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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
[Table view] |
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consumers Power G BSlade
<nrieral Manarrr PawERING~
.MICHlliAN"S PROliRESS
.Palisades Nuclear Plant: 27780 Blue Star Memoriai Highway, Cowrt. Mt. 49043
. September 22, 1993 Nuclear Regulatory Commission Document Control Desk Washington, DC 2055~
DOCKET 50-255 - LICENSE DPR PALISADES PLANT - SUPPLEMENT TO THE JULY 6, 1993 RESPONSE TO INSPECTION REPORT No.*~3010:
- On July 6, 1993, Consumers Power.Company submitted a response to Inspection R~port 93010. That response.was discussed during.a cdnference telephone call p~tween with members of the NRC Region III and Palisades staffs on July 22, 19~3. During that* conference call, the* NRC membefs clarified several items where:they f~lt that the CPCo response did not ful.ly an~~e~ the ~nderlyi~9.
ccincerns: *It was agreed that a supplemental response to the I~sp~ction Report would be subm.itted within 60 days*; thjs *letter provides that s*upp.lement.
We understand .the NRC cpncerns associated with ~he subject inspection report, which w~re not fully. covere_d by our July.6, 1993*response, to be as follows:
1.. : The.Emergency Operating Procedures (EOPs} m~st provjde cl*ar and timely instruction for entry fotoOnce-thr*ough Cooling.-(OTC}; ~hen*it :is required.
Operation with closed Power Operated .Reli*ef Valve (PORV.) block valves, which are not environmentally qualified. for th.e environment following*
the**~vent, combined with delay in initi~ting OTC ~ould eliminate the ability to cool the core.*
The Primary Coolant System (PCS} conditions caused-'bY the event may "uncouple" the Steam Generators (SGs} from the Reactor and *inhibit
- initiation of natural circulation.
- 2. Simulator modeling of containment response does not a*gree. with previous analyses. (Open Item 93010-05}
- Simulator model*ihg res~lts, which di.ffe~ fro~ ~afety a~alys~s results,
.may provide the operators with improper expectaticms. of plant responses to this and other events which release large amounts of energy to the containment.
\_ . , A CMS ENERGY COMPANY
- Training on the event should be enhanced. (Open Item 93010-03)
No operator training has been provided on the specific event
. concerned, steam line break inside the containment with a concurrent failure of th~ Main Steam Isolation Valve (MSIV) on the opposite steam line (hereinafter, the "event"). . . * * .
No operator training has been provided on determining operability of instrumentation which may be adversely affected by its environment.
- 4. An additional concern, not specifically mentioned in the inspection report, is. that while the EOPs. provide information for correcting the narrow range SG level instruments for potential errors due to-adverse containment environment, no similar correcti~ns are provided for the wide rang~.
instruments. The wide range instruments are those *used to verify that adequate level is available for cooli~g the core with the steam generators.
Several corrective actions were initiated as a result of the iubject inspection report. These actions were discussed in our July 6, 1993 response but, as evidenced by the NRC's continued concerns, not in sufficient detail.
Analysis and fvaluations The following analyses and evaluations have been completed in order to assure that EOP guidance and training inaterials.for the subject event are appropriate:
- 1. *The cbrrective.actions related *to appropriat~ EtiP guidance, event specifi~
operator training, and verification of simulator modeling all must be -based on a clear understanding of the expected plant response to the event.
Analyses have been* pe~formed, using a Consumer~ Power Company version of the MAAP code, CPMAAP, to provide a best engineering estimate .of PCS,. SG,
- ind *Cori ta i nme-nt response to t.he event.
- These analyses ~ere performed with several varied parameters to determine which parameters had significant *.
effects on the plant response and which did not~* Examples of these variations include: Immediate tripping, delayed tripping, and continuous running of the PCPs 1 variations in the amourit of containment ~ooling
- equipment available, and variations in the modeled break size. In addition*
cases were run, using the sami p~rameters, comparing .a larg* steam line break with and without the MSIV failure. These different analyses, all run
- with the same code changing only a single parameter, allow direct comparisons between cases. That comparison would not be valid if made between analyses done using different codes or ~ifferent basic assump~ions.
