ML20205E469
| ML20205E469 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 07/31/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1048, NUREG-1048-S06, NUREG-1048-S6, NUDOCS 8608180366 | |
| Download: ML20205E469 (84) | |
Text
_ _ _ . . - -_. - -. ._ _____
, Supplement No. 6
' Safety Evaluation Report related to the operation of Hope Creek Generating Station
, Docket No. 50-354 Public Service Electric and Gas Company Atlantic City Electric Company i
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation July 1986 fa "%,,
a E i D DO K 5000354 E PDR
%m NOTICE Availability of Reference Materials Cited in NRC Puc!ications Most documents cited in NRC publications will be available from one of the following sources:
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Washington, DC 20555
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NUREG-1048 Supplement No. 6 l
Safety Evaluation Report related to the operation of Hope Creek Generating Station Docket No. 50-354 Public Service Electric and Gas Company Atlantic City Electric Company U~.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation July 1986 f a ....,,
/
ABSTRACT Supplement No. 6 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company on its own behalf as co-owner and as agent for the other co-owner, the Atlantic City Electric Company, for a license to operate Hope Creek Generating Station has been prepai'ed by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. The facility is located in Lower A110 ways Creek Township in 5alem County, New Jersey.
This supplement reports the status of certain items that had not been resolved at the time of the publication of the Safety Evaluation Report. This supple-ment supports the issuance of a full power license to operate Hope Creek Generating Station.
Hope Creek SSER 6 iii
l TABLE OF CONTENTS P, age ABSTRACT ........................................................ iii 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT .............. 1-1 1.1 Introduction .......................................... 1-1
- 1. 7 Outstanding Issues .................................... 1-2 1.8 Confirmatory Issues ........... ....................... 1-2
- 1. 9 License Condition Items ............................... 1-2 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, EQUIPMENT, AND COMP 0NENTS.............................................. 3-1 3.5 Missile Protection..................................... 3-1 3.5.1 Missile Selection and Description............... 3-1
- 3. 5.1. 3 Turbine Missiles....................... 3-1 3.5.1.3.3 Summary and Conclusions..... 3-1 3.11 Environmental Qualification of Mechanical and Electrical Equipment............................. ..... 3-1 3.11.5 Conclusion...................................... 3-1 7 INSTRUMENTATION AND CONTROLS ............................... 7-1 7.2 Reactor Protection (Trip) System....................... 7-1 7.2.2 Specific Findings............................... 7-1 7.2.2.8 Anticipated Transients Without Scram... 7-1 7.3 Engineered Safety Feature Systems ..................... 7-2 7.3.2 Specific Findings .............................. 7-2 7.3.2.5 Solid-State Logic Modules ............. 7-2 9 AUXILIARY ~YSTEMS........................................... 9-1 9.5 O the r Auxi l i a ry Sys tems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.5.1 Fire Protection.................................. 9-1 Hope Creek SSER 6 v
1 1
TABLE OF CONTENTS (Continued)
Pag 14 INITIAL TEST PR0 GRAM......................................... 14-1 14.2 Initial Plant Test Program - Final Safety Analysis Report................................................. 14-1 l
15 S A F ETY AN A LYS I S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-1 15.1 Decrease in Core Coolant Temperature ................... 15-1 16 TECHNICAL SPECIFICATIONS .................................... 16-1 APPENDIX A CONTINUATION OF CHRON0 LOGY APPENDIX B BIBLIOGRAPHY-APPENDIX C UNRESOLVED SAFETY ISSUES APPENDIX D ACRONYMS AND INITIALISMS APPENDIX E PRINCIPAL STAFF CONTRIBUTORS APPENDIX S REVIEW 0F POWER ASCENSION PROGRAM ACCELERATION APPENDIX T TECHNICAL SPECIFICATION CHANGES BETWEEN LOW-POWER LICENSE ISSUANCE AND FULL-POWER LICENSE ISSUANCE APPENDIX U PROBABILITY OF MISSILE GENERATION IN GENERAL ELECTRIC NUCLEAR TURBINES LIST OF TABLES 1.1 Outstanding Issues (Revised Table 1.1 From Supplement No. 5) ......................................................
1-3 1.2 Confirmatory Issues (Revised Table 1.2 From Supplement No. 5) ......................................................
1-5 1.3 License Conditions (Revised Table 1.3 From Supplement 1-8 No. 5) ......................................................
7.1 Generic Letter 83-28 Review for Hope Creek .................. 7-6 16.1 Technical Speci fication Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16-2 l
Hope Creek SSER 6 vi
1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 Introduction In October 1984, the U.S. Nuclear Regulatory Commission (NRC) staff issued its Safety Evaluation Report (SER) (NUREG-1048) on the application filed by Public Service Electric and Gas Company (PSE&G) (hereinafter referred to as the licensee) on its own behalf as co-owner and as agent for the other co-owner, the Atlantic City Electric Company, for a license to operate the Hope Creek Generating Station (Docket No. 50-354). At that time, the 3taff identified items that were not yet resolved with the licensee. Supplement Nos. 1, 2, 3, 4, and 5 to the SER were issued in March 1985, August 1985, October 1985, December 1985, and April 1986, respectively. Supplement No. 5 was issued in support of the issuance of the low power license. On April 11, 1986, this license, NPF-50, was issued to allow Hope Creek to operate at power levels not in excess of 5% of rated power. Except for one issue, SER Outstanding Issue 5 (solid-state logic modules), all SER confirmatory and outstanding (open) issues were resolved in Supplement Nos. 1 through 5. The purpose of this supplement to the SER is to provide the staff evaluation of the resolution of this open issue, to document Technical Specification changes, and to report on the modi-fication of certain license conditions. This supplement supports the issuance of a full power license for Hope Creek.
Each of the following sections or appendices of this SER supplement is numbered the same as the corresponding SER section or appendix that is being updated.
Appendix A is a continuation of the chronology of the staff's actions related to the processing of the Hope Creek application and lists letters between the NRC staff and the licensee in chronological order. Appendix B is a list of refer-ences cited in this report.* Appendix C documents the resolution of Unresolved Safety Issue A-43, " Containment Emergency Sump Reliability." Appendix D is a list of acronyms used herein. Appendix E identifies principal contributors to this SER supplement. Appendix S contains the staff's review of the power ascension program acceleration. Appendix T documents the staff's review of certain Technical Specification changes requested by the licensee. Appen-dix U contains the staff's evaluation of the General Electric Company report,
" Probability of Missile Generation in General Electric Nuclear Turbines."
Copies of this SER supplement are available for inspection at the NRC Public Document Room at 1717 H Street, N.W., Washington, D.C., and at the Pennsville Public Library, 190 South Broadway, Pennsville, New Jersey. They are also avail-able for purchase from the sources indicated on the inside front cover of this report.
The NRC Project Manager assigned to the operating license application for Hope Creek is Mr. David H. Wagner. Mr. Wagner may be contacted by writing to
- Availability of all material cited is described on the inside front cover of this raport.
Hope Creek SSER 6 1-1
Mr. David H. Wagner Division of BWR Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555 1.7 Outstanding Issues The staff identified certain outstanding issues in the SER that had not been resolved with the licensee. The status of these issues is listed in Table 1.1 and discussed further in the indicated sections of this report. I 1
1.8 Confirmatory Issues The staff identified confi matory items in the SER that required additional in-formation to confirm preliminary conclusions. The status of these items is listed in Table 1.2.
1.9 License Condition Items There are certain issues for which a license condition may be desirable to en-sure that staff requirements are met by a specified date (Table 1.3). These conditions will be in the form of a condition in the body of the operating license.
E Hope Creek SSER 6 1-2
i Table 1.1 Outstanding issues (revised Table 1.1 from Supplement No. 5) i Issue Status SER section(s)
(1) Riverborne missiles Closed 3.4.1 (Supplement 4)
(2) Equipment qualification Closed 3.10, 3.11 (also 3.10, Supplements 2 and 3)
(3) Preservice inspection program Closed 5.2.4, 6.6 (also Supple-ment 3)
(4) GDC 51 compliance Closed 6.2.7 (Supplement 2)
(5) Solid-state logic modules Closed 7.3.2.5 (6) Postaccident monitoring Closed 7.5.2.3 instrumentation (Supplement 2)
(7) Minimum separation between Closed 8.3.3.3.3 non-Class 1E conduit and (Supplement 3)
Class 1E cable trays (8) Control of heavy loads Closed 9.1.5 (Supplement 1)
(9) Alternate and safe shutdown Closed 9.5.1.4 (also Supple-ment 2)
(10) Delivery of diesel generator Closed 9.5.4.2 fuel oil and lube oil (Supplement 1)
(11) Filling of key management Closed 13.1 positions (Supplement 5)
(12) Training program items (a) Initial training programs Closed 13.2.1.1 (Supplement 2)
(b) Requalification training Closed 13.2.1.2 programs (Supplement 2)
(c) Replacement training Closed 13.2.1.3 programs (Supplement 2)
(d) TMI issues I.A.2.1, I.A.3.1, Closed 13.2.1.4 and II.B.4 (Supplement 2)
Hope Creek SSER 6 1-3
l l
Table 1.1 (Continued)
Issue Status SER section(s) l l
(e) Nonlicensed training Closed 13.2.2 ,
programs (Supplement 2)
(13) Emergency dose assessment Closed 13.3.2.9 computer model (Supplement 5)
(14) Procedures generation package Closed 13.5.2 (Supplement 5)
(15) Human factors engineering Closed 18.1, 18.2 (Supplement 5)
Hope Creek SSER 6 1-4
Table 1.2 Confirmatory issues (revised Table 1.2 from Supplement No. 5)
Issue Status SER section(s)
(1) Feedwater isolation check valve Clc ad 3.6.2 analysis (Supplement 3)
(2) Plant-unique analysis report Closed 3.9.3.1, (also Suople-ment 3)
(3) Inservice testing of pumps and Closed 3.9.6 valves (Supplement 4)
(4) Fuel assembly accelerations Closed 4.2 (Supplement 2)
(5) Fuel assembly liftoff Closed 4.2 (Supplement 2)
(6) Review of stress report Closed 5.2.1.1 (Supplement 3)
(7) Use of Code cases Closed 5.2.1.2 (Supplement 2)
(8) Reactor vessel studs and Closed 5.3.1.5 fasteners (Supplement 3)
(9) Containment depressurization Closed 6.2.1.4 analysis (Supplement 5)
(10) Reactor pressure vessel shield Closed 6.2.1.5 annulus analysis (Supplement 3)
(11) Drywell head region pressure Closed 6.2.1.5 response analysis (Supplement 3)
(12) Drywell-to-wetwell vacuum Closed 6.2.1.7 breaker loads (Supplement 3)
(13) Short-term feedwater system Closed 6.2.3 analysis (Supplement 3)
(14) Loss-of-coolant-accident Closed 6.3.5, 15.9.3 analysis (Supplement 2)
(15) Balance-of plant testability Closed 7.2.2.3 analysis (Supplement 4)
(16) Instrumentation setpoints Closed 7.2.2.5 (Supplement 4)
Hope Creek SSER 6 1-5
Table 1.2 (Continued)
Issue Status SER section(s)
(17) Isolation devices Closed 7.2.2.6 (Supplement 5)
(18) Regulatory Guide 1.75 Closed 7.2.2.7 (Supplement 5)
(19) Reactor mode switch Closed 7.2.2.9 (Supplement 3)
(20) Engineered safety features Closed 7.3.2.6 reset controls (Supplement 5)
(21) High pressure coolant injection Closed 7.3.2.9 initiation (Supplement 3)
(22) IE Bulletin 79-27 Closed 7.4.2.1 (Supplement 3)
(23) Bypassed and inoperable status Closed 7.5.2.4 indication (Supplement 4)
(24) Logic for high pressure coolant Closed 7.6.2.1 injection interlock circuitry (Supplement 4)
(25) End-of-cycle recirculation pump Closed 7.6.2.4 trip (Supplement 3)
(26) Multiple control system failires Closed 7.7.2.1 (Supplement 3)
(27) Relief function of safety / relief Closed 7.7.2.2 valves (Supplement 3)
(28) Main steam tunnel flooding Closed 8.3.3.1.4 analysis (Supplement 3)
(29) Cable tray separation testing Closed 8.3.3.3.2 (Supplement 3)
(30) Use of inverter as isolation Closed 8.3.3.3.4 device (Supplement 3)
(31) Core damage estimate procedure Closed 9.3.2 (Supplement 3)
(32) Continuous airborne particulate Closed 12.3.4.2 monitors (Supplement 3)
(33) Qualifications of senior Closed 12.5.1 radiation protection engineer (Supplement 2)
Hope Creek SSER 6 1-6
f~
Table 1.2 (Continued)
Issue Status SER section(s)
(34) Onsite instrument information Closed 12.5.2 (Supplement 3)
(35) Airborne iodine concentration Closed 12.5.2 instruments (Supplement 3)
(36) Emergency Plan items Closed 13.3 (Supplement 5)
(37) TMI Item II.K.3.18 Closed 15.9.3 (Supplement 2)
(38) Independent design verification Closed 17.5 (Supplement 5)
Hope Creek SSER 6 1-7
Table 1.3 License conditions (revised Table 1.3 from Supplement No. 5)
License condition Status SER section (1) Turbine system maintenance program Removed 3.5.1.3.3 (2) NUREG-0803 implementation Removed 4.6 (Supplement 5) 1 (3) Inservice inspection Revised 5.2.4.3 and 6.6.3 (Supplement 5)
(4) Postaccident sampling system Removed 9.3.2 (Supplement 3)
(5) Solid waste process control Revised 11.4 program (Supplement 5)
(6) Partial feedwater heating Revised 15.1 (7) Cask drop accident Removed 15.7.5 (Supplement 3)
(8) Inservice testing of pumps and valves 3.9.6 (Supplement 4)
(9) Environmental qualification Removed 3.11.5 (10) Fire protection Revised 9.5.1 (11) Emergency planning 13.3.3 (Supplement 5)
(12) Initial startup test program 14.2 (Supplement 5)
(13) Detailed control room design review 18.1 (Supplement 5)
(14) Fuel storage and handling 9.1 (Supplement 5)
(15) Safety parameter display system 18.2 (Supplement 5)
(16) Solid-state logic modules 7.3.2.5 Hope Creek SSER 6 1-8
3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, EQUIPMENT, AND COMP 0NENTS 3.5 Missile Protection 3.5.1 Missile Selection and Description 3.5.1.3 Turbine Missiles 3.5.1.3.3 Summary and Conclusions License Condition 2.C.(3) of Facility Operating License NPF-50 for Hope Creek requires that PSE&G submit a turbine system maintenance program based on the turbine manufacturer's calculations of missile generation probabilities no later than April 11, 1989. In a letter dated July 7, 1986 (C. McNeill, PSE&G, to E. Adensam, NRC), the licensee submitted a turbine system maintenance program based on the proprietary General Electric Company (GE) report, " Probability of Missile Generation in General Electric Nuclear Turbines." The GE report was submitted by the licensee in a letter dated July 11, 1986 (C. McNeill, PSE&G, to E. Adensam, NRC).
