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Category:Letter
MONTHYEARML24022A1172024-01-23023 January 2024 Acceptance of Requested Licensing Action Amendment Request to Revise Renewed Facility Operating License and Permanently Defueled Technical Specifications to Support Resumption of Power Operations ML24012A2422024-01-16016 January 2024 Acceptance of Requested Licensing Action License Transfer Request ML23236A0042023-12-27027 December 2023 Issuance of Amendment 274 Re Changes to Perm Defueled Emergency Plan and Perm Defueled Emergency Action Level Scheme ML23355A1242023-12-26026 December 2023 Withdrawal of an Amendment Request Re License Amendment Request to Revise License Condition to Eliminate Cyber Security Plan Requirements ML23192A0772023-12-26026 December 2023 Letter Exemption from the Requirements of 10 CFR 140.11(a)(4) Concerning Offsite Primary and Secondary Liability Insurance ML23191A5222023-12-22022 December 2023 Exemption Letter from the Requirements of 10 CFR 50.54(W)(1) Concerning Onsite Property Damage Insurance (EPID - L-2022-LLE-0032) ML23263A9772023-12-22022 December 2023 Exemption from Certain Emergency Planning Requirements and Related Safety Evaluation ML23354A2602023-12-21021 December 2023 Reference Simulator Inspection Request for Information L-23-019, Proof of Financial Protection 10 CFR 140.152023-12-18018 December 2023 Proof of Financial Protection 10 CFR 140.15 PNP 2023-030, License Amendment Request to Revise Renewed Facility Operating License and Permanently Defueled Technical Specifications to Support Resumption of Power Operations2023-12-14014 December 2023 License Amendment Request to Revise Renewed Facility Operating License and Permanently Defueled Technical Specifications to Support Resumption of Power Operations PNP 2023-035, Withdrawal of License Amendment Request - Revise License Condition to Eliminate Cyber Security Plan Requirements2023-12-12012 December 2023 Withdrawal of License Amendment Request - Revise License Condition to Eliminate Cyber Security Plan Requirements PNP 2023-028, Application for Order Consenting to Transfer of Control of License and Approving Conforming License Amendments2023-12-0606 December 2023 Application for Order Consenting to Transfer of Control of License and Approving Conforming License Amendments L-23-012, Master Decommissioning Trust Agreement Changes for Indian Point Nuclear Generating Units 1, 2 and 3, Pilgrim Nuclear Power Station, Palisades Nuclear Plant and the Non-Qualified Trust for Big Rock Point2023-11-13013 November 2023 Master Decommissioning Trust Agreement Changes for Indian Point Nuclear Generating Units 1, 2 and 3, Pilgrim Nuclear Power Station, Palisades Nuclear Plant and the Non-Qualified Trust for Big Rock Point ML23291A4402023-11-0303 November 2023 Acceptance of Requested Licensing Action Request for Exemption from 10 CFR 50.82(a)(2) to Support Reauthorization of Power Operations IR 05000255/20230032023-10-0404 October 2023 NRC Inspection Report No. 05000255/2023003(DRSS)-Holtec Decommissioning International, LLC, Palisades Nuclear Plant ML23275A0012023-10-0202 October 2023 Request for Withholding Information from Public Disclosure for Palisades Nuclear Plant PNP 2023-025, Request for Exemption from Certain Termination of License Requirements of 10 CFR 50.822023-09-28028 September 2023 Request for Exemption from Certain Termination of License Requirements of 10 CFR 50.82 PNP 2023-026, Pre-Submittal Meeting Presentation - Palisades Nuclear Plant License Transfer Application to Support Resumption of Power Operations2023-09-28028 September 2023 Pre-Submittal Meeting Presentation - Palisades Nuclear Plant License Transfer Application to Support Resumption of Power Operations PNP 2023-023, Special Report High Range Noble Gas Monitor Inoperable2023-08-0909 August 2023 Special Report High Range Noble Gas Monitor Inoperable ML23215A2302023-08-0303 August 2023 Notice of Organization Change - Chief Nuclear Officer IR 05000255/20230022023-07-19019 July 2023 NRC Inspection Report 05000255/2023002 DRSS-Holtec Decommissioning International, LLC, Palisades Nuclear Plant ML23087A0362023-05-0202 May 2023 PSDAR Review Letter ML23117A2172023-05-0101 May 2023 Safety Evaluation for Quality Assurance Program Manual Reduction in Commitment PNP 2023-018, 2022 Annual Non-Radiological Environmental Operating Report2023-04-25025 April 2023 2022 Annual Non-Radiological Environmental Operating Report L-23-004, HDI Annual Occupational Radiation Exposure Data Reports - 20222023-04-24024 April 2023 HDI Annual