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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20196K4821999-07-0606 July 1999 Safety Evaluation Concluding That Because of Permanently Shutdown & Defueled Status of Myaps Facility,Confirmatory Orders No Longer Necessary for Safe Operation or Maint of Plant ML20206H1611999-05-0505 May 1999 Safety Evaluation Supporting Amend 164 to License DPR-36 ML20206G5731999-05-0303 May 1999 Safety Evaluation Supporting Amend 163 to License DPR-36 ML20205D5261999-03-26026 March 1999 SER Accepting Util Rev 1 to CFH Training & Retraining Program for Maine Yankee.Rev 1 to CFH Training & Retraining Program Consistent with Current Licensing Practice for Facilities Undergoing Decommissioning ML20204C4631999-03-16016 March 1999 Safety Evaluation Supporting Amend 162 to License DPR-36 ML20197C8231998-09-0303 September 1998 Safety Evaluation Supporting Request for Exemption from Certain Requirements of 10CFR50.54(q),10CFR50.47(b) & (C) & App E to 10CFR50 Re Emergency Planning ML20236U8451998-07-24024 July 1998 Safety Evaluation Accepting Rev 14 to Maine Yankee Atomic Power Co Operational QA Program ML20217H5241998-03-30030 March 1998 Safety Evaluation Supporting Amend 161 to License DPR-36 ML20202D3181997-11-26026 November 1997 Safety Evaluation Supporting Amend 160 to License DPR-36 ML20216F1511997-08-0808 August 1997 Safety Evaluation Supporting Amend 159 to License DPR-36 ML20141G6511997-05-19019 May 1997 Safety Evaluation Supporting Amend 158 to License DPR-36 ML20138G3541997-05-0202 May 1997 Safety Evaluation Supporting Amend 157 to License DPR-36 ML20058E2641993-11-0505 November 1993 Safety Evaluation Supporting Amend 143 to License DPR-36 ML20056F7831993-08-23023 August 1993 Safety Evaluation Supporting Amend 142 to License DPR-36 ML20127D5061993-01-11011 January 1993 Safety Evaluation Supporting Amend 136 to License DPR-36 ML20126H6211992-12-29029 December 1992 Safety Evaluation Supporting Amend 135 to License DPR-36 ML20059K9991990-09-20020 September 1990 SER Accepting Methodology Re Statistical Combination of Uncertainties for RPS Setpoints ML20059H5901990-09-12012 September 1990 Safety Evaluation Re Facility Response to Station Blackout Rule.Issue of Conformance to Station Blackout Rule Will Remain Open at Facility Until Identified Nonconformances Resolved ML20059D0501990-08-30030 August 1990 Safety Evaluation Supporting Amend 117 to License DPR-36 ML20055C2911990-02-20020 February 1990 Safety Evaluation Accepting Util Assessment of Asymmetric LOCA Loads Problem ML20246F6281989-07-10010 July 1989 Safety Evaluation Supporting Amend 113 to License DPR-36 ML20245J0571989-04-25025 April 1989 SER Re Accepting Facility Emergency Response Capability in Conformance to Reg Guide 1.97,Rev 3,w/exception of Instrumentation Re Variables Accumulator Tank Level & Pressure & Containment Sump Water Temp ML20245D4481989-04-24024 April 1989 Safety Evaluation Supporting Amend 111 to License DPR-36 ML20245D4351989-04-24024 April 1989 Safety Evaluation Supporting Amend 112 to License DPR-36 ML20155B4631988-09-27027 September 1988 Safety Evaluation Supporting Amend 107 to License DPR-36 ML20154A1411988-09-0707 September 1988 Safety Evaluation Supporting Amend 106 to License DPR-36 ML20150D7681988-07-0707 July 1988 Safety Evaluation Approving Util Inadequate Core Cooling Instrumentation Sys Contingent on Completion of Emergency Procedures & Operator Training & Submittal of Applicable Tech Specs ML20153A9641988-06-28028 June 1988 Safety Evaluation Accepting Util Proposed Reflood Steam Cooling Model ML20196G6871988-06-23023 June 1988 Safety Evaluation Supporting Amend 105 to License DPR-36 ML20151W5861988-04-26026 April 1988 Safety Evaluation Supporting Amend 104 to License DPR-36 ML20153B3381988-03-16016 March 1988 Safety Evaluation Supporting Util 831110,840216,0412,1214, 850618 & 0820 Responses to Generic Ltr 83-28,Item 2.2.2, Based on Stated Util Commitments,Insp Rept 50-309/86-07 & Actions Described in Procedure for Vendor Interface ML20149H2831988-02-17017 February 1988 Safety Evaluation