ML20247A550
| ML20247A550 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 08/31/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0991, NUREG-0991-S09, NUREG-991, NUREG-991-S9, NUDOCS 8909120142 | |
| Download: ML20247A550 (46) | |
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LI jilmerick Generatitig Station, Cnits: 1 and 2 LOocke't Nos.L50-352:and 50-353 i ;a
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q "M ; AVAILABIL11Y NOTICEl U r
D@' 2 Availability of RefNence Materials' Cited'in NRC Publications M . .?. .. . , ' K: .. ..
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i Most;documerats.citedin. NRC' publications will be available from one'of thelfoilowingl h % Tsources: , ?.
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i NUREG-0991 Supplement No. 9 Safety Evaluation Report related to the operation of Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Philadelphia Electric Company U.S. Nuclear Regulatory Commission OlHce of Nuclear Reactor Regulation August 1989 j## "%
i ABSTRACT !
i In August 1963 the staff of the Nuclear Regulatory Commission issued its f Safety Evaluation Report (NUREG-0991) regarding the application of the ;
Philadelphia Electric Compa:y (the licensee) for licenses to operate the !
Limerick Generatitig_ Station, Units 1 and 2, located on a site in Montgomery and Chester Counties, Pennsylvania.
Supplement 1 to ff0 REG-0991 was issued in December 1983. Supplements 2 and 3 were issued in October 1984. License NpF-27 for the low-power operation of Limerick Unit I was issued on October 26, 1984. Suppleuent 4 was issued in l
May 1985, Supplement 5 was issued in July 1985, and Supplement 6 was issued in August 1985. These supplements addressed further issues that required resolution before Unit 1 proceeded beyond the 5-percent power level. The full-power operating license for Limerick Unit 1 (NPF-39) was issued August 8, 1985, and the unit has completed two cycles of operation.
Supplement 7 was issued April 1989 to address some of the few significant design differences between Units I and 2, the resolution of issues that remained open when the Unit I full-power license was issued and an assessment of some of the issues that required resolution before issuance of an operating license for Unit 2.
Supplement 3, issued in June 1989 resolved all the issues necessary to support the issuance of a low power license for Unit 2. Operating license NPF-83, authorizing Unit 2 to load fuel and conduct pre-criticality testing, was
' issued on June 22, 1989. Operating license NPF-84, authorizing continued testing and operation of Limerick Unit 2 at power levels up to five percent (5%), was issued on July 10, 1989.
This document, the ninth supplement to the SER (SSER-9), also primarily relates to Unit 2. This supplement addresses the remaining issues that required resolution before issuance of a full power license for Unit 2.
Limerick SSER 9 iii
TABLE OF CONTENTS Page ABSTRACT............................................................. 111 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT................... 1-1 1.1 I n t rodu c t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 1 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMP 0NENTS......... 3-1 3.9 Mechanica l Systems and Components. . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.9.3 ASME Code Class 1, 2 and 3 Components, Component Supports, and Core Support Structures.......... 3-1
" 3.9.3.5 Bulletin 88-05......................... 3-1 l
3.10 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment................................... 3-1 3.11 Environmental Qualification of Electrical Equipment i
important to Safety and Safety-Related Mechanical E q u i p me n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 2 3.11.1 Introduction..................................... 3-2 3.11.2 B a c k g ro u n d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 2 3.11.2.1 Pu rp o s e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -3 3.11.2.2 Scope.................................. 3-3 3.11.3 Staff Evaluation................................. 3-3 3.11.4 Conclusions...................................... 3-4 4 REACT 0R......................................................... 4-1 1 4.4 T h e rma l - Hy d r a u l i c D e s i g n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 -1 4.4.4 Thermal-Hydraulic Stability...................... 4-1 5 REACTOR COOLANT SYSTEMS......................................... 5-1 l 5.3 Rea c t o r Ve s se l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 -1 5.3.1 Rea ctor Vessel Ma teri a l s. . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.3.1.2 Fractu re Toughne ss. . . . . . . . . . . . . . . . . . . . . 5-1 5.3.2 Pres sure-Temperatu re Limits. . . . . . . . . . . . . . . . . . . . . . 5-2 6 ENGINEERED SAFETY FEATURES................................ 6-1 6.2 C o n t a i nme n t Sy s tem s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 1 6.2.5 Combustible Ga s Control . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 7 INSTRUMENTATION AND CONTR0LS.................................... 7-1 7.4 Systems Required for Safe Shu tdown. . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.4.2 Specific Findings.................................... 7-1 7.4.2.1 Capability for Safe Shutdown Following Loss of Electrical Power to Instrumentation and Controls............. 7-1 7.4.2.1.1 Common Power Source Failure A n a ly s i s . . . . . . . . . . . . . . . . . . . . . . . . . . 7 -2 7.4.2.1.2 Common Sensor Failure Analysis.... 7-2 7.4.2.1.3 HELB and Affected NSCS Component......................... 7-2 7.4.2.1.4 Conclusions....................... 7-3 Limerick SSER 9 -v-
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l TABLE OF CONTENTS (Continued) 7.7 Control Systems............................................. 7-3 7.7.2 Specific Findings................................. 7-3 7.7.2.1 High-Energy-Line-Break and Consequential Control Systems Failures (IE Information Notice 79-22) and Multiple Control Systems failures...................... 7-3 15 ACCIDENT ANALYSES............................................... 15-1 15.8 An t icipa ted Tra ns ient !!i thou t S cram. . . . . . . . . . . . . . . . . . . . . . . . 15-1 15.8.1 Introduction..................................... 15-1 15.8.2 Review Criteria.................................. 15-2 15.8.3 ARI and RPT System Description................... 15-2 15.8.4 Evaluation of ARI System. . . . . . . . . . . . . . . . . . . . . . . . . 15-3 15.8.4.1 Safety Related Requirements (IEEE Standard 279)..........................15-3 15.8.4.2 Redundancy............................. 15-4 15.8.4.3 Diversi ty f rom Exist ing RTS. . . . . . . . . . . . 15-4 15.8.4.4 Physical Separation from Existing RTS........................... 15-4 15.8.4.5 Environmental Qualification............ 15-4 15.8.4.6 Seismic Qualification.................. 15-5 15.8.4.7 Quality Assurance...................... 15-5 15.8.4.8 Safety Related (IE) Power Supply... . .. .15-5 15.8.4.9 Te stabi li ty at Power. . . . . . . . . . . . . . . . . . . 15-5 15.8.4.10 Inadvertent Actuat ion. . . . . . . . . . . . . . . . . . 15-5 15.8.4.11 Manual Initiation...................... 15-5 15.8.4.12 Information Readout.................... 15-6 15.8.4.13 Completion of Protective Action Once Initiated......................... 15-6 15.8.4.14 Maintenance Bypass..................... 15-6 15.8.4.15 Conclusion............................. 15-6 15.8.5 Evalu ation of ATWS/RPT System. . . . . . . . . . . . . . . . . . . . 15-6 15.8.5.1 Safety Related Requ i rements. . . . . . . . . . . . 15-6 15.8.5.2 Redundancy ............................ 15-6 15.8.5.3 Diversity from Existing RTS ........... 15-6 15.8.5.4 Physical Separatien from Existing RTS .......................... 15-7 15.8.5.5 Environmental Qualification . . . . . . . . . . . 15-7 15.8.5.6 Seismic Qualification.................. 15-7 15.8.5.7 Guality Assurance...................... 15-7 15.8.5.8 Safety Relate.d (IE) Power Supply.......15-7 15.8.5.9 Testabili ty a t Power. . . . . . . . . . . . . . . . . . . 15-7 15.8.5.10 Ina dv ertent Actua tion. . . . . . . . . . . . . . . . . . 15-7 15.8.5.11 Conclusion on ATWS RPT System.......... 15-8 15.8.6 Evaluation of SLCS............................... 15-8 15.8.6.1 Safety Related Requirements............ 15-8 15.8.6.2 Evaluation............................. 15-8 15.8.6.3 Conclusion on SLCS. . . . . . . . . . . . . . . . . . . . . 15-8 15.8.7 Technical Specifications......................... 15-9 15.8.8 Conclusions...................................... 15-9 Limerick SSER 9 -vi-
TABLE OF CONTENTS (Continued) 16 TECHNICAL SFECIFICAT0NS......................................... 16-1 17 QUALITY ASSURANCE............... ............................... 17-1 17.6 Readiness Verification Program............................. 17-1 17.6.1 Independent Construction Assessment (ICA)... ... . ...17-1 17.6.2 Independent Design Assessment (IDA)................ 17-4 17.6.3 Conclusions........................................ 17-5 APPENDICES A CHRON0 LOGY H PRINCIPAL STAFF CONTRIBUTORS U ERRATA TO THE SAFETY EVALUATION REPORT FOR THE LIMERICK GENERATING STATION Limerick SSER 9 -vii-
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1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 Introduction In August 1983, the Nuclear Regulatory Commission (hereinaf ter referred to as NUREG-0991, theNRCorthestaff)issueditsSafetyEvaluationReport(SER)[
regarding the application of the Philadelphia Electric Company hereinafter referred to as PECo or the licensee) for licenses to operate the Limerick Generating Station, Units 1 and 2, Docket Nos. 50-352 and 50-353. Supplement 1 to the SER was issued in December 1983, Supplements 2 and 3 were issued in October 1984, and Operating License NPF-27, authorizing power up to 5 percent, was issued on October 26, 1984. Supplement 4 to the SER was issued in May 1985, Supplement 5 was issued in July 1985, and Supplement 6 was issued in August 1985. These supplements addressed issues that required further resolution before Unit 1 proceeded beyond the 5-percent power level. A full-power operating license (NPF-39) was issued for Limerick Unit 1 on August 8, 1985.
As noted above, the staff's SER assessed operation of both Limerick Units 1 and 2. Construction of Unit 2 was halted in January 1984 by Order of the Pennsylvania Public Utility Commissien. At the time, construction was about 30 percent complete. Construction of Unit 2 resumed in February 1986 with PECo's agreement to accept a cost containment cap of about $3.1 billicn for construction and certain operational incentive programs. On May 3, 1988, the Commission modified Construction Permit CPPR-107 to extend the earliest and latest completion dates to May 1,1989, and January 1,1992, respectively.
Supplement 7 to the SER was issued April 1989 and primarily related to Unit 2.
SSER-7 addressed some of the few significant design differences between Units 1 and 2, the resolution of issues that remained open when the Unit I full-power license was issued and an assessment of some of the issues that required resolu-tion before issuance of a low-power operating license for Unit 2. Supplement 8 to the SER was issued June 1989 and also addressed primarily Unit 2 issues that required resolution before issuance of a low power license, Operating License NPF-83, authorizing Unit 2 to load fuel and conduct pre-criticality testing, was issued on June 22, 1989. Operating license NPF-84, authorizing continued testing and operation for Limerick Unit 2 at power levels up to five percent (5%), was issued on July 10, 1989.
This document, the ninth supplement to the SER (SSER-9), also primarily relates to Unit 2. This supplement addresses the remaining issues that require resolu-tion before issuance of a full power operating license for Unit 2.
Limerick SSER 9 1-1
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[Eachofthe.sectionsandappendicesofthissupplementLis,numberedthesame
.as the related portion of the SER. Each section complements the-discussion inL
- the .SER and Supplements. I through 8, unless. otherwise noted. Appendix A is a continuation of the chronology of.this safety review. Appendix H lists the
. principal contributors. Appendix U is an errata which.provides minor corrections of SSER-8.
