2.2 (continued)
| |
| REQUIREMENTS shift technical advisor or reactor engineer). I1 I Su.vcllanc.s pcrfor...d at the 12 hour. nd hu Insert 3 Frcquoncics arc intended to pr..id, app....iat. a.suranc.
| |
| that cach opcrating shift is awano of and vonif; compliancc withthec Speclal OperFationsr LCGO roguiromonets.
| |
| REFERENCES 1. USAR, Section 7.2.
| |
| : 2. USAR, Section 15.4.1.1.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.10.2-5 Revision 0
| |
| | |
| Single Control Rod Withdrawal - Hot Shutdown B 3.10.3 BASES ACTIONS A.1 (continued) switch to the shutdown position. A second Note has been added, which clarifies that this Required Action is only applicable if the requirements not met are for an affected LCO.
| |
| A.2.1 and A.2.2 Required Actions A.2.1 and A.2.2 and are alternative Required Actions that can be taken instead of Required Action A.1 to restore compliance with the normal MODE 3 requirements, thereby exiting this Special Operations LCO's Applicability. Actions must be initiated immediately to insert all insertable control rods. Actions must continue until all such control rods are fully inserted. Placing the reactor mode switch in the shutdown position will ensure that all inserted rods remain inserted and restore operation in accordance with Table 1.1-1. The allowed Completion Time of 1 hour to place the reactor mode switch in the shutdown position provides sufficient time to normally insert the control rods.
| |
| SURVEILLANCE SR 3.10.3.1. SR 3.10.3.2. and SR 3.10.3.3 REQUIREMENTS The other LCOs made applicable in this Special Operations LCO are required to have their Surveillances met to establish that this Special Operations LCO is being met. If the local array of control rods is inserted and disarmed while the scram function for the withdrawn rod is not available, periodic verification in accordance with SR 3.10.3.2 is required to preclude the possibility of criticality. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. Electrically, the control rods can be disarmed by disconnecting power from all four directional control valve solenoids. SR 3.10.3.2 has been modified by a Note, which clarifies that this SR is not required to be met if SR 3.10.3.1 is satisfied for LCO 3.10.3.d.1 requirements, since SR 3.10.3.2 demonstrates that the alternative LCO 3.10.3.d.2 requirements are satisfied. Also, SR 3.10.3.3 verifies that all control rods other than the lnsert 3 control rod being withdrawn are fully inserted. l-h9UF Frcquoncy ic acccptablo b..au.. of the admitr*: .ative hon tFho on oRnto*l rod withdrawal , the potoctoAn 1ffordod by the LCs involycd, anid hardwar- i~ter~loklcthat prculudc additional conRtrol rod withdrawals.
| |
| (continued)
| |
| NMP2 B 3.10.3-4 Revision 0
| |
| | |
| Single Control Rod Withdrawal - Cold Shutdown B 3.10.4 BASES (continued)
| |
| SURVEILLANCE SR 3.10.4.1, SR 3.10.4.2, SR 3.10.4.3, and SR 3.10.4.4 REQUIREMENTS The other LCOs made applicable by this Special Operations LCO are required to have their associated Surveillances met to establish that this Special Operations LCO is being met.
| |
| If the local array of control rods is inserted and disarmed while the scram function for the withdrawn rod is not available, periodic verification is required to ensure that the possibility of criticality remains precluded. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves.
| |
| Electrically, the control rods can be disarmed by disconnecting power from all four directional control valve solenoids. Verification that all the other control rods are fully inserted is required to meet the SDM requirements.
| |
| Verification that a control rod withdrawal block has been inserted ensures that no other control rods can be inadvertently withdrawn under conditions when position indication instrumentation is inoperable for the affected
| |
| /Insert 3 n r The 2 hour F*..qucn.y ic a. .,ptablobc.au.. of L...~theoadmiicitrativo controlc on conRol1 rod withdrawals, tho protcction affordcd by the LCG~s involyed, and hardwarc Dintorlocks to prccludo an additional control rod withdrawal.
| |
| SR 3.10.4.2 and SR 3.10.4.4 have been modified by Notes, which clarify that these SRs are not required to be met if the alternative requirements demonstrated by SR 3.10.4.1 are satisfied.
| |
| REFERENCES 1. USAR, Section 15.4.1.1.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| NMIP2 B 3.10.4-5 Revision 0
| |
| | |
| Single CRD Removal - Refueling B 3.10.5 BASES (continued)
| |
| ACTIONS A.1. A.2.1, and A.2.2 If one or more of the requirements of this Special Operations LCO are not met, the immediate implementation of these Required Actions restores operation consistent with the normal requirements for failure to meet LCO 3.3.1.1, LCO 3.9.1, LCO 3.9.2, LCO 3.9.4, and LCO 3.9.5 (i.e., all control rods inserted) or with the allowances of this Special Operations LCO. The Completion Times for Required Action A.1, Required Action A.2.1, and Required Action A.2.2 are intended to require these Required Actions be implemented in a very short time and carried through in an expeditious manner to either initiate action to restore the CRD and insert its control rod, or initiate action to restore compliance with this Special Operations LCO.
| |
| Actions must continue until either Required Action A.2.1 or Required Action A.2.2 is satisfied.
| |
| SURVEILLANCE SR 3.10.5.1, SR 3.10.5.2, SR 3.10.5.3, SR 3.10.5.4. and REQUIREMENTS SR 3.10.5.5 Verification that all the control rods, other than the control rod withdrawn for the removal of the associated CRD, are fully inserted is required to ensure the SDM is within limits. Verification that the local five by five array of control rods other than the control rod withdrawn for the removal of the associated CRD, is inserted and disarmed, while the scram function for the withdrawn rod is not available, is required to ensure that the possibility of criticality remains precluded. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. Electrically, the control rods can be disarmed by disconnecting power from all four directional control valve solenoids. Verification that a control rod withdrawal block has been inserted ensures that no other control rods can be inadvertently withdrawn under conditions when position indication instrumentation is inoperable for the withdrawn control rod. The Surveillance for LCO 3.1.1, which is made applicable by this Special Operations LCO, is required in order to establish that this Special Operations LCO is being met. Verification that no other CORE ALTERATIONS are being made is required to ensure the assumptions of the safety analysis are satisfied.