Occupational Radiation Exposure Data Reports - 2022 PNP 2023-007, and Big Rock Point, 2022 Annual Radioactive Effluent Release and Waste Disposal Reports2023-04-19019 April 2023 and Big Rock Point, 2022 Annual Radioactive Effluent Release and Waste Disposal Reports PNP 2023-008, 2022 Radiological Environmental Operating Report2023-04-18018 April 2023 2022 Radiological Environmental Operating Report L-23-003, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations2023-03-31031 March 2023 Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations PNP 2023-002, 6 to Updated Final Safety Analysis Report2023-03-31031 March 2023 6 to Updated Final Safety Analysis Report ML23088A0382023-03-29029 March 2023 Stations 1, 2, & 3, Palisades Nuclear Plant, and Big Rock Point - Nuclear Onsite Property Damage Insurance PNP 2023-006, Report of Changes to Security Plan, Revision 202023-03-29029 March 2023 Report of Changes to Security Plan, Revision 20 PNP 2023-012, Presentation on Regulatory Path to Reauthorize Power Operations2023-03-16016 March 2023 Presentation on Regulatory Path to Reauthorize Power Operations ML23038A0982023-03-15015 March 2023 Request for Withholding Information from Public Disclosure ML23095A0642023-03-14014 March 2023 American Nuclear Insurers, Notice of Cancellation Rescinded PNP 2023-001, Regulatory Path to Reauthorize Power Operations2023-03-13013 March 2023 Regulatory Path to Reauthorize Power Operations PNP 2023-004, Report of Changes to Palisades Nuclear Plant Technical Specification Bases2023-03-0808 March 2023 Report of Changes to Palisades Nuclear Plant Technical Specification Bases PNP 2023-005, Response to Palisades Nuclear Plant - Request for Additional Information Related to the Post-Shutdown Decommissioning Activities Report2023-03-0101 March 2023 Response to Palisades Nuclear Plant - Request for Additional Information Related to the Post-Shutdown Decommissioning Activities Report ML22361A1022023-02-24024 February 2023 Reactor Decommissioning Branch Project Management Changes for Some Decommissioning Facilities and Establishment of Backup Project Manager for All Decommissioning Facilities ML23052A1092023-02-17017 February 2023 FEMA Letter to NRC, Proposed Commission Paper Language for Palisades Nuclear Plant Emergency Plan Decommissioning Exemption Request ML23032A3992023-02-0101 February 2023 Regulatory Path to Reauthorize Power Operations IR 05000255/20220032022-12-28028 December 2022 NRC Inspection Report No. 05000255/2022003(DRSS); 07200007/2022001 (Drss) Holtec Decommissioning International, LLC, Palisades Nuclear Plant L-22-042, Oyster, Pilgrim, Indian Point, Palisades and Big Rock Point - Proof of Financial Protection 10 CFR 140.152022-12-14014 December 2022 Oyster, Pilgrim, Indian Point, Palisades and Big Rock Point - Proof of Financial Protection 10 CFR 140.15 PNP 2022-037, Report of Changes to Security Plan, Revision 192022-12-14014 December 2022 Report of Changes to Security Plan, Revision 19 ML22321A2852022-11-17017 November 2022 LLC Master Decommissioning Trust Agreement for Palisades Nuclear Plant IR 05000255/20224012022-11-0909 November 2022 Decommissioning Palisades Nuclear Plant - Decommissioning Security Inspection Report 05000255/2022401 PNP 2022-036, Response to Request for Additional Information Regarding License Amendment Request for Proposed Permanently Defueled Emergency Plan and Permanently Defueled Emergency Action Level Scheme2022-11-0808 November 2022 Response to Request for Additional Information Regarding License Amendment Request for Proposed Permanently Defueled Emergency Plan and Permanently Defueled Emergency Action Level Scheme PNP 2022-035, International - Notification of Commitment Cancellations for Remaining Activities Related to Beyond-Design-Basis Seismic Hazard Reevaluations2022-11-0202 November 2022 International - Notification of Commitment Cancellations for Remaining Activities Related to Beyond-Design-Basis Seismic Hazard Reevaluations PNP 2022-024, Request for Exemption from 10 CFR 140.11(a)(4) Concerning Primary and Secondary Liability Insurance2022-10-26026 October 2022 Request for Exemption from 10 CFR 140.11(a)(4) Concerning Primary and Secondary Liability Insurance PNP 2022-026, Request for Exemption from 10 CFR 50.54(w)(1) Concerning Onsite Property Damage Insurance2022-10-26026 October 2022 Request for Exemption from 10 CFR 50.54(w)(1) Concerning Onsite Property Damage Insurance ML22292A2572022-10-25025 October 2022 Permanently Defueled Emergency Plan License Amendment RAI Letter 2024-01-23