The intent of the turbine system maintenance program is to provide early detec-tion of cracking in the low pressure turbine wheels, and thereby reduce the probability of turbine wheel failure. The maintenance program submitted by the licensee will include disassembly of the low pressure turbines in stages over i a 6 year period (during plant shutdowns) so that all low pressure turbine wheels '
are inspected within the 6 year period. The program includes a complete inspec-tion of all normally inaccessible parts such as couplings, coupling bolts, tur-bine shafts, low pressure turbine buckets, and the low pressure turbine wheels.
The low pressure turbine wheels will be subject to volumetric examination.
The staff has reviewed the licensee's turbine system maintenance program and concludes that the low pressure turbine wheel inspection interval of 6 years is supported by the missile generation probabilities derived from the GE report,
" Probability of Missile Generation in General Electric Nuclear Turbines."
Furthermore, License Condition 2.C.(3) of the Hope Creek license (NPF-50) has been satisfied, and the staff recommends deletion of this license condition from the full power license. (The staff's evaluation of the GE report sub-mitted by the licensee's July 11, 1986, letter, is contained in Appendix U to this supplement.)
- 3.11 Environmental Qualification of Mechanical and Electrical Equipment 1
i 3.11.5 Conclusion License Condition 2.C.(5)a of Facility Operating License NPF-50 for Hope Creek requires that, before startup following the first refueling outage, the quali-fled life of the electrical equipment under Purchase Order M-48 shall be recal-culated on the basis of the actual temperatures monitored at the equipment lo-cations during the first cycle of operation, with adequate consideration of margin.
Hope Creek SSER 6 3-1 l
l l
l
The staff recommended this license condition be included in the license because during the environmental qualification audit, it was noted that the licensee used the calculated average temperature for one Purchase Order M-48 item (pri-mary containment instrument gas compressor).
In a letter dated June 2, 1986 (C. McNeill, PSE&G, to E. Adensam, NRC), the licensee indicated that it has reevaluated the issue and has revised qualified life values for electrical equipment purchased under Purchase Order M-48 for the maximum normal service temperature. The new values have been incorporated in the Hope Creek maintenance and surveillance program to reflect the revised replacement schedule for the components.
The staff has reviewed the information provided in the licensee's June 2, 1986, letter and concludes that the intent of the license condition has been met.
Accordingly, the license condition can be deleted.
License Condition 2.C.(5)b of Facility Operating License NPF-50 for Hope Creek requires that, before initial criticality, the 53 Tobar Model 32, Series 2 trans-mitters included in the harsh environment qualification program be replaced with qualified Rosemount Model 1153 B transmitters. In a letter dated May 27, 1986 (C. McNeill, PSE&G, to E. Adensam, NRC), PSE&G confirmed that this had been done. Accordingly, License Condition 2.C.(5)b of the low power license is satisfied and no longer required.
Hope Creek SSER 6 3-2
7 INSTRUMENTATION AND CONTROLS
- 7. 2 Reactor Protection (Trip) System 7.2.2 Specific Findings 7.2.2.8 Anticipated Transients Without Scram On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed to open on an automatic reactor trip signal from the reactor protection system. This incident occurred during the plant startup, and the reactor was tripped manually by the operator about 30 sec after the initiation of the automatic trip signal. The failure of the circuit breakers has been determined to be related to the sticking of the undervoltage trip attachment. Before this incident, on February 22, 1983, at Unit 1 of the Salem Nuclear Power Plant, an automatic trip' signal was generated based on steam gen-erator low-low level during plant startup. In this case, the reactor was tripped manually by the operator c1most coincidentally with the automatic trip. Follow-ing these incidents, on February 28, 1983, the NRC Executive Director for Opera-tions directed the NRC staff to investigate and report on the generic implica-tions of these occurrences at Unit 1 of the Salem Nuclear Power Plant. The results of the staff's inquiry into the generic implications of the incidents at the Salem unit are reported in NUREG-1000, " Generic Implications of ATWS
[ Anticipated Transient Without Scram] Events at the Salem Nuclear Power Plant."
As a result of this investigation, the Commission (NRC) requested (by Generic Letter 83-28 dated July 8, 1983) all licensees of operating reactors, appli-cants for an operating license, and holders of construction permits to respond to certain generic concerns.
The licensee responded to these concerns by letters dated March 30 and Decem-ber 17, 1984; February 28, May 21, July 15, August 7, and October 30, 1985; and January 24, 1986.
In a letter dated June 9, 1986 (E. Adensam, NRC, to C. McNeill, PSE&G), the status of the staff's review of these responses was sent to the licensee.
Table 7.1 presents the Hope Creek status of all Generic Letter 83-28 issues applicable to boiling-water reactors. The licensee's responses to all items, except for Items 2.1, 2.2, 4.5.2, and 4.5.3, have been reviewed and found acceptable. As noted in Table 7.1, the review for Items 2.1, 2.2, 4.5.2, and 4.5.3 is ongoing. These four items are primarily administrative in nature and do not affect the plant design (or the implementation of that design).
The review status of these items is similar to that of many operating plants.
Therefore, the staff does not consider the completion of these items a pre-requisite for authorizing full power operation. Should the staff's review conclude that additional actions by the licensee are required to be in com-pliance with Generic Letter 83-28, the licensee will be notified.
Hope Creek SSER 6 7-1
7.3 Engineered Safety Feature Systems 7.3.2 Specific Findings 7.3.2.5 Solid-State Logic Modules With regard to logic functional testing of Bailey solid-state logic modules (SSLMs), the staff recommended in Supplement No. 5 that the licensee review the existing test method (performed every 18 months under cold conditions) to see if the test, with necessary modifications, can be used at power (without challenging plant systems and creating undue competing risks) to demonstrate SSLM reliability. Additionally, the licensee was asked to investigate other methods by which testability frequency can be increased. ;
1 In a letter dated May 23, 1986 (C. McNeill, PSE&G, to E. Adensam, NRC), the licensee provided (1) a listing of Class 1E Bailey 862 SSLMs, which perform logic functions, that are operated or surveilled in the performance of Hope Creek Technical Specification surveillances, (2) a description of Technical !
Specification surveillance testing performed on the Bailey 862 SSLMs, and (3) a description of four options (described below) available to perform in situ, at power surveillance testing of the surveilled Bailey 862 SSLMs identi-fled in Item (1) above and the effect associated with performing each of the four options. Also contained in the May 23, 1986, letter was a proposal from the licensee to develop a program by which reliability data associated with the Bailey 862 SSLMs will be gathered to demonstrate the reliability of the Bailey 862 SSLMs.
(1) Option 1 Option 1 would maintain the current Hope Creek configuration and, where possible, utilize existing procedures, or portions of these procedures, that were developed to comply with Technical Specification surveillance requirements. These surveillance tests would be performed on a more fre-quent basis than currently required (once every 18 months) by the Techni-cal Specifications.
The effects of performing Option 1 would include (a) additional person-hours and cost to perform the surveillance testing (b) potential to enter limiting conditions for operation for those sys-tems rendered inoperable as a result of testing at power (c) utiliza**on of jumpers and lifted leads to prevent undesired isolations/actuations and to simulate various plant conditions (d) disabling Bailey 862 SSLMs logic outputs, thus disabling equipment required to be available to operate or possibly initiate safety functions or permissives on equipment already in operation (e) challenging the ability to control and cool down the reactor in the event of an actual loss-of-coolant accident (LOCA) or loss of power Hope Creek SSER 6 7-2
(f) negating the purpose of performing logic functional testing because portions of the decision logic would be off line (2) Option 2 Option 2 would maintain the current Hope Creek configuration and develop new surveillance test procedures to allow in situ testing while at power. l The specific intent of these new procedures would be to verify function-ally the SSLMs used within a system.
The effects of performing Option 2 would include (a) additional person-hours and extensive expense associated with develop-ing unique procedures for in situ testing of logic modules (b) potential to enter limiting conditions for operation for those systems rendered inoperable as a result of testing at power (c) utilization of jumpers and lifted leads to prevent undesired isolations/
actuations and to simulate various plant conditions (d) disabling Bailey 862 SSLM logic outputs, thus disabling equipment required to be available to operate or possibly initiate safety func-tions or permissives on equipment already in operation.
(e) challenging the ability to control and cool down the reactor in the event of an actual LOCA or loss of power (f) negating the purpose of performing logic functional testing because portions of the decision logic would be off line (3) Option 3 Option 3 would maintain the current Hope Creek configuration, develop new surveillance test procedures to allow in situ testing while at power, and test 100% of a total test population of 577 Bailey 862 SSLMs during each cycle of operation, as outlined in letters dated February 14 and 24, 1986 (C. McNeill, PSE&G, to E. Adensam, NRC).
The effects of performing Option 3 would include (a) additional person-hours and expense to perform the testing (b) disabling Bailey SSLM logic outputs, thus disabling equipment required to be available to operate or possibly initiate safety functions or permissives on equipment already in operation (c) challenging the ability to control and cool down the reactor in the event of an actual LOCA or loss of power (d) negating the purpose of performing logic functional testing because portions of the decision logic would be off line.
Hope Creek SSER 6 7-3
(4) Option 4 Option 4 would maintain the current Hope Creek configuration and testing in accordance with current Technical Specification requirements. There are no effects or operational perturbations associated with performing Option 4.
On the basis of its review of the above.information, the staff has concluded that the tests to be performed under Options 1, 2, or 3 described above will result in disabling of equipment required to be available to operate or possibly J will inhibit safety function or permissives on equipment already in operation. '
With such a plant status at full power, if an emergency condition (LOCA or loss of offsite power) should occur, the plant's ability to cope with such a con-dition could be challenged. Recognizing this limitation in on-line testing, Regulatory Guide 1.22 exempts on-line testing of those systems / components whose testing at power adversely affects the safety or operability of the plant and allows them to be tested while the plant is shut down.
The licensee has stated that the percentage of SSLMs at Hope Creek that are tested monthly, quarterly, and semiannually is 21%, 36%, and 0.72%, respec-tively. The licensee has also indicated that the Bailey 862 SSLMs in operation at Colorado Ute, Craig Station, Unit 3, for the past 36 months had not experi-enced any failures.
Solid-state components and circuits are being increasingly used in safety-related systems in nuclear power plants. On the basis of the current operating experience, the staff believes that the solid-state systems in general are performing acceptably. Because the Bailey SSLMs have not been used in nuclear pou r plant application, it is the staff's determination that the licensee for Hope Creek should demonstrate the reliability of these modules.
On the basis of the above, the staff has concluded that the licensee should adopt Option 4 for SSLMs at Hope Creek.
In letters dated June 13 and 24, 1986 (C. McNeil, PSE&G, to E. Adensam, NRC),
the licensee provided details of.its reliability program to demonstrate the reliability of the Bailey 862 SSLMs. The reliability program would consist of the following three subprograms:
(1) The first subprogram would be an in plant reliability program to monitor the performance of the Bailey 862 SSLMs installed at Hope Creek. This subprogram would obtain reliability data, failure characteristic informa-tion, and root cause of failure of both safety-related and non-safety-related Bailey 862 SSLMs for a period of at least 18 months. This will indicate the actual in plant performance of the Bailey 862 SSLMs.
(2) The second subprogram would consist of an accredited laboratory performing physical testing of a statistical sample of the Bailey 862 SSLMs. These tests would simulate inputs and outputs as near to plant conditions as possible. The tests would include the effects of aging. The test results would be analyzed to verify that the modules will perform their intended safety functions under service conditions. The laboratory would generate a report to provide the results of the reliability testing and reliability analysis.
Hope Creek SSER 6 7-4 L____.__ - - - - - -
(3) The third subprogram would be the collection of reliability data by Bailey Controls over a period of at least 18 months from other iridustrial install-ations of Bailey 862 SSLMs to provide an additional overall reliability basis on the modules.
The reliability program will have specific recommend.ations on qualified life, surveillance requirements, testing frequency, identification of specific param-eters for degradation monitoring, preventive maintenance requirements, limiting conditions for operation, and any corrective actions.
Before the end of the first refueling outage, the licensee should provide an analysis of the results of the reliability program to demonstrate that the SSLMs can perform their safety functions in a reliable manner.
The licensee will submit the details of the reliability test program for staff review by August 15, 1986. The licensee in a letter dated June 24, 1986, has committed to submit for staff review the results of the above-stated reliability program before the end of the first refueling outage. The Hope Creek full power license will be so conditioned.
On the basis of its review of the technical approach, methodology, acceptance criteria, and quality assurance procedures, the staff has concluded that the licensee's proposed reliability program is a viable approach co demonstrate the reliability of SSLMs at Hope Creek.
Because of the generic backfit implications of a reinterpretation of General Design Criterion 21 (Appendix A to 10 CFR 50) regarding testability of protec-tion systems at power, the staff has categorized the on-line testability of protection systems as Generic Issue 120. Any requirements that emanate from the resolution of this generic issue will be applied to the protection system at Hope Creek.
Because Bailey 862 SSLMs at Colorado Ute, Craig Station, Unit 3, had no observed failures in the past 36 months of operation, a certain percentage of SSLMs at Hope Creek will be periodically exercised during power operation, and the sur-veillance requirements of the SSLMs are in accordance with the Standard Tech-nical Specifications, the staff concludes that there is reasonable assurance that the SSLMs at Hope Creek will perform the required safety functions during the first cycle of operation.