~
Copies'of-this. supplement are available for inspection at the NRC Public Document Room, 2120 L-Street, N.W., Washington, D.C. and at the local Public.
Document Room at the Pottstown Public Library, 500 High Street, Pottstown, Pennsylvania L19464.
The'.NRC Project Manager for Limerick Units 1 and 2:is Richard J. Clark.. He
- may be contacted by. telephone at (301) 492-3041 or by niail at the follo' wing .
address:
Office of Nuclear Reactor Regulation-U.S. Nuclear Regulatory Comission Washington, DC 20555 Limerick SSER 9 1-P
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-l 3 DESIGN CRITERIA FOR. STRUCTURES, SYSTEMS, AND COMPONENTS .l l
'3.9 ' Mechanical Systems and Components j 3.9.3 ASME. Code Class 1, 2 and 3 Components, Component Supports, and Core. !
Support Structures j 3.9.3.5 Bulletin 88-05 !
In the safety evaluation transmitted on June 20, 1989 and in SSER 8, the staff i provided the results of our review of-the licensee's submittals of March 31 .
and June 2,1989 for Limerick 2.. In each of these, we indicated that the I licensee had' conducted stress analysis on 52 installed safety-related items 1 which were found to have tensile strengths below 66 ksi (396 L converted to '
137 BHN). These52itemswereidentifiedas46carbonsteelf9angesand6 !
stainless steel flanges. Actually, stress analysis was only conducted on the !
46 carbon steel flanges. The 6 stainless steel flanges were found to be acceptable based on compliance with the criteria in Sectioti 5 of NUMARC's l Generic Testing Program Response of October 1988, as previously indicated in l section 3.3 of our June 20', 1989 safety evaluation. l H
3.10 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment By letter dated March 7, 1989, the licensee submitted a report to provide !
confirmation that the Equipment Qualification Programs used for dynamic and !
. environmental qualification of Limerick Generating Station Unit 2 are ;
consistent with the programs used for Unit 1. I I
While the programs are consistent, some equipment items in Unit 2 are found to !
be different from those considered in the original Unit I qualification program. l
.The licensee has identified 17 and 34 equipment types.in the Nuclear Steam Supply System (NSSS) and Balance of Plant (80P), respectively, that are different.
The extent of the differences was judged to be of limited scope. Because of '
this and the fact that the Unit I seismic qualification program was well implemented, the staff has determined that a plant site audit of Unit 2 equipment seismic qualification was not warranted.
For NSSS equipment the differences are due to changes in equipment design or location. For BOP equipment the differences are mainly due to design modifica-tions or procurement of similar equipment from different manufacturers. A complete list of equipment differences is provided in the submittal. For Unit 2 equipment found similar to Unit 1, qualification is based on and documented using Unit 1 qualification documentation and SQRT (Seismic Qualification Review Team) forms. For equipment that is different, qualification was performed by developing rationale / calculations for location differences, providing evaluation for qualification using the new input spectra, reviewing qualification reports, and preparing SQRT forms for unique Unit 2 equipment. Plant modifications i accomplished by Project Change Notices (PCNs) and affectino Unit 2 equipment I qualification have also been evaluated. For modifications similar to Unit 1 Limerick SSER 9 3-1
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i modifications, Unit 1 evaluations are used as-the basis. For modifications unique to Unit 2, calculations.have been generated in support of Unit 2 qualification.
Finally, walkdowns were performed according to an approved procedure for Unit 2 equipment to ensure that installed equipment is consistent with qualification documentation. These walkdowns were performed on a sampling basis.
At the time of the March 7, 1989 submittals, all Unit 2 NSSS equipment and about cighty-five percent of DOP equipment had been evaluated and determined to be acceptable for the required loading combinations. SQRT reports had been prepared for all items as is required. Preparation of the New Load Evaluations Final Summary Report was also in progress. These reports were completed and revised prior to fuel load and incorporated the results of the evaluation of discrepancies identified in the final walkdown. The licensee's evaluations indicated that none of the identified discrepancies was significant and did not affect the validity of the Unit 2 equipment seismic qualification.
Our review of the above information verified that the seismic and dynamic qualification of Unit 2 equipment was performed by an extension of the Unit 1 program. We conclude that the program as identified in Limerick Final Safety Analysis Report (FSAR) Sections 3.9 and 3.10 is applicable to both units, and the. staff acceptance of the Unit 1 program as stated in the SER and its supple-ments is also valid for the Limerick 2 licensing application.
3.11 Environmental-Qualification of Electrical Equipment Important to Safety and Safety-Related Mechanical Equipment 3.11.1 Introduction Equipment that is used to perform a necessary safety function must be demonstrated to be capable of maintaining functional operability under all service conditions postulated to occur during its installed life for the time it is required to operate. This requirement, which is embodied in General Design Criteria 1 and 4 of Appendix A and criteria III, XI, and XVII of Appendix B to 10 CFR 50, is' applicable to equipment located inside as well as outside containment. More detailed requirements and guidance relating to the methods and procedures for demonstrating this capability for electrical equipment have been set forth in 10 CFR 50.49, " Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants"; NUREG-0588,
" Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," which supplements IEEE Standard 323; and various NRC Guides and industry standards.
3.11.2 Background NUREG-0588 was issued in December 1979 to promote a more orderly and systematic implementation of equipment qualification programs by industry and to provide guidance to the NRC staff for its use in ongoing licensing reviews. The positions contained in that report provide guidance on (1) how to establish environmental service conditions, (2) how to select methods that are considered appropriate for qualifying equipment in different areas of the plant, (3) other areas such as safety margin, aging and documentation for each item of safety-related Limerick SSER 9 3-2 l
L Lelectrical eouipment, and (4) to identify the degree to which their qualification programs complied with the staff positions discussed in NUFCG-0588.
IE Bulletin 79-01B. " Environmental Qualification of Class IE Equipment," issued January 14, 1980, and its supplements dated February 29, September 30, and October 24, 1980, established environmental qualification requirements for cterating reactors. This tulletin.and its supplements were provided to OL~
applicants for consideration.
Af ule on environmental qualification of electrical equipment important to y for nuclear power plants becane effective on February 22, 1983. This rule, Section 50.49 of 10 CFR 50, specifies the requirements to be met for demonstrating the environmental qualification of electrical equipment important to !.afety located in a harsh environment. In accordance with 10 CFR 50.49, electrical equipment for Limerick Generating Station, Unit 2 may be qualified in accordance with the acceptance criteria specified in Category II of NUREG-0588.
In order to' document the degree to which the environmental qualification program complies with the NRC's environmental qualification requirements and criteria, the applicant provided equipment qualification information by letters dated March 7, 1989, April 10, 1989 and May 9,1989. ;
3.11.2.1 Purpose The purpose of the staff's review was to evaluate the adequacy of the Limerick Generating Station, Unit 2 environmental qualification program for safety-related mechanical equipment and electrical equipment important to safety as defined in 10 CFR 50.49.
3.11.2.2 Scope ,
The scope of the staff's review was limited to an evaluation of the safety-related mechanical equipment and electrical equipment important to safety at j Limerick Unit 2 that is different from equipment at Unit 1. This mechanical i and electrical equipment must function in order to mitigate the consequences of a design basis accident, inside or outside containment, while sub,iected to the hostile environment associated with this type of accident.
Safety-related mechanical equipment and electrical equipment important to safety at Limerick, Unit 2 that are identical to equipment at Unit I were addressed in Supplement 2 (NUREG-0991).
3.11.3 Staff Evaluation By letters dated March 7, April 10, and May 9,1989, the applicant identified the followin9 items of electrical equipment as specific Unit 2 equipment different from Unit 1.
Equipment Item Manufacturer Model Number 600V Power Cable Rockbestos XLPE Ins.
Flow Transmitter Rosemount 1153 Series B Differential Pressure Transmitter Rosemount 1153 Series B Limerick SSER 9 3-3
Equipment Item' Manufacturer Model Number Level Transmitter Rosemount 1153 Series B-Pressure Transmitter Rosemount 1153' Series B Motor Operator - Limitorque SMB-00-10 Motor Operator _
Limitorque SMB-1-60 Pilot Solenoid Velve- ASCO. 206-832-30-3RU Pilot Solenoid Valve ASCO NP8316A74E Electric Conduit Seal Patel Engineering -
Solenoid Valve Valcor V526-5000-Series Transformer Westinghouse 750 KVA The applicant also providea a summary description.for extension of the Limerick Generation Station (LGS) Unit 1 mechanical equipment qualification (MEQ)_ program (environmental)'to Unit 2. The sunnary was provided for staff review in order'to (1) show that the MEQ licensing commitments and progrcm description, as identified in the LGS Final Safety Analysis Report (FSAR).
Section 3.11, is applicable to both units, (2) that the implementation for Unit 2 is an extension of the Unit 1 program, (3) identify minor differences in components or in items that affect qualification between Units I and 2 and document the-qualification acceptability of those' differences, and'(4) establish that the conclusion reached in the SER for the acceptance of the LGS Unit 1 qualification program is also valid for the application of the LGS program to Unit 2.
As a result of a review of the information presented by the applicant, the~ staff finds this approach to identification and qualification of
'both electrical ana mechanical equipment acceptable.
l 3.11.4 Conclusions The staff has reviewed the summary information provided by the applicant for thel Limerick Unit 2 program for environmental qualification of electrical equipment'important to safety and safety-related mechanical rouipment. l As noted above, this review is limited to Limerick Unit 2 safety-related mechanical equipment and electrical equipment important to safety as defined in 10 CFR 50.49 that is different from equipment in Limerick, Unit 1. The purpose of the review was to assess the qualification status of such equipment and to determine the adequacy of the qualification program.
Based on the results of our review and evaluation of the information provided j by the applicant, the staff concludes that the applicant has demonstrated compliance with the requirements of 10 CFR 50.49, the relevant parts of General Design Criteria 1 and 4 of Appendix A, criteria III, XI and XVII of Appendix B to 10 CFR 50, and the criteria specified in UUREG-0588.
Limerick SSER 9 3-4
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4 REACTOR 4.4 Thermal-Hydraulic Design 4.4.4 Thermal-Hydraulic Stability In Supplement 4 to the Staff's SER (NUREG-0991), the staff concluded that the Technical Specifications proposed for Limerick Unit I were consistent with the recommendations in General Electric Company Service Information Letter (SIL)-380 and acceptably resolve the thermal-hydraulic stability concern for Limerick Units 1 and 2, assuming long-term single-loop operation is not permitted. Should i such operation be requested in the future, the staff will evaluate Limerick l Units 1 and 2 Technical Specifications to determine if additional modifications are required.
On January 23, 1986, the staff issued Generic Letter No. 86-02, " Technical Resolution of Generic Issue B-19-Thermal Hydraulic Stability," On March 31, 1986, the staff issued Generic Letter No. 86-09, " Technical Resolution of Generic Issue No. B-59 (N-1) Loop Operation in BWRs and PWRs." On June 15, 1988, the staff issued NRC Bulletin No. 88-07; " Power Oscillations in Boiling Water Reactors" which requires adoption of certain operating procedures.
Supplement No. I to this Bulletin was issued December 30, 1988. The licensee responded to the Bulletin and Supplement 1 by letters dated September 5, 1988, March 7, 1989 and March 31, 1989.