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARPNP 2023-005, Response to Palisades Nuclear Plant - Request for Additional Information Related to the Post-Shutdown Decommissioning Activities Report2023-03-0101 March 2023 Response to Palisades Nuclear Plant - Request for Additional Information Related to the Post-Shutdown Decommissioning Activities Report PNP 2022-036, Response to Request for Additional Information Regarding License Amendment Request for Proposed Permanently Defueled Emergency Plan and Permanently Defueled Emergency Action Level Scheme2022-11-0808 November 2022 Response to Request for Additional Information Regarding License Amendment Request for Proposed Permanently Defueled Emergency Plan and Permanently Defueled Emergency Action Level Scheme PNP 2022-012, Response to Request for Additional Information Regarding License Amendment Request to Revise Facility Operating License and Technical Specifications for a Permanently Defueled Condition2022-04-21021 April 2022 Response to Request for Additional Information Regarding License Amendment Request to Revise Facility Operating License and Technical Specifications for a Permanently Defueled Condition CNRO-2021-00002, Entergy Operations, Inc. - Basis for Concluding the Terms of Confirmatory Order EA-17-132/EA-17-153 Are Complete, Element L2021-01-28028 January 2021 Entergy Operations, Inc. - Basis for Concluding the Terms of Confirmatory Order EA-17-132/EA-17-153 Are Complete, Element L ML20272A1662020-09-30030 September 2020 Attachment 3 - Framatome Document No. ANP-3876, Revision 1Q1NP, Response to NRC Request for Additional Information of Palisades Relief Request Number RR 5-8, Repair of Reactor Pressure Vessel Head Penetration, Inservice Inspection Program, CNRO-2019-00030, Response to Confirmatory Order EA-17-132/EA-17-153, Element K 2019 Summary2019-12-30030 December 2019 Response to Confirmatory Order EA-17-132/EA-17-153, Element K 2019 Summary PNP 2019-034, Response to Request for Additional Information Regarding License Amendment Request Resubmittal to Adopt TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control- Risk Informed Technical Specification...2019-08-23023 August 2019 Response to Request for Additional Information Regarding License Amendment Request Resubmittal to Adopt TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control- Risk Informed Technical Specification... ML19149A3042019-05-28028 May 2019 Enclosure Attachment 1 (Continued) to Pnp 2019-028: Operating License Page Change Instructions and Retyped Renewed Facility Operating License Pages ML19149A3022019-05-28028 May 2019 Enclosure to Pnp 2019-028: Response to Request for Additional Information - License Amendment Request to Revise Existing Facility Operating License Conditions Regarding NFPA 805 Modifications ML19149A3032019-05-28028 May 2019 Enclosure Attachment 1 to Pnp 2019-028: Renewed Facility Operating License Page Markups PNP 2019-003, Response to Request for Additional Information for License Amendment Request to Revise Emergency Diesel Generator Degraded Voltage Surveillance Requirement2019-02-0707 February 2019 Response to Request for Additional Information for License Amendment Request to Revise Emergency Diesel Generator Degraded Voltage Surveillance Requirement PNP 2018-059, Response to Request for Additional Information for Relief Request No. RR-5-7, Proposed Alternative to ASME Section XI Code Requirements for Repair of Reactor Pressure Vessel Head Penetrations2018-12-0303 December 2018 Response to Request for Additional Information for Relief Request No. RR-5-7, Proposed Alternative to ASME Section XI Code Requirements for Repair of Reactor Pressure Vessel Head Penetrations PNP 2018-023, Response to Second Request for Additional Information - Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel2018-04-30030 April 2018 Response to Second Request for Additional Information - Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel PNP 2018-018, Response to Request for Additional Information - Proposed Changes to the Emergency Plan to Reflect a Permanently Shut Down and Defueled Reactor Vessel2018-04-16016 April 2018 Response to Request for Additional Information - Proposed Changes to the Emergency Plan to Reflect a Permanently Shut Down and Defueled Reactor Vessel PNP 2018-014, Response to Request for Additional Information - Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel2018-03-27027 March 2018 Response to Request for Additional Information - Alternative to the Reexamination Frequency for a Relevant Condition