The staff recommends that the following license condition be added to the full-power license:
PSE&G shall implement a reliability program, to demonstrate solid state logic module reliability, as described in its letters dated June 13 and 24, 1986. The results of the reliability program shall be submitted to the staff prior to the end of the first refueling outage.
Hope Creek SSER 6 7-5 1
Table 7.1 Generic Letter 83-28 review for Hope Creek Date of letter transmitting safety evaluation Item Status to licensee 1.1 Review is complete. January 22, 1986 1.2 Review is complete. February 3, 1986 2.1 Review is ongoing.
2.2 Review is ongoing.
3.1.1 Review is complete. June 25, 1986 l 3.1.2 Review is complete. June 25, 1986 3.1.3 Review is complete. January 22, 1986 3.2.1 Review is complete. June 25, 1986 3.2.2 Review is complete. June 25, 1986 3.2.3 Review is complete. January 22, 1986 4.5.1 Review is complete. June 25, 1986 4.5.2 Review is ongoing.
4.5.3 Review is ongoing.
Hope Creek SSER 6 7-6
9 AUXILIARY SYSTEMS 9.5 Other Auxiliary Systems 9.5.1 Fire Protection In a letter dated May 13, 1986 (C. McNeill, PSE&G, to E. Adensam, NRC), the licensee requested the deletion of Fire Protection Program (FPP) elements from the Hope Creek Generating Station Technical Specifications when the full power license is issued. The FPP requirements defined in the Technical Specifica-tions would be embodied within a periodic testing and surveillance program at Hope Creek and entail a shifting of testing requirements from surveillance procedures to periodic test procedures. In preparing the request, the li-censee used the guidance contained in Generic Letter 86-10, " Implementation of Fire Protection Requirements."
In response to the licensee's request, the staff performed an inspection at the Hope Creek site to verify that the Technical Specification surveillances are incorporated into station procedures. The results of this inspection are reported in NRC Office of Inspection and Enforcement (IE) Inspection Re-port 50-354/86-29, dated June 25, 1986. In summary the staff found:
(1) The existing Final Safety Analysis Report (FSAR) sections of the FPP and the licensee's May 13, 1986, request, satisfy the guidance in Generic Letter 86-10 for incorporating the FPP into the FSAR.
(I') The deletion of the Technical Specification sections is in accordance with the guidance in Generic Letter 86-10.
(3) The existing fire protection Technical Specification requirements are in-corporated into equivalent plant procedures, and equivalent administrative controls exist to control these activities.
(4) Adequate administrative controls exist to determine if a proposed FPP change would adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
The staff has reviewed the licensee's May 13, 1986, request and concludes that the incorporation of the Technical Specification surveillance procedures into periodic test procedures will result in an equivalent level of fire protection as would have resulted had the FPP elements remained in the Technical Specifi-cations. Therefore, the licensee's request is acceptable.
In the Hope Creek low power license, NPF-50, the fire protection license condi-tion, 2.C.(8), stated, in part, PSE&G shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility through Amendment No. 13 and as Hope Creek SSER 6 9-1
described in submittals dated August 12 and 27, November 21 and 29, December 3 and 16, 1985, January 3 and April 8, 1986, and as ap-proved in the SER dated October 1984 (and Supplements 1 through 5)....
This license condition must be revised to incorporate the latest changes to the FPP and reference the licensee's May 13, 1986, request. FSAR Amendments 14 and 15 incorporated the submittals referenced in the license condition. Accord- l ingly, these submittal dates may be replaced with the proper incorporation of Amendments 14 and 15 into the license condition. Additionally, the license l condition is revised to reflect the licensee's commitments made in the May 13, i 1986, request to delete FPP elements from the Technical Specifications. Ac-cordingly, the new fire protection license condition reads as follows:
PSE&G shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility through Amendment No. 15 and as described in its submittal dated May 13, 1986, and as approved in the SER dated October 1984 (and Supplements 1 through 6) subject to the following provision:
The licensee may make changes to the approved fire pro-tection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
Hope Creek SSER 6 9-2
14 INITIAL TEST PROGRAM 14.2 Initial Plant Test Program - Final Safety Analysis Report At the request of the licens,ee, a meeting was held on August 1, 1985, to dis-cuss the proposed acceleration of the Hope Creek power ascension program (PAP).
(The meeting summary is dated August 19, 1985.) At the meeting, the licensee identified five categories of methods by which the PAP could be accelerated:
(1) Replace testing with Technical Specification surveillances.
(2) Delete nonessential testing.
(3) Simplify certain tests.
(4) Replace tests with data from other tests.
(5) Delete certain Regulatory Guide 1.68 testing.
In letters dated August 21, 1985 (R. Mitt 1, PSE&G, to W. Butler, NRC), Septem-ber 20, 23, and 30, 1985 (R. Mitti, PSE&G, to W. Butler, NRC), October 4 and 17, 1985 (R. Mitt 1, PSE&G, to W. Butler, NRC), November 6,1985 (R. Mitt 1, PSE&G, to W. Butler, NRC), and December 9,1985 (C. McNeill, PSE&G, to E. Adensam, NRC), the licensee submitted PAP test modifications for NRC review. In total, modifications for 26 PAP tests were submitted. In letters dated January 22, February 4, and March 20, 1986 (R. Bernero, NRC, to C. McNeill, PSE&G), the staff provided a safety evaluation for each proposed test modification. These letters are included as Appendix S to this supplement.
Hope Creek SSER 6 14-1
15 SAFETY ANALYSIS 15.1 Decrease in Core Coolant Temperature By letter dated May 8, 1986, the licensee requested that the license condition,
" Partial Feedwater Heating," be changed and that this change be incorporated into the full power license. The license condition currently reads:
PSE&G shall not operate the facility (other than for normal start-up or shutdown) with the feedwater inlet temperature less than 424.5 F.
The licensee informed the staff that, during normal operation at 75% power, the feedwater temperature is estimated to be 398 F. In this particular case, the licensee would be in violation of the license condition as it is currently written. Therefore, the licensee requests a revised license condition for the full power license.
The staff understands that reduced feedwater temperature (RFT) operation is possible during normal operation. Feedwater temperature normally drops with decreasing power level while all feedwater heaters are in service and could be reduced during maintenance on the feedwater system.
The staff has reviewed analyses performed for a BWR/6 using staff-approved models tojustifysteady-stateoperationwithRFTs. Temperatures ranging from 420 F to 250 F were considered. The anticipated operating transients were reevaluated to determine the required operating limit minimum critical power ratio (MCPR) limits.
The operating limit MCPR values were found to increase by 0.01 for operation between feedwater temperatures of 370 F and 320 F and by 0.03 for feedwater tem-peratures between 320 F and 250 F.
The staff has reviewed the application of these findings to Hope Creek opera-tions. Although plant-specific RFT analyses were not performed for Hope Creek, the staff believes operation with RFT, for the first cycle only, would be accept-able, provided that the Hope Creek Technical Specification operating limit MCPR is increased by 0.03 for a feedwater temperature between 400 F and 320"F and by 0.06 for a feedwater temperature between 320 F and 250 F. The additional mar-gin is provided to account for possible differences in transient response between Hope Creek, a BWR/4, and the analyzed BWR/6.
The staff believes that in lieu of the license condition as currently written, the following license condition will permit greater operational flexibility while ensuring operation in analyzed regions:
The facili'.y shall not be operated with reduced feedwater tempera-ture for the purpose of extending the normal fuel cycle. After the first operating cycle, steady state operation with reduced feedwater temperature (relative to the Final Safety Analysis Report analysis value) during the norraal fuel cycle shall be prohibited until plant-specific analyses justifying such operation are provided by PSE&G and approved by the staff.
Hope Creek SSER 6 15-1
In a letter dated June 23, 1986 (E. Adensam, NRC, to C. McNeill, PSE&G), the staff proposed this license condition to the licensee and requested that the licensee notify the staff if the proposed license condition and MCPR modifica-tions were acceptable. In a letter dated July 7, 1986 (C. McNeill, PSE&G, to E. Adensam, NRC), the licensee informed the staff that the proposed license condition was acceptable, with certain clarifications. Additionally, the li-censee proposed a revision to Technical Specification 3.2.3 to incorporate the MCPR modifications. This proposed Technical Specification revision included the addition of a graph of feedwater temperature versus power level to Technical Specification 3.2.3. The staff concluded that the addition of this graph to the Technical Specifications was not acceptable, since the graph had not been pre-viously reviewed and approved by the staff.
On the basis of discussions with the licensee, the licensee and staff both agree that the following license condition, accompanied by an MCPR modification table added to the Technical Specifications, will meet the intent of the staff's June 23, 1986, letter to the licensee. Accordingly, the staff recommends that license condition, " Partial Feedwater Heating," be modified to state:
The facility shall not be operated with reduced feedwater temperature for the purpose of extending the normal fuel cycle. After the first operating cycle, the facility shall not be operated with a feedwater heating capacity that would result in a rated thermal power feedwater temperature less than 400 F unless analyses supporting such operation are submitted by the licensee and approved by the staff.
Hope Creek SSER 6 15-2
16 TECHNICAL SPECIFICATIONS The Technical Specifications in a license define certain features, characteris-tics, and conditions governing the operation of the facility that cannot be changed without prior approval of the NRC staff. The Hope Creek Technical Specifications are included as Appendix A to the Hope Creek license. Included in the Technical Specifications are sections covering definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls.
In letters dated May 13 and 30, June 4 and 13, and July 7 and 24, 1986 (C. McNeill, PSE&G, to E. Adensam, NRC), the licensee requested certain changes to the Technical Specifications issued as part of the Hope Creek license. The licensee requested these changes as a result of low power operating experience to date. The requested changes and their status are identified in Table 16.1.
Hope Creek SSER 6 16-1
L I T
Table 16.1 Technical Specification changes e
Date of PSE&G
=
letter request Change description
- Status May 13, 1986 The licensee requested daletion of See Section 9.5.1 of fire protection program elements this supplement. This
. from the Technical Specifications, change required modi-in accordance with guidance con- fication of a license tained in Generic Letter 86-10. condition.
May 30, 1986 The licensee requested changes to Review will be completed the Technical Specificatians to per- after full power license mit operation with one recirculation is issued.
loop out of service.
T June 4, 1986 The licensee requested changes to Review will be completed the Technical Specifications to per- after full power license mit operation with drywell and sup- is issued.
pression chamber purge system valves under administrative control.
June 13, 1986 The licensee requested changes to 13 Technical Specification sections based on operational experience achieved since issuance of the low-power license. These changes are:
(1) The licensee requested changes Review will be completed
- to Technical Specifica- after full power license tion 3/4.8.2 permitti,ng one is issued.
channel of dc power to be in-operable for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, instead of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, before requiring
[ plant shutdown.
( (2) The licensee requested changes See Appendix T, Item (1),
- to Technical Specifica- of this supplement.
tions 4.6.5.3 and 4.7.2 to state that the filtraiion, re-circulation, and venLilation system (FRVS) and control room emergency filtration system (CREFS) heaters will be on rather than just operable for the required surveillance.
- For a more complete discussion, see the referenced letter.
Hope Creek SSER 6 16-2
W Table 16.1 (Continued)
Date of PSE&G lett::r request Change description
- Status June 13, 1986 (3) The licensee requested changes Changes are editorial to the Technical Specification and have been incorpo-inder to enhance the index and rated into the Technical make the Technical Specifica- Specifications.
tions easier to use.
(4) The licensee requested changes Review will be completed g to Technical Specification after full power license 3.8.1.1 to permit 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, is issued.
rather than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, to shut down the plant if two diesel generators are inoperable.
(5) The licensee requested a Change is editorial and change to Technical Specifica- has been incorporated tion 6.9.1.8 revising the into the Technical address to which routine re- Specifications.
ports of operating statistics and shutdown experience are sent.
(6) The licensee requested changes Changes are editorial to Technical Specification and have been incorpo-Tables 2.2.1-1, 3.3.1-1, rated into the Techni-3.3.1-2, and 4.3.1.1-1 to cal Specifications.
correct typographical errors.
(7) The licensee requested Changes are editoral changes to Technical Specifi- and have been incorpo-cation 3.9.2 deleting the re- rated into the Techni-quirement that thermal power cal Specifications.
be less than 1% when one of three required source range monitors may be retracted.
(8) The licensee requested See Appendix T, Item (5),
changes to Technical Speci- of this supplement.
fication 4.4.6.1.4 allowing the reactor vessel flange and head flange temperature to be reduced.
- For a more complete discussion, see the referenced letter.
Hope Creek SSER 6 16-3
Table 16.1 (Continued)
Date of PSE&G letter request Change description
- Status June 13, 1986 (9) The licensee requested Changes are editorial changes to Technical Speci- and have been incorpo-fication Tables 3.3.3-1, rated into the Technical 3.3.7.5-1, and 4.3.7.5-1 and Specifications.
Section 3.11.3 to delete vari-ous provisions that allowed completion of preoperational tests after initial fuel load but before 5% power is exceeded.
(10) The licensee requested changes See Appendix T, Item (4),
I to Technical Specification of this supplement.
Table 3.6.3-1 to delete oper-ability testing of excess flow check valve BB-XV-3649.
(11) The licensee requested changes Changes are editorial to Technical Specification and have been incorpo-Table 3.6.3-1 to correct the rated into the Techni-identification of the Remote cal Specifications.
Manual Isolation Valves Group 26 - RHR Suppression Pool Re-turn Valves as Other Primary Containment Isolation Valves Group 39 - RHR System, since these valves are manual, not remote manual valves.
(12) The licensee requested changes See Appendix T, Item (2),
to Technical Specifications of this supplement.
4.6.5.3.e.3 and 4.7.2.e.4 to clarify the method of verify-ing humidity control instru-ments are operable by using a channel calibration for the FRVS and CREFS and to change the FRVS heater dissipation from 100 1 5 kW to 100 t 10 kW.
(13) The licensee requested changes See Appendix T, Itein (3),
to Technical Specification of this supplement.
Table 3.3.7.1-1 to delete Pro-vision (b), which requires that the offgas post-treatment radiation monitor be operable.
- For a more complete discussion, see the referenced letter.
Hope Creek SSER 6 16-4
Table 16.1 (Continued)
Date of PSE&G letter request Change description
- Status July 7, 1986 The licensee requested changes See Section 15.1 of this to Technical Specification 3.2.3 supplement. This change in response to the staff's required modification of June 23, 1986, letter. a license condition.