By application dated November 4,1988, the licensee requested approval of changes to the Unit 1 Tachnical Specifications (16) to permit extended single loop operation. The licensee's letter of March 29, 1989 submitted the necessary analyses to support single loop operation of Unit 2. By letter dated June 30, 1989, the staff issued Amendment No. 30 to Facility Operating License No. NpF-39 for Unit 1. The amendment and letter approved extended single loop operation for Limerick Units 1 and 2. As part of our evaluation, we reviewed the operating restrictions proposed by the licensee in response to Bulletin 88-07 and Supplement 1. As discussed in our letter of June 30, 1989 and the accompanying safety evaluation, we advised the licensee that the responses to Bulletin 88-07 and Supplement 1 satisfactorily resolved thermal hydraulic stability concerns for Limerick, Units 1 and 2.
In the letter of March 31, 1989, the licensee revised the response of March 7, 1989 to Bulletin 88-07, Supplement 1 to indicate that one exception will be taken to the implementation of the GE interim stability recommendations during the initial startup testing. As discussed in Chapter 14 of the revised
' Limerick FSAR, during Test Condition 4 Limerick Unit 2 will conduct recirculation pump trip testing which is specifically intended to identify any concerns with transients of the nature described in Bulletin 88-07 and any possible resulting instabilities. Because of the increased awareness of the possibility of instabilities and the increased monitoring by plant staff and augmented test instrumentation during the initial startup testing programs, the staff finds this one exception acceptable.
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l 5 REACTOR COOLANT SYSTEMS 5.3 Reactor Vessel 5.3.1 Reactor Vessel Materials E.3.1.2 Fracture Toughness In the original SER, the NRC staff reviewed the fracture toughness of the ferritic reactor vessel and reactor coolant pressure bnundry (RCPB), and the materials surveillance program for the reactor vessel belt-line according to SRP 5.2.3.11.3.a and SRP 5.3.1.II.5, II.6, and 11.7.
GDC 31 requires, in part, that the RCPB be designed with sufficient margin to ensure that when the boundary is stressed under operating, maintenance, testing, and anticipated transient conditions, it behaves in a ronbrittle manner and the probability of rapidly propagating fracture is minimited.
GDC 32 requires, in part, that the RCPB be designed to permit an appropriate material surveillance program for the RCP8. Materials selection, toughness requirements, and extent of material testing were reviewed in accordance with the above criteria, subject to the rules and requirements of 10 CFR 50.55a, and Appendices G and H to 10 CFR Part 50.
At the time of the original review, the applicant had submitted only Unit I material information. Since that time, Unit 2 material information has been submitted both in a letter on June 14, 1983, and in FSAR Revision 22.
Hcwever, on July 12, 1988, the NRC issued Generic Letter 88-11. "NRC Position ,
en Radiaticn Embrittlement of Reactor Vessel Materials and its impact on Plant '
Operations." This Generic Letter identified to licensees and applicants that Revision 2 of Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials," became effective in May 1988 and that it would be used by the NRC to review submittals regarding pressure temperature (P-T) that require en estimate of the embrittlement of reactor vessel beltline materials.
The licensee provided a response to Generic Letter 88-11 on November 23, 1988 that included the results of a Limerick Unit I fracture toughness analysis for the reactor vessel utilizing Revision 2 of Regulatory Guide 1.99. These results concluded that:
- 1. The Rev. 2 adjusted reference temperature (ART) values at 32 effective full power years (EFPY) for Unit I are below 200 F, which is the allowable limit in 10 CFR Part 50, Appendix G. Therefore, implementation of Rev. 2 will not result in any additional analysis, testing or provisions for thermal annealing.
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- 2. The ART value that applies to the pressure-temperature (P-T) curves in the Technical Specifications is 56*F at 32 EFPY. The maximum Rev. 2 ART value identified is 85.6 F at 32 EFPY. Therefore, the Rev. 1 32 EFPY P-T l curves are less conservative than 32 EFPY curves that would be generated with Rev. 2. However, the current P-T curves are applicable up to 10 EFPY if ART is calculated acccrding to Rev. 2 methods. The licensee-has committed to submit a Technical Specification amendment by December 29, 1989 to specify the revised time period for which the current curves are valid.
- 3. The worst case low pressure coolant injection (LPCI) nozzle is also included in this beltline region analysis due to its predicted neutron fluence value at 32 EFPY. Since it has a Rev. 2 ART at 32 EFPY that is less than the 40*F RT applicable to the limiting vessel discontinuity curves,theLPCInozzNTis bounc'ed by the limiting vessel discontinuity curves. Therefore, the discontinuity limits shown on the P-T curves of the Technical Specifications need not be' adjusted as a result of the implementation of Rev. 2 of Regulatory Guide 1.99.
The staff fir.c's these conclusions and commitments acceptable until such time as the review of the licensee's analysis is complete.
-The licensee's November 23, 1988 response also indicated that a Unit 2 analyses would be submitted by April 28, 1989, and that any required Technical Specification revisions would be submitted by December 29, 1989. The applicant submitted the Limerick Unit 2 analysis on March 31, 1989, which reached the following conclusions:
- 1. The Rev. 2 ART values at 32 EFPY for Unit 2 are below 200*F, which is the allowable limit in 10 CFR 50, Appendix G. Therefore, implementation of Rev. 2 will not result in any additional analysis, testing or provisions for thermal annealing.
and C') in the original draft Technical Specifications is 75*F at 32 EFPY. Therefore, these Rev. 1 32 EFPY P-T curves are less conservative than 32 EFPY curves that would be generated with Rev. 2.
The applicant has generated new minimum reactor pressure vessel metal temperature versus reactor vessel pressure curves for use in the Technical Specifications that are based on Rev. 2 of Regulatory Guide 1.99. Since these curves are more conservative than those based on Rev. 1, the staff finds their use acceptable until such time as the review of the licensee's analysis is complete. The staff review is being conducted with the assistance of our Contractor, EG&G Idaho, and is nearly complete pending review of data on the i nickel content on 11 beltline weld materials, weld surveillance material l nickel content and other data.
5.3.2 Pressure-Temperature Limits This topic is discussed in Section 5.3.1.2 above.
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6 ENGINEERED SAFETY TEATURES 6.2 Containment Systems l
6.2.5 Combustible Gas Control i
Inerting the containment for the LGS-2 plant is required by 10 CFR 50.44. In 10 CFR 50.44, " Standards for Combustible Gas Control System in Light Water Cooled Power Reactors," Section 50.44(c)(3)(i) states in part that,
- Effective May 4, 1982 or 6 months after initial criticality, whichever is later, an inerted atmosphere shall be provided for each boiling light-water nuclear power reactor with a Mark I or Mark Il type containment". By letter dated December 5, 1988, the licensee requested an exemption from 10 CFR 50.44(c)(3)(i) to extend the permitted time of operation with a non-inerted containment to accommodate completion of the Power Ascension Test Program (PATP). The Limerick Unit 2 PATP is based on maintaining the containment in a non-inerted condition until the successful completion of the 100-hour warranty run, a condition that normally would be expected to occur within approximately 120 effective full power days of core burn-up.
l The proposed exemption from the regulation is requested in order to complete the balance of the PATP in accordance with the licensee's test plan. No changes are being made in the maximum number of full power abys of core burn-up cormally expected before inerting is required. To assure this, the maximum expected value of 120 effective full power days is made Pcrt of the proposed action.
It is desirable to operate the reactor without inerting during the PATP, as an uninerted containment would permit unscheduled inspections or identification of possible problems during this period. The anticipated high frequency of containment entries during the PATP period and the required deinerting and re-inerting time (about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) would tend to discourage early and frequent containment entries for identifying and correcting any potential safety problems.
Further, the NRC staff believes that to require inerting before the PATP tests have been completed could result in less assurance of safety, because of the added time and/or decreased ability to directly examine and evaluate components and systems inside containment while the PATP tests are under way. Completing the PATP tests with en uninerted containment (exemption granted) vould reduce the likelihood of development of an event requiring protective safety actions during the period of exemption. Because of a low fission product inventory during the PATP period, and the short duration anticipated for the exemption, there is an extremely low likelihood that the inerting systen would be required.
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Eased on the information provided by the licensee and experience at other BWRs, the staff concludes that there will be no in:rease in risks of operation through completion of the PATP tests with'the proposed limited exemption recarding initial inerting over the risks from postulated accidents with an inerted containment.
Therefore, since there is no perceived increase in risk by the mere fact of extending the time allowed for completion of the PATP tests under uninerted conditions, the NRC staff finds that operation would be as safe under the conditions proposed by the exemption as it would have been had the test been completed in the shorter calendar time of six months Lfter initial testing.
Based on the considerations discussed above, we have concluded that the proposed temporary exemption from 10 CFR 50.44(c)(3)(i) is authorized by law, will not endanger life or property or the common defense and is otherwise in the public interest and shculd be granted.
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7 INSTRUMENTATION AND CONTROLS l
.4 Systems Reonired for Safe Shutdown l
7.4.2 Specific Findings 7.4.2.1 capability for Safe Shutdown Nwer to Instrumentation Following(Loss and Controls IE Bulletinof Electrical 79-?7)
Supplement 2 of the staff's Safety Evaluation Report related to the operation of Limerick Generating Station, Units 1 and 2 (NUREG-0991) evaluated PECO's various reports on the subject analysis. The reports included analysis of multiple control system failures due to High Energy Line Break (HELB) and common sensors instrument line or power source failures. The combined effects of an HELB and control system malfunctions were compared with the transient and accident analyses contained in the FSAR. The consequences of the identified failures were bounded by the FSAR analyses, and the staff found the shared design configuration of certain power sources, sensors and instrument ,
lines acceptable. '
l By letter dated February 17, 1989, PEco submitted a supplement to previous reports that identified areas where the current LGS Unit 2 design did not conform with the previously reviewed design. The impact of the current changes in Unit 2 design on the previous report's conclusions are determined by using
" comparative methodology" (Unit 2, 1988 versus Unit 1, 1983 plant design). By l this method, failures in the Unit 2 eouipment changes in the shared design .
configuration of certain power sources, sensors and instrument lines, and the j consequential failure of multiple control systems is compared with the previously reviewed failures of Unit 3 shared design configuration and the j 6ffected multiple control system. In each case, the differences were found to y be insignificant and bounded by the curre.nt LGS FSAR accident analysis.
The staff has evaluated the licensee's description of their assumptions, comparative methodology and determination that combined failure affects of HELB for the changed configurations with Unit 2 are bounded by the FSAR Chapter 15 analysis.
The submittal includes three sets of analyses to determine if a failure in a power source, sensor or instrument line or an HELB common to multiple Non-Safety Related Control System (NSCS) components will adversely affect the primary reactor parameters, i.e., vessel water level, pressure or reactivity, i
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7.4.2.1.1 Common Power Source Failure Analysis The current Unit 2 design of the station electrical distribution system has minor differences in bus structures from that previously reviewed.' The differences have caused minor changes in bus loading and their combined failure effects. The changes are in 125 Vdc buses that are the preferred source of l power to the instrument power supply inverters and several ac buses of different voltages. The licensee's analysis indicates that even with the changed config-uration of buses, the effects of a loss-of-bus event were not significantly different from those previously reviewed for Unit 1. The critical buses chosen to fail for this analysis were those non-safety related buses that supply power to two or more major reactor non-safety related control system (NSCS) components.
The effect of the loss of these strategic components on system operation and reactor primary parameters was insignificant and did not significantly deviate from Unit I analysis results.
7.4.2.1.2 Common Sensor Failure Analysis Similar to Unit 1 design, the identified common sensor lines and loads were mainly associated directly with the nuclear boiler applications. The sensor line support services both safety related system sensors and NSCS sensors.