Foreign Material Lodged in the Reactor Pressure Vessel PNP 2017-075, Response to Request for Additional Information - Proposed Changes to Administrative Controls Section of the Technical Specifications for Permanently Defueled Condition2017-12-19019 December 2017 Response to Request for Additional Information - Proposed Changes to Administrative Controls Section of the Technical Specifications for Permanently Defueled Condition PNP 2017-020, Response to Request for Additional Information - Relief Request Number RR 4-25 Impracticality - Limited Coverage Examinations During the Fourth 10-year Inservice Inspection Interval2017-04-0505 April 2017 Response to Request for Additional Information - Relief Request Number RR 4-25 Impracticality - Limited Coverage Examinations During the Fourth 10-year Inservice Inspection Interval PNP 2016-055, Response to Generic Letter 2016-01, Monitoring of Neutron Absorbing Materials in Spent Fuel Pools.2016-10-25025 October 2016 Response to Generic Letter 2016-01, Monitoring of Neutron Absorbing Materials in Spent Fuel Pools. PNP 2016-053, Supplement to License Amendment Request: Control Rod Drive Exercise Surveillance2016-09-0808 September 2016 Supplement to License Amendment Request: Control Rod Drive Exercise Surveillance PNP 2016-047, Voluntary Response to NRC Regulatory Issue Summary 2016-09: Preparation and Scheduling of Operator Licensing Examinations2016-07-26026 July 2016 Voluntary Response to NRC Regulatory Issue Summary 2016-09: Preparation and Scheduling of Operator Licensing Examinations PNP 2016-037, Response to Request for Additional Information Regarding the License Amendment Request for Implementation of an Alternate Repair Criterion on the Steam Generator Tubes (CAC No. MF74352016-06-0707 June 2016 Response to Request for Additional Information Regarding the License Amendment Request for Implementation of an Alternate Repair Criterion on the Steam Generator Tubes (CAC No. MF7435 ML16071A4412016-03-0707 March 2016 Entergy Fleet Relief Request No. RR-EN-15-1-Proposed Alternative to Use ASME Code Case N-789-1 - E-mail from G.Davant to R.Guzman - Response to Second RAI (MF6340 - MF6349) PNP 2016-016, Reply to Request for Information EA-16-0112016-03-0303 March 2016 Reply to Request for Information EA-16-011 CNRO-2016-00005, Response to Request for Additional Information Pertaining to a Change to the Entergy Quality Assurance Program Manual (QAPM)2016-02-25025 February 2016 Response to Request for Additional Information Pertaining to a Change to the Entergy Quality Assurance Program Manual (QAPM) CNRO-2016-00002, Entergy - Relief Request Number RR EN-15-1, Rev. 1 - Proposed Alternative to Use ASME Code Case N-789-1, Alternative Requirements for Pad Reinforcement of Class 2 and 3 Moderate Energy Carbon Steel Piping for Raw Water Service, Secti2016-01-29029 January 2016 Entergy - Relief Request Number RR EN-15-1, Rev. 1 - Proposed Alternative to Use ASME Code Case N-789-1, Alternative Requirements for Pad Reinforcement of Class 2 and 3 Moderate Energy Carbon Steel Piping for Raw Water Service, Section Xl, CNRO-2015-00002, Entergy Operations, Inc. - Response to RAI Questions and Submittal of RR EN-15-2, Rev. 12015-12-0404 December 2015 Entergy Operations, Inc. - Response to RAI Questions and Submittal of RR EN-15-2, Rev. 1 PNP 2015-069, Response to Request for Additional Information Regarding Relief Request No. RR 5-22015-09-0909 September 2015 Response to Request for Additional Information Regarding Relief Request No. RR 5-2 PNP 2015-063, Supplemental Information for the Response to the First Request for Additional Information Regarding the License Amendment Request to Implement 10 CFR 50.61a2015-08-14014 August 2015 Supplemental Information for the Response to the First Request for Additional Information Regarding the License Amendment Request to Implement 10 CFR 50.61a PNP 2015-059, Response to Request for Supplemental Information for Relief Request Number RR 4-2 - Proposed Alternative, Use of Alternate ASME Code Case N-770-1 Baseline Examination2015-07-31031 July 2015 Response to Request for Supplemental Information for Relief Request Number RR 4-2 - Proposed Alternative, Use of Alternate ASME Code Case N-770-1 Baseline Examination ML18344A4421993-11-30030 November 1993 Reply to NRC Request for Information Regarding the Pressurizer Safe End Crack Critical Flaw Size and Margin to Failure Analysis. Response to Items 10 and 11 of the Nrc'S October 8, 1993 Information Request ML18346A2931993-09-22022 September 1993 CPC Letter of 7/6/1993, Responding