July 24, 1986 The licensee requested changes to two Technical Specification sec-tions based on operational experi-ence achieved since issuance of the low power license. These changes are:
(1) The licensee requested a Change is editorial and change to Technical Speci- has been incorporated fication 4.8.1.1.2.f.1.b into the Technical changing the maximum Say- Specification.
bolt Second Universal (SSU) viscosity of new diesel gen-erator fuel oil from 45 SSU to 40.1 SSU.
(2) The licensee requested that The requested addition a Section 6.9.3 be added to has been incorporated the Technical Specifications into the Technical which would require that vio- Specifications.
lations of the Fire Protec-tion Program be reported via a Licensee Event Report within 30 days.
- For a more complete discussion, see the referenced letter.
Hope Creek SSER 6 16-5
APPENDIX A CONTINUATION OF CHRON0 LOGY August 1, 1985 Meeting between applicant and staff to discuss the proposed compression of the Hope Creek power ascension program.
March 20, 1986 Letter to applicant forwarding staff's safety evaluation on the last set of power ascension program test modifications.
April 8, 1986 Letter from applicant documenting a teleconference with the staff concerning the effect of a fire on high pressure /
low pressure interfaces.
April 10, 1986 Letter from applicant requesting withdrawal of April 8, 1986, letter regarding clarification / definition of initial criticality and applicability to plant.
April 11, 1986 Letter from applicant responding to NRC letter of March 14, 1986, regarding violations in Inspection Report 50-354/86-03.
Revision 1 to Test Procedure TE-SU.ZZ-041 (Q) issued on March 8,1986, will properly address identified deficiencies.
April 11, 1986 Letter to applicant forwarding License NPF-50, environmental protection plan, Amendment 11 to Indemnity Agreement B-74, Supplement No. 5, and Federal Register notice of issuance.
April 11, 1986 License NPF-50 issued for Hope Creek Generating Station.
April 14, 1986 Letter from licensee forwarding 1985 Annual Report for PSE&G, Atlantic City Electric Co. , Delmarva Power and Light Co., and Philadelphia Electric Co.
April 14, 1986 Letter to licensee forwarding Inspection Report 50-354/
80-14. Utility-sponsored SAFETEAM program satisfactorily identified and resolved employee concerns during final stages of plant construction and preoperational testing.
April 16, 1986 Letter to licensee forwarding final Systematic Assessment of Licensee Performance Report 50-354/85-99 for November 1984-October 1985, per February 27, 1986, meeting and review of March 17, 1986, comments.
April 16, 1986 Letter to licensee expressing appreciation for assistance in training effort during site visit on March 19, 1986, while preparing to load fuel in Unit 1.
April 17, 1986 Letter to licensee forwarding Safety Inspection Report 50-354/86-21.
Hope Creek SSER 6 1 Appendix A
April 18, 1986 Letter to licensee expressing appreciation for recipient and M. Headrick participation in NRC Operator Licensing Subject Matter Expert Panel meetings during February and March 1986.
April 23, 1986 Letter to licensee forwarding Safety Inspection ;
Report 50-354/86-19. j April 23, 1986 Letter to licensee forwarding Safety Inspection '
Report 50-354/86-18.
April 24, 1986 Generic Letter 86-10 to all power reactor licensees and applicants for power reactor licenses regarding implementa-tion of fire protection requirements.
April 28, 1986 Letter to licensee forwarding initial Operating License (OL) Review Report 50-354/86-25.
April 28, 1986 Letter to licensee requesting results of investigation regarding potential improprieties of quality control super-visors and proposed corrective actions within 30 days.
May 2, 1986 Letter to licensee forwarding Examination Report 50-354/
86-16 regarding examination administered during week of February 24, 1986.
May 8, 1986 Letter from licensee forwarding application for amendment to License NPF-50, revising License Condition 2.C.12 regarding partial feedwater heating. Facility will not be operated with partial feedwater heating for purpose of extending normal fuel cycle without prior written staff consent.
May 9, 1986 Letter to licensee acknowledging receipt of April 11, 1986, letter informing staff of steps taken to correct violations noted in Inspection Report 50-354/86-03. Corrective actions documented in Inspection Reports 50-354/86-22 and 50-354/86-23.
May 12,-1986 Letter from licensee forwarding Licensee Event Reports (LERs) 86-002-00 and 86-003-00.
May 13, 1986 Letter from licensee informing staff that, as of May 3, 1986, offgas system was successfully preoperationally tested, allowing tensioning of reactor pressure vessel head closure bolts in support of low power testing activi-ties per Supplement No. 5 and Appendix R.
May 13, 1986 Letter from licensee discussing current program for engi-neering expertise on shift, shift technical advisor degree requirement, and proposed modifications to Final Safety Analysis Report (FSAR) regarding engineering expertise on shift and licensed operator staffing, per Generic Letter 86-04 and NUREG-0737.
Hope Creek SSER 6 2 Appendix A
May 13, 1986 Letter from licensee requesting deletion of fire protec-tion program elements from Technical Specifications when t
full power operating license is issued.
May 14, 1986 Letter to licensee requesting results of investigation and proposed corrective action within 30 days regarding quality concerns with electrical supports at facility listed in April 29, 1986, anonymous letter.
May 20, 1986 Letter to licensee forwarding Safety Inspection Report 50-354/86-20 and notice of violation and acknow-ledging receipt of April 2, 1986, response to violations cited in Inspection Report 50-354/86-06.
May 22, 1986 Letter from licensee forwarding affidavit certifying distribution of Amendment 15 to FSAR.
May 22, 1986 Letter to licensee confirming June 5, 1986, meeting in King of Prussia, PA, to discuss operating experience and lessons learned since receipt of low power operating license.
May 23, 1986 Letter from licensee forwarding additional information on testability of Bailey 862 solid-state logic modules, in response to April 8,1986, commitment.
May 27, 1986 Letter from licensee forwarding plant specific responses to NRC Office of Inspection and Enforcement (IE) Bul-letin 85-03, " Motor-0perated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings,"
for Salem Units 1 and 2 and Hope Creek, for review.
May 27, 1986 Letter from licensee forwarding Addendum 3 to Revision 4 to ' Environmental Qualification Summary Report," reflecting replacement of 53 Tobar Model 32 Series 2 transmitters with Rosemount Model 1153B transmitters, in support of initial criticality and fulfillment of Supplement No. 5.
May 29, 1986 Letter from licensee informing staff that preoperational tests for diesel generator D, in support of OL Condi-tion 2.D(a) and continuation of low power testing activi-ties, were successfully completed on May 16, 1986.
May 30, 1986 Letter from licensee forwarding application for amendment to License NPF-50, supporting operation with one recircu-lation loop out of service. General Electric report, " Hope Creek Single Loop Operation Analysis," justifies proposed change and will be incorporated as Appendix 15C to FSAR.
June 2, 1986 Letter from licensee notifying staff that preoperational tests for traversing in-core probe system were successfully completed on May 23, 1986.
Hope Creek SSER 6 3 Appendix A
June 2, 1986 Letter from licensee forwarding Addendum 2 to Revision 4 to " Environmental Qualification Summary Report," reflecting reevaluation of electric equipment under Purchase Order M-48, per Supplement No. 5.
June 2, 1986 Letter to licensee forwarding Safety Inspection Report 50-354/86-22.
June 3, 1986 Letter from licensee advising staff of contract with l
Chem-Nuclear Systems, Inc., to provide temporary radwaste processing services instead of Westinghouse-Hittman Nuclear Services.
June 4, 1986 Letter from licensee requesting approval to operate dry-well and suppression chamber purge system per enclosed proposed revision to Technical Specifications.
June 5, 1986 Letter from licensee forwarding LERs 86-014-00 and 86-015-00.
June 6, 1986 Letter to licensee forwarding Safety Inspection Report 50-354/86-24.
June 6, 1986 Letter to licensee forwarding Safety Inspection Report 50-354/86-23.
June 9, 1986 Letter from licensee forwarding LERs 86-016-00 and 86-017-00.
June 9, 1986 Letter to licensee requesting that enclosed table detailing status of NRC review of responses to Generic Letter 83-28,
" Required Actions Based on Generic Implications of Salem ATWS Events," be completed with dates for ireplementation of required actions.
June 13, 1986 Letter from licensee forwarding application for amendment to License NPF-50, revising Technical Specifications be-cause of pending issuance of full power license.
June 13, 1986 Lettr..r from licensee regarding Bailey 862 solid-state logic module reliability program.
June 18, 1986 Letter to licensee forwarding status of Generic Letter 83-28 review items.
June 23, 1986 Letter to licensee discussing proposed revision to the license condition, " Partial Feedwater Heating."
June 24, 1986 Letter from licensee summarizing the June 23, 1986, tele-conference with the staff regarding the solid-state logic module reliability test program.
Hope Creek SSER 6 4 Appendix A
l July 7, 1986 Letter from licensee requesting revision to the Hope Creek l Technical Specifications in response to the staff's June 23, t 1986, letter.
July 7,1986 Letter from licensee submitting the Hope Creek turbine main-tenance program based on the turbine manufacturer's proba-bilities of missile generation.
July 11, 1986 Letter from licensee forwarding a copy of the General Elec-tric proprietary report, " Probability of Missile Generation in General Electric Nucicar Turbines." The licensee re-quested that this report be withheld from public disclosure.
July 24, 1986 Letter from licensee regarding certain revisions to the Hope Creek Technical Specifications.
l 1
l i
l l
i l l
Hope Creek SSER 6 5 Appendix A i
)
,, , ------e , , - , . , ,
f APPENDIX B BIBLIOGRAPHY U.S. Nuclear Regulatory Commission, Generic Letter 83-28, " Required Actions Based on Generic Implications of Salem ATWS Events," July 8, 1983.
-- , Generic Letter 86-10, " Implementation of Fire Protection Requirements,"
April 24, 1986.
-- , NUREG-0803, " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping," August 1981.
-- , NUREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant," April- 1983.
-- , Office of Inspection and Enforcement, IE Bulletin.79-27, " Loss of Non-Class 1E Instrumentation and Control Power System Bus During Operation,"
November 30, 1979.
-- , IE Inspection Report 50-354/86-29, June 25, 1986.
i a
4 4
i Hope Creek SSER 6 1 Appendix B
I APPENDIX C r
UNRESOLVED SAFETY ISSUES In Appendix C to the SER, the staff presented its review of unresolved safety issues (USIs) and their application to Hope Creek. Since that time, USI A-43 has been resolved. The staff's discussion of this USI is given below.
A-43 Containment Emergency Sump Reliability This concern is related to the potential for degraded emergency core cooling system performance as a result of thermal insulation debris that may be blown into the suppression pool during a loss-of-coolant accident and cause blockage of the pump suction lines. In the SER, the staff concluded that Hope Creek could be operated before ultimate resolution of this generic issue without endangering the health and safety of the public.
Since the issuance of the SER, the staff has completed its analysis of this issue, which is contained in NUREG-0869, Revision 1, "USI A-43 Regulatory Analysis." The staff has concluded, as a result of this analysis, that no new requirements need to be imposed on licensees and construction permit holders (see Generic Letter 85-22, " Potential for Loss of Post-LOCA Recirculation Capability Due to Insulation Debris Blockage," December 3, 1985). There-fore, this issue is considered resolved for Hope Creek.
I l
Hope Creek SSER 6 1 Appendix C
APPENDIX D ACRONYMS AND INITIALISMS i
ATWS anticipated transient (s) without scram BWR boiling-water reactor CFR Code of Federal Regulations CREFS control room emergency filtration system EllC electrohydraulic control FATT fracture appearance transition temperature FPP Fire Protection Program FRVS filtration, recirculation, and ventilation system FSAR Final Safety Analysis Report GDC general design criterion (a)
GE General Electric Company HEPA high efficiency particulate air IE Office of Inspection and Enforcement LER Licensee Event Report LOCA loss-of-coolant accident MCPR minimum critical power ratio MHC mechanical hydraulic control NRC U.S. Nuclear Regulatory Commission OL operating license PAP power ascension program PSE&G Public Service Electric and Gas Company
- ' RFT reduced feedwater temperature RG regulatory guide SER Safety Evaluation Report SSLM solid state logic module TMI Three Mile Island USI unresolved safety issue UT ultrasonic testing Hope Creek SSER 6 1 Appendix 0 1
. _ _ . _ _ . . _ - . - - - . ~ - - - - - ----- ---- - ~ ^ ~ - "
l l
APPENDIX E PRINCIPAL STAFF CONTRIBUTORS This supplement to the Safety Evaluation Report is a product of the NRC staff and its consultants. The NRC staff members listed below were principal contrib-utors to this report.
Name Title Branch Richard Becker Reactor Engineer Facility (perations Jay Lee Health Physicist Plant Systems Felix Litton Mechanical Engineer Engineering ,
Barry Marcus Electrical Engineer Electrical, Instrumentation l and Control Systems Carl Schulten Reactor Engineer Facility Operations George Thomas Reactor Engineer Plant Systems John Tsao Mechanical Enginaer Engineering Frank Witt Mechanical Engineer Plant Systems 4
1 4
Hope Creek SSER 6 1 Appendix E
APPENDIX S REVIEW OF POWER ASCENSION PROGRAM ACCELERATION f
I l
Hope Creek SSER 6 Appendix S
'o,, UNITED STATES
[ g NUCLEAR REGULATORY COMMISSION g ,a w AsHINGTON, D. C. 20555
..... January 22, 1986 Docket No. 50-354 Mr. Corbin A. McNeill, Jr., Vice President-Nuclear Public Service Electric & Gas Company Nuclear Administration Building P. O. Box 236 Hancocks Bridge, New Jersey 08038
Dear Mr. Mctleill:
SUBJECT:
HOPE CREEK POWER ASCENSION PROGRAM TEST MODIFICATI0t:S By letters dated August 21, September 20,23 and 30, 1985, PSE&G submitted for staff review a number of proposed test modifications to the Hope Creek Power Ascensi.on Program. The proposed test modifications were submitted for staff review as part of a program to accelerate power ascension testing for Hope Creek. Additionally, the Hope Creek accelerated power ascension program is the lead plant in an effort by General Electric to similarly accelerate the power ascension program on a number of other boiling water reactors.