Thus, a common NSCS sensor line failure could cause changes in both the safety related and non-safety related actions. The Unit 2 nuclear boiler P& ids and various NSCS schematic diagrams were res ,ewed and the rientified sensors were compared with Unit I common sensor line loads. The differences were analyzed and found insignificant and bour.ded by the FSAR Chapter 15 analysis. For example, the reactor vessel taps, one each for lines 6 and 7, were reanalyzed after a difference was identified. A break in these lines will cause a reactor low water level (LWL) scram. A plug in these lines will inhibit a reactor LWL trip only by one set of channels. However, the redundant set of channels will not be affected by the plugged line and will trip the reactor when needed.
7.4.2.1.3 HELB and Affected NSCS Component HELB zone and the associated NSCS component are subject to the plant construc-tion and layout difference between Units 1 and 2. The licensee performed a visual walkdown of the critical zones and component locations in Unit 2 similar to that performed at Unit 1. The walkdown identified physical differences in j some zone layouts and NSCS component contents between the Unit 1 zone layout and NSCS component contents. The HELB failure analysis for such zones did not identify any new combined failure effects events. These events were similar to those in Unit I and were found bounded by FSAR Chapter 15 analysis.
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7.4.2.1.4 Conclusions
- Based on the above evaluations, it-is concluded that each of the three' i analyses adequately identified the differences between Unit I and Unit 2 ;
electrical. distribution systems, sensor-lines, and HELB zone layout and NSCS '!
- component. contents. The. consequential common failure of. control systems in Unit P. was adequately. compared with those.in-Unit 1 to arrive at the
. conclusion that' the- failure events were bounded by the FSAR Chapter.15 analysis. !
17.7 Control Systems
.7.7.2 Specific Findings ~ -
7.7.2.1 HigsEnergy-Line-Break and Consequential Control Systems Failures (IE Information Notice-79-22) and Multiple' Control Systems Failures .;
This topic.is addressed in section 7.4.2.1. !
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15 ACCIDENT ANALYSIS
.15.8 ' Anticipated Transients Without Scram
-15.8.1 Introduction The original SER provided a review of the applicants action plans with regard l to an anticipated transient without scram (ATWS) and found them acceptable, i but indicated that the Commission would, through rule-making, determine any 1 future modifications necessary to resolve the ATWS concerns. !
On July 26, 1984, the Code of Federal Regulations (CFR) was amended to include Section 10 CFR 50.61, " Requirer.ents for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power !
plants" (known as the "ATWS Rule"). The'ATWS Rule requires specific improve- i ments in the design and operation of commercial nuclear power facilities to reduce the likelihood of failure to shutdown the reactor following anticipated transients, and to mitigate the consequences of an ATWS event.
-For each boiling water *eactor, three systems are required to mitigate the consequences of an ATi svent.
- 1. 'It must have an alternate rod injection (ARI) system that is diverse f (from the reactor trip system) from sensor output to the final actuation !
devices. The ARI system must have redundant scram air header exhaust valves. The ARI system must be designed to perform its function in a reliable manner and be independent (from the existing reactor trip system) from senar output to the final actuation device.
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- 2. It must have a standby liquid control system (SLCS) with a minimum flow i capacity and boron content equivalent in control capacity to 86 gallons !
per minute of 13 weight percent sodium pentaborate solution. The SLCS and its injection location must be designed to perform its function in a reliable manner.
- 3. It must have equipment to trip the reactor ccclant recirculating pumps automatically (recirc pump trip or RPT) under conditions indicative of an ATUS. This equipment must be designed to perform its function in a reliable manner.
This evaluation addresses the ARI system (Item 1), the SLCS (Item 2) and the ATWS/RPT system (Item 3).
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' 15.8.2 Review Criteria The systems and equipment renuired by 10 CFR 50.62 do rot have to meet all of the stringent requirements normally applied to safety-related equipment.
However, this equipment' . is part of the broader class of structures, systems,
- and components important to safety defined in the introduction to 10 CFR 50,
' Appendix A, General Design Criteria (GDC). CDC-1 requires that " structures, systems, and components important to safety. shall be designed, fabricated, erected, and tested to quality standards commensurate with the.importance of the safety functions to be performed." Generic Letter 85-06 " Quality Assurance Guidance for ATWS Equipment' that is nct Safety Related" details the quality assurance that must be applied to this equipment.
In general, the equipment to be installed in accordance with the ATWS Rule.is required to be diverse from the existing RTS, and must be testable ct power.
This equipment is intended to provide needed diversity (where only minimal diversity currently exists in the RTS) to reduce the potential for common mode failures that could result in an ATWS leading to unacceptable plant conditions.
The criteria used in evaluating the licensee's submittal include 10 CFR 50.62
" Rule Considerations Regarding Systems and Equipment Criteria" published in Federal Register Volume 49, No.124 dated June 26, 1984, and Generic Letter 85-06 " Quality Assurance Guidance for ATUS Equipment that is not Safety Related."
15.8.3 ARI & RPT System Description The Limerick Generating Station has installed a Redundant Reactivity Control System (RRCS) to mitigate the potential consequences of an anticipated transient without scram event. The RRCS consists of reactor pressure and reactor water level sensors, logic, power supplies, control room cabinets, and instrumentation to initiate the protective actions to mitigate an ATWS event.
The protective actions include:
- a. Alternate Rod Injection (ARI),
- b. . Recirculation Pump Trip (RPT),
- c. Feedwater Runbhck, and
- d. Standby Liquid Control System (SLCS),
The RRCS is independent ' rom the reactor trip system. It is a two divisional safety related system. Either division is capable of initiating protective actions when both input channels A and B within a division are tripped. The 1 RRCS output will energize the devices to start the protective actions. The j system can be manually initiated by depressing two push buttons (tripping both !
Channels A and B) in the same division. l i
The ARI. logic will cause the immediate energization of the alternate rod l injection inst rt valves when either the reactor vessel high pressure trip j setpoint or the low water level-2 trip setpoint is reached, or the manual i
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push buttons are armed and depressed. The ARI valves and bleed paths are sized to allow insertion of all control rods to begin within 15 seconds. The status of the ARI system is indicated in the main control room.
The function of the RPT is to reduce the severity of thermal transients on fuel elements by tripping the recirculation pumps early in the transient events (such as turbine trip, or load rejections). The rapid core flow reduction increases void content and thereby introduces negative reactivity in the reactor to reduce the thermal power. There are two separate and independent i systens to trip the recirculation pumps. One is the reactor trip system end-l of-cycle recirculation pump trip (E0C/RPT), which detects turbine control valve fast closure and main sto control system (ATWS/RPT)p whichvalve closure.
detects The other high reactor is the or pressure redundant low reactor reactivity water level. The Limerick design has two breakers in series for each reactor coolant recirculation pump. Each breaker has two independent trip coils; one receives a trip signal from the reactor trip system and the other receives a trip signal from the redundant reactivity control system. Both trip coils are l Class IE qualified. The Class IE RTS and RRCS trip coils are separated frcr each other. Etch trip coil is capable of tripping the associated breaker independently of the other.
The RRCS detects high reactor pressure. After a 25 second time delay, it initiates the feedwater runback - provided the APRM (nuclear instrument average power range monitor) down-scale signal is not present. After a 100 second time delay, it isolates the reactor water cleanup system and automatically initiates the standby liquid control system.
The RRCS recirculation pump trip and feedwater runback are not initiated by manual initiation of the RRCS. However, these may be manually initiated at the respective system control panels.
The RRCS is continually checked by a solid state microprocessor based self-test system. This self-test system checks the RRCS sensors, logic, and actuated devices. The RRCS sensors, logic and actuated devices and the APRM permissive circuits are Class IE, independent of the RTS, and environmentally qualified. The ARI function can be reset by the ARI reset switches after a 30 second time delay to ensure that the ARI scram goes to completion. The other RRCS functions can be reset by the RRCS reset switches, provided the high reactor pressure or the low water level signal no longer exists.
15.8.4 Evaluation of ARI System 15.8.4.1 Safety Related Requirements (IEEE Standard-279)
The ATWS Rule does not require the ARI system to be safety grade, but the implementation must be such that the existing protection system continues to meet all applicable safety related criteria. The licensee stated that the ARI system (a subsystem of the RRCS) is classified as a Class 1E system. It is electrically diverse and independent from the reactor trip system, and it meets IEEE Standard 279-1971 in all applicable areas. The RPCS interfaces Limerick SSER 9 15-3
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!with control systems through' the qualified isolation devices. Any electrical failures in the control systems will not propagate into the RRCS to prevent
. ARI system.from performing its protective functions. The staff finds this acceptable.
15.0.4.2 . Redundancy
'The ATWS Rule' requires that the ARI system must have redundant scram air header exhaust valves, but the ARI system itself does not need to be redundant.
Limerick's ARI system has redundant scram air header exhaust valves.. The I..
initiation and' control circuits are redundant. All vent paths will allow insertion'of all; control rods to begin within 15 seconds.and to be completed
. system. The' staff 1 finds this acceptable.
15.8.4.3 Diversity from Existing RTS The ATWS Rule requires the ARI system to be diverse from the existing reactor trip system. . Limerick's ARI system uses energize-to-function valves instead of deenergize-to-function valves. It has DC powered valves and logic instead of AC. powered valves and logic. Four reactor high pressure sensors and four
' low reactor reactor vessel water level sensors are dedicated for use to detect the ATWS events. The detection logic circuitries, power supplies and final actu'ated devices are independent from the reactor trip system. The built-in continuous self-testing feature will provide an additional assurance of reliability for the ARI system. The staff finds this acceptable.
15.8.4.4 Physical Separation from Existing RTS The ATilS Rule guidance states that the implementation of the ARI system must be such that separation criteria applied to the existing protection system are not violated.
The Limerick ARI system sensors, transmitters, trip units and associated circuits are Class 1E. The ARI system is separated and' independent from the reactor trip system and has redundant divisions from sensor to the actuation of ARI valves. Either division can perform the protective action. The separation between two redundant divisions satisfies the guidance provided in Regulatory Guide 1.75. The staff finds this acceptable.
15.8.4.5 Environmental Qualification
.The. ATWS-Rule guidance states that the qualification of the ARI system is for anticipated operational occurrences only, not for accidents.
The Limerick ARI system is a Class 1E system. It is qualified to the anticipated operatforal occurrence conditions. The staff finds this acceptable.
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15.8.4.6- Seismic Qualification No seismic qualification is required for ARI system hardware.
15.8.4.7 Quality Assurance NRC Generic Letter 85-06 dated April 16, 1985, provides quality assurance guidance for the ARI system. The licensee is required to follow this guidance.
15.8.f.8 Safety Related (IE) Power Supply The ATWS Rule guidance states that the ARI system must be capable of performing its safety functions with loss of offsite power, and that the power source should be independent from. the existing reactor trip system. The Limerick ARI systems '
l are powered from Class IE 125 Vdc power sources that are independent from ;
existing reactor trip. system power sources. Division I RRCS is powered by 125 Vdc from bus A, Division I. Division II RRCS is powered by 125 Vdc from bus B, Division II. These DC buses are backed up by station batteries. The staff finds that the Ai!I system'is capable of performing its safety functions with loss of offsite power. The API power sources are independent from the existing RTS power source, and therefore this power supply arrangement is acceptable.
4 15.8.4.9 Testability at Power The ATWS Rule guidance states that the ARI system should be testable at power.