to Inspection Report 93010 & Subsequent Conference Call of 7/22/1993, Letter Submit Supplemental Information to Inspection Report within 60 Days ML18344A2651993-08-16016 August 1993 Response to Request for Additional Information Recent Fuel Failure Event ML18354A6531990-05-30030 May 1990 Information Required by the November 9, 1989 Technical Evaluation Report - NUREG 0737, Item Ii.D.L, Performance Testing of Relief and Safety Valves, Palisades Plant to Close Items Not Fully Resolved ML18354A6151988-01-15015 January 1988 Updated Response to IE Bulletin 87-03 Dated 11/15/1985, Entitled, Motor-Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings ML18348A8881979-05-15015 May 1979 Rapid Response to Additional Information Request on Three Mile Island ML18348A3721978-07-0707 July 1978 Provide Additional Information Related to Diesel Generators Control Circulatory, as Requested ML18348A3741978-07-0606 July 1978 Provide Requested Information of Additional Analysis Specific to Determine Consequences of Potential Boron Dilution Incidents ML18348A7441978-05-23023 May 1978 Response to Request for Additional Information Reactor Vessel Material Surveillance ML18346A1121978-01-24024 January 1978 Response to Request for Additional Information Relating to Water Hammer in Feed-Water Lines and Feed-Water Spargers ML18353B1571977-12-22022 December 1977 Response to Request for Specific Information Re Potential Problem of Post-LOCA Ph Control of Containment Sump Water of IE Bulletin 77-04 ML18347A1711977-09-26026 September 1977 Additional Information Relating to Power Increase Request ML18348A3961977-07-29029 July 1977 Response to Request for Specific Information Concerning Reactor Vessel Materials & Associated Surveillance Programs ML18348A4151977-07-12012 July 1977 Response to Request for Additional Information Re IE Bulletin 77-01, Relating to Use of Pneumatic Time Delay Relays in Safety-Related Systems ML18348A6911977-05-16016 May 1977 Response to Request for Additional Information Alarm and Diesel Generator Control Circuitry ML18348A6921977-05-12012 May 1977 Response to Request for Additional Information Proposed Emergency Dose Assessment System ML18348A8521977-05-0404 May 1977 Advising Exxon Concluded Three of Six Documents Do Not Contain Proprietary Information, Which Was Identified by AEC Letter of 3/1/1977, & Forwarding Affidavit as Additional Information Re Proprietary Documents ML18348A7051977-03-23023 March 1977 Response to Request for Additional Information Environmental Qualification of Electrical Equipment and the Effects of Its Submergence ML18348A8681977-03-0808 March 1977 Letter Reactor Vessel Overpressurization ML18348A8631977-03-0808 March 1977 Response to Request for Additional Information Reactor Vessel Overpressurization ML18348A7121977-02-15015 February 1977 Response to Request for Additional Information Power Defect Value Stated in Report No. XN-76-52 2023-03-01
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General Offices: 212 West Michigan Avenue, Jackson, Michigan 49201
- Area Code 517 788-0550 March 8, 1977 Director of Nuclear Reactor Regulatio Att: JY.ir Albert Schwencer, Chief Operating Reactor Branc.h No 1 US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT - REACTOR VESSEL OVERPRESSURIZATION As requested in your January 10, 1977 letter, one of two proposed implementation options for providing overpressure protection must be selected. As indicat§:d in our Decem-ber 6, 1976 letter, plant modifications will be completed. during our next refueling outage, presently scheduled to start in August 1977. This
- plan is contingent upon a plant specific analysis that indicates that the exist-ing PORVs provide acceptable relief capacity for all potential overpressuriza-tion inc id.en ts.
The design criteria developed during the November 3, 1976 mee-cing with the CE uwn-er' s group and published with* your January 10, 1977 letter seems to be in slight disagreeme:::i.t with our notes of the meeting. The "Single Failure Criteria" and "Seismic Design and IEEE-279 Criteria" are requirements that even the staff were in disagreement on. These requirements if followed completely may lengthen the time to complete long--term fixes. The results of our plant specific analysis and proposed modifications will consider the criteria and provide justification for any variances.
Attacb.."Ilent A is a summary of the questions/answers requested in your LTanuary 10, 1977 letter. Eome of the detailed information cannot be provided prior to com-pletion of the plant specific a~alysis.