The staff has completed its review of the following test modifications proposed in the above referenced letters:
- 1. Automatic Load Following
- 2. Test No. 17 - Core Performance
- 3. " Test No. 27 - Recirculation Flow Control System
- 4. Test No.12 - RCIC System
- 5. Test No. 13 - HPCI System
- 6. Test No.14A - Selected Process Temperatures
- 7. Test Condition 4 - Natural Circulation Operation
- 8. Test No. 20 - Pressure Regulator
- 9. Test No. 21A - Feedwater System Response Testing The safety evaluation detailing each review is attached. For those test modifications which the staff has accepted (Items 1,4,5,6,8 in part and 9 in part), we request that the Hope Creek FSAR be amended.
Sincerely, h _=_-s Robert Bernero Director Division of BWR Licensing Office of Nuclear Reactor Regulation See next page Hope Creek SSER 6 1 Appendix 5
Mr. C. A. McNeill Public Service Electric & Gas Co. Hope Creek Generating Station cc:
Gregory Minor Susan C. Remis Richard Hubbard Division of Public Interest Advocacy Dale Bridenbaugh New Jersey State Department of MHB Technical Associates the Public Advocate 1723 Hamilton Avenue, Suite K Richard J. Hughes Justice Comples San Jose, California 95125 CN-850 Trenton, New Jersey 08625 Troy B. Conner, Jr. Esquire Office of Legal Counsel Conner & Wetterhahn Department of Natural Resources 1747 Pennsylvania Avenue N.W. and Environmental Control Washington, D.C. 20006 89 Kings Highway P.O. Box 1401 l Dover, Delaware 19903 Richard Fryling, Jr., Esquire Mr. K. W. Burrowes, Project Engineer Associate General Solicitor Bechtel Power Corporation Public Service Electric & Gas Company 50 Beale Street P. O. Box 570 TSE P. O. Box 3965 Newark, New Jersey 07101 San Francisco, California 94119 Manager - Licensing Regulation Resident Inspector c/o Public Service Electric & Gas U.S.N.R.C. Bethesda Office Center, Suite 550 P. O. Box 241 4520 East-West Highway Hancocks Bridge, New Jersey 08038 Bethesda, Maryland 20814 Ms. Rebecca Green Richard F. Engel New Jersey Bureau of Radiation Deputy Attorney General Protection Division of Law 380 Scotch Road Environmental Protection Section Trenton, New Jersey 08628 Richard J. Hughes Justice Complex CN-112P Trenton, New Jersey 08625 Mr. Robert J. Touhey, Mr. Anthony J. Pietrofitta Acting Director General Manager DNREC - Division of Power Production Engineering Environmental Control Atlantic Electric 89 King's Highway 1199 Black Horse Pike P. O. Box 1401 Pleasantville, New Jersey 08232 Dover, Delaware 19903 Regional Administrator, Region I Mr. R. S. Salvesen U. S. Nuclear Regulatory Comission General Manager-Hope Creek Operation 631 Park Avenue
, Public Service Electric & Gas Co. King of Prussia, Pennsylvania 19406 P.O. Box A Hancocks Bridge, New Jersey 08038 Hope Creek SSER 6 2 Appendix 5
l Public Service Electric & Gas Co. Hope Creek Generating Station 1
cc:
Mr. B. A. Preston Public Service Electric & Gas Co.
Hope Creek Site MC12Y Licensing Trailer 12LI Foot of Button wood Road Hancock's Bridge, New Jersey 08038 4
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1 Hope Creek SSER 6 3 Appendix S
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\...../. SAFETY EVALUATION HOPE CREEK GENERATING STATION POWER ASCENSION TEST PROGRAM ACCELERATION This safety evaluation describes the staff's review of a number of Hope Creek Generating Station - Power Ascension Program (PAP) test modifications proposed by Public Service Electric and Gas Company (PSE&G). These proposed test modifications were submitted for staff review as part of a program to accelerate power ascension testing for Hope Creek. The Hope Creek accelerated power ascension testing program is also the lead plant for generic changes for acceleration of the traditional General Electric Boiling Water Reactor power ascension test-program.
The proposed test modifications discussed in this safety evaluation were sub-mitted by letters dated August 21, September 20, 23, and 30, 1985, (R. Mitti, PSE8G to W. Butler, NRC). This safety evaluation discusses the following tests:
- 1. Automatic Load Following
- 2. Test No. 17 - Core Performance
- 3. Test No. 27 - Recirculation Flow Control System )
- 4. Test No. 12 - RCIC System
- 5. Test No. 13 - HPCI System
- 6. Test No. 14A - Selected Process Temperatures
- 7. Test Condition 4 - Natural Circulation Operation
- 8. Test No. 20 - Pressure Regulator
- 9. Test No. 21A - Feedwater System Response Testing Discussion of the above tests follows:
- 1. DELETION OF AUTOMATIC LOAD FOLLOWING (ALF) TESTING l
The applicant proposed deleting all testing of the ALF function from the power ascension test program. The ALF portion of the Recirculation ;
- i Flow Control System is a non-safety related function which is not intended
! to be used at present. Because the Automatic function is not going to be 2
used, it is unnecessary to test this function of the Recirculation Flow
! Control System. If at some future time the ALF function is to be used.
l it can be tested at that time. The Automatic mode of operation should 1 be physically disconnected from the Master Flow Controller and wired so l
that the system is in manual control in both Master Flow Controller 1
, Positions to prevent inadvertent use of the Automatic function. A similar deletion has been granted at other plants (e.g., Grand Gulf).
Therefore, if the Automatic function of the Master Flow Controller is i physically disabled, deletion of testing of the ALF function is acceptable.
1 l
Hope Creek SSER 6 4 Appendix S
- 2. SUBSTITUTION OF SURVEILLANCE TEST FOR CORE PERFORMANCE STARTUP TEST (TEST NO. 17)
The applicant proposed substitution of the Core Performance Technical Specification Surveillance test for the Core Performance Startup test to reduce duplication of testing.
The applicant maintains that adherence to plant technical specifications meets the objectives of startup test No.17 and the intent of Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, paragraph 5.b. The staff believes that the performance of the fuel thermal margins technical
~
specification surveillances simply allows operational personnel to determine that the plant is operating within its required limiting condition for operation and does not constitute a determination "that steady-state core performance is in accordance with design" (Reg. Guide 1.68). The objective of Startup Test No.17 and Regulatory Guide 1.68 is that an engineering evaluation be perfomed to evaluate the actual thennal and hydraulic parameters against design expectations for these parameters and that it, in turn, receives the appropriate level of manage-ment attention. Accordingly, the applicant's proposed substitution of the Core Performance Technical Specification surveillance test for the Core Performance Startup test is not acceptable.
An alternative to the applicant's proposed substitution exists. It is the intent of Regulatory Guide 1.68, Revision 2, to utilize plant procedures wherever possible, both to validate the procedures and provide training for plant personnel in their use. In Section C, Regulatory Position, Part 2, of Regulatory Guide 1.68, it states, "The overall test program should also include surveillance tests necessary to demonstrate the proper operation of interlocks, setpoints, and other protective features, systems, and equipment required by the technical specifications." On the basis of the regulatory guidance in place, the staff feels that both the duplication of tests can be avoided and the more technically extensive review and approval required for startup tests can be accomplished by writing the detailed core performance startup test procedure such that the surveillance procedure is incorporated to direct data taking and recording. By this method, both regulatory intent to utilize plant procedures and the administrative review, evaluation, and approval controls for startup terts will be included. Because this regulatory position already exists, staff approval of this proposed change would not be necessary.
Hope Creek SSER 6 5 Appendix S
- 3. SIMPLIFICATION OF THE RECIRCULATION FLOW CONTROL SYSTEM (TEST NO. 27)
The applicant proposed deleting testing of the ALF portion of this system and utilizing experience from other plant startups to modify the cali-bration and testing of the recirculation flow control system to reduce the number of test points used to optimize control setting and check stable operation.
The staff has discussed and accepted (in Item 1) the delttice of ALF testing in the Recirculation Flow Control System. The selection of the number of test points and their distribution in testing space is a level of detail implementation which does not appear in the FSAR test abstract, but may be reviewed at the Regional level in the detailed test procedure.
Because the requested utilization of experience for bench calibration of controllers to reduce the number of system test points does not change the test objectives, summary test method or acceptance criteria of the test abstract, the staff finds the proposed changes conceptually acceptable.
As long as the applicant can demonstrate compliance with the acceptance criteria throughout the appropriate system control range, test simplifi-cation is encouraged.
The applicant proposed deleting tuning of the HPCI and RCIC controllers during the low power testing. Currently, tuning of the controllers for*
the RCIC and HPCI is perfomed at both low pressure and near rated reactor pressure. The applicant indicates that testing experience with recently comercialized BWRs has demonstrated that a best estimate controller setting may be selected from previous test data and used to bench calibrate the controllers during preoperational testing. This preliminary bench controller tuning is sufficient for low pressure testing and final tuning near rated reactor pressure. Experience at previous plants has shown that controller tuning at low pressure condition does not result in optimum performance at higher pressures. The applicant is requesting deletion of controller tuning at the low reactor pressure condition, relying on controller tuning or optimization at or near rated reactor pressure. Testing at low reactor pressure with the rated reactor pressure controller setting will be completed to confirm acceptable, but, perhaps, less than optimized system performance at low pressure conditions.
The staff has reviewed the applicant's requested change to the method of testing these two systems. The staff recognizes that it would be unlikely that operation of these two systems could be optimized over the entire pressure range for which they are expected to operate. Selecting rated pressure conditions for the point of optimization is reasonable and acceptable to the staff because acceptable operation at low reactor pressure will be confimed with the rated pressure controller setting.
Hope Creek SSER 6 6 Appendix S
- 6. SELECTED PROCESS TEMPERATURES (TEST NO. 14A)
The applicant is requesting the substitution of the Technical Specification surveillance test for s,tartup of an idle reactor recirculation loop for portions of the selected process temperatures startup procedure and deletion of detennination of the icw pump speed limit by utilizing a fixed value established from results of this test on previous plants.
The staff has reviewed the applicant's proposed changes and notes that the startup test monitors process temperatures to assure the technical specifi-cations are not exceeded. The staff believes that the technical specifi-cation surveillances accomplish the same purpose. Additionally, the staff notes that testing performed at Limerick and Susquehanna has demonstrated ample margins for stratification.
Accordingly, the staff finds this requested substitution acceptable. The Region staff will confirm during future test procedure review that the surveillance test has been properly incorporated into the procedure and that all necessary test data will be obtained.
The staff found that the request to delete the determination of the low pump speed limit in favor of using a fixed limit derived from previous plant tests was not supported by sufficient information. The applicant supplied additional information on the startup testing for the Brunswick Steam Electric Plant, Unit I and Edwin I. Hatch, Unit 2 which was reviewed by the staff. This information was contained in the Final Sumary Report -
Edwin I. Hatch Unit 2 - Startup Test Results, NED0-24734, R. W. Turkowski and W. Yee, October 1979, and Brunswick Unit 1 - Startup Test Results -
Final Sumary Report, NED0-24562, J. 9. Poppel, November 1977. This additional information and regulatinn-mandated startup reports of other plants were reviewed and allowed the staff to confirm that a fixed low pump speed limit could be selected (based on previous plant testing) which assures adequate coolant mixing and, therefore, acceptable temperature differentials in the lower plenum.
7 NATURAL CIRCULATION OPERATION - TEST CONDITION NO. 4 The applicant requested deletion of all testing at Test Condition 4 (approximately 50% power and'30% core flow) based on the stated premise that the applicant does not intend to operate in this domain of the power / core flow map and that the only way this domain would be approached would be an abnormal operational transient (trip of two recirculation pumps) and resulting natural circulation. The plant technical specifications require that with no reactor coolant system recirculation loops in operation, the operator must take imediate action to reduce thermal power to less than or equal to that allowed by the Core Thermal Power to Core Flow Map within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and, further, the operator must take action to place the plant in at least startup within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Therefore, the applicant states that testing need not be done at Test Condition 4.
Hope Creek SSER 6 7 Appendix 5
l The staff felt that the applicant's justification for deleting testing at Test Condition 4 was inadequate. The reactor manufacturer, GE, submitted additional information to supplement the applicant's justification by letter from J. F. Klapproth, GE, to R. A. Becker, NRC, dated November 6, 1985. The staff included this information in addition to the applicant's submission in its review.
The staff finds that the justification for deleting testing at Test Condition 4 is insufficient. The staff considers that the technical specifications define the "possible operating modes," whether or not the applicant foresees their utiltzation at the commencement of the plant's 40-year life. The length of time allowed at or near natural circulation is long compared to a phenomenon which could generate a safety concern such as stability. The staff agrees with the applicant that the generic natural circulation and stability have been sufficiently confirmed in recent domestic and foreign commercialized BWRs. However, the staff believes the testing encompassed by Regulatory Guide 1.68, Appendix A, Section 5, third introductory paragraph refers to plant specific confirma-tory testing and Test Condition 4 represents one of the " extremes of possible operating modes" referred to in the Regulatory Guide. Therefore, thorough testing dictates equipment specific testing at the natural circu-lation. Test Condition 4 and the staff finds the elimination of testing at Test Condition 4 to be unacceptable.
- 8. PRESSURE REGULATOR (TEST NO. 20)
The applicant proposed to modify the pressure regulator test to:
(a) delete perfonning this test at Test Condition 4 (TC4), (b) delete the backup pressure replator takeover testing at TC5, and (c) delete ALF mode tests at TC3 and TC6.
Regarding item (a) above, the staff evaluation of Natural Circulation Operation - Test Condition No. 4 (Item No. 7 in this safety evaluation) concluded that elimination of testing at TC4 is unacceptable. Accordingly, the proposal to delete testing of the pressure regulator at TC4 is not acceptable.
Regarding item (b) above, the applicant requested the deletinn of backup pressure regulator takeover testing at TC5. The system design includes redundant pressure regulators set at slightly different operating pressures.
If the pressure regulator operating at the lower pressure were to fail, the higher pressure regulator would assume pressure control. This design function is verified over the spectrum of operating conditions. Currently, this is performed at TC1 through TC6 which cover the power range from low to rated steam flow.