Supporting Amend 103 to License DPR-36 ML20196C4301988-02-0909 February 1988 Safety Evaluation Supporting Amend 102 to License DPR-36 ML20235M4661987-09-29029 September 1987 Safety Evaluation Supporting Amend 101 to License DPR-36 ML20236F6441987-07-29029 July 1987 Safety Evaluation Supporting Util Response to Item 2.1 of Generic Ltr 83-28 ML20234B1781987-06-25025 June 1987 Safety Evaluation Supporting Amend 100 to License DPR-36 ML20216J0371987-06-25025 June 1987 Safety Evaluation Supporting Amend 99 to License DPR-36 ML20215E9301987-06-15015 June 1987 Safety Evaluation Supporting Amend 98 to License DPR-36 ML20214W3841987-06-0404 June 1987 SER Supporting Util 831110 Response to Generic Ltr 83-28, Item 2.2 (Part 1) Re Requirement for Program Description to Ensure All Components of safety-related Sys Identified as safety-related on Informational Matls ML20214M1491987-05-21021 May 1987 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Item 2.1 (Part 2) Re Established Interface W/Nsss or Vendors of Each Component of Reactor Trip Sys ML20214J6001987-05-20020 May 1987 Safety Evaluation Supporting Amend 97 to License DPR-36 ML20210C6631987-04-23023 April 1987 Safety Evaluation Re Adequacy of Offsite Power Sys.Further Evaluation Cannot Continue Until All Requested Info Received.Surowiec Line Not Acceptable Substitute for Mason Line During Startup or Normal Operations ML20206U9231987-04-0909 April 1987 Safety Evaluation Supporting Amend 96 to License DPR-36 ML20197D4421987-04-0707 April 1987 Safety Evaluation Supporting 850807 Request for Relief from ASME Code,Section XI Inservice Testing Requirements for Pump & Valves ML20205M6391987-03-26026 March 1987 Safety Evaluation Accepting Upgraded Seismic Design Program Contingent on Licensee Commitment to Upgrade Items in Table 3 ML20197D4241987-03-26026 March 1987 Safety Evaluation Supporting Amend 94 to License DPR-36 ML20212N7091987-03-0404 March 1987 Safety Evaluation Supporting Amend 93 to License DPR-36 ML20211G4361987-02-14014 February 1987 Safety Evaluation Supporting Util 831110 Response to Generic Ltr 83-28,Item 4.5.2 Re on-line Testing for Reactor Trip Sys Reliability ML20210V1831987-02-0909 February 1987 Safety Evaluation Supporting Amend 92 to License DPR-36 ML20207N2181987-01-12012 January 1987 Safety Evaluation Supporting Util Response to NRC 860721 Request for Addl Info Re Reactor Coolant Pump Shaft Integrity,Per IE Info Notice 86-019 1999-07-06
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20211M4841999-08-31031 August 1999 Replacement Pages 2-48,2-49 & 2-50 to Rev 14 of Defueled Sar ML20211D7111999-08-0909 August 1999 Rev 17 to Maine Yankee Defueled Safety Analysis Rept (Dsar) ML20196K4821999-07-0606 July 1999 Safety Evaluation Concluding That Because of Permanently Shutdown & Defueled Status of Myaps Facility,Confirmatory Orders No Longer Necessary for Safe Operation or Maint of Plant ML20206H1611999-05-0505 May 1999 Safety Evaluation Supporting Amend 164 to License DPR-36 ML20206G5731999-05-0303 May 1999 Safety Evaluation Supporting Amend 163 to License DPR-36 ML20205D5261999-03-26026 March 1999 SER Accepting Util Rev 1 to CFH Training & Retraining Program for Maine Yankee.Rev 1 to CFH Training & Retraining Program Consistent with Current Licensing Practice for Facilities Undergoing Decommissioning ML20204C4631999-03-16016 March 1999 Safety Evaluation Supporting Amend 162 to License DPR-36 ML20206D7491998-12-31031 December 1998 Co Annual Financial Rept for 1998. with ML20155G9591998-11-0303 November 1998 Rev 1 to Post-Shutdown Decommissioning Activities Rept ML20155D8651998-10-28028 October 1998 Public Version of, Maine Yankee Emergency Preparedness Exercise ML20197C8231998-09-0303 September 1998 Safety Evaluation Supporting Request for Exemption from Certain Requirements of 10CFR50.54(q),10CFR50.47(b) & (C) & App E to 10CFR50 Re Emergency Planning ML18066A2771998-08-13013 August 1998 Part 21 Rept Re Deficiency in CE Current Screening Methodology for Determining Limiting Fuel Assembly for Detailed PWR thermal-hydraulic Sa.Evaluations Were