The Limerick ARI system is continually self-tested by a microcomputer based self-test system that tests the signal, trip setpoint and logic. An analog trip module (ATM) failure, out of calibration condition, or a lack of system continuity condition will be annunciated. The ARI system uses a redundant 2-out-of-4 logic arrangement. Each reactor vessel level and pressure instrument can be tested during plant operation without initiating the ARI system, because two' level or two pressure signals must be present in the same division to initiate the action. The staff finds this acceptable.
15.8.4.10 Inadvertent Actuation The ATWS Rule guidance states that inadvertent ARI actuation that challenges other safety systems should be minimized.
The Limerick ARI system has redundant chanrels in each division. Both channels A and B must be tripped to initiate the protective actions. A marcal initia-tion also requires arming the switch and depressing two push buttons to initiate the action. As a result, inadvertent actuation is minimized. The staff finds this acceptabic.
15.8.4.11 Manual Initiation
- The Limerick ARI system has two sets of manual initiation switches (two switches in each division) in the control room. The operator must first rotate the Limerick SSER 9 15-5 1
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push button collar to arm the switches, then depress both switches to initiate
-the protective actions. The staff finds this acceptable.
15.8.4.12 Information Readout
.The Limerick RRCS system provides status indications in the control room for.
potential ATWS, confirmed ATWS, ARI initiated, RRCS ready for reset and RRCS system related' malfunctions. With continuous self-testing capability, the operator always has current status of the RRCS. The staff finds this information presentation is adequate.
15.8.4.13 Completion of Protective Action Once. Initiated The Limerick RRCS has a seal-in feature to ensure the completion of the protective action once initiated. After initial conditions return to normal, deliberate operator action is required to reset the safety system logic to normal. The staff finds this acceptable.
15.8.4.14 Maintenance Bypass
.There is no maintenance (manual) bypass of the PRCS. The staff finds this acceptable.
15.8.4.15 Conclusion Based on its review, the staff concludes that the ARI system design basis requirements identified above are in compliance with ATWS Rule 10 CFR 50.62 paragraph (C)(3) and guidance published in Federal Register Volume 49 No. 124 dated June 26, 1984, and is therefore acceptable.
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- 15. 8. 5 -- Evaluation of ATWS/RPT System 15.8.5.1 Safety Related Requirements The ATWS/RPT system is a subsystem of the RRCS and is classified as a Class IE system. It is electrically diverse and independent from the reactor trip system, and meets IEEE Standards 279-1971 in all applicable areas. The staff finds this acceptable.
15.8.5.2 . Redundancy stem has two trains. The ATWS/RPT function is redundant to the The trip ATWS/RPT function sy(end-of-cycle RPT).
The staff firds this acceptable.
15.8.5.3 Diversity from Existing RTS The ATWS/RPT system uses energize-tc-function logic; instead of deenergize-to-function logic used in the RTS. The sensors, trip units, and power supplies of ATWS/RPT are diverse and independent from the RTS. The staff finds this acceptable.
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15.8.5.41 Physical Separation from Existing PTS The ATWS/RPTf system sensors, transmitters, trip units and associated circuits Jare Class'IE. They are separate and independent'from the reactor trip system
/ components. The staff finds this acceptable, n
15.8.5.5-_ Environmental Qualification
' The ATWS/RPT system is a Class IE system.
It is qualified to anticipated operational occurrence conditions. The staff finds this acceptable.
15.8.5.6 Seismic Qualification No seismic qualification is required for the ATWS/RPT hardware.
15.8.5.7 Quality Assurance NRC Generic Letter 85-06 dated April 16, 1985, provides quality assurance guidance for the ATWS/RPT system. The licensee is required to follow this guidance.
15.8.5.8 Safety Related (IE)~ Power Supply-The ATWS/RPT system is' powered from the Class 1E 125 Vdc power sources, which are independent from the existing reactor trip system. The DC buses are backed up by station batteries; therefore, the ATWS/P,PT system is capable of performing its safety functions with a loss of offsite power. The staff finds this acceptable.
15.8.5.9 Testability at Power The ATWS/RPT system uses a redundant 2-out-of-4 logic arrangement. Each level
~and pressure instrument can be tested during plant operation. The ATWS/RPT system is continuously self-tested by a microcomputer based self-test system that tests the signal, trip setpoint and logic. An analog trip module failure, an out-of-calibration. condition, or a lack of system continuity condition will be annunciated. The staff finds this acceptable.
15.8.5.10 Inadvertent Actuation The ATWS/RPT system has redundant channels in each division. Both channels (A and B) must be tripped to initiate the protective actions. The ATWS/RPT actuation setpoint on reactor vessel high pressure is 1093 psig; the setpoint for reactor water low level is -38 inches. The RTS actuation setpoints on high reactor vessel pressure is 1037 psig; the setpoint for retctor water low level is 12.5 inches. Therefore, the ATWS/RPT actuation will not challenge the RTS. The staff finds this acceptable.
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15.8.5111' Conclusion on ATUS RPT System Based.on its review, the staff concludes that the ATWS/RPT design basis requirements identified above are in compliance with ATUS Rule 10 CFR 50.62 paragraph (C)(5) anc . guidance published in Federal Register Volume 49 tio.124 dated June 26, 1984. The staff finds this acceptable.
15.8.6 -Evaluation of SLCS 15.8.6.1 Safety Related Requirements The standby liquid control system (SLCS) must have a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution.
15.8.6.2 Evaluation-The SLCS design information given by the licensee has been reviewed by the staff against the requirements of the ATWS Rule (10 CFR 50.62), and Generic
. Letter 85-03 " Clarification of Equivalent Control Capacity for Standby Liquid Control System," ~ dated January 28, 1985. The Limerick design to meet the safety related requirement calls for two of the three installed SLCS pumps operating at a total conbineo flow rate of greater than or equal to 41.2 CPM, with the corresponding solution concentration, to meet the above equivalency l requirement. Operation at this minimum requirement requires a solution l concentration of not less than 13.6%. The flow capacity and solution concen-tration provided by the licensee exceeds the ATWS Eule requirement of 86 GPM of 13 weight percent sodium pentaborate. This is acceptable.
The licensee's plan to periodically test only cne SLCS system pump at a time is also acceptable. This is based on the licensee's statement that tests performed on Limerick 1 during stortup verified that the SLCS is capable of operating under the increased pressures associated with more than one pump operation.
15.8.6.3 - Conclusion on SLCS The license's design for the SLCS is acceptable, because it will deliver an equivalent boron concentration of 13 weight percent sodium pentaborate at 86 GPM as required by 10 CFR 50.62.
i' Limerick SSER 9 15-8 1
15.8.7. . Technical Specifications
- The equipment required by the ATWS Rule to reduce: the risk associated with an ATWS event must be designed to perform ~its function in a reliable manner. A method acceptable .to-the-staff-for demonstrating that the equipment saticiies the- reliability requirements of the ATWS Rule is to provide. equipment technical specifications,-including operability and surveillance requirements.
The Limerick plant technical specifications have incorporated requirements
.for the A-~AS/RPT and the SLCS.. The staff has not required technical specifications for the ARI system.
15.8.8 Conclusions-The staff has reviewed the design information provided by the licensee and concluded that the ARI design, the ATWS/RPT-design and the SLCS design comply-with the-requirements of.10 CFR 50.62 and the guidance published in the Federal Register on June 26, 1984-(49 FR 26036). The design of each of these
~
systems' is acceptable. Portions of this Safety Evaluation were previously.
transmitted to the licensee on November 3, 1987, and on June 8 (Amendment 22
'to LGS-1). The staff has also reviewed the results of-the surveillance tests-performed by the licensee and has verified that the. systems function as ~ intended..
Limerick SSER 9 15-9
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16 TECHNICAL SPECIFICATIONS Technical Specifications for Unit 2 were issued with the fuel load license and were re-issued unchanged with the low power license. These eriginal Technical Specifications were developed to be identical to those of Unit 1, where possible, and amendments to the Unit 1 Technical Specifications were incorporated as discussed in SSER-8. Since the original issuance of the Unit 2 Technical Specifications, additional Unit 1 amendments were approved.
These amendments, listed in Table 16.1, and their safety evaluations have been determined to be applicable to Unit 2 and have been incorporated into the Unit 2 Technical Specifications to be issued with the full power operating license.
Some other minor typographical / editorial items have also been revised, but these items have no impact on the previous safety evaluations.
By letter dated May 31, 1989, Philadelphia Electric Company (PECo) (the licensee) provided a markup of the current Final Safety Analysis Report (FSAR) for the Limerick generating Station (LGS), Units 1 and 2. The markup of FSAR pages was made to incorporate the extended load line limit analysis (ELLLA), increased core flow (ICF) and partial feedwater heating (PFH) into the FSAR, so these modes of operation could be included in the draft Technical Specifications (TSs) for Unit 2. To support these operational conditions and the draft revision to the FSAR, PECo also provided two reports prepared by General Electric (GE). These reports are: NEDC-31577P " Extended Load Line Limit Analysis for Limerick Generating Station Unit 2, Cycle 1", dated March 1989, and NEDC-31578P " Increased Core Flow and Partial Feedwater Heating Analysis for Limerick Generating Station Unit 2, Cycle 1", dated March 1989 with Errata and Addenda No. 1, dated May 31, 1989. These markup changes were discussed with NRC staff on May 11, 1929 and are consistent with the Final Draft version of the Unit 2 Technical Specification transmitted by NRC letter dated May 19, 1989. The analyses supporting these modes of operation for Unit 2 are identical to the Unit 1 analyses that the NRC has previously accepted by safety evaluations dated February 17, 1987 (ICF and PFH), and August 14, 1987 (ELLLA).
The PECo submittal proposes extensions to standard operating regions in the GESTAR II standard category of " Operating Flexibility or Margin Improvement Options". The selected options are ELLLA, ICF, final feedwater temperature reduction (FFWTR) and feedwater heater out of service (FH005). These have become commonly selected and approved options for a number of reactors in recent years. These options are described and discussed in the GE topical reports for Limerick Unit 2, referenced above, which provide generic analyses of transients and accidents.
The proposed ELLLA changes the Average Power Range Monitor (APRM) rod block and scram lines on the power-flow map, and permits operation along the new APRM rod block line (0.58W + 50%) up to the intersection with the 100 percent power line, which occurs at a core flow of 87 percent. These are standard changes for ELLLA. For ICF the approved flow increase is to 105 percent of rated core flow at 100 percent power. The increased flow is allowed throughout the cycle and af ter normal end-of-cycle (with or without FFWTR) with reactivity coast down. FFWTR involves feedwater temperature reduction up to 60%F (to 360 F at full power) and is proposed only for operation after a normal end-of-cycle. Limiting events have been analyzed for cycle extension to the exposure attainable using FH005, ICF and FFWTR at full power.
Limerick SSER 9 16-1
l i
for the ELLLA extension, the topical reports discuss a full range of transient and accident events relevant to the region extension, and presents results of calculations or previously approved conclusions. The transient analyses demonstrate that the licensing basis results (e.g., 100 percent flow, 100 percent power for pressurization transients) bound the ELLLA region results (e.g., 87 percent flow, 100 percent power). These conclusions apply to all relevant minimum critical power ratio.(MCPR) events such as pressurization, rod withdrawal and flow runout events. Changes to MCPR TS are not required because of ELLLA adoption. Other relevant areas, such as overpressure protec-l tion, LOCA and containment analysis have also been examined, and the analyses indicate that results are within allowable design limits. Thermal-hydraulic stability will be verified by appropriate surveillance. The analyses have been done with approved methodologies and the results are similar to previously approvec ELLLA extensions. Thus, operation within the ELLLA region is acceptable for cycle 1 operation of Limerick Unit 2.