It shouid be emphasized that Consumers Power is making every effort to comply with the Tu"R.C req_uests on potential overp:ressuriz.3.tion incidents and has taken corrective actions to eliminate such occurrences by modifying plant operating
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- corrective actions to eliminate such occurrences by modifying plant operating procedures and providing extensive training of plant operating personnel.
David P Hoffman Assistant Nuclear Licensing Administrator CC: JGKeppler, USNRC
-1 ATTACHMENT A The staff considers it essential that all plant operators (ie, reactor operators, equipment operators, Instrument & Control personnel) be made awe.re of the details of the pressure transients which have taken place at all PWR facilities.
POSITION: Formal discussions should be held with the operator to review the causes of past pressure transients that have occurred at other operating PWR facilities. Your discussions should include the plant conditions at the time, the mitigating action that could have been or was taken; and the preventive measures that could have been taken to avoid the event and the steps taken to prevent similar, further occurrences. Plant similarities and distinctions should be identified along with how these relate to plant start-up, shutdown, and test-ing operations. With regard to this position, you are requested to provide the following information:
- a. If you have not already completed the required formal discussion, when will you do so?
- b. How will the discussion be held?
- c. Of the past PWR Appendix G violations that have occurred at PWR facilities and which are described in Licensee Event Reports, identify which are not credible in your plant due to equipment differences. Provide a description of the distinctions.
- d. Describe, in detail, how you are reducing the likelihood of the other remain-ing credible events. Furnish schematics, diagrams or procedural summaries necessary to support the effectiveness and reliability of these measures.
Answer l Informal discussions were held with shift supervisors and licensed operators at the time procedural changes (November l976) were made to minimize overpressuriza-tion incidents. Beginning in February l977 formal training sessions, as part of the normal operator requalification program, were started to review all past PWR overpressurization incidents which might be applicable to the Palisades Plant.
In addition, particular Palisades Plant conditions and operations which might be susceptible to overpressurization incidents were identified and preventive measures discussed. The training sessions will be completed by March l5, 1977 and will have included Palisades Plant auxiliary operators, instrumentation
- technicians, licensed operators and shift supervisors.
i
- The "Generic Report - Overpressurization Protection for Operating CE NSSS" dated December 3, 1976 has identified potential credible events which might cause over-pressurization events in the Palisades Plant. The potential of these events was significantly reduced by procedural changes identified in our November 5, 1976 letter. The plant specific analysis, now in progress, will more completely identify events that the Palisades Plant might be susceptible to.
Question 2 The majority of the reported pressure transients f=Vents have occurred while the plants were operating in a water solid condition.
POSITION: The staff will require that operations <luring which the plant is maintained in a water solid condition be minimized or, if possible, eliminated.
Those operations in which the plant is in a water solid condition must be fully justified. Accordingly, please provide the following information:
- a. Describe the procedures, evolutions or situations that require the plant be maintained in a water solid condition. Also, provide reasons why a nitrogen, air or steam bubble cannot be maintained in these situations .
- b. Include sufficient background or supplementary information such as system diagrams, procedure summaries and descriptions of equipment operation to justify your need for operating the plant in a water solid condition.
Answer 2 The Palisades Plant operating procedures were reviewed and modified on November 1, 1976 to minimize initiating conditions that could lead to overpressurization. As identified in the CE generic report dated December 3, 1976, the following actions lead to the most severe transients:
- a. A primary coolant pump (PCP) start with hot steam generators.
- b. Spurious safety injection signal.
- c. Inadvertent start of a high-pressure safety injection (HPSI) pump.
As previously identified, the Palisades Plant Operating Procedures required that one or more PCP be operated until the steam generators are in equilibrium with the rest of the primary coolant system (PCS). The safety injection bottles are isolated and the fuses for the HPSI pump motors are removed during plant cooldown
- from hot standby (at 1400 psia on PCS).
2
- Alternatives to water solid operation have been identified in the CE generic report and will be further addressed in the final fixes. At present, operation with a nitrogen or air bubble does not appear to be feasible due to chemistry and waste gas handling problems. The plant specific analysis will identify those conditions for which a steam bubble can be maintained.
Question 3 The inadvertent operation of SIS components during the cold shudown conditions has been responsible for a major portion of the overpressure incidents.
POSITION: Based on the licensee submittals, the recent November 3-5, 1976 meet-ings, and discussions with NSSS vendors, the staff will require the de-energizing of SIS pumps and closure of SI header/discharge valves during cold shutdown oper-ations.