These analyse*s allow for verification that EOP strategy is*appropriate,-
_provide a basis for operator training,* and may be used as an alternate calculation.method for comparison with simulator modeling. Typical safety analyses are not always appropriate for these uses since their function is simply t6 demonstrate that the results of particular events ~ill Femain within design or regulatory limits. The simplifying and bounding assumptions made in typical safe~y analyses, while conservative with respect to the analytical goal, often make the results far different from expected plant response.
3*
The containment temperature and pressure results of these best engineering estimate analyse~ have been compared to the environmental qualification t~sting of the PORVs, the PORV block valves, and the wide range SG level transmitters.* The PORVs were successfully tested at conditions exceeding.
those in the calculated containment response; testing on the block valves and level transmitters, combined with thermal lag calculations imply that there is a high probability of this equipment re~aining operable. These items were chosen as the prime equipment, ~ithin the cont*inment, to assu~e core cooling with either the SGs or OTC.
- 3. The ability to cool the core using delayed once through cooling has.been analyzed. This ~nalysis concluded that OTC would be successf~l with one*
PORV flow path ~pen and either two charging pumps or one HPSI pump in .
series with one spray pump. The analysis assumed that OTC was initiated with the SG dry and the PCS above saturation temperature .for the set point .
of the secondary safety valves, about 545°F. With two PORV paths or two HPSI pumps available, the initial PCS temperature could be signifitantly hotter.
- 4. The ability to cool the core by usfng AFW, even after SG "dryout", has been analyz~d. The results show that a single AFW pump can provide enough makeup flow to maintain natural circulation in the primary coolant system.*
Again, the availability of additional pumps provides additional margin and more rapid cooling. *
- 5. . The ~otential er~brs in wide range indicated SG level *are included in the existing EOPs, although not through use of a ~orrection curve. - Instead, the specified instrument reading which corresponds to the minimum acceptable level for secondary cooling, -84%, has the maximum predicted error included. Engineering Analysis EA-GAW.:.89-EQ-Ol, Revision 1, concludes that .the maximum expected error for .the wide range SG level
- in~truments would b~ 36% ... When ~ 36% error is added to the minimum actual level of ~120%~ fhe specified minimum indicated level, -84%, is ~ttained.
EOP Strategy Review:
The strategy of the EOPs, with respect to initiation of OTC, has been rev.iewed. The existing strategy is, very briefly,_ to allow automatic initiation of feedwater, *ensured by manual action,_ with acceptable cool i.ng.
verified by SG level *and PCS conditions. If continued use of the SGs for decay heat removable is not possible, OTC.would be initiated.
This current strategy was compared to an alternate strategy of immediately initiating OTC upon observing the symptoms of a steam line break concurrent with the lack of full closed indication on an MSIV. The overriding consideration, bf course, is that the ch6seri strategy be capable of.
assuring adequate core cooling. Our conclu~ion is that either method would result i~ continued core cooling. Additional considerations, discussed
- below, result in our decision to retain the current strategy.
4
- -* _. Immediate Initiation of Once Through Cool inq:
The considerations which tend~d to favor the alternate appr~ach 6f
. invnediate initiation of OTC were as follows:
a} Early analyses of use of OTC following a loss of all feedwater event, which assumed flow through an area equal to two smaller PORVs similar to those formerly installed at Palisades~*concluded that i nit i at ion of OTC must occur before SG dryout. Dryout was ,
predicted to occur about 20 ~inutes into the event. This ~esult, that OTC must be initiated within 20 minutes to be assured of succ~ss, was often considered to apply to other events requiring OTC. .
this consideration is rio longer appropriate.for Palisades. lhe PORVs now installed at Palisades are significantly larger than those used in the earlier study.* As mentioned above, analyses of.
~he currently installed PORVs and.flow paths show more capability with a single flow path than the former analyses did with both.
The blowdown of both SGs could cause the containment e~vironment to exceed the environmental qualification envelope for electrical/
equipment preventing verification that AFW is functioning and
- preventing opening of the PORVs and block valves to achieve OTC._
- This consid~ration is now less important than it w~uld have been prior to the installation of the new PORVs~ which are qualified
- for the containment :environmental conditions resulting from design events, .and the new SGs which incorporate flow r~strictors in their outlet nozzles. As discussed in the analy~is section above, the environmental testing envelope for the PORVs exceeds
. the predicted containment response for the."event"; the
- combination -0f environmental testing and thermal. lag cal~ulations
- indicate that the block valves and tha wide range SG le~el . *
- t~ansmitters should ~urvive the "event".* Therefore it is hi~hly probable that delaying the initiation of OTC while. the succes$ful initiation of AFW is being verified will not add significantly:to the risk ~f failing to maintain core cooling.