Nuclear transient data LOCA analyses and thermal hydraulic stability analyses consistent with the analyses previously performed for Unit I were developed to
. include the combination of ELLLA with PFH and ICF. Lower initial operating pressure and steam flow rate (due to lower feedwater temperature) provide more overpressure margin for the limiting MSIV closure flux scram event.
Hence, it is concluded that pressure barrier integrity is. maintained under PFH conditions. The licensee has analyzed the overpressurization limiting transient (MSIY closure) for increased core flow (ICF) without PFH. The analysis of this bounding transient predicted a peak vessel pre:;sure of 1273 psig, which is below the ASME code limit of 1375 psig; the analysis results are therefore acceptable.
The fuel loading error accident, rod drop accident, and rod withdrawal error have been evaluated by the licensee for ICF and/or PFH operation. The rod withdrawal error transient is limited by a rod block system. The addition of a "high flow clamped" trip setpoint limit of 106 percent and allowable value of 109 percent of rated flow for the rod block monitor upscale alarm ensures that the rod block trip value currently in the TS will not be exceeded. The reactor coolant system recirculation flow upscale trip setpoints and allowable 1 values, and the values for the recirculation pump MG set scoop tube mechanical j and electrical stops are increased. These changes are necessary to accommodate increase core flow operation and are acceptable. The licensee has stated that the fuel loading error and rod drop accident are not adversely affected by the proposed changes. For the fuel loading error event, the licensee reported in their letter dated January 2,1987, a maximum increase in CPR of 0.04 from the value of 0.11 stated in the FSAR for this event at rated conditions. Thus the fuel loading error remains a non-limiting event. With regard to the rod drop accident, the LGS uses a banked position withdrawal sequence (BPWS) for control, rod movement. Based on prior staff review of BPWS as presented in Section S.2.5.1.3 of the General Electric Standard Application for' Reactor Fuel (Supplement for US), May 1986 (NEDE-24011-P-A-8-US, as amended), the staff agrees that a fuel loading error event is not adversely affected by the proposed changes.
l A loss of coolant accident (LOCA) with ICF and PFH was addressed in NEDC-31578P. The LOCA analyses with ICF alone bound operation with ICF and PFH. Since the peak clad temperature for ICF increases by less than 10 F for the limiting break compared to the rated core flow condition, the calculated Limerick SSER 9 16-2
peak clad temperature (PCT) of approximately 2100*F remains below the 10 CFR 50.46 cladding temperature limit. ilo changes to the current maximum average planar linear heat generation rates (MAPLHGR) are required. In NEDC-31578P, GE stated that PCT changes throughout the remainder of the large break spectrum will be of a similar magnitude (less than 10 F).
Consideration was given to the break spectrum range of 60 to 100 percent DBA for the separate effects of ICF for several classes of BWR plants. The conclusion is that increased core flow results in a peak clad temperature increase of less than 10*F throughout the large break spectrum.
The separate effect of reduced feedwater temperature is to reduce the calculated peak clad temperature. A discussion was presented for both reduced feedwater temperature and increased core flow conditions, which bound the conditions described in the proposed amendment. Based on the staff's review of the information provided by the licensee, the staff agrees with the conclusion in NEDC-31578P that the effect of ICF will not alter the limiting break size. The calculated peak clad temperature remains below the 10 CFR 50.46 cladding temperature limits and is acceptable.
The impact of the proposed operating mode on containment LOCA response was considered by the licensee. A conservative analysis resulted in a peak drywell deck downward differential pressure 2.6 psi higher than the value of 26.0 psid in the LGS FSAR. However, this is still below the design limit of 30.0 psid reported in the FSAR. It was also stated that the peak suppression pool temperatures, chugging loads, condensation oscillations and pool swell bounding loads were all found to be bounded by the rated power analysis in FSAR Chapter 6. We find this acceptable.
NEDC-31578P included a discussion of thermal-hydraulic stability (THS) for the LGS. The proposed LGS Unit 2 technical specifications implement a generic set of operating recommendations (General Electric Service Information Letter No.
380, Revision 1, February 10,1984) to assure acceptable plant performance in the least stable portion of the power / flow map, and to provide operator instructions for the detect-and-suppress mode of operation. The THS compliance for all licensed GE BWR core fuel is demonstrated on a generic basis by NEDE-22277-P-1 and has been approved by the staff (NRC Safety Evaluation Report Approving Amendment B to NEDE-24011-P contained in Appendix US-C). PECo also committed in their letters of March 7, 1989 and March 31, 1989, to implement GE recommendations for thermal-hydraulic stability actions as outlined in NRC Bulletin No.88-07 supplement 1: " Power Oscillations in 1988. The staff concludes BoilingWaterReactors(BWR)",datedDecember30(AlsoseeSection4.4.4) that acceptable THS provisions have been made.
We have reviewed the information provided by the Philadelphia Electric Company relative to the proposed operation of the Limerick Generating Station Unit 2 in the ELLLA region, combined with partial feedwater heating and increase core flow. Based on the results of the evaluation, the staff concludes that the proposed operations are acceptable. This information was previously transmitted to the licensee on June 14, 1989.
Limerick SSER 9 16-3
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Table 16.1 Additional Limerick Unit 1' Technical Specification Amendments and Safety Evaluations Applicable to Limerick Unit 2 Original Amendment Amend. Submittal Issue No. Date Date Subject 29 2/14/86 6/22/89 Clarification of TS.
30 11/4/88 6/22/89 Single Loop Operation.
31 6/10/89 7/24/89 CRD Accumulator Testing.
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17 QUALITY ASSURANCE 17.6 Readiness Verification Program In SSERs 7 & 8, the staff described the extensive Readiness Assessment and Readiness Verification Programs (RVP) being conducted by PECo to assess the design, construction and operational aspects of Limerick Unit 2. A major feature of he RVP was an independent design and construction assessment (IDCA). The IOCA consisted of two major programs - an independent design assessment (IDA) ar.d an independent construction assessment (ICA). PECo's performance of the IDCA and the NRC's inspection of the programs are complete.
The following provides the firel status report on both programs.
17.6.1 IndependentConstructionAssessment(ICA)
In SSER-8, the staff described the review of PECo's February 10, 1989 submittal and the follow up on-site inspection documented in Inspection Report 50-353/89-200. As indicated, the team was favorably impressed with the licensee's efforts to determine the scope of deficiencies identified by SWEC (Stone & Webster Engineering Corporation) and the NRC. However, several issues remained opened following the NRC inspection. The issues which required additional information from the licensee or additional review by the NRC are:
- 1) verification by the licensec that the wire size used for motor leads on the operator for valve HV-52-2F001C is adequate for its application, 2) a clarification by the licensee of its construction quality assurance program as it relates to the licensee's response to COR-056, 3) NRC review of additional information regarding resolution of grouted-in anchors that did not meet minimum embedment depths, and 4) an NRC-identified weakness associated with improperly performed quality control inspections.
On June 21, 1989, PECo responded to Inspection Report 50-353/89-200. PECo addressed the wire size issue in item 1 of Attachment 1 and addressed both the clarification of its construction QA program and the QC inspections in item 2 of Attachment 2. As indicated in the Inspection Report, the licensee provided the team with additional information regarding the grouted-in anchors immediately after the inspection.
As noted above, the licensee provided additional information about the wiring in the operator for valve HV-52-2F001C (core spray suction primary containment isolation) in the June 21, 1989 letter. The licensee stated in the letter that although the wire's continuous current rating is based on 30 degrees C and the maximum ambient temperature is 125 degrees F (51.7 degrees C), the wire insulation is rated for 125 degrees C. In addition, the licensee indicated that the insulation is made of flame retardant cross-link polyethylene. The staff found that the motor lead wire used in the operator for valve HV-52-2F001C is adequate based on its insulation flame resistance and high temperature rating. The staff considers this open item to be adequately resolved.
In its June 21, 1989, letter, the licensee responded to items 2) and 4) above relating to quality assurance as one item, but the staff will discuss them separately here. The licensee described the Limerick quality program in detail and identified the various levels of reviews and inspections performed by the Limerick SSER 9 17-1
l' I constructor (Bechtel) and the licensee. The licensee also.related this description-of their construction quality program to Bechtel's references to
/the~ quality program, which the NRC found to be narrow, in Bechtel's response to COR 056.' The staff found pEco's response to this open item adequate and considers this open item to be-resolved. 3 Region. I' issued the final open item as one of two violations in the flotice of Violation's accompanying combined inspection report.50-352/89-10 and
~50-353/89-16,.which was issued June 30, 1989. The licensee responded to the Notice of Violations'in its letter of July 28, 1989, and addressed the NRC's concern:about a trend of inadequate QC inspections. The licensee acknowledged-the violation from the NRC's ICA irispection with the clarification that the examples cited by the NRC were the result of improperly conducted inspections.
For corrective. action, the-licensee trained construction and quality control engineers to emphasize.the importance of technical manual and drawing reviews before starting field installations and inspections. In addition, the licensee. reviewed the seven . specific examples from the violation and>
determined that.(1) no' adverse trends existed, (2) the equipment would have-functioned properly, (3) sufficient formal training was provided to personnel, and:(4)~ independent reviews of the. plant's design and construction found the design to be adequate and construction to be of high cuality. The staff found' the. licensee's response to the violation from the NRC ICA inspection to be
-adequate.-
As. discussed in Inspection Report 50-353/89-200, the inspectors-found that some of..the grouted-in anchors had embedment lengths less than specified on the design drawings. The. staff has performed an engineering evaluation of the edditional information provided by the applicant regarding safety significance of the deviation from the design specification. Two kinds of anchorage were used: threaded rods without heads which were grouted in the concrete structure and used to support steel platforms and threaded bolts with nuts, ASME A-36 material, used as equipment anchorage. This evaluation addresses adequacy of both types of these anchors.
The staff reviewed Observation Report No. COR-034, Rev. O dated October 13, 1988 and the enclosed Non-Conformance Reports (NCRs) (NCR No. 14047 dated November 11, 1988-and eleven start-up NCRs dated between October 26, 1988 and November 14,1988). The staff also reviewed the supplemental information provided to the inspection team as noted in our letter of May 17, 1989. The 11 etter indicated that the factor of safety for the grouted-in rods which were used for steel platforms, is so high that the reduction in capacity for pullout due to the less than the minimum required embedment is insignificant.
More specifically, a test program was undertaken in 1978 at the Limerick jobsite, to establish the shear and tension values for grouted-in anchors for
-various embedments. (Limerick Generating Station " Tensile Test Report on Grouted in Anchors in Concrete Walls and Slabs," Bechtel Power Corporation, August 17, 1987 and Limerick Generating Station, Units I and 2, Job 8031,
" Shear Load Testing of Grouted-in Anchors, Specification 8031 C-51," dated September 21,1978). The results indicate that for 1 inch diameter anchors with 6 inch embedment the shear capacity was 20.8 kips. Similar results were obtained for other size anchors. The specimens were threaded rods and the compressive strength of concrete was 5.0 ksi. Since the concrete strength at Limerick was 4.0 ksi, the corresponding shear value was reduced accordingly, in Limerick SSER 9 17-2 l
l l proportion to the square root of the respective concrete strengths. This was compared with the allowable value of.approximately 2 kips, which would have an I ample factor of safety against the test results. Through discussions with Bechtel and PECo personnel, the staff established that the tests reflected the installation conditions at the site.
In order to assess the safety significance of the reduced embedments of the grouted-in anchors which were used for equipment supports a statistical analysis was performed.