In your November 5, 1976 letter to NRC staff, Paragraph 2, you state that the plant HPSI pumps will be disabled and returned to service at the same time as the safety injection tanks. Based on our review of this proposal, and the position stated above, the following additional information is requested:
b.
c.
A schematic diagram of the SIS showing the flow paths into the RCS .
The head flow characteristics of each of the SIS pumps.
Identify on the schematic diagrams the pumps and the valves to be closed and disabled.
- d. How you intend to "disable" a component, and*if this is not de-energized, why power cannot be removed.
- e. Your time schedule for implementing these administrative and operating pro-cedural changes to meet this position.
- f. Indicate all circumstances for which the SIS pumps and valves may not be isolated and disabled, and for those situations, describe the manner in which SIS injection would be prevented.
- g. If it is your intention to place component control switches in the "lockout" or "pullout" position, the location of these switches. Can the components be controlled from any other location? How?
- h. The location of the component supply breakers, and the places from which they can be controlled.
- i. Describe the position indication and status signals which would be lost if the components were (1) de-energized or (2) disabled by placing control switch in "pullout" or "lockout."
3
- j. Describe in detail; the administrative procedures which will be used to assure proper equipment alignment and the supervisory personnel respon-sible for maintaining control.
- k. Indicate the RCS temperature/pressure conditions for which the accumulator isolation valve will be closed. .How will this valve be disabled and if the breaker is not opened, why it cannot be. The locations from which the breaker can be controlled.
- 1. Describe the impact on overall plant operations if you routinely lowered accumulator nitrogen pressure when in a cold shutdown.condition.
Answer 3 Although our November 5, 1976 letter stated that our HPSI pumps are disabled during cold shutdown, our procedures, in fact, require that the fuses for the HPSI pump motors be pulled. This became part of the Palisades Plant Operating Procedures.on November 1, 1976. Detailed answers to Question 3 will be provided with the plant specific analy~is.
Question 4
- The staff has noted that several Appendix G violations have occurred during component or system tests while in cold and shutdown conditions. In this regard, please address the following questions.
- a. What components or systems that could cause overpressure transients are routinely tested while in a cold shutdown condition?
- b. What extra measures are taken to prevent in overpressure event during these tests?
Answer 4 The Palisades Plant's required Safety Injection System (SIS) tests are performed while the plant is at power or in a hot standby condition. The testing includes tests for the correct operation of *all SIS pumps and their control circuits, all
- actuator-operated valves and their control circuits, and operability of numerous check valves as described in Palisades Plant Operating Procedure B3.5. The performance of SIS testing while at power or in a hot standby condition eliminates potential overpressurization incidents during component and system testing .
- Question 5 The staff believes that a high-pressure alarm used during low RCS temperature operations is an effective means to attract the operator's attention to a transient in progress.
POSITION: The staff will require that if it is not presently installed, such an alarm must be installed as soon as possible. Accordingly, please furnish the following information:
- a. Your method to provide the alarm, and the associated time schedule,_ or your justification for why this cannot be done.
- b. A synopsis of the system modifications that are necessar~.
- c. The alarm set point, mode of annunciation and sensor.
- d. How you ensure that the alarm is available and operating properly during all water solid operations and how you minimize its downtime for all other cold shutdown conditions.
Answer 5
- The Pali.sades Plant does not currently have fhe capability of providing a low-pressure alarm during low PCS temperature operation. It is intended as part of the.final design modification for pressure protection to provide a single set
- point alarm at approximately 400 psi when the PCS is below 200°F. The final design details will be provided With the completion of the plant specific analysis.
Question 6 The RHR (or SCS) is normally connected to the RCS and operating when the plant is in a cold shutdown condition. The inadvertent isolation of the RHR system while water solid has caused a number of overpressure transients, and the RHR safety valve has actually terminated others. The RHR (or SCS), therefore, plays an important part in the initiation and possible mitigation of the PWR overpres-surizations. Accordingly, we retj_uest the following additional information.
- a. RHR (or SCS) design pressure.
- b. A description of the system isolation valves and their arrangement (eg, number and configuration of valves installed, and pneumatic or motor operated).
- c. Interlocks, interlock set points and alarms associated with each isolation valve.
5
e.
Nominal stroke time of isolation valves.
The set point ~nd capacity of RHR (or SCS) relief and safety valves.
- f. All pressure alarms, set points and associated annunciation for the system.
Answer 6 The Palisades Plant does not have a pressure interlock on increasing pressure while operating in the shutdown cooling mode. The following information is provided for the shutdown cooling system.