Performed for Affected Plants to Determine Effect of Deficiency ML20236X1751998-08-0303 August 1998 Rev 16 to Defueled Sar ML20236U8451998-07-24024 July 1998 Safety Evaluation Accepting Rev 14 to Maine Yankee Atomic Power Co Operational QA Program ML20248D3011998-05-26026 May 1998 Rev 14,page 16 of 17,Section II of QA Program ML20247J0211998-05-11011 May 1998 Revised Page 16 of 17 of Section II of QA Program,Rev 14 ML20247D3451998-05-0606 May 1998 Rev 15 to Defueled SAR, Replacing List of Effective Pages ML20217J9811998-04-28028 April 1998 Part 21 Rept Re Incorrect Description of Drift Specification for Model 1154,gauge Pressure Transmitters,Range Code 0 in Manual Man 4514,Dec 1992.Cause Indeterminate.Will Issue & Include Errata Sheets in All Future Shipments to Users ML20217H5241998-03-30030 March 1998 Safety Evaluation Supporting Amend 161 to License DPR-36 ML20217D9691998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Maine Yankee.W/ ML20216D7681998-02-25025 February 1998 Rept to Duke Engineering & Services,Inc,On Allegations of Willfulness Related to Us NRC 971219 Demand for Info ML20203A3011998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Maine Yankee ML20202F3771998-01-31031 January 1998 Annual Rept of Facility Changes & Relief & Safety Valve Failures & Challenges ML20202E0541998-01-30030 January 1998 Rev 14 to Myaps Defueled Safety Analysis Rept ML20199K3211998-01-27027 January 1998 Rev 13 to Maine Yankee Atomic Power Co,Qa Program ML20199K3471998-01-22022 January 1998 Rev 14 to Maine Yankee Atomic Power Co QA Program ML20199C2281997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for Maine Yankee ML20217R2301997-12-31031 December 1997 Myap Annual Financial Rept for 1997 ML20203F8551997-11-30030 November 1997 Monthly Operating Rept for Nov 1997 for Maine Yankee ML20202D3181997-11-26026 November 1997 Safety Evaluation Supporting Amend 160 to License DPR-36 ML20198R6371997-11-0606 November 1997 Yankee Mutual Assistance Agreement ML20155G9511997-10-31031 October 1997 Rev 1 to M01-1258-002, Decommissioning Cost Analysis for Myaps ML20198P9431997-10-31031 October 1997 Monthly Operating Rept for Oct 1997 for Maine Yankee ML20217K5191997-10-24024 October 1997 Part 21 Rept Re Five Valves That May Have Defect Related to Possible Crack within Forging Wall at Die Flash Line.Caused by Less than Optimal Forging Temperatures.Newer Temperature Monitoring Devices at Forging Area Heating Ovens Procured ML20211N0571997-10-0707 October 1997 Revised Pages to Jul/Aug 1994 SG Insp Summary Rept ML20198K0201997-09-30030 September 1997 Monthly Operating Rept for Sept 1997 for Maine Yankee ML20199H1861997-09-25025 September 1997 Rev 12 to Maine Yankee Atomic Power Co,Qa Program ML20217A9251997-08-31031 August 1997 Monthly Operating Rept for Aug 1997 for Maine Yankee ML20217R1051997-08-27027 August 1997 Myaps Post Shutdown Decommissioning Activities Rept ML20216F1511997-08-0808 August 1997 Safety Evaluation Supporting Amend 159 to License DPR-36 ML20210M4451997-07-31031 July 1997 Monthly Operating Rept for July 1997 for Myaps ML20151M1351997-07-21021 July 1997 Rev 0 to Technical Evaluation 172-97, Cable Separation Safety Assessment Rept ML20141H2691997-06-30030 June 1997 Monthly Operating Rept for June 1997 for Maine Yankee ML20141B9951997-05-31031 May 1997 Monthly Operating Rept for May 1997 for Maine Yankee.W/ ML20141G6511997-05-19019 May 1997 Safety Evaluation Supporting Amend 158 to License DPR-36 ML20138G3541997-05-0202 May 1997 Safety Evaluation Supporting Amend 157 to License DPR-36 ML20141G2871997-04-30030 April 1997 Monthly Operating Rept for Apr 1997 for Maine Yankee ML20137Z1971997-04-14014 April 1997 Forwards to Commission Results of Staff Evaluation of Performance of Licensees W/Ownership Structure Similar to Plant ML20137N7541997-03-31031 March 1997 Rev 11 to Operational Quality Assurance Program ML20138B1491997-03-31031 March 1997 Monthly Operating Rept for Mar 1997 for Maine Yankee 1999-08-09