| |
| (continued)
| |
| NMP2 B 3.10.5-4 Revision 0
| |
| | |
| Single CRD Removal - Refueling B 3.10.5 BASES SURVEILLANCE SR 3.10.5.1. SR 3.10.5.2, SR 3.10.5.3, SR 3.10.5.4, and REQUIREMENTS SR 3.10.5.5 (continued)
| |
| Periodic verification of the administrative controls established by this Special Operations LCO is prudent t preclude the possibility of an inadvertent criticality.
| |
| 24hourFoucncy a ptabl, given "F the adm r cot,,,- on controi rod romo-al And hardwarI ito-,eh-kl to block -.n additional control rod withdrawal.
| |
| REFERENCES 1. USAR, Section 15.4.1.1.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.10.5-5 Revision 0
| |
| | |
| Multiple Control Rod Withdrawal - Refueling B 3.10.6 BASES (continued)
| |
| APPLICABILITY Operation in MODE 5 is controlled by existing LCOs. The exceptions from other LCO requirements (e.g., the ACTIONS of LCO 3.9.3, LCO 3.9.4 or LCO 3.9.5) allowed by this Special Operations LCO are appropriately controlled by requiring all fuel to be removed from cells whose "full in" indicators are allowed to be bypassed.
| |
| ACTIONS A.1, A.2, A.3.1, and A.3.2 If one or more of the requirements of this Special Operations LCO are not met, the immediate implementation of these Required Actions restores operation consistent with the normal requirements for refueling (i.e., all control rods inserted in core cells containing one or more fuel assemblies) or with the exceptions granted by this Special Operations LCO. The Completion Times for Required Action A.1, Required Action A.2, Required Action A.3.1, and Required Action A.3.2 are intended to require that these Required Actions be implemented in a very short time and carried through in an expeditious manner to either initiate action to restore the affected CRDs and insert their control rods, or initiate action to restore compliance with this Special Operations LCO.
| |
| SURVEILLANCE SR 3.10.6.1, SR 3.10.6.2, and SR 3.10.6.3 Insert 3 REQUIREMENTS Periodic verification of the administrative controls established by this Special Operations LCO is prud nt to preclude the possibility of an inadvertent criticality.
| |
| 24 hour Fraequon'; is acccptablo, given the administrativo 69Atr9l6 on fu*l assembly., and -rl,, roed romal, and ta1-Finto account othcr iniations of control rod stau ilabli t c l ro mIo I REFERENCES 1. USAR, Section 15.4.1.1.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.10.6-3 Revision 0
| |
| | |
| SDM Test - Refueling B 3.10.8 BASES (continued)
| |
| SURVEILLANCE SR 3.10.8.1. SR 3.10.8.2, and SR 3.10.8.3 REQUIREMENTS LCO 3.3.1.1, Functions 2.a, 2.d, and 2.f, made applicable in this Special Operations LCO, are required to have applicable Surveillances met to establish that this Special Operations LCO is being met (SR 3.10.8.1). However, the control rod withdrawal sequences during the SDM tests may be enforced by the RWM (LCO 3.3.2.1, Function 2, MODE 2 requirements) or by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff (e.g., a qualified shift technical advisor or reactor engineer). As noted, either the applicable SRs for the RWM (LCO 3.3.2.1) must be satisfied according to the applicable Frequencies (SR 3.10.8.2), or the proper movement of control rods must be verified (SR 3.10.8.3). This latter verification (i.e., SR 3.10.8.3) must be performed during control rod movement to prevent deviations from the specified sequence. These Surveillances provide adequate assurance that the specified test sequence is being followed.
| |
| SR 3.10.8.4 Periodic verification of the administrative controls Insert 3 established by this LCO will ensure that the reactor is operated within the bounds of the safety analysis. Z ascuranco;P that oach operating shift is aw~aroe of andd vcerificc, comnplianco with those Special Gperations LCO roquiromonte.1 SR 3.10.8.5 Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control rod is withdrawn to the "full-out" notch position or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved as well as operating experience related to uncoupling events.