It has been determined that 919 grouted-in anchors have been installed in Unit !
2 to date on Se.ismic Category I large pipe supports. A sample, consisting of i 136 (15 percent) of the. installed grouted-in anchors, was examined. Out of the 136 specimens, 31 rods were found to not meet the minimum embedment requirements as specified on the pertinent drawings. At the request of the plicant submitted documentation which reports on the statistical staff a the ap(Calculation for Capacity of Grout-in Rods with Reduced Embedment, nalysis.
Calculation #114.11.13, Rev. 1.) The sample of 136 anchors was tabulated and separated into populations by diameter. The quality of each population was specified by computing a lower bound, called tolerance limits. These tolerance limits were calculated so that 95 percent of the population should fall above the limit at a 95 percent confidence level, i.e., the probability that 95 percent of a population falls above the tolerance limit is 0.95. By comparing the tolerance limits and the actual embedment lengths it can be established that out of the sample of 136 anchors examined two were outside of tolerance limits (one - 5/8 in, diameter and one 3/4 in. diameter). We consider this to be acceptable.
This statistical evaluation was used as the basis for determination of the design allowable values of the threaded rods.
The design loads for the anchors which did not meet the minimum embedment requirements were compared with the corresponding design allowable values using the interaction formula. The results of this comparison was reviewed by the staff and found acceptable.
A question was raised in connection with the applicant's analysis of seismic qualification of steel platforms using the grouted-in rods as anchorages. The loads on grouted-in rods were obtained from the equivalent static analysis using peak acceleration but without the factor of 1.5 as required by the Standard Review Plan Section 3.7. These loads were compared with those obtained by the dynamic analysis using the response spectrum analysis technique and the BSAP (Bechtel Structural Analysii Program) computer code. The reactions on grouted-in rods obtained from the equivalent static analysis described above were compared with the corresponding loads obtained from the dynamic analysis. Review of the results indicates that the equivalent static analysis without the factor of 1.5 provides a conservative estimate as compared with the actual loads obtained from the dynamic analysis. It is therefore ' concluded that the use of equivalent static analysis without the 1.5 factor is acceptable for determination of the seismic loads on the platforms. 1 On the basis of the review described above, the staff concluded that the deficiency in embedment of the grouted-in rods as observed during the special ,
inspection at Limerick Unit 2, conducted in March 1989, is not significant ]
Limerick SSER 9 17-3
I enough'to cause. a risk to public safety and that' the construction can therefore be considered as acceptable. The above conclusion is based on the
.following:
-(a) The applicant conducted tests on similar type of grouted-in rods with and without nuts and the test results indicate that there is an ample factor of safety against the allowable loads, indicating considerable conservatism ,
inherent in the design of the : ads. !
(b) The steel platforms supported by the grouted-in rods were investigated j by computerized dynamic analysis as well as by the equivalent static analysis and the results compared. The loads in the analysis from both were computed using the straight line interaction equations and the results are slightly more conservative than those obtained by the computer analysis.
(c) A statistical analysis of the sampled. embedded rod length data was made which indicated that out of the 136 grouted-in rods only two are outside of 95 percent tolerance limits. We consider that this is insignificant.
Based on the above the staff concludes that for the rods / anchors that did not meet the minimum embedment depths specified on the drawings that: the deviations were not sufficient to present a concern with respect to integrity of the construction.of the plant or to cause a concern regarding their intended functions and the as-found condition of the grouted-in rods / anchors
.is acceptable.
The staff has reviewed the above information and the additional information j provided in Attachment 2 to the licensee's June 21, 1989 submittal. The staff finds these respenses and the licensee's May 16, 1989 confirmation that the 1
action items resulting from the construction portion of the IDCA are complete l as sufficient documentation to close the independent construction assessment review of Limerick Unit 2. The team concludes that the Limerick Unit'2 IDCA and the staff's reviews have confirmed that the Limerick Unit 2 construction program has been satisfactorily implemented.
17.6.2 IndependentDesignAssessment(IDA)
In SSER-8, the staff described the review of PECo's April 12, 1989 submittal and the follow-up inspection conducted the week of April 24, 1989. (Inspection Report 50-353/89-201, not yet issue). As indicated, the team concluded that the IDA provided the needed additional design assurance that Limerick Unit 2 '
has met its licensing commitments. However, this was contingent upon PECo's providing acceptable responses to six (6) items. As stated in SSER-8, two of ,
the six items were satisfactorily resolved, except for the supplement hazards report. The licensee submitted the Hazards Program Evaluation Supplement on May 25, 1989, and included additional information which provided acceptable confirmations of the actions requested in item (6), SSER-8. The Hazards '
Program and SWEC's evaluation indicated that the program and its implementation are acceptable. The staff concludes that the Limerick Unit 2 IDCA and the staff's reviews have confirmed that the Limerick Unit 2 design program has been satisfactorily implemented.
Limerick SSER 9 17-4 '
17.6.3 Conclusions PECo has initiated and completed on independent assessment of the design and construction processes for Limerick Generating Station Unit 2. The staff has reviewed this assessment and found it to be appropriately conducted and its results valid. The results conclude:
The design of safety related systems and structures for Limerick 2 complies with licensing commitments and is technically adequate.
- The construction of safety related systems and structures for Limerick 2 is satisfactory and is generally in accordance with drawings and specifications.
The design and construction process employed for Limerick 2 is an acceptable process.
i Limerick SSER 9 17-5
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APPEND]X A'
' CHRONOLOGY LIMERICK GENERATING STATION, UNIT 2 May 31,'1989 Letter from applicant certifying that facility designed, constructed'and tested in compliance with l 10 CFR 20, 50, 51 and 100 with exception of specific L exemptions requested per Commission regulations, L Section 50.12 and Commission May 5, 1989 order.
1 May 31, 1989' Letter from applicant forwarding marked-up FSAR pages, incorporating extended load line region, ,
increased core' flow, partial feedwater heating and' L proprietary GE Reports NEDC-31577P and NEDC-31578P.
Reports withheld (reference 10 CFR 2.790).
.May 31, 1989 Letter to licensee forwarding Amendment 21 to License
'EPF-39 and safety evaluation. Amendment revises Technical Specifications to permit use of increased pore size of filters used during testing of diesel generator fuel oil.
1 June 2, 1989 Letter from applicant responding to NRC May 19, 1989 request for utility to review and certify final draf t version of Unit 2 Technical Specifications. One major change that is required to be made to final draft is incorporation of changes to standby liquid control system Technical Specifications.
June 2, 1989 Letter from applicant submitting follow-up activity in which four fuel assemblies containing fuel rods with alternate clad surface processes will be inserted into facility initial core. Ho new materials introduced to reactor environment.
June 2, 1989 Letter from applicant confirming that rod sequence control system deleted and rod worth minimizer setpoins 1cwered, per NRC February 7,1989 approval of November 9, 1988 request for subject changes.
June 2, 1989 Letter from applicant forwarding Revision I to "PECO Response to NRC Bulletin 88-005 for Limerick Generating Station Unit 2," incorporating comments and providing additional information per April 27,
-1989 and May 18, 1989 requests. ,
1 Limerick SSER 9 Appendix A l
June 5, 1989 Letter from licensee forwarding ar 'icant motion for clarification of Connission deleg, in of authority and for issuance of operating license or alternatively, 4 for exemption from any requirement that license for Unit 2 cannot be issued.
June 5, 1989 Letter from applicant requesting that NRC be prepared l to issue license authorizing unit fuel loading and operation up to 5% of rated power as early as Juno 16, i 1989. Testing activities that may not be completed at initial fuel loading enclosed.
June 7, 1989 Letter from licensee forwarding signed affidavit cf C. A. McNeill, per applicant June 5, 1989 letter to be substituted for copy attached to applicant motion for clarification of Commission delegation of authority and for issuance of operating license or for exemption.
l June 8, 1989 Letter to licensee forwarding Amendment 22 to License ]
NPF-39 and safety evaluation. Amendment revises d Technical Specifications regarding standby liquid control system to ensure compliance with Paragraph (c)(4) of ATWS rule (10 CFR 50.62) to simplify and {
improve specifications for system. 1 June 9, 1989 Letter from licensee requesting temporary wafver of compliance for facility to allow tima for processing 1 of emergency Technical Specificatica change request.
Depressurization of accumulators through accumulator check valves not significant is, current operating condition.
June 9, 1989 Letter to licensee granting above temporary waiver of compliance June 10, 1989 Application for amendment to License NPF-39, consisting of Technical Specification Change Request 89-05, revising Surveillance Requirement 4.1.3.5.b.2 regarding control rod accumulator check valve measuring and recrrding times.
June 14, 1989 Letter to licensee forwarding Amendment 23 to License HPF-39 and safety evaluation. Amendment changes Technical Specifications to reflect completion and tie-in of standby gas treatment system and refueling area HVAC system.
Limerick SSER 9 Appendix A
June 14, 1989 Letter from licensee advising that plant operations review committee chairman will be responsible for ensuring that individual who satisfies Regulatory Guide 1.8, Revision 1-R qualifications for manager will participate in committee activities, per May 17, 1989 discussion.
June 14, 1989 Letter to licensee forwarding safety evaluation regarding increased core flow analysis, partial feedwater heating analysis and extended load line limit analysis.
June 15, 1989 Letter from applicant confirming that utility will be read; for NRC to issue operating license to permit fuel load to begin on June 21, 1989, per 10 CFR 50.50, which permits initial loading of fuel to begin on June 21, 1989 and provides clarification of information contained in June 5, 1989 letter.
June 16, 1989 Letter to licensee forwarding Amendment 25 to Lic m e NPF-39 and safety evaluation. Amendment revises Technical Specifications to increase minimum level of water that must be maintained in spray pond to support operation of Unit 2.
June 19, 1989 Letter from licensee confirming withdrawal of request for exemption from 10 CFR 50.44(c)(3)(ii)(B).
June 19, 19_89 Letter from licensee providing status of modifications regarding four outstanding control room human engineering discrepancies (HEDs) identified during Unit 1 CRDR program. HED TA-03 resolved and HED Al-02 reevaluated and priority lowered from priority 2 to 4.
June 19, 1989 Letter to licensee forwarding Amendnent 26 to License NPF-39 and safety evaluation. Amendment revises effluent dose limits to per site basis.
Jur.c 20, 1989 Letter to litersee forwarding Amendment 28 to License NPF-39 and safety evaluation. Anendment revises Technical Specifications to delete requirements that APRMs be operable when plant is in cold shutdown condition.
Limerick SSER 9 Appendix A
l June 20, 1989 Letter to licensee forwarding Amendment 27 to License NPF-39 and safety evaluation. Anendment revises Technical Specifications regarding RHR and emergency service water system to reflect unit operation.
June 20, 1989 Letter to applicant forwarding Safety Evaluation Report (SER) accepting utility March 31, 1989 and June 2, 1989 responses to NRC Bulletin 88-005, .
" Nonconforming Materials Supplied by Piping Supplies, Inc. at Folsom, NJ and West Jersey lifg Co. at Williamstown, NJ." j June 20, 1989 Letter to applicant approvirs utility June 21, 1988 power ascension test program for facility, per ,
Regulatory Guide 1.68, fe,r execution. ,
1 June 21, 1989 Letter to licensee forwarding SER accepting utility September 6, 1983, November 10, 1983 and May 8, 1984 responses to Generic Letter 83-28, Item 4.5.3 "
regarding on-line functional tes ing of reactor trip system, per BWR Owners Group Report NECD-30844.