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SAFETY EVALUATION BY OFFICE OF NUCLEAR REACTOR REGULATION MAINE YANKEE ATOMIC POWER COMPANY MAINE YANKEE ATOMIC POWER STATION DOCKET NO. 50-309 LARGE BREAK ECC5 EVALUATION MODEL MODIFICATIONS RELATED TO AXIAL POWER SHAPE ISSUE PHASE I
- 1. INTRODUCTION In a telecon on September 2, 1986, the licensee, Maine Yankee Atomic Power Company (MYAPCo) informed us of a potential non-conservatism in their large break LOCA analysis. In a meeting on September 9, 1986, Yankee Atomic, which performs Maine Yankee's fuel reload analysis, described how they had discovered that the highly peaked axial power distribution used in Maine Yankee's large break LOCA analysis since 1977 is not the bounding shape.
Appendix K to 10 CFR Part 50 requires that the axial shape which results in the most severe consequences should be used in the ECCS evaluation model LOCA calculations. It was determined by Yankee Atomic that the bounding shape is a flattened shape which results in higher peak cladding temperatures, especially later in the life of the reload. For the balance of Cycle 9, administrative limits on power peaking factors are being implemented to assure that results of LOCA analyses done with the current model comply with the limits of 10 CFR 50.46. For Cycle 10, MYAPCo has proposed changes in their approved ECCS evaluation model (EM) to recover the margin lost by utilizing the more severe flattened axial shape. The modifications, review, and analysis will be a two-step process. This SE addresses the changes in Phase I which are applicable to the analysis to be performed for the first half of Cycle 10.
The changes involve the selection of appropriate power shapes, and modification of the injection AP penalty.
l 2.0 POWER SHAPE SELECTION l In Reference ?, the licensee proposed a method for selecting appropriate l radial and axial power distributions to be used in LOCA analyses to assure l compliance with paragraph I. A. of Appendix K which states:
A range of power distribution shapes and peaking factors representing power distributions that may occur over the core lifetime shall be studied and the one selected should be that which results in the most severe calculated consequences, for the spectrum of postulated breaks and single failures analyzed.
Power shapes must also be selected so as not to violate Specified Acceptable Fuel Design Limits (SAFDLs) and they must also be calculated based on nuclear design parameters and possible operating conditions for the particular cycles under consideration. The radial power distribution (maximum pin power) is 10 hD 519 870106 p OCK 05000309 PDR
i determined for a variety of cycle conditions for thermal margin considerations.
, The maximum value is selected for the hot-test pin in the LOCA analysis. The corresponding radial power for the hot assembly is selected, and all uncer-tainties are included to maximize hot pin, hot assembly, and average core power
- which are used in LOCA analysis.
With the pin power established for all LOCA analyses as a single maximum value, several axial shapes are next selected for thermal margin based on core power and symmetric offset considerations. This results in a symetric offsetcurveoflinearheatgenerationrate(LEGR)versuscoreheight.
This curve results in fairly limiting kw/ft values at elevations below the mid plane.
Using a representative selection of axial power shapes determined for the thermal analysis, several large break LOCA analyses are performed to assure compliance with 10 CFR 50.46 and Appendix K. It is verified that these shapes include flatter, more symetric shapes than were previously considered. LOCA analyses performed with this set of axial power shapes result in a LOCA limit curve which is more limiting than the thermal margin curve at higher. elevations.
At lower elevations the thermal margin curve is so clearly limiting that bottom peaked LOCA analyses are not required. This is particularly true since periods of core uncovery for both large and small breaks are much longer at higher elevations, thus making top skewed or symmetric power shapes more limiting than bottom skewed for LOCA analysis.