| |
| (continued)
| |
| NMP2 B 3.10.8-5 Revision 0, 1
| |
| | |
| SDM Test - Refueling B 3.10.8 BASES SURVEILLANCE SR 3.10.8.6 REQUIREMENTS (c:ontinued) CRD charging water header pressure verification is performed to ensure the motive force is available to scram the control rods in the event of a scram signal. Since the reactor is depressurized in MODE 5, there is insufficient reactor pressure to scram the control rods. Verification of charging water header pressure ensures that if a scram were required, capability for rapid control rod insertion would exist. The minimum pressure of 940 psig is well below the expected pressure of 1050 psig to 1100 psig while still Insert 3.ntsuring sufficient pressure for rapid control rod insaeutlb Th 7 ddiat ' Fabcuincy ho bcntr shown tr bm ae .,wt mn ne,,,nlbl i n in th Gen.,e, *
| |
| * Fee**,-*,,,* ,* ,
| |
| I REFERENCES 1. USAR Section 15.4.9.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| NMP2 B 3.10.8-6 Revision 0
| |
| | |
| ATTACHMENT 5 License Amendment Request Nine Mile Point Nuclear Station, Unit 2 Docket No. 50-410 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
| |
| TSTF-425 (NUREG-1433/1434) vs. NMP Unit 2 Cross-Reference
| |
| | |
| LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-410 Page 1 of 10 TSTF-425 vs. NMP Unit 2 Cross-Reference Technical Specification Section Title/Surveillance Description* TSTF-425 NMP Unit 2 (NUREG1433 Control Rod Operability 3.1.3 3.1.3 Control rod position 3.1.3.1 3.1.3.1 Notch test - fully withdrawn control rod one notch 3.1.3.2 Notch test - partially withdrawn control rod one notch 3.1.3.3 3.1.3.3 Control Rod Scram Times 3.1.4 3.1.4 Scram time testing 3.1.4.2 3.1.4.2 Control Rod Scram Accumulators 3.1.5 3.1.5 Control rod scram accumulator pressure 3.1.5.1 3.1.5.1 Rod Pattern Control 3.1.6 3.1.6
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| [BPWS] Analyzed rod position sequence 3.1.6.1 3.1.6.1 Standby Liquid Control (SLC) System 3.1.7 3.1.7 Volume of sodium pentaborate [Level of pentaborate in SLC tank] 3.1.7.1 3.1.7.1 Temperature of sodium pentaborate solution 3.1.7.2 3.1.7.2 Temperature of pump suction piping 3.1.7.3 3.1.7.3 Continuity of explosive charge 3.1.7.4 3.1.7.4 Concentration of boron solution (sodium pentaborate) 3.1.7.5 3.1.7.5 Manual/power operated valve position 3.1.7.6 3.1.7.6 Pump flow rate 3.1.7.7 Flow through one SLC subsystem 3.1.7.8 3.1.7.8 Heat traced piping is unblocked 3.1.7.9 3.1.7.9 Scram Discharge Volume (SDV) Vent & Drain Valves 3.1.8 3.1.8 Each SDV vent & drain valve open 3.1.8.1 3.1.8.1 Cycle each SDV vent & drain valve fully closed/fully open position 3.1.8.2 3.1.8.2 Each SDV vent & drain valve closes on receipt of scram 3.1.8.3 3.1.8.3 Average Planar Linear Heat Generation Rate (APLHGR) 3.2.1 3.2.1 APLHGR less than or equal to limits 3.2.1.1 3.2.1.1 Minimum Critical Power Ratio (MCPR) 3.2.2 3.2.2 MCPR greater than or equal to limits 3.2.2.1 3.2.2.1 Linear Heat Generation Rate (LHGR) 3.2.3 3.2.3 LHGR less than or equal to limits 3.2.3.1 3.2.3.1 Average Power Range Monitor (APRM) Gain & Setpoints 3.2.4 MFLPD is within limits 3.2.4.1 APRM setpoints or gain are adjusted for calculated MFLPD 3.2.4.2 Reactor Protection System (RPS) Instrumentation 3.3.1.1 3.3.1.1 Channel Check 3.3.1.1.1 3.3.1.1.1 Channel Check (OPRM upscale) - 3.3.1.1.2 Absolute diff. between APRM channels & calculated power 3.3.1.1.2 3.3.1.1.3 Adjust channel to conform to calibrated flow 3.3.1.1.3 Table 3.3.1.1-1 2b Channel Functional Test (12 hours after entering Mode 2) 3.3.1.1.4 3.3.1.1.4 Channel Functional Test (weekly) 3.3.1.1.5 Verify SRM and IRM overlap - 3.3.1.1.5
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-410 Page 2 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 NMP Unit 2 (NUREG1433 Verify IRM and APRM overlap 3.3.1.1.6 Calibrate local power range monitors 3.3.1.1.6 3.3.1.1.7 Channel Functional Test 3.3.1.1.7 3.3.1.1.8 Calibrate trip units 3.3.1.1.8 3.3.1.1.9 Channel Calibration (APRMs/WRNMs) 3.3.1.1.9 3.3.1.1.10 Perform a Channel Calibration 3.3.1.1.11 Channel Functional Test 3.3.1.1.10 3.3.1.1.12 Channel Calibration 3.3.1.1.11 3.3.1.1.13 Verify APRM Flow Biased STP - High 3.3.1.1.12 Table 3.3.1.1-1 2b Logic System Functional Test 3.3.1.1.13 3.3.1.1.14 Verify TSV/TCV closure/Trip Oil Press-Low Not Bypassed 3.3.1.1.14 3.3.1.1.15 Verify RPS Response Time 3.3.1.1.15 3.3.1.1.17 Verify APRM OPRM Upscale is not by-passed 3.3.1.1.16 Source Range Monitor (SRM) [Wide Range Neutron Monitor 3.3.1.2 3.3.1.2 (WRNM)] Instrumentation Channel Check 3.3.1.2.1 3.3.1.2.1 Verify Operable SRM Detector [WRNM Detector] 3.3.1.2.2 3.3.1.2.2 Channel Check 3.3.1.2.3 3.3.1.2.3 Verify count rate 3.3.1.2.4 3.3.1.2.4 Channel Functional Test (Mode 5) 3.3.1.2.5 3.3.1.2.5 Channel Functional Test (Modes 2, 3, 4, 5) 3.3.1.2.6 3.3.1.2.6 Channel Calibration 3.3.1.2.7 3.3.1.2.7 Control Rod Block Instrumentation 3.3.2.1 3.3.2.1 Channel Functional Test (RWM) 3.3.2.1.2 3.3.2.1.1 3.3.2.1.3 3.3.2.1.2 Channel Functional Test (RBM) 3.3.2.1.1 3.3.2.1.3 3.3.2.1.4 3.3.2.1.4 RWM not bypassed 3.3.2.1.5 3.3.2.1.5 Channel Functional Test (Reactor Mode Switch -Shutdown position) 3.3.2.1.6 3.3.2.1.6 Channel Calibration 3.3.2.1.7 3.3.2.1.7 Feedwater & Main Turbine High Water Level Trip Instrumentation 3.3.2.2 3.3.2.2 Channel Check 3.3.2.2.1 3.3.2.2.1 Channel Functional Test 3.3.2.2.2 3.3.2.2.2 Channel Calibration 3.3.2.2.3 3.3.2.2.3 Logic System Functional Test 3.3.2.2.4 3.3.2.2.4 Post Accident Monitor (PAM) Instrumentation 3.3.3.1 3.3.3.1 Channel Check 3.3.3.1.1 3.3.3.1.1 Calibration 3.3.3.1.2 3.3.3.1.2 Remote Shutdown System 3.3.3.2 3.3.3.2 Channel Check 3.3.3.2.1 3.3.3.2.1 Verify control circuit and transfer switch capable of function 3.3.3.2.2 3.3.3.2.2 Channel Calibration 3.3.3.2.3 3.3.3.2.3 End-of-Cycle-Recirculation Pump Trip (RPT) Instrumentation 3.3.4.1 3.3.4.1 Channel Functional Test 3.3.4.1.1 3.3.4.1.1