The staff believes that the only variable of concern is total steam flow or, indirectly, plant power level because the pressure regulator maintains fixed pressure for various total flows. For a fixed power level, variation Hope Creek SSER 6 8 Appendix S
l of core flow would have no impact on the pressure regulator. Examination
! of the operational power / flow map indicates that TC3 covers a similar but larger power span than TCS and, therefore, provides a redundant test point for testing of the backup pressure regulator takeover. Because of the redundant nature of this portion of the pressure regulator test at TCS and TC3, the staff believes that testing of the takeover of the backup pressure regulator at TCS may be deleted without affecting the objectives of the test or loss of confirmation of system performance over the entire operating range.
Regarding item (c) above, the applicant requested deleting all testing associated with the ALF mode of plant operation. This request was addressed and accepted in our review contained in Automatic Load Following (Item No. 1 of this safety evaluation).
The applicant (a) deleting proposed testing at TC4to modify (b) deleting ALF testing.
and Regarding item (a), the staff evaluation of Natural Circulation Operation -
Test Condition 4 (Item No. 7 in this safety evaluation) concluded that elimination of testing at TC4 is unacceptable. Accordingly, the proposal to delete feedwater system response testing at TC 4 is not acceptable.
Regarding item (b), the applicant has requested deleting all testing associated with the ALF mode of plant operation. This request was addressed and accepted in our review contained in Automatic Load Following (Item No. 1 of this safety evaluation),
llope Creek SSER 6 9 Appendix S
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/** *% UNITED STATES NUCLEAR REGULATORY COMMISSION 3 .g wassincrow. p. c. rosss
\,, DUE Docket No. 50-354 Mr. Corbin A. McNeill, Jr., Vice President-Nuclear Public Service Electric & Gas Company Nuclear Adninistration Building P. O. Box 236 Hancocks Bridge, New Jersey 08038
Dear Mr. McNeill:
Subject:
Hope Creek Power Ascension Program Test Modifications By letters dated August 21, October 4 and 17 November 6, and December 9,1985, PSE&G submitted, for staff review, a number of proposed test modifications to the Hope Creek Power Ascension Prooram. The proposed test modifications are part of a program to accelerate power ascension testing for Hope Creek. Additionally, Hope Creek is the lead plant in ar effort by General Electric to promote acceler-ated power ascension programs on a number of other boiling water reactors.
The staff has completed its review of the following test modifications proposed in the above referenced letters:
- 1. Test No. 24 - Relief Valves
- 2. Test No. 28E - Recirculation System Cavitation
- 3. Test No. 31 - Drywell Piping Vibration 4 Test No. 25 - Turbine Trip and Generator Load Rejection
- 5. Test No. 28D - Recirculation Pump Runback Test
- 6. Test No. 3 - Elimination of Fuel Chambers During Fuel Loading
- 7. Test No. 11 - Process Computer Test Simplification
- 8. Test No. 16 - TIP Uncertainty
- 9. Test No. 1 - Chemical Radiochemical Test Simplification
- 10. Test No. 32 - Reactor Water Cleanup System
- 11. Test No. 288 - Two Pump Recirculation Pump Trip Test The safety evaluation detailing each review is attached. For those test modifications which the staff has accepted (Items 1, 2, 4, 5, 6, 7, 8, 9 and 10 in part), we request that the Hope Creek FSAR be amended to reflect these changes.
Sincerely.
3 s
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Robert Bernero, Director
~
Division of BWR Licensing 1 s Office of Nuclear Reactor Regulation cc w/ enclosure See next page Hope Creek SSER 6 10 Appendix 5
Mr. C. A. McNeill Public Service Electric & Gas Co. Hope Creek Generating Station I cc:
- Gregory Minor Susan C. Remis j Richard Hubbard Division of Public Interest Ad<ocacy
! Dale Bridenbaugh New Jersey State Department of MHB Technical Associates the Public Advocate 1723 Hamilton Avenue, Suite K - Richard J. Hughes Justice Comples San Jose, California 95125 CN-850 Trenton, New Jersey 08625 1roy B. Conner, Jr. Esquire Office of Legal Counsel Conner & Wetterhahn Department of Natural Resources 1747 Pennsylvania Avenue N.W. and Environmental Control Washington, D.C. 20006 89 Kings Highway P.O. Box 1401 Dover, Delaware 19903 Richard Fryling, Jr., Esquire Mr. K. W. Burrowes, Project Engineer j Associate General So:icitor Bechtel Power Corporation 4 Public Service Electric & Gas Company 50 Beale Street P. O. Box 570 T5E l P. O. Box 3965 Newark, New Jersey 07101 San Francisco, California 94119 Manager - Licensing Regulation Resident inspector c/o Public Service Electric & Gas U.S.N.R.C, Bethesda Office Center, Suite 550 P. O. Box 241 4520 East-West Highway Hancocks Bridge, New Jersey 08038 Bethesda, Maryland 20814 I Ms. Rebecca Green Richard F. Engel New Jersey Bureau of Radiation Deputy Attorney General Protection Division of Law 380 Scotch Road Environmental Protection Section Trenton, New Jersey 08628 Richard J. Hughes Justice Complex CN-112P Trenton, New Jersey 08625 Mr. Robert J. Touhey, Mr. Anthony J. Pietrofitta Acting Director General Manager DNREC - Division of Power Production Engineering Environmental Control Atlantic Electric 89 Kings Highway 1199 Black Horse Pike P. O. Box 1401 Pleasantville, New Jersey 08232 Dover, Delaware 19903 Regional Administrator, Region I Mr. R. S. Salvesen V. S. Nuclear Regulatory Comission General Manager-Hope Creek Operation 631 Park Avenue Public Service Electric & Gas Co. King of Prussia, Pennsylvania 19406 P.O. Box A Hancocks Bridge, New Jersey 08038 liope Creek SSER 6 11 Appendix 5
Public Service Electric & Gas Co. Hope Creek Generating Station cc:
Mr. B. A. Preston Public Service Electric & Gas Co.
Hope Creek Site MC12Y Licensing Trailer 12LI ,
Foot of Button wood Road l Hancock's Bridge, New Jersey 08038
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l Hope Creek SSER 6 12 Appendix S
UNITED STATES 3
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+, .....# SAFETY EVALUATION HOPE CREEK GENERATING STATION POWER ASCENSION TEST PROGRAM ACCELERATION This safety evaluation describes the staff's review of a number of Hope Creek Generating Station - Power Ascension Program (PAP) test modifications proposed by Public Service Electric and Gas Company (PSE&G). These proposed test modifications were submitted for staff review as part of a program to accelerate power ascension testing for Hope Creek. Hope Creek is also the lead plant for generic changes for acceleration of the traditional General Electric Boiling Water Reactor power ascension test program.
The proposed test modifications discussed in this safety evaluation were submitted by letters dated August 21, October 4 and 17, Ncvember 6 and December 9, 1985. This safety evaluation discusses the following tests:
- 1. Test No. 24 - Relief Valves
- 2. Test No. 28E - Recirculation System Cavitation
- 3. Test No. 31 - Drywell Piping Vibration
- 4. Test No. 25 - Turbine Trip and Generator Load Rejection
- 5. Test No. 28D - Recirculation Pump Runback Test
- 6. Test No. 3 - Elimination of Fuel Chambers During Fuel Leading
- 7. Test No.11 - Process Computer Test Simplification
- 8. Test No. 16 - TIP Uncertainty
- 9. Test No.1 - Chemical Radiochemical Test Simplification
- 10. Test No. 32 - Reactor Water Cleanup System
- 11. Test No. 28B - Two Pump Recirculation Pump Trip Test Discussion of the above tests follows:
- 1. TEST NO. 24 - RELIEF VALVES The purpose of Test number 24 - Relief Valves, is to demonstrate the opera-bility of the relief valves, i.e., that they can be opened and closed manually, that they reseat properly, and that there are no blockages in the relief valve piping. This testing is currently planned to be perfonned at low (250-500 psig) pressure during heatup and at rated reactor pressure between Test Conditions 2 and 3. By letter dated October 17, 1985, the applicant proposed to delete the operability test at low pressure and between Test Conditions 2 and 3.
Alternatively, the applicant proposed verifying operability between 10% and 201 power at Test Condition 1.
The Hope Creek Safety Relief Valves (SRVs) are manufactured by. Target Rock Corporation. Actuation of the relief valves at low pressure has been identified as a contributor to valve seat damage caused by the valves' reseating against abnonnally low pressure. Therefore, operation of the valves at low pressure should be avoided whenever possible. Furthermore, from an operational point of view, conducting the test at a steam flow greater than the capacity of a Hope Creek SSER 6 13 Appendix 5
relief valve (typically 5-7%) will significantly enhance plant pressure control during the transient. Finally, protection from the effects of over-pressure transients which may occur prior to relief valve operability testing will be assured through compliance with Technical Specification 4.5.1.d which requires that all automatic depressurization system (ADS) valves be manually opened within twelve hours of reaching a steam dome pressure of 100 psig.
The ADS valves by themselves provide sufficient relief capacity to mitigate the relatively mild overpressure transients which could occur at less than 201 power. For the reasons stated above, the proposed change to test the SRVs at Test Condition 1 is acceptable.
- 2. TEST NO. 28E - RECIRCULATION SYSTEM CAVITATION Test Number 28E Recirculation System Cavitation, verifies that no recircu-lation system cavitation will occur in the operable region of the power-flow map. Currently, this test is planned to be performed at Test Condition 3 by lowering power at high recirculation flow until a recirculation pump run-back occurs, or the plant is at approximately 18% power and no runback has occurred, or there is indication of cavitation. By letter dated October 17, 1985, the applicant proposed that the testing be simplified by temporarily bypassing the cavitation interlock to prevent runback of the recirculation pumps.
Acceptable response of the system near the cavitation region is determined by analyzing test data and comparing to acceptance criteria which define the
. required system performance. For the recirculation system cavitation test,
! the recirculation runback logic is required to be demonstrated to have settings l adequate to prevent operation in areas of potential cavitation. This may be
- demonstrated without requiring that recirculation runback occur by temporarily l bypassing the runback logic during the test. Power can then be reduced by inserting control rods when cavitation is indicated. The cavitation interlock logic can be monitored to ensure that it is actuated at the correct setpoint.
With appropriate placement of the jumper on the cavitation interlock, no other recirculation pump runback logic feature will be affected. In addition, should a feedwater transient occur during the perfonnance of this test, the operators can manually run back recirculation pump speed as necessary to prevent cavitation.
Cavitation interlock setpoints are designed to allow maximum operation in the power-flow mp but are conservatively set to assure that no recirculation system cavitation occurs. The proposed simplified testing will verify that cavitation does not occur at or above the cavitation interlock setpoint while eliminating unnecessary core flow reductions during the test. Similar testing simplification was used at Limerick and Susquehanna. The proposed change to the Recirculation System Cavitation test is acceptable.
Hope Creek SSER 6 14 Appendix S
- 3. TEST NO. 31 - DRYWELL PIPING VIBRATION Test Number 31 was intended to evaluate the dynamic response of the main steam and recirculation piping to various transients in terms of vibrational charac-teristics and dynamic deflection. By letter dated October 17, 1985, the applicant proposed eliminating two of these transients from this test series. These trans-ients are the recirculation p' ump trips and restarts. The applicant's justification for eliminating these tests is that previous startup results on similar plants indicate that vibration and deflection measurements of recirculation piping during recirculation pump trips and restarts are always well below the prescribed limits.
As part of this justification, the applicant submitted a quantitative comparison of recirculation piping system parameters on Hope Creek and Susquehanna Unit 1.
Table 2 in the applicant's October 17, 1985, letter shows that these parameters are almost identical. In addition, Table 1 of the applicant's letter shows that measured vibrations for the recirculation pump trip / restart transients on Susquehanna Unit I were well below the allowable.
The basic objective of all piping preoperational testing is to confirm the as-built condition of piping systems in each plant. Although the design of such systems may be identical from plant to plant, fabrication, installation, inspection and quality assurance of the piping and the pump could introduce differences or defects which may affect the dynamic characteristics of the piping systems. Therefore, the staff's position is that all of the pre-operational piping tests which the applicant, in the FSAR, agreed to conduct shall be completed and that the proposed modifications to these tests are not acceptable.
- 4. TEST NO. 25 - TURB!NE TRIP AND GENERATOR LOAD REJECTION By letter dated August 21, 1985, the applicant proposed a) changing the generator load rejection test at Test Condition 2 to a turbine trip test and b) deleting the turbine trip test at Test Condition 3.
The staff finds these proposed changes acceptable. Regarding changing the generator load rejection test at Test Condition 2 to a turbine trip test at Test Condition 2 the staff notes that the purpose of this test segment, namely to test turbine bypass system performance, can be accomplished by either a turbine trip test or a generator load rejection test. Accordingly, this modification is accep-table. Regarding deleting the turbine trip test at Test Condition 3, the applicant indicates that performing the already planned turbine generator full load rejection test at 100% power required by Regulatory Guide 1.68 is more bounding with respect to severity than the turbine trip test at Test Condition 3. The applicant provided comparative date for these two tests for five boiling water reactors to support that position.
The staff has reviewed the evaluation and supporting data and agrees with the applicant that the regulatory guidance outlined in Regulatory Guide 1.68, Revision 2 related to plant testing involving the turbine generator is satisfied by perfonning a full load rejection / turbine trip test at Test Condition 6 (100%
power). Therefore, the staff finds it acceptable to delete the turbine trip test at Test Condition 3.
Hope Creek SSER 6 15 Appendix S
- 5. TEST NO. 28D - RECIRCULATION PUMP RUNBACK TEST By letter dated October 4,1985, the applicant requested that the recircu-lation pump runback test, currently planned as a separate test at Test Condition 3, be combined with the feedwater pump trip test at Test Condition 6.
The staff finds this change acceptable for the following reasons. The recircu-lation pump runback feature is included in the design to avoid unnecessary scrams due to low water level when a feedwater pump inadvertently trips. When performed at Test Condition 3, the runback is initiated by a simulated feedwater pump trip signal. If performed at Test Condttion 6 as part of the feedwater pump trip test, demonstration of the runback circuit would be obtained as an actual integrated test. Therefore, some testing advantage is obtained without compromising any safety objectives.