Appropriate monitoring is performed to verify that the radial and axial components of power distribution are maintained below the limiting values as described above, 3.0 INJECTION AP PENALTY Section I.D.4 of Appendix K to 10 CFR Part 50 requires that the thermalhydraulic interaction between steam and all emergency core cooling water shall be taken into account when calculating core reflooding rates. The currently approved ECCS evaluation rodel for the Maine Yankee plant utilizes an additional frictional pressure drop (AP penalty) to account for the steam-water inter-
, action effect. A AP penalty of 1.5 psid is utilized during the accumulator injection period; a penalty of 0.8 psid is used during the pumped injection
- phase.
In its November 10, 1986 letter, MYAPCo proposed to modify the ECCS evaluation model for the Maine Yankee plant by reducing the AP penalty from 0.8 psid to 1
0.15 psid during the pumped injection phase. This section presents the staff's findings on the proposed modification.
i 3.1 EVALUATION OF AP PENALTY Since the time that the AP penalty used in the ECCS evaluation model for Maine l Yankee was approved, tests have been performed by EPRI to examine the effects of steam-water interaction in the cold legs of a PWR. These tests were per-
'i formed with 1/14 and 1/3 scale cold leggeometries and are discussed in EPRI reports EPRI-294-2, " Mixing of Emergency Core Cooling Water With Steam: 1/14 Scale Testing Phase," dated January 1975, and EPRI-294-2, " Mixing of Emergency Core Cooling Water With Steam: 1/3 Scale Test and Sumary," dated June 1975.
The testing program examined the cold leg pressure drop during both the accumulator injection phase and the pumped injection phase of the reflood portion of a large break LOCA.
In its November 10, 1986 letter, MYAPCo reported the results from all the EPRI tests performed to examine the pumped injection phase. Using the measured cold leg pressure drop data, the AP penalty associated with the steam-water inter-action effect was derived by subtracting out the piping frictional pressure drop. MYAPCo's examination of the data indicated that the AP penalty could be bounded by 0.15 psid. Only one of the 131 data points was above 0.15 psid, and it did not replicate; the duplicate run for that data point had a pressure loss of 0.06 psid.
4 MYAPCo also examined the calculated reflooding parameters for the Maine Yankee plant and compared them to the EPRI test ranges. All the calculated conditions -
were within the range of conditions examined during the EPRI tests.
The staff has previously reviewed the use of the EPRI test data for determining the AP penalty as part of the staff's review of the Exxon Nuclear Company ENC-WREM-IIA ECCS evaluation model and found use of the EPRI data acceptable.
Since the test conditions examined by the EPRI tests envelop the calculated Maine Yankee plant reflooding conditions and the EPRI test results indicate that the AP penalty can be bounded by a value of 0.15 psid, the staff finds the proposed model modification acceptable, f
4.0 CONCLUSION
S Selection of the neximum allowable pin power for use in LOCA analyses is always conservative and ac:eptable. A sufficient sample of axial shapes has been proposed for LOCA an.alyses. This in combination with the envelope of axial shapes for limiting thermal margin is acceptable for detemining axial power distribution and is in compliance with Section I.A of Appendix K to 10 CFR Part 50. For each reload, the licensee should address break spectrum con-siderations especially as relates to axial shape as required by Appendix K.
In particular, the possibility exists that small breaks may be limiting for the maximum kw/ft at the top 1.5 ft. of the core. Also since a full spectrum a
is not always analyzed for each axial shape selected and a completely consis-tent calculation is not always done for each break analyzed, the spectrum chosen and methods used must be justified for each reload.
Based upon the above, the staff finds that the proposed AP penalty of 0.15 psid during the pumped injection period meets the requirements of Section I.D.4 of Appendix K to 10 CFR Part 50. Accordingly, the staff finds the proposed modification to the ECCS evaluation model for the Maine Yankee plant acceptable.
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5.0 REFERENCES
- 1. Letter from G. D. Whittier (PYAPCo) to A. 1. Thadani (NRC), " Maine Yankee LOCA Analysis," GDW-86-212, dated September 15, 1986
- 2. Letter from G. D. Whittier (MYAPCo) to A. T. Thadani (NRC) " Maine Yankee LOCA Analysis," GDW-86-267, dated November 10, 1986 Principal Contributors:
N. Lauben R. Jones
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