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-410 Page 3 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 NMP Unit 2 (NUREG1433 Calibrate trip units 3.3.4.1.2 Channel Calibration 3.3.4.1.3 3.3.4.1.2 Logic System Functional Test 3.3.4.1.4 3.3.4.1.3 Verify TSV/TCV Closure/Trip Oil Press-Low Not Bypassed 3.3.4.1.5 3.3.4.1.4 Verify EOC-RPT System Response Time 3.3.4.1.6 3.3.4.1.5 Determine RPT breaker interruption time 3.3.4.1.7 3.3.4.1.6' Anticipated Trip Without Scram-RPT Instrumentation 3.3.4.2 3.3.4.2 Channel Check 3.3.4.2.1 3.3.4.2.1 Channel Functional Test 3.3.4.2.2 3.3.4.2.2 Calibrate trip units 3.3.4.2.3 3.3.4.2.3 Reactor Vessel Steam Dome Pressure 3.3.4.2.4 Channel Calibration 3.3.4.2.4 3.3.4.2.5 Logic System Functional Test 3.3.4.2.5 3.3.4.2.6 Emergency Core Cooling System (ECCS) Instrumentation 3.3.5.1 3.3.5.1 Channel Check 3.3.5.1.1 3.3.5.1.1 Channel Functional Test 3.3.5.1.2 3.3.5.1.2 Calibrate trip units 3.3.5.1.3 3.3.5.1.3 Channel Calibration 3.3.5.1.4 3.3.5.1.4 Channel Calibration 3.3.5.1.5 3.3.5.1.5 Logic System Functional Test 3.3.5.1.6 3.3.5.1.6 Verify ECCS Response Time 3.3.5.1.7 Reactor Core Isolation Cooling (RCIC) System Instrumentation 3.3.5.2 3.3.5.2 Channel Check 3.3.5.2.1 3.3.5.2.1 Channel Functional Test 3.3.5.2.2 3.3.5.2.2 Calibrate trip units 3.3.5.2.3 3.3.5.2.3 Channel Calibration (Quarterly) 3.3.5.2.4 Channel Calibration(24 Months) 3.3.5.2.5 3.3.5.2.4 Logic System Functional Test 3.3.5.2.6 3.3.5.2.5 Primary Containment Isolation Instrumentation 3.3.6.1 3.3.6.1 Channel Check 3.3.6.1.1 3.3.6.1.1 Verify Actual Ambient Temperature 3.3.6.1.2 Channel Functional Test (Quarterly) 3.3.6.1.2 3.3.6.1.3 Channel Functional Test (semi annually) 3.3.6.1.5 Calibrate trip units 3.3.6.1.3 3.3.6.1.4 Channel Calibration (Quarterly) 3.3.6.1.4 Channel Calibration (24 Months) 3.3.6.1.6 3.3.6.1.5 Logic System Functional Test 3.3.6.1.7 3.3.6.1.6 Verify Isolation Response Time 3.3.6.1.8 3.3.6.1.7 Secondary Containment Isolation Instrumentation 3.3.6.2 3.3.6.2 Channel Check 3.3.6.2.1 3.3.6.2.1 Channel Functional Test 3.3.6.2.2 3.3.6.2.2 Calibrate trip units 3.3.6.2.3 3.3.6.2.3 Channel Calibration (Quarterly) 3.3.6.2.4 -
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| Channel Calibration (24 months) 3.3.6.2.5 3.3.6.2.4
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-410 Page 4 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 NMP Unit 2 (NUREG1433 Logic System Functional Test 3.3.6.2.6 3.3.6.2.5 Verify Isolation Response Time (Refueling Floor Exhaust Radiation -High) 3.3.6.2.7 Low-Low-Set (LLS) Instrumentation 3.3.6.3 Channel Check 3.3.6.3.1 Channel Functional Test 3.3.6.3.2 Channel Functional Test 3.3.6.3.3 Channel Functional Test 3.3.6.3.4 Calibrate trip units 3.3.6.3.5 Channel Calibration 3.3.6.3.6 Logic System Functional Test 3.3.6.3.7 Main Control Room Environmental Control (MCREC) [Main 3.3.7.1 3.3.7.1 Control Room Emergency Ventilation (MCREV)] Instrumentation Channel Check 3.3.7.1.1 3.3.7.1.1 Channel Functional Test 3.3.7.1.2 3.3.7.1.2 Calibrate trip units 3.3.7.1.3 3.3.7.1.3 Channel Calibration 3.3.7.1.4 3.3.7.1.4 Logic System Functional Test 3.3.7.1.5 3.3.7.1.5 Mechanical Vacuum Pump Isolation Instrumentation 3.3.7.2 Channel Check 3.3.7.2.1 Channel Functional Test 3.3.7.2.2 Channel Calibration 3.3.7.2.3 Logic System Functional Test 3.3.7.2.4 Loss of Power (LOP) Instrumentation 3.3.8.1 3.3.8.1 Channel Check 3.3.8.1.1 Channel Functional Test 3.3.8.1.2 3.3.8.1.1 Channel Calibration 3.3.8.1.3 3.3.8.1.2 Logic System Functional Test 3.3.8.1.4 3.3.8.1.3 RPS Electric Power Monitoring 3.3.8.2 3.3.8.2 Channel Functional Test 3.3.8.2.1 3.3.8.2.1 Channel Calibration (RPS MG set/alt. power supply monitoring) 3.3.8.2.2 3.3.8.2.2 System functional test 3.3.8.2.3 3.3.8.2.3 RPS - Scram Solenoid 3.3.8.3 Channel Functional Test 3.3.8.3.1 Channel Calibration 3.3.8.3.2 Logic System Functional Test 3.3.8.3.3 Recirculation Loops Operating 3.4.1 3.4.1 Recirc loop jet pump flow mismatch with both loops operating 3.4.1.1 3.4.1.1 Flow Control Valves (See Note 1) 3.4.2 3.4.2 FCV fails as is on loss of Hydraulic Pressure 3.4.2.1 3.4.2.1 Average rate of FCV movement 3.4.2.2 3.4.2.2 Jet Pumps 3.4.2 3.4.3 Criteria satisfied for each operating recirc loop 3.4.2.1 3.4.3.1 Safety/Relief Valves (SRVs) [and Safety Valves (SVs)] 3.4.3 3.4.4 Safety function lift setpoints 3.4.3.1 SRV actuates on actual or simulated signal 3.4.3.2