- 6. TEST NO. 3 - ELIMINATION OF FUEL LOADING CHAMBERS DURING FUEL LOADING By letter dated December 9,1985, the applicant proposed simplifying the fuel loading procedure by replacing the fuel loading chambers (FLCs) with source range monitor (SRM) instrumentation. Additionally, the startup sources will be positioned in their alternate locations (to be closer to the SRMs) and the fuel loading sequence will be modified such that initial fuel loading will begin between an SRM detector and a neutron source. Fuel will be loaded in a spiral pattern around the SRM until the core is fully loaded.
Test No. 3 is conducted in the fuel loading phase of initial operations. It has usually been necessary in past fuel loadings to use FLCs in addition to SRM detec-tors to achieve Technical Specification required count rates with fuel in the core.
A number of utilities have, in the past, requested and been granted reload fuel loading operations in which a small number of fuel assemblies are loaded before
- the usual required count rate on SRMs or FLCs are achieved. This initial loading l is sufficient to provide the needed count rate. Additionally, the pemitted (small) number of assemblies can not become critical, even with all control rods removed.
The proposed modifications to Hope Creek Test No. 3 are to permit the loading of
! 16 assemblies without (necessarily) meeting the usual required 0.7 counts /sec for the SRMs or FLCs. This is based on analysis (by General Electric) which shows that this array would not be critical (with rods out) and would provide the necessary count rate. This change would make it unnecessary to use FLCs, which interfere with operations, and would pemit the use of the standard SRMs alone.
The procedure places the sources in alternate locations, close to an SRM and uses a spiral loading pattern around the initial source-SRM and 16 assembly locations rather than around the core center.
These procedures are compatible with a number of previously approved reload procedures. The criticality calculations are done with standard methodology.
j They are consistent with other analyses reviewed in this area. Tests on other reactor startups indicate that required count rates should be achieved. The Technical Specifications will require (after loading 16 assemblies) the usual count rate on at least one SRM (the SRM near the initial loading). The other SRMs will be checked with a source, as has been approved for other reactors, until they reach a suitable count rate.
l Hope Creek SSER 6 16 Appendix S l
The staff notes that the Technical Specifications should be revised to specify that one of the operable SRMs must be in an area that has fuel loaded around l it on one side. Specifically, we require that two SRMs be operable and continuously indicating in the control room. One of the SRMs will be in the quadrant where fuel is being loaded and the other will be in an adjacent quadrant. One of these two SRMs will be in an area in which fuel has been loaded. The minimum count rate must be met for at least one of the SRMs. With this additional requirement, the applicant's proposal is acceptable.
- 7. TEST NO. 11 - PROCESS COMPUTER TEST SIMPLIFICATION By letter dated November 6,1985, the applicant proposed simplifying Test No.11.
Test No. 11 involves the testing of the Process Computer and its programs. OD-11 is one of these programs and deals with the area of fuel pellet-clad interaction monMoring (Preconditioning Interim Operating Management Recomendations (PCIOMR)).
The program assists in implementing PCIOMR to prevent this type of fuel failure mechanism during operation. However,withbarrierfuel(asusedatHopeCreek),
General Electric Company (GE) (the fuel facricator) has removed the PCIOMR pro-cedures from the operation plans since they are no longer needed. Accordingly, the applicant has proposed the removal of the 0D-11 test from the startup program.
Our review has indicated that there is no need for PCIOMR monitoring for this fuel. The removal of OD-11 monitoring is acceptable.
- 8. TEST NO. 16 - TIP UNCERTAINTY By letter dated November 6,1985, the applicant proposed deleting Test No.16 from the power ascension program. Test No. 16 measures the Traversing Incore Probe (TIP) uncertainty. The uncertainty is composed of geometry effects and random noise. These are determined by comparing symetric pairs of TIP readings and repeated traverses of comon TIP tubes. The criterion for first cycle TIP uncertainty tests is that uncertainty should be less than 6 percent. This is the value which, if used in the uncertainty analyses for GETAB (rather than the 2.6 percent value nonnally used in first cycle), would increase the power density value sufficiently to increase the safety limit minimum critical power ratio (MCPR) by 0.01. Previous tests in other reactors (including LaSalle 1 and Susquehanna 1) have always provided a TIP uncertainty well below 6 percent.
Furthermore, the uncertainty is lower when using the recently introduced TIP gama detector rather than the usual neutron detector since the gama system is less sensitive to geometry errors. For these reasons, the applicant has proposed deleting this test.
TIP operability is determined in preoperational testing and during power ascension power distribution measurements and tests of the Process Computer.
Previous tests in other reactors have indicated no problem in the TIP uncertainty area, and the gama detectors to be used have lower uncertainty parameters. The Hope Creek system should be well below the criteria. The safe operation of the plant will not be affected by deleting this test. Accordingly, the proposed deletion of Test No. 16 is acceptable.
Hope Creek SSER 6 17 Appendix 5
- 9. TEST NO. 1 - CHEMICAL RADIOCHEMICAL TEST SIMPLIFICATION By letter dated December 9,1985, the applicant proposed to substitute plant surveillance procedures for the chemistry and radiochemistry monitoring requirements. Additionally, the applicant' proposed deleting the integrated performance testing of the reactor water cleanup system (RWCU) and condensate demineralizer system at Test Condition 3.
The purpose of Test No.1 is to demonstrate that the plant water chemistry and radiochemistry are within limits during the power ascension test program and also to demonstrate the design capability of the plant chemistry system. It is proposed by the applicant to substitute the plant surveillance procedure CH-TI.ZZ-012(Q), Chemistry Sampling and Surveillance Procedure, for the chemistry and radiochemistry monitoring requirements of Test No. 1. This surveillance procedure, based on BWR Water Chemistry Guidelines (BWR Owners Group /EPRI report dated April 1, 1984), is used to ensure that plant water chemistry meets fuel l warranty limits. The improved water chemistry can enhance fuel performance and minimize radiation field buildup on out-of-core surfaces. The surveillance procedure insures that plant water chemistry meets the limits specified by i Technical Specification 3.4.4 and the General Electric fuel warranty. The surveillance procedure limits are at least as restrictive as those of Test No.1.
i Since the surveillance procedure will be used during normal operation it would l be prudent to also use this procedure during power ascension testing. It is therefore acceptable to substitute the plant surveillance procedure CH-TI-ZZO12(Q) for Test No.1, Chemical and Radiochemical, for monitoring plant water chemistry and radiochemistry during power ascension testing. Although this substitution is acceptable, we encourage the applicant to review the results of the surveillance with the same management attention which would have been given the review of the startup test.
In Test No.1, the RWCU and condensate demineralizer systems are perfomance tested to demonstrate that they meet design specifications at Test Conditions 3 (Iow power) and 6 (rated power / flow). Performance testing the RWCU and condensate demineralizer systems at full power and flow. Test Condition 6, will demonstrate the ability of these systems to adequately control coolant chemistry at the most demanding plant operating condition. Performance testing at a lower power, Test Condition 3, can verify procedures and provide preliminary data, but is not required to demonstrate meeting design specifications. Therefore, it is acceptable to delete the RWCU and condensate domineralizer systems inte-grated perfomance testing requirement of Test No.1 at Test Condition 3 as proposed by the applicant.
- 10. TEST NO. 32 - REACTOR WATER CLEANUP SYSTEM By letter dated. October 17, 1985, the applicant proposed modifying Test No. 32 -
Reactor Water Cleanup System. The purpose of Test No. 32, is to demonstrate i the operability of reactor coolant system purification and cleanup systems l during low power testing. It is proposed b l
RWCU non-regenerative heat exchanger (NRHX)y the applicant to a) delete the flowtestintheblowdownmode,b) delete the bottom head flow rate calibration from the power ascension test program, c) perform the reactor water cleanup system (RWCU) pump net positive Hope Creek SSER 6 18 Appendix 5
i suction head (NPSH) test under cold conditions during preoperational testing, and d) perform the non-regenerative heat exchanger (NRHX) flow test in the
- normal mode during Test Condition 1 instead of during Test Condition Heatup.
The RWCU system temperature and flow measurements will be obtained during l the normal operating mode to demonstrate the heat exchange capability of the NRHX. However, during the blowdown mode, the regenerative heat exchanger capacity is decreased as a result of partially bypassing a portion of the reactor coolant system return flow to the main condenser or radwaste system.
This could automatically isolate the RWCU system on NRHX high outlet temperature.
RWCU isolation is not desirable during heatup when water chemistry is critical i and when excess reactor coolant needs to be discharged. Therefore, the RWCU NRHX flow test in the blowdown mode should remain in Test No. 32.
The applicant proposed that the bottom head flow rate calibration be performed after completion of power ascension testing. This test is not critical to demon-strating the performance of the RWCU system. Therefore, it is acceptable to delete the bottom head flow calibration from Test No. 32 and postpone it until after completion of power ascension testing.
, The applicant proposed to determine RWCU pump NPSH during preoperational testing under cold conditions. Calculations can be performed to extrapolate NPSH from preoperational cold conditions to operational conditions to demon-strate compliance with acceptance criteria. Therefore, it is acceptable to determino RWCU pump NPSH during preoperational testing.
The applicant proposed to perform the RWCU flow test for the NRHX at Test Condition 1 instead of at Test Condition Heatup. RWCU operation in the blow-down mode is important during heatup since during this phase, part of the reactor coolant will be bypassed to the main condenser or radwaste system.
Thereture, the RWCU flow test for the NRHX should remain in Test Condition l Heatup in Test No. 32. The applicant's proposal is not acceptable.
, 11. TEST NO. 28B - TWO PUMP RECIRCULATION PUMP TRIP TEST By letter dated October 4, 1985, the applicant requested deleting the two pump trip and flow coastdown at Test Condition 3 and using the data obtained on the integrated systems Generator Load Rejection Test at Test Condition 6 to obtain the two pump trip and flow coastdown data to satisfy Regulatory Guide 1.68 Revision 2. The staff found this change to be unacceptable for the following reason. By letter dated August 21, 1985, the applicant requested to delete the turbine trip test at Test Condition 3 (see Test No. 25 - Turbine Trip and Generator Load Rejection, Item 4 in this safety evaluation). The turbine trip test deletion at Test Condition 3 was found acceptable based, in part, on the implicit condition i that safety and accident mitigation features were tested before full power was
- attained. Explicitly, this conclusion was based, in part, on the fact that the two recirculation pump trip and flow coastdown would be performed at Test Condition 3 prior to the load rejection test at full power. Therefore, the two pump trip and flow coastdown should be performed at Test Condition 3 to test this accident mitigation feature before full power is attained.
j llope Creek SSER 6 19 Appendix S
- 'o
~,, UNITED STATES
[3 -ag g
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555
%,...../
MAR 2 0 W Docket No. 50-354 Mr. Corbin A. McNeill, Jr., Vice President-Nuclear Public Service Electric & Gas Company Nuclear Administration Ruilding P. O. Box 236 Hancocks Bridge, New .lersey 08038
Dear Mr. McNeill:
Suh, lect : Hope Creek Power Ascension Procram Test Modifications By letters dated August 21, September 20, October 17, and December 9,1985, PSEAG submitted, for staff review, a number of proposed test modifications to the Hoce Creek Power Ascension Drogram. The proposed test modifications are part of a program to accelerate power ascensinn testino for Hope Creek. Additionally, Hooe Creek is the lead plant in an effort by General Electric to promote accelerated Dower ascension programs on a number of other boiling water reactors.
The staff has comoleted its review of the following test modifications proposed in the above referenced letters:
- 1. Test No. 19 - Core Power / Void Mode Response
- 2. Test No. 9 - LPRM Calibration
- 5. Test No. 5 - Control Red Scram Time Testing at Hot Standby with Full Reactor Scram Control Rod Time Testina
- 6. Test No. 23R - MSIV Full Isolation The safety evaluation detailing each review is attached. For those test modifi-cations which the staff has accepted (Items 1, 3 in part,;and 4), we request that the Hope Creek FSAR be amended to reflect those changes. This concludes our review of the submitted Hcpe Creek Power Ascension Program Test Modifications.
Sincerely.
Robert Rernero, Director Division of RWR Licensing Office of Nuclear Reactor Reculation
Enclosure:
As stated cc w/ enclosure:
See next page llope Creek SSER 6 20 Appendix S
Mr. C. A. McNeill Public Service Electric & Gas Co. Hope Creek Generating Station ec:
Gregory Minor Susan C. Remis Richard Hubbard Division of Public Interest Advocacy Dale Bridenbaugh New Jersey State Department of MHB Technical Associates the Public Advocate 1723 Pamilton Avenue, Suite K Richard J. Hughes Justice Comples San Jose, California 95125 CN-850 Trenton, New Jersey 08625 Troy B. Conner, Jr. Esquire Office of Legal Counsel Conner & Wetterhahn Department of Natural Pesources 1747 Pennsylvania Avenue N.W. and Environmental Control Washington, D.C. 20006 89 Kings Highway P.O. Box 1401 Dover, Delaware 19903 Richard Fryling, Jr., Esquire Mr. K. W. Burrowes, Pro.iect Engineer Associate General Solicitor Bechtel Power Coronration Public Service Electric & Gas Company 50 Beale Street P. O. Box 570 TSE P. O. Box 3965 Newark, New Jersey 07101 San Francisco, California 94119 Manager - Licensing and Regulation Resident Inspector c/o Public Service Electric & Gas U.S.N.R.C. Bethesda Office Center, Suite 550 P. O. Box P41 4520 East-West Highway Hancocks Bridge, New Jersey 08038 Bethesda, Maryland 20814 Ms. Rebecca Green Richard F. Engel New Jersey Bureau of Radiation Deputy Attorney General Protection Division of Law 380 Scotch Road Environmental Protection Section Trenton, New Jersey 08628 Richard J. Huohes Justice Complex CN-112P Trenton, New Jersey 086?S Mr. Robert J. Touhey, Mr. Anthony J. Pietrofitta Acting Director General Manager DNREC - Division of Power Production Engineering Environmental Control Atlantic Electric 89 Kings Highway 1199 Black Horse Pike P. O. Box 1401 Pleasantville, New Jersey 08232 Dover. Delaware 19903 Regional Administrator, Region !