June 21, 1989 Letter from applicant forwarding response to items requiring additional information and revised or supplemental information to Independent Construction Assessment Inspection Report 50-353/89-200.
June 22, 1989 Letter to applicant forwarding License NPF-83 !
authorizing fuel loading and precriticality testing of utility and FR notice. Amendment 4 to Indemnity Agreement B-101 also enclosed.
June 22, 1989 Summary of June 15, 1989 meeting with Office of Executive Director, Nuclear Reactor Regulation, Office of General Counsel and Office of Nuclear Regulatory Research regarding progress of work on severe accident mitigation design alternative issue for plant.
June 22, 1989 Letter to licensee forwarding Amendment 29 to License NPF-39 and safety evaluation. Amendment revises Technical Specifications to achieve consistency, remove outdated material and make minor text changes and correct errors.
I Limerick SSER 9 Appendix A
y- )
, n h, .
,S
' June 22,'1989- Letter to licensee forwarding Amendment 29 to License NPF-39 and safety evaluation. Amendment revises-Technical Specifications to ar'11 eve consistency, remove outdated material and make minor text' changes and correct errors.
~~ June.23, 1989. Letter from licensee advising that utility did.not purchase any safety-related components and/or parts for. Planned Maintenance System, Inc. from July 1, 1985'to present.
June 23, 1989 Letter from licensee respon' ding to reques't for additional information regarding' consideration of severe accident mitigation design alternatives.
Tables. listing current estimated core-damage frequency per reactor year and dominant population dose sequences enclosed.
June 28, 1989 Generic Letter 89-10 to all licensees-of operating nuclear power plants and holders of cps for nuclear.
power plants regarding safety-related motor-operated valve testing and surveillance.
June 28, 1989 Letter to licensee forwarding corrected Page 3/4 7-5 and overleaf Page 3/4 7-6 for Technical. Specifications to increase minimum level of water that must be maintained in spray pond to support operation of units per Amendment 25 to license NPF-39.
June 28, 1989 Letter to licensee forwarding corrected Pages 3/4 3-21 and 3/4 3-22 for Amendment 29 to License NPF-39.
. June 29, 1989 Letter from licensee forwarding Amendment 95 to OL application for Licenses NPF-39 and NPF-83, consisting of Revision 58 to FSAR (filed in Category K) and Revision 12 to fire protection evaluation report and Revision 19 to emergency plan (filed in Category F).
June.29, 1989 Letter from licensee forwarding Revision 58 to FSAR (filed in PDR Category K).
i l
l Limerick SSER 9 Appendix A
1 June 30, 1989 Generic Letter 89-11 to all holders of OLs or cps for BWRs regarding resolution of Generic Issue 101, "BWR Water Level Redundancy."
July 3, 1989 Letter to applicant forwarding Technical Specifications, NUREG-1360 and SSER 8 regarding application for OL for Unit 2. Without enclosure.
July 10, 1989 Letter to licensee issuing License NPF-84 and forwarding Technical Specifications, NUREG 1371.
July 24, 1989 Letter to licensee forwarding Amendment 31 to License NPF-39.
Limerick SSER 9 Appendix A
i APPENDIX H Principal Staff Contributors
-Supplement 9 to the SER is a product of the NRC staff. The NRC staff members.
listed below were principal contributors to this report.
[ Name Unit Iqbal Ahmed Instrumentation & Controls-Walter R. Butler. Project Directorate I-2 Richard J. Clark Prcject-Directorate I-2 Kulin D. Desai- Reactor Systems Michele Evans Resident Inspector 1- Roy L. Fuhrmeister Resident Inspector Mark Hartzman Mechanical Engineering Thomas J.- Kenny Senior Resident Inspector Hulbert C. Li Instrumentation & Controls Margaret B. O'Brien Project Directorate I-2 Ronald W. Parkhill Special Inspections Howard J. Richings Reactor Systems Larry L. Scholl Resident Inspector Carl S. Schulten Technical Specifications Steven R. Stein Speciel Inspections.
George Thomas Reactor Systems Ed Trottier Project Directorate I-2 John C. Tsao Materials Engineering l
Limerick SSER 9 Appendix H
APPENDIX U ERRATA to the Safety Evaluation Report for the Limerick Generating Station L- Location Current Wording Revision Pg vii, line 15 "other' "Other" line 25/26 -
Add "H PRINCIPAL STAFF CONTRIBUTORS" l Pg 1-1, line 6 " licensees" " licenses" i line 42 "Public" "Public" Pg 1-2, line 24 "
Add Stairway" " Add Stairway" line 28 " Reactor" " Rod" Pg 3-1, line 35 "P "P is" Pg 3-2, line 32 "(E,is" S Inc.)" "(hSI)"
i' line 33 "(WJMC)" "(WJM)"
line 34 "(CLMM)" "(CLM)"
line 39 "PS Inc., WJMC, " PSI, WJM and CLM" and CLMM" i l Pg 4-2, line 8 "(Ref. 3)" delete ;
i Pg 6-1, line 4 "SGRS" "SGTS" line 27 "GDC-56; and" "GDC-56 and" Pg 7-1, line 4 "7.2.27" "7.2.2.7" line 7 "Amendnment 45" " Amendment 45" Pg 7-2, line 27 "NEDE-24011-P~ A" "NEDE-24011-P-A" Pg 7-3, line 6 " bypassing" " Bypassing" line 15 " staff" " staff's" Pg 10-1, line 13/14 -
Add blank line Pg 15-1, line 19 "2.1.1, and" "2.1.1 and" Pg 16-2, line 32/33 "Diffe "/"rential" " Differ "/"ential" Pg 17-1, line 22 " Limerick Unit 1" " Limerick Unit 2" line 29 " function An" " function. Ai "
Pg 17-3, line 10 "commtments" " commitments" line 15 "supplemtnal" " Supplemental Pg 18vi, line 5 " Nuclear" delete Pg 18-2, line 29 "an displays" "and displays" Pg 18-3, line 26 " Nuclear" delete Pg 19-1, line 35 " operation" Pg 22-1, line 8 ""50.33(k)(a operating")" "50.33(k)(2)"
Pg 22-2, line 5 "(20 in" "(2) in" line 8 "50.12(a)(v)," "50.12(a)(2)(v),"
Appendix H missing Included with SSER-9 I Appendix U Pg 1, line 18 "100 percent" "100-percent" line 25 "Page 20-3," '"Page 10-3,"
Limerick SSER 9 Appendix U
J APPENDIXLU ERRATA to the' Safety Evaluation Report for the Limerick Generating Station Iocation Current Wording Revision Pg vii, line 15 "other' "Other" line 25/26 --
Add "H PRINCIPAL STAFF CONTRIBUTORS" Pg 1-1, line 6' " licensees" " licenses" line 42- "Public" "Public" Pg 1-2, line 24- "
Add Stairway" " Add Stairway" line 28 " Reactor" " Rod" Pg 3-1, line 35 "P , is" "P is" Pg 3-2, lir.e 32. "(kSInc.)" "(hSI)"
line 33 "(WJMC)" "(WJM)"
line 34 "(CLMM)" "(CLM)"
line 39 "PS Inc., WJMC, " PSI, WJM and CLM" and CLMM" Pg 4-2, line 8 "(Ref. 3)" delete Pg 6-1, line 4 "SGRS" "SGTS"
'line 27 "GDC-56; and" - "GDC-56 and" Pg 7-1, line 4 "7.2.27" "7.2.2.7" line 7 "Amendnment 45" " Amendment 45" Pg 7-2, line 27 "NEDE-24011-P~ A" "NEDE-24011-P-A" Pg 7-3, line 6 " bypassing" " Bypassing" line 15 " staff" " staff's" Pg 10-1, line 13/14 -
Add blank line Pg 15-1, line 19 "2.1.1. and" "2.1.1 and" Pg 16-2, line 32/33 "Diffe "/"rential" " Differ "/"ential" Pg 17-1, line 22 " Limerick Unit 1" " Limerick Unit 2" line 29 " function An" " function. An" Pg 17-3, line 10 "commtments" " commitments" line 15 "supplemtnal" " Supplemental" Pg 18-1, line 5 " Nuclear" delete Pg 18-2, line 29 "an displays" "and displays" Pg 18-3, line 26 " Nuclear" delete Pg 19-1, line 35 " operation" Pg 22-1, line 8 ""50.33(k)(a operating")" "50.33( k)(2 )"
Pg 22-2, line 5 "(20 in" "(2) in" line 8 "50.12(a)(v)," "50.12(a)(2)(v),"
Appendix H missing Included with SSER-9 Appendix U Pg 1, line 18 "100 percent" "100-percent" line 25 "Page 20-3," "Page 10-3,"
l Limerick SSER 9 Appendix U
-- =- - - - - - - - -- - -- - - -- -- -- - - - - - - - - - - - - - - - - -- -
NRC sOAM 335 U.S. NUCLE AR REGUL ATORY COMMISSION l 1. Rt PORT NUMEt 4 20 uet I af 22 'J5"u"2.d,,f4'li ""-
mm BIBLIOGRAPHIC DATA SHEET an ansttar,or:, an une rea,n NUREG-0991
- 7. Tn LE ANo buto n Lt Supplement No. 9 Safety Evaluation Report related to the operation of Limerick oATE REroRT eveusm0 Generating Station, Units 1 and 2 a m,c ,, o,4 AUGUST I 1989
- 4. F IN OR GR ANT NUMBE R t AUTHORtS6 6 TYPE OF REPOR1 f
I L VL RLOD COVL Rt D nnetume Darna i
- * ',,"!L",,.',t?,0?
,, T 2^r"'**""'~^*'"""^"""'"""'""*""*""""'"""*""*""'**""""""*""'""*"'"""'"'"*"'"*""""'"'""'"'"
Division of Reactor Projects I/11 l
Office of Nuclear Reactor Regulation {
i I
U.S. Nuclear Regulatory Commission l
Washington, D.C. 20555 9 SVON50RtNG ORGANtlATION - N AME AND ADDRESS tot Nnc. syre mnw a arme or entra c soa.p.. wren nec Dowwon. onacr or nnen u s Nackar k.guieror, cor,,nawon, one oneo,ornt noorna Same as 8 above.
10 SUPPLE ME NT ARY NOTEL Pertains to Docket Nos. 50-352 and 50-353.
- 11. ABST R ACT (73O wurm or eu In August 1983 the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-0991) regarding the application of the Philadelphia Electric Company (the licensee) for the licenses to operate the Limerick Generating Station, Units 1 and 2, located on a site in Montgomery and Chester Counties, Pennsylvania.
Supplement I was issued in December 1983. Supplement 2 was issued in October 1984. Supplement 3 was issued in October 1984. Supplement 4 was issued in May 1985. Supplement 5 was issued in July 1985. Supplement 6 was issued in August 1985 and Supplement 7 was issued in April 1989. Supplement 7 addresses the major design differences between Units 1 and 2, the resolution of all issues that remained open when the Unit 1 full-power license was issued, the staff's assessment regarding the application by the licensee to operate Unit 2 and issues that require resolutic, efore issuance of an operating license for Unit 2.
Supplements 8 and 9 add, is further issues that require resolution prior to issuance of an operating license.
- 12. KE Y WORDS/DL5CR:P f 0H5 ft et worm pr arwen reser =<t/ mme rnme ermr an iorm"ny u.e moort i u A & As L Abit 3 5 blal t ML NI Unlimited 14 ht LUHi l V LLAWt rLai sun Safety Evaluation Report g,,,,-
Limerick Generating Station, Units I and 2 Unclassified u,,,,u,,,,-,,
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