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-410 Page 5 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 NMP Unit 2 (NUREG1433 Reactor Coolant System (RCS) Operational Leakage 3.4.4 3.4.5 RCS unidentified and total leakage increase within limits 3.4.4.1 3.4.5.1 RCS PIV Leakage 3.4.5 3.4.6 PIV Leakage within limits 3.4.5.1 RCS Leakage Detection Instrumentation 3.4.6 3.4.7 Channel Check 3.4.6.1 3.4.7.1 Channel Functional Test' 3.4.6.2 3.4.7.2 Source Check of required drywell atmospheric monitoring system 3.4.7.3 Channel Functional Test of required drywell atmospheric monitoring 3.4.7.4 system Channel Calibration of required leak detection instrumentation 3.4.6.3 3.4.7.5 RCS Specific Activity 3.4.7 3.4.8 Dose Equivalent 1-131 specific activity 3.4.7.1 3.4.8.1 Residual Heat Removal (RHR) Shutdown Cooling - Hot Shutdown 3.4.8 3.4.9 One RHR Shutdown cooling subsystem operating 3.4.8.1 3.4.9.1 RHR Shutdown Cooling - Cold Shutdown 3.4.9 3.4.10 One RHR Shutdown cooling subsystem operating 3.4.9.1 3.4.10.1 RCS Pressure/Temperature Limit 3.4.10 3.4.11 RCS pressure, temperature, heatup and cooldown rates 3.4.10.1 3.4.11.1 RPV flange/head flange temperatures (tensioning head bolt stud) 3.4.10.7 3.4.11.7 RPV flange/head flange temperatures (after RCS temp < 80 0 F) 3.4.10.8 3.4.11.8 RPV flange/head flange temperatures (after RCS temp < 1 00°F) 3.4.10.9 3.4.11.9 Reactor Steam Dome Pressure 3.4.11 3.4.12 Verify reactor steam dome pressure 3.4.11.1 3.4.12.1 ECCS - Operating 3.5.1 3.5.1 Verify injection/spray piping filled with water 3.5.1.1 3.5.1.1 Verify each valve in flow path is in correct position 3.5.1.2 3.5.1.2 Verify ADS nitrogen pressure 3.5.1.3 3.5.1.3 Verify RHR (LPCI) cross tie valve is closed and power removed 3.5.1.4 Verify LPCI inverter output voltage 3.5.1.5 Verify ECCS pumps develop specified flow 3.5.1.7 Verify HPCI flow rate (Rx press < 1020 [1053], > 920 [940]) 3.5.1.8 Verify HPCI flow rate (Rx press < 165 [175]) 3.5.1.9 -
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| Verify ECCS actuates on initiation signal 3.5.1.10 3.5.1.5 Verify ADS actuates on initiation signal 3.5.1.11 3.5.1.6 Verify each ADS valve opens [actuator strokes] when manually 3.5.1.12 3.5.1.7 actuated Verify ECCS Response Time 3.5.1.8 ECCS - Shutdown 3.5.2 3.5.2 Verify suppression pool water level 3.5.2.1 3.5.2.1 Verify, for CS, suppression pool water level and CST water level 3.5.2.2 3.5.2.2 Verify ECCS piping filled with water 3.5.2.3 3.5.2.3 Verify each valve in flow path is in correct position 3.5.2.4 3.5.2.4 Verify ECCS actuates on initiation signal 3.5.2.6 3.5.2.6 Verify ECCS Response Time 3.5.2.7
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-410 Page 6 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 NMP Unit 2 (NUREG1433 RCIC System 3.5.3 3.5.3 Verify RCIC piping filled with water 3.5.3.1 3.5.3.1 Verify each valve in flow path is in correct position 3.5.3.2 3.5.3.2 Verify RCIC flow rate 3.5.3.3 3.5.3.3 Verify RCIC flow rate (Rx press < 165) 3.5.3.4 3.5.3.4 Verify RCIC actuates on initiation signal 3.5.3.5 3.5.3.5 Primary Containment 3.6.1.1 3.6.1.1 Verify drywell to suppression chamber differential pressure [bypass 3.6.1.1.2 3.6.1.1.2 leakage]
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| Verify drywell to suppression chamber bypass and combined 3.6.1.1.3 leakage at initial diff. pressure of > 3 psid Primary Containment Air Lock 3.6.1.2 3.6.1.2 Verify only one door can be opened at a time 3.6.1.2.2 3.6.1.2.2 Primary Containment Isolation Valves (PCIVs) 3.6.1.3 3.6.1.3 Verify purge valve is closed except one valve in a penetration 3.6.1.3.1 -
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| Verify each 18 (20) inch primary purge valve is closed 3.6.1.3.2 3.6.1.3.1 Verify each manual PCIV outside containment is closed 3.6.1.3.3 3.6.1.3.2 Verify continuity of traversing incore probe (TIP) shear valve 3.6.1.3.5 3.6.1.3.4 Verify PCIV isolation times (except MSIVs) 3.6.1.3.6 i Perform leak rate testing for valves with resilient seals 3.6.1.3.7 3.6.1.3.6 Verify automatic PCIV actuates to isolation position 3.6.1.3.9 3.6.1.3.8 Verify sample of Excess Flow Check Valves actuate to isolation 3.6.1.3.10 3.6.1.3.9 position Test explosive squib from each shear valve 3.6.1.3.11 3.6.1.3.10 Drywell/Containment Pressure 3.6.1.4 3.6.1.4 Verify drywell pressure is within limit 3.6.1.4.1 3.6.1.4.1 Drywell Average Air Temperature 3.6.1.5 3.6.1.5 Verify drywell average air temperature is within limit 3.6.1.5.1 3.6.1.5.1 RHR Drywell Spray (See Note 1) 3.6.1.7 3.6.1.6 Verify RHR drywell spray valves is not locked is in its correct position 3.6.1.7.1 3.6.1.6.1 Verify each RHR pump is operable (Admin. Means) 3.6.1.6.2 LLS Valves 3.6.1.6 Verify each LLS valve opens when manually actuated 3.6.1.6.1 Verify LLS system actuates on initiation signal 3.6.1.6.2 Reactor Building - Suppression Chamber Vacuum Breakers 3.6.1.7 Verify each vacuum breaker is closed 3.6.1.7.1 Perform functional test on each vacuum breaker 3.6.1.7.2 Verify opening setpoint for each vacuum breaker 3.6.1.7.3 Suppression Chamber - Drywell Vacuum Breakers 3.6.1.8 3.6.1.7 Verify each vacuum breaker is closed 3.6.1.8.1 3.6.1.7.1 Perform functional test on each vacuum breaker 3.6.1.8.2 3.6.1.7.2 Verify opening setpoint for each vacuum breaker 3.6.1.8.3 3.6.1.7.3 Main Steam Isolation Valve (MSIV) Leakage Control System 3.6.1.9 Operate each MSIV LCS blower 3.6.1.9.1 Verify continuity of inboard MSIV LCS heater element 3.6.1.9.2