Initiation of Once Through Cooling onlv upon failure Coolirig using the Steam Generator: * *
- Those considerations which favor *the. current EOP strategy are as follows:
- a) The ~urrent strategy applies to any Ex~essive Steam Demand Ev~nt
~nd does not req~ire a special procedure, or special steps, for the subject event. The design concept of new EOPs is to avoid event ba~~d actions with special procedures for ea~h possible event.
. ' 6 Class.room trainfog on the "event" is also currently scheduled to be completed prior to fhe end of 1993-. -Class room training will foclude discussion on the followi~g:
a) .How a blowdown of both SGs could occur b} ~n explan~tion of why there are differences between safety analyses and* simulator modeling of some events
- c} Discussions of symptoms, expected plant response, the*EOP paths invblved, and the potential for significant error ~r failure of i~strumentation located* in the containment.
d} Discussions* on verificati~n of instrument.reading validity and use of alternate instrumentatiOn* for* this a.nd other events which degrade the
- containment environment.**
- When the neces~~ry simulator modeling corrections are c.ompleted, the
- details of those EOPs ~ssociated with the "event" will be* validated.
Add~tional simulator training on* a full r~nge oi steam line break sizes,
- ~ith and*withput a co~iurrent failure of a MSJV,* will then* be included in the tr~ining curriculum. The "event,"~ l~rge break in one main st~ani line with a concurrent failure of the opposite MSIV, ,ca.nnot be exactly modeled *
. on the simulator~ What ~ari be.modeled, closely simulating th~ "event," is a combination of the following: a.large *break in one main stea~ line, a somewhat smaller break (to eniulate'the piping fl-0w losses} in the other steam line, and a fail~re of closed.indication on one MSIV .
. *.'I( . * ' -
Plant General Manager CC: Regfon III Administratot .
Palisades -Resident Inspector
5
- . b) The current strategy utilizes the defense in depth concept, by relying on the automatically initi~ted AFW system first. If cooling by_ AFW cannot be assured, then other cooling methods are
_employ~d; condensate pumps if available~ and then OTC.
The. initiation of OTC would inunediately reduce the PCS to saturation pressure, forming voids in the SG tubes and reactor vessel head, and reduce PCS inventory. *A subsequent failure of the PORVs or of the HPSI pumps would necessitate returning to .
cooling by natural circulation and SGs. Reduced-PCS inventory,*
additional PCS voids, and restarting or realignment of equipment used with the SGs add additional failur~ possibilities which are not.encountered with the ~urrent strategy.
A very simplistic fault tree an:alysis of the two choices_iinplied that there would be a reduction in risk of loss of core cooling.
- for th~ "event" of about an ordei of magnitude using the current*
strategy.* It is assumed that the early initiation of ~TC would adversely affect the failure pro_babil i ty of secondary coo 1i rig by a factor of 10, and that ~elayed initiation of OTC would adversely affect the failure probability of the block valves by
- the same amount (since the block valv_es typically fail as is,
- failures occurring after they are op~n are inconsequential).
Unaffected failure rates were set at 10*3 and degraded rates* at 1~ 2
- These ~aies were chosen'.simply to examine the effects of the two alternate choice~. The chosen rates have no analytical
~ ..-._ basis, but they.are not atypical, either. Since either cooling method transfers the decay heat to the containment, other failure rates, including that of the PORVs would b~ unaffected;
'.,.:.
- c) The current st~ategy avoid~ co~po~nding a steam line break event with a Loss of Cool*ant event. Such a ~ompound eveht is not within-the d~sign base of the plant~ **.
Jn sunvnary, the current EOP strategy of using OT( only if cooling using S.Gs cannot be verified is preferred to inunediate initiation of OTC. This choice reduces the risk of loosing the ability to cool the core, does not further compound an already co~plicated event, and conforms to the_appro~ed guidance for CE plant EOPs.
Actions Planned to. Address Training Issues of Ooen Items:
Several actions have been assigned to assure that appropriate op~rator training on the event is provided.
- A comparison is being* completed between simulator, CPMAAP, and Safety Analysis calculations for containment response to the event. Those corrections necessary for proper simulator modeling_are currently-scheduled to be completed prior to the end of 1993.