Shutdown Cooling System
- 1. Maximum operating pressure = 270 psia and maximum operating temperature =
325°F per Plant Operating Procedures, Section B3.4.l.2.b. Also, see FSAR Table 6-3 on Page 6-16 for design data on SDC heat exchangers. See P&ID Drawing M-204 for SDC system schematic.
- 2. SDC relief valves.
NOTE: SDC uses LPSI pumps for coolant circulation, returns cooled water to the PCS through safety injection lines.
- a. Suction relief valve:
- RV3164 on P&ID M-204, drawing location D&E-2.
Set to relieve @ 300 psig.
- Relief flow capacity - 133 gpm .
. b. Discharge relief valve (this also LPSI relief valve):
- RV3162 on P&ID M-203, drawing location C&D-1.
- Set to relieve @ 500 psig.
- Relief flow capacity =5 gpm.
- 3. Yes* - The SDC suction valves have an isolation interlock. This interlock only operates when pressure is decreasing, not when it is increasing. There are two such valves in'series, MOV-3015 and MOV-3016, as shown on P&ID drawing M-204 at drawing location D&E-1&2.
- a. The suction valve interlock is automatic and it operates off of PS-0103 which is on the pressurizer per P&ID drawing M-201 @ location F-3.
- b. The isolation valve interlock pressure set point is 260 psig on PS-0103.
This interlock only operates on decreasing pressure.
- c. The closure time of the MOVs is 45 seconds .
6
5.
The LPSI pumps shutoff head (psig) = 410 ft= 177.7 psig:
- Per FSAR Table 6-1 on Page 6-14.
The containment spray pumps shutoff head (psig) = 500 ft = 216.8 psig:
- Per CP Co File M8, Sheets 23, 25 and 27, which are from the preoperation pump test characteristic *curves.
Question 7 We have reviewed your letters dated October 7 and November 5, 1976, and the interim measures described. Paragraph 1 of your November 5 letter states your 1
requirement to run at least one PCP until the loops are essentially isothermal and the steam generator shell pressure is essentially atmospheric. Regarding this procedure:
- a. What is the status of primary coolant flow in the other loops if only one PCP is in operation?
- b. Provide diagrams and schematics to show that the steam generator shell side cannot be subsequently heated after completion of cooldown. Signify the valves that are closed, pumps that are disabled, and other flow or tempera-ture limiting devices .
d.
What instrumentation is available to measure the RCS temperature conditions?
What methods are used to prevent a nonisothermal RCS from developing?
Provide system diagrams and schematics to illustrate the technique, and include procedure descriptions or summaries.
Answer 7 The actual primary to secondary temperature and pressure differences for a PCP start during water solid conditions will be addressed in the plant spec~fic analysis.
As relates to PCP operation, the following information is provided:
- a. With all 4 PCPs operating, the average flow per loop/pump is approximately 85,000 gpm.
- b. With only 1 PCP operating, the flow through the operating pump is approximately 123,000 gpm. The nonoperating pump on the loop with the operating pump ex-periences approximately 20,000 gpm reverse flow resulting in a flow of ap-proximately 103,000 gpm through the steam generator. The backflow through
l t'-
- each of the other nonoperating pumps is about 4,700 gpm.
through the nonoperating steam generator is_ about 9, 400 Therefore, the flow gp~. The flow paths, equipment and instrumentation are detailed on Palisades Plant P&ID Numbers M-20l ~nd_M-207.
Question 8 We have reviewed your proposed reliance on the Shutdown Cooling System(SCS) relief valve during an inadvertent .isolation of the letdown valve. The following additional information is requested.
- a. What alarms are available to the control room.personnel that indicate the SCS relief's opening?
- b. What are the testing requirements of this valve to ensure its proper opera-tion?
- c. What is the anticipated RCS pressure overshoot while the safety valve is opening and the solid system is being pressurized?
Answer 8
- As indicated in our November- 5, 1976 letter to the NRC, a pressure transient caused by an inadvertent isolation of the letdown control valve while operating in a shutdown cooling mode would be limited to approximately 300 psi. Relief capacity of 133 gpm, equivalent to the capacity of all three charging pumps, is provided by the Shutdown .Cooling Relief Valve (RV-3164). This relief valve is a totally enclosed, bellows type valve and the set point is checked and reset, if necessary, as part of the Palisades Plant Inservice Testing of Plant Valves (Engineering Manual Procedure EM-09-02 dated October 29, 1976).
Additional analysis of PCS conditions during a water solid transient will be provided with the plant specific analyses .