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-410 Page 7 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 NMP Unit 2 (NUREG1433 Perform functional test of each MSIV LCS subsystem 3.6.1.9.3 Suppression Pool Average Temperature 3.6.2.1 3.6.2.1 Verify suppression pool average temperature within limits 3.6.2.1.1 3.6.2.1.1 Suppression Pool Water Level 3.6.2.2 3.6.2.2 Verify suppression pool water level within limits 3.6.2.2.1 3.6.2.2.1 RHR Suppression Pool Cooling 3.6.2.3 3.6.2.3 Verify each valve in flow path is in correct position 3.6.2.3.1 3.6.2.3.1 RHR Suppression Pool Spray 3.6.2.4 3.6.2.4 Verify each valve in flow path is in correct position 3.6.2.4.1 3.6.2.4.1 Drywell - Suppression Chamber Differential Pressure 3.6.2.5 Verify differential pressure is within limit 3.6.2.5.1 Drywell Cooling System Fans 3.6.3.1 Operate each fan > 15 minutes 3.6.3.1.1 Verify each fan flow rate 3.6.3.1.2 Primary Containment Oxygen Concentration 3.6.3.2 3.6.3.2 Verify oxygen concentration is within limits 3.6.3.2.1 3.6.3.2.1 Containment Atmosphere Dilution (CAD) System 3.6.3.3 Verify CAD liquid nitrogen storage 3.6.3.3.1 ------
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| Verify each CAD valve in flow path is in correct position 3.6.3.3.2 Secondary Containment 3.6.4.1 3.6.4.1 Verify SC vacuum is > 0.25 inch of vacuum water gauge 3.6.4.1.1 3.6.4.1.1 Verify all SC equipment hatches closed and sealed 3.6.4.1.2 3.6.4.1.2 Verify one SC access door in each opening is closed 3.6.4.1.3 3.6.4.1.3 Verify SC drawn down using one SGTS 3.6.4.1.4 3.6.4.1.4 Verify SC can be maintained using one SGTS 3.6.4.1.5 3.6.4.1.5 Secondary Containment Isolation Valves 3.6.4.2 3.6.4.2 Verify each SC isolation manual valve is closed 3.6.4.2.1 3.6.4.2.1 Verify isolation time of each SCIV 3.6.4.2.2 3.6.4.2.2 Verify each automatic SCIV actuates to isolation position 3.6.4.2.3 3.6.4.2.3 Standby Gas Treatment (SGT) System 3.6.4.3 3.6.4.3 Operate each SGT subsystem with heaters operating 3.6.4.3.1 3.6.4.3.1 Verify each SGT subsystem actuates on initiation signal 3.6.4.3.3 3.6.4.3.3 Verify each SGT filter cooler bypass damper can be opened 3.6.4.3.4 3.6.4.3.4 Residual Heat Removal Service Water (RHRSW) System [High 3.7.1 ----
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| Pressure Service Water (HPSW) System]
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| Verify each RHRSW [HPSW] valve in flow path in correct position 3.7.1.1 Plant Service Water (PSW) System and Ultimate Heat Sink (UHS 3.7.2 3.7.1 Verify water temperature of intake tunnel 3.7.1.1 Verify water level in Cooling tower basin 3.7.2.1 Verify water level in pump well of pump structure 3.7.2.2 3.7.1.2 Verify average water temperature of heat sink 3.7.2.3 3.7.1.3 Verify each required SW pump is in operation 3.7.1.4 Verify current of deicer heater within limits 3.7.1.5 Verify each PSW valve in flow path is in correct position 3.7.2.5 3.7.1.6
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-410 Page 8 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 NMP Unit 2 (NUREG1433 Verify PSW actuates on initiation signal 3.7.2.6 3.7.1.7 Verify current of deicer heater within limits 3.7.1.8 Operate each cooling tower fan 3.7.2.4 -
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| DG (1 B) SSW System 3.7.3 Verify valves are in the correct position 3.7.3.1 Ensure SSW system pump automatic start 3.7.3.2 MCREC [MCREV] System [CREF for NMP] 3.7.4 3.7.2 Operate each MCREC [CREF for NMP] subsystem 3.7.4.1 3.7.2.1 Verify each subsystem actuates on initiation signal 3.7.4.3 3.7.2.3 Maintain positive pressure 3.7.4.4 Control Room Air Conditioning System 3.7.5 3.7.3 Verify each subsystem has capability to remove heat load 3.7.5.1 3.7.3.1 Main Condenser Offgas 3.7.6 3.7.4 Verify gross gamma activity rate of the noble gases 3.7.6.1 3.7.4.1 Main Turbine Bypass System 3.7.7 3.7.5 Verify one complete cycle of each main turbine bypass valve 3.7.7.1 Perform system functional test 3.7.7.2 3.7.5.1 Verify Turbine Bypass System Response Time within limits 3.7.7.3 3.7.5.2 Spent Fuel Storage Pool Water Level 3.7.8 3.7.6 Verify spent fuel storage pool water level 3.7.8.1 3.7.6.1 AC Sources - Operating 3.8.1 Verify correct breaker alignment 3.8.1.1 3.8.1.1 Verify each DG starts from standby conditions/steady state 3.8.1.2 3.8.1.2 Verify each DG is synchronized and loaded 3.8.1.3 3.8.1.3 Verify each day tank level 3.8.1.4 3.8.1.4 Check for and remove accumulated water from day tank 3.8.1.5 3.8.1.5 Verify fuel oil transfer system operates 3.8.1.6 3.8.1.6 Verify each DG starts from standby conditions 3.8.1.7 3.8.1.2 Verify transfer of power from offsite circuit to alternate circuit 3.8.1.8 3.8.1.14 Verify DG rejects load greater than single largest load 3.8.1.9 3.8.1.7 Verify DG maintains load following load reject 3.8.1.10 3.8.1.8 Verify on loss of offsite power signal 3.8.1.11 3.8.1.9 Verify DG starts on ECCS initiation signal 3.8.1.12 3.8.1.10 Verify DG automatic trips bypassed on ECCS initiation signal 3.8.1.13 3.8.1.11 Verify each DG operates for > 24 hours 3.8.1.14 3.8.1.12 Hot Restart 3.8.1.15 3.8.1.13 Verify each DG synchronizes with offsite power 3.8.1.16 3.8.1.14 Verify ECCS initiation signal overrides test mode 3.8.1.17 3.8.1.15 Verify interval between each timed load block 3.8.1.18 3.8.1.16 Verify on LOOP in conjunction with ECCS initiation signal 3.8.1.19 3.8.1.17 Verify simultaneous DG starts 3.8.1.20 3.8.1.18 Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 3.8.3 Verify fuel oil storage tank volume 3.8.3.1 3.8.3.1 Verify lube oil inventory 3.8.3.2 3.8.3.2 Verify each DG air start receiver pressure 3.8.3.4 3.8.3.4
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-410 Page 9 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 NMP Unit 2 (NUREG1433 Check/remove accumulated water from fuel oil storage tank 3.8.3.5 3.8.3.5 DC Sources - Operating 3.8.4 3.8.4 Verify battery terminal voltage 3.8.4.1 3.8.4.1 Verify no visible corrosion or battery connection resistance 3.8.4.2 Verify battery cells, plates, racks show no physical damage 3.8.4.3 Remove visible corrosion and coat connections 3.8.4.4 Verify battery connection resistance 3.8.4.5 Verify each battery charger supplies amperage 3.8.4.2 3.8.4.6 Verify battery capacity is adequate to maintain emergency loads 3.8.4.3 3.8.4.7 Verify battery capacity during performance discharge test 3.8.4.8 Battery Parameters 3.8.6 3.8.6 Verify battery meets Category A limits 3.8.6.1 Verify battery meets Category B limits 3.8.6.2 Verify electrolyte temperature 3.8.6.3 Verify battery float current 3.8.6.1 Verify battery pilot cell voltage 3.8.6.2 Verify battery connected cell electrolyte level 3.8.6.3 Verify battery pilot cell temperature 3.8.6.4 Verify battery connected cell voltage 3.8.6.5 Verify battery capacity is adequate to maintain emergency loads 3.8.6.6 3.8.4.8 Inverters - Operating 3.8.7 3.8.7 Verify correct inverter voltage, frequency and alignment 3.8.7.1 3.8.7.1 Inverters - Shutdown 3.8.8 Verify correct inverter voltage, frequency and alignment 3.8.8.1 Distribution System - Operating 3.8.9 3.8.8 Verify correct breaker alignment/power to distribution subsystems 3.8.9.1 3.8.8.1 Distribution System - Shutdown 3.8.10 3.8.9 Verify correct breaker alignment/power to distribution subsystems 3.8.10.1 3.8.9.1 Refueling Equipment Interlocks 3.9.1 3.9.1 Channel Functional Test of refueling equip interlock inputs 3.9.1.1 3.9.1.1 Refuel Position One-Rod-Out Interlock 3.9.2 3.9.2 Verify reactor mode switch locked in refuel position 3.9.2.1 3.9.2.1 Perform Channel Functional Test 3.9.2.2 3.9.2.2 Control Rod Position 3.9.3 3.9.3 Verify all control rods fully inserted 3.9.3.1 3.9.3.1 Control Rod Operability - Refueling 3.9.5 3.9.5 Insert each withdrawn control rod one notch 3.9.5.1 3.9.5.1 Verify each withdrawn control rod scram accumulator press 3.9.5.2 3.9.5.2 Reactor Pressure Vessel (RPV) Water Level - Irradiated Fuel 3.9.6 3.9.6 Verify RPV water level 3.9.6.1 3.9.6.1 Reactor Pressure Vessel (RPV) Water Level - New Fuel 3.9.7 3.9.7 Verify RPV water level 3.9.7.1 3.9.7.1 RHR - High Water Level 3.9.8 3.9.8 Verify one RHR shutdown cooling subsystem operating 3.9.8.1 3.9.8.1 RHR - Low Water Level 3.9.9 3.9.9
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-410 Page 10 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 NMP Unit 2 (NUREG1433 Verify one RHR shutdown cooling subsystem operating 3.9.9.1 3.9.9.1 Reactor Mode Switch Interlock Testing 3.10.2 3.10.2 Verify all control rods fully inserted in core cells 3.10.2.1 3.10.2.1 Verify no core alterations in progress 3.10.2.2 3.10.2.2 Single Control Rod Withdrawal - Hot Shutdown 3.10.3 3.10.3 Verify all control rods in five-by-five array are disarmed 3.10.3.2 3.10.3.2 Verify all control rods other than withdrawn rod are fully inserted 3.10.3.3 3.10.3.3 Single Control Rod Withdrawal - Cold Shutdown 3.10.4 3.10.4 Verify all control rods in five-by-five array are disarmed 3.10.4.2 3.10.4.2 Verify all control rods other than withdrawn rod are fully inserted 3.10.4.3 3.10.4.3 Verify a control rod withdrawal block is inserted 3.10.4.4 3.10.4.4 Single Control Rod Drive (CRD) Removal - Refueling 3.10.5 3.10.5 Verify all control rods other than withdrawn rod are fully inserted 3.10.5.1 3.10.5.1 Verify all control rods in five-by-five array are disarmed 3.10.5.2 3.10.5.2 Verify a control rod withdrawal block is inserted 3.10.5.3 3.10.5.3 Verify no core alterations in progress 3.10.5.5 3.10.5.5 Multiple CRD Withdrawal - Refueling 3.10.6 3.10.6 Verify four fuel assemblies removed from core cells 3.10.6.1 3.10.6.1 Verify all other rods in core cells inserted 3.10.6.2 3.10.6.2 Verify fuel assemblies being loaded comply with reload sequence 3.10.6.3 3.10.6.3 ShutDown Margin Test - Refueling 3.10.8 3.10.8 Verify no other core alterations in progress 3.10.8.4 3.10.8.4 Verify CRD charging water header pressure 3.10.8.6 3.10.8.6 Recirculation Loops - Testing 3.10.9 Verify LCO 3.4.1 requirements suspended for < 24 hours 3.10.9.1 Verify Thermal power < 5% RTP during Physics Test 3.10.9.2 Training Startups 3.10.10 Verify all operable IRM channels are <25/40 div. of full scale 3.10.10.1 Verify average reactor coolant temperature < 200 F 3.10.10.2--------
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| Programs (Surveillance Frequency Control Program [SFCP]) 5.5.15 5.5.14
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| * The Technical Specification Section Title/Surveillance Description portion of this attachment is a summary description of the referenced TSTF-425 (NUREG-1433 and 1434)/NMP Unit 2 TS Surveillances which is provided for information purposes only and is not intended to be a verbatim description of the TS Surveillances.
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| Note 1: Numbering correspond to NUREG 1434 Technical Specifications and surveillance requirements.
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| ATTACHMENT 6 License Amendment Request Nine Mile Point Nuclear Station, Unit 2 Docket No. 50-410 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
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| Proposed No Significant Hazards Consideration Description of Amendment Request: This amendment request involves the adoption of approved changes to the standard technical specifications (STS) for General Electric Plants, BWR/4 (NUREG-1433), to allow relocation of specific TS surveillance frequencies to a licensee-controlled program. The proposed changes are described in Technical Specification Task Force (TSTF) Traveler, TSTF-425, Revision 3 (ADAMS Accession No. ML090850642) related to the Relocation of Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b and are described in the Notice of Availability published in the Federal Register on July 6, 2009 (74 FR 31996).
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| The proposed changes are consistent with NRC-approved Industry/ TSTF Traveler, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b."
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| The proposed changes relocate surveillance frequencies to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP). The changes are applicable to licensees using probabilistic risk guidelines contained in NRC-approved NEI 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. 071360456).
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| Basis for proposed no significant hazards consideration: As required by 10 CFR 50.91 (a),
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| the Exelon analysis of the issue of no significant hazards consideration is presented below:
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| : 1. Do the proposed changes involve a significant increase in the probability or consequences of any accident previously evaluated?
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| Response: No The proposed changes relocate the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program (SFCP). Surveillance frequencies are not an initiator to any accident previously evaluated.
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| As a result, the probability of any accident previously evaluated is not significantly increased.
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| The systems and components required by the technical specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation
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| LAR - Adoption of TSTF-425, Revision 3 Attachment 6 Docket No. 50-410 Page 2 of 2 function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.
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| Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
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| : 2. Do the proposed changes create the possibility of a new or different kind of accident from any previously evaluated?
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| Response: No No new or different accidents result from utilizing the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.
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| Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
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| : 3. Do the proposed changes involve a significant reduction in the margin of safety?
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| Response: No The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Exelon will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1 in accordance with the TS SFCP.
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| NEI 04-10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.
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| Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
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| Based upon the reasoning presented above, Exelon concludes that the requested changes do not involve a significant hazards consideration as set forth in 10 CFR 50.92(c), Issuance of Amendment.}}
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