Mr. R. S. Salvesen U. S. Nuclear Regulatcry Consnission General Manager-Hope Creek Operation 631 Park Avenue Public Service Electric & Gas Co. King of Prussia, Pennsylvania 19406 P.O. Box A Hancocks Bridge, New Jersey 08038 liope Creek SSER 6 21 Appendix 5 i
Public Service Electric & Gas Co. Hope Creek Generating Station cc:
Mr. 8. A. Preston Public Service Electric & Gas Co.
Hope Creek Site MC12Y Licensing Trailer 12LI Foot of Buttonwood Road Hancocks Bridge, New Jersey 08038 Hope Creek SSER 6 22 Appendix S
l l
l l
SAFETY EVALUATION HOPE CREEK GENERATING STATION POWER ASCENSION TEST PROGRAM ACCELERATION This safety evaluation describes the staff's review of a number of Hope Creek Generating Station - Power Ascension Program (PAP) test modifications proposed by Public Service Electric and Gas Company (PSE4G). Thes<r proposed test modi-fications were submitted for staff review as part of a program to accelerate power ascension testing for Hope Creek. Hope Creek is also the lead plant for generic changes for acceleration of the traditional General Electric Boiling Water Reactor power ascension test program.
The proposed test modifications discussed in this safety evaluation were sub-mitted by letters dated August 21. September 20. October 17. and December 9,1985.
This safety evaluation discusses the following tests:
- 1. Test No. 19 - Core Power-Void Mode Response Calibration
- 2. Test No. 9 - Local Power Range Monitor
- 3. Test No. 10 - Average Power Range Monitor Calibration
- 4. Test No. 23A - MSIV Functional Tests
- 5. Test No. 5 - Control Rod Scram Time Testing at Hot Standby with Full Reactor Scram Control Rod Time Testing
- 6. Test No. 238 - MSIV Full Isolation Discussion of the above tests follows:
- 1. TEST NO. 19 - CORE POWER-VOID MODE RESPONSE In a letter dated August 21, 1985, the applicant proposed deletion of the Core Power-Void Mode Response test. This is a test where the core power response to both the insertion and removal of a high reactivity worth control rod and small step changes in reactor pressure is obtained and examined for stable behavior. The applicant states that this test may be eliminated based largely on experience gained during other plant startups, and that the Pressure Regulator Test (Test No. 20) perfomed at identical test con-ditions provides better information to detemine dynamic stability in large boilingwaterreactors(BWRs)(.
and safety evaluation report SER)The applicant for the submitted Licensing the staff's Topical Report. acceptance
" Thermal Hydraulic Stability Amendment to GESTAR !! " Rev. 6. Amendment 8, NEDE-24011 as supporting technical information.
- The significant issue with this test is whether it produces significant i testing not duplicated by the Pressure Regulator Test. Results of these two tests in three recently declared comercial BWRs were examined. It was possible to confirm from examination of the testing matrix and the test descriptions that identical pressure response testing was accomplished Ilope Creek SSER 6 23 Appendix S
-2 in both core power-void mode response and pressure regulator tests. How-ever, beyond confirmation of duplication of testing, the test reports were of little value because they only included discussion of meeting the accep-tance criteria for the respective tests and did not provide comparative data. One plant which did include some reduced data indicated very little response to control rod movement but significant response to pressure changes. The results of both tests met the acceptance criteria and confirmed stability.
The supporting technical material related to Topical Report NEDE-24011 was reviewed. From the review of this supporting material, the staff finds that (a) a considerable body of BWR special stability testing exists in addition to the successful stability testing of the newly commercial BWR plants and I (b) small perturbations of reactor pressure have been demonstrated to be, a simple testing approach to determine reactor stability (see L. A. Carmit.hael 1 and R. O. Niemi, " Transient and Stability Tests at Peach Bottom Atomic Power Station, Unit 2, End of Cycle 2," Electric Power Research Institute, 1978 (EPRI Np. 564)). Therefore, the staff concludes that control rod oscilla-tion testing to establish reactor stability is unnecessary because it is duplicative of testing performed at similar plants and produces lesser quality stability confirmation than reactor pressure perturbation tech-niques. Further, the staff concludes that it is unnecessary to duplicate the pressure cycling tests of the Pressure Regulator Test and it is, therefore, acceptable to delete the Core Power-Void Mode Response Test.
- 2. TEST NO. 9 - LOCAL POWER RANGE MONITOR CALIBRATION in a letter dated September 20, 1985, the applicant requested that their planned response checks to control rod movements during heatup and Test Condition 1 be deleted for the LPRM calibration test. This testing verifies that the LPRMs are responding to neutron flux and provides con-fidence that when the APRMs are calibrated under the constant rate heatup calibration, a representative calibration is obtained.
The staff notes that the requested change is a level of detailed implemen.
I tation which does not appear in the FSAR test abstract. The staff recog-i nizes that early confirmation of control rod movement and proper instalia-tion of the LPRM detectors are valuable when the detector output is in its useful range, and that, until the threshold of usefulness is attained,
! response checks of this type are of little value. However, the low-power license for the facility will be conditioned to prohibit operation above i
i 5% of full thermal power. The only way to assure this power level is not exceeded is based on the APRM constant rate heatup calibration, which j
requires proper LPRM calibration. For this reason the staff believes that the LPRM response checks should remain in the test.
l 3. TEST No. 10 - AVERAGE POWER RANGE MONITOR CALIBRATION The objective of the APRM Calibration Startup Test is to calibrate the l APRM system with respect to the plant heat balance. During initial heatup, at power levels too low for a heat balance, the first APRM adjustment is accomplished by adjusting the APRM gains so the APRMs indicate slightly Hope Creek SSER 6 24 Appendix S
L .
l higher than an estimated power level, calculated by using the heatup rate
! and other plant parameters. By letter dated October 17, 1985, the appli-cant proposed using preselected gains based on startup experience at other l plants to initialize the low power adjustment of the APRM system. The staff notes that this is a level of detailed implementation which does not appear in the FSAR test abstract. It is the staff's belief that the startup experience from other facilities does not necessarily demonstrate that preselected APRM gains will provide conservative indication of reactor power levels. It is the staff's position that these preselected gains should not be used.
The Technical Specifications require surveillance testing for APRM system calibration at conditions where the APRM startup testing calibration is performed. The applicant requested combining the startup test and surveil-lance test for calibration of the APRM system into a single test to eliminate redundant testing. The objectives of both tests are identical. The staff discussed the combining of surveillance and startup testing in the safety evaluation for Test No.17, transmitted by letter dated January 22, 1986. As stated in the January 22, 1986 letter, the staff finds the combination of surveillance and startup testing to be acceptable if (a) the test objectives are the same; (b) it is consistent with regulatory guidance, and (c) if combined as discussed in the January 22, 1986 safety evaluation, the same level of review, approval and data evaluation of a startup test procedure is retained.
- 4. TEST NO. 23A - MSIV FUNCTIONAL TESTS In a letter dated October 17, 1985, the applicant proposed changing Test No.
23A- MSIV Functional Tests. The proposed test change deletes determining the maximum power condition for MS!V functional testing. The applicant stated that future full closure tests of the MSIVs to satisf specification requirements will be done at cold conditions as is (y plant done at technical Limerick). The testing at cold conditions is conservative because the MSIVs, by their design, close faster during the rated conditions (when steam is flowinginthesteamlines). Therefore, the testing to detemine the maximum power level for subsequent testing is not required. A one time test of full MSIV closure will be performed. Power Ascension Test No. 238, MSIV full closure test is scheduled to be performed at full power to satisfy the regulatory requirements. The proposed change to delete the testing of MS!Vs to determine the maximum power level for subsequent surveillance testing is acceptable.
- 5. TEST N0. 5 - CONTROL R00 SCRAM TIME TESTING AT HOT STANDBY WITH FULL REACTOR 5 CRAM CONTROL ROD TIME TESTING In a letter dated December 9, 1985, the applicant proposed modifying Test No. 5. Test No. 5 is concerned with testing of control rod drives. As part of the test, rods are scram time tested after fuel loading at cold shut-down. During reactor heatup, four selected rods are time tested at various reactor pressures. Nomally, all rods are time tested at rated pressure and low power, and again four rods are selected for testing during power ascension. The applicant has proposed that this sequence be llope Creek SSER 6 25 Appendix 5
I modified such that the test of all rods at rated pressure and low power be replaced with a test of four rods, and a test at approximately 20 percent power of all withdrawn rods as part of the scram in the Loss of Offsite Power test. This would result in the 37. " Control Cell Core" (CCC) rods ,
not being fully tested since they would not be fully withdrawn. '
General Electric has analyzed the reactivity worth of the scram over the insertion range of interest to transient analyses and found only a small difference in reactivity insertion between the scram of all rods and a scram in which the CCC rods and 8 other (assumed) extreme, inoperable rods do not insert. (The CCC rods are low worth rods.) They have also reexamined the transient analyses of events with this reduced scram function and determined 1 a minimum critical power ratio (MCPR) penalty to be applied for the assumption '
that these rods do not participate in the scram. The operating limit MCPR (OLMCPR) is increased by no more than 0.01 for either the ODYN option A or for the curve of OLMCPR vs measured scram time for 00YN option B. The pro-posed ODYN scram time would involve only those rods fully measured in the scram time test.
, In the December 9, 1985 letter, the applicant proposed Technical Specification changeswhichwould(1)excludeforfirstcycleonly,(the37CCCrodsfrom2)increasetheO scram time tests in Specification 4.1.3.2 and .3 arid by 0.01 in Specification 3/4.2.3 and the related figure.
, Based on the above information, the staff finds that (1) all rods will have i been scram tested at cold conditions (2) four rods will have been tested i during heatup, including full pressure, and thus indicate any departure j from normal trends, (3) analysis indicates little reactivity worth change i during times important for transient analyses, and transients have been t
reexamined assuming no scram at all for CCC rods and 8 inoperable rods and suitable OLMCPR ad;ustments have been made to account for changes, and (4) suitable Technical Specifications have been proposed to exempt CCC ,
( rods and to increase OLMCPR. However, the applicant has not addressed i
! the affect the proposed test modification will have on the shutdown margin technical specification. Additionally, performing the first full-core
,l scram at rated conditions (temperature and pressure) during a loss of offsite power test is not acceptable. Because this would be the first scram under operating conditions, there is no assurance that a successful full-core scram will be realized. Additionally, should the scram not occur, the platit would be in an anticipated transient without scram situation, i Although the applicant has provided sound technical justification from I
a core physics perspective why the proposed test modification should be accepted, the above concerns are of such significance that the proposed test modification, in its current fom, is unacceptable.
l 6. TEST No. 238 - MSIV FULL ISOLATION In a letter dated October 17, 1985, the applicant proposed substitution of i an inadvertent M51V closure for the planned MSIV closure test. The staff i finds the proposed substitution of an inadvertent full M$1V closure for the planned full MSIV closure test to be unacceptable for the following flope Creek SSER 6 26 Appendix S
l 5
reason. Although precedent exists for the substitution of inadvertent transients for planned tests, the acceptability of the substitution has been determined only efter thorough analysis to assure that the proper quantity and quality of data was obtained which demonstrated that the in-advertent transient fulfilled all the objectives of the planned test. The evaluation of an inadvertent full MSIV closure can only be performed after the transient has occurred. Therefore, pre-approval to substitute an in-advertent transient for a planned test is inappropriate, and the proposal is unacceptable.
Hope Creek SSER 6 27 Appendix S
APPENDIX T TECHNICAL SPECIFICATION CHANGES BETWEEN LOW-POWER LICENSE ISSUANCE AND FULL-POWER LICENSE ISSUANCE l Since the issuance of the low-power license, the licensee has requested certain i
changes to the Technical Specifications. These changes and the staffs' conclu-sions regarding their acceptablity are discussed below. Some of these changes modify license conditions.
(1) Technical Specifications 4.6.5.3.b and 4.7.2.b In a letter dated June 13, 1986, the licensee requested changes to Technical Specifications 4.6.5.3 and 4.7.2 to state that the filtration, recirculation, and ventilation system (FRVS) and control room emergency filtration system (CREFS) heaters will be on, rather than just operable, for the required sur-veillance. Surveillance Requirements 4.6.5.3.b and 4.7.2.b currently state:
At least once per 31 days by initiating, from the control room, flow through the HEPA (high efficiency particulate air) filters and char-coal adsorbers and verifying that the subsystem operates for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters and humidity control instrumentation OPERABLE.
The licensee requested that these requirements be changed to read:
At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on in order to reduce the buildup of moisture on the carbon adsorbers and HEPA filters.
The primary purpose of the required periodical test operation of engineered safety feature flitration systems (10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> for every 31 days with heaters on) is to reduce the potential buildup of moisture on the charcoal adsorbers and HEPA filters. The only function of the humidity control instrumentation (humidity controllers) is to activate the heaters when relative humidity in the incoming air to the charcoal adsorber and HEPA filter exceeds 70%. The humidity controllers are channel calibrated every 18 months, as specified separately in Surveillance Requirements 4.6.5.3.e.3 and 4.7.2.e.4. Since the surveillance requirement will necessitate having the heaters operable, the licensee's requested changes are reasonable and consistent with the intent of this Technical Specification section. Therefore, the staff finds the requested changes acceptable.
(2) Technical Specifications 4.6.5.3.e.3 and 4.7.2.e.4 in a letter dated June 13, 1986, the licensee requested changes to these Tech-nical Specifications to clarify the method of verifying humidity control oper-ability and to change the FRVS heater dissipation rating.
Hope Creek SSER 6 1 Appendix T
The requested changes clarify the surveillance requirements for the humidity controllers by stipulating the performance requirements for humidity controller testing in the last sentences in Technical Specifications 4.6.5.3.e.3 and 4.7.2.e.4 as follows:
Verifying humidity is maintained less than or equal to 70% relative humidity through the carbon adsorbers by performance of a channel calibration of humidity control instrumentation.
As discussed in Item (1) above, this change requires a channel calibration of the humidity controllers every 18 months, independent of periodic heater opera-tion, and constitutes clarification of one surveillance requirement for the humidity controllers. The staff finds this change acceptable because channel calibration is an acceptable means of verifying the operability of the controllers.
In addition, the licensee requested a change to the heater dissipation rate specified in Technical Specificatien 4.6.5.3.e.3 from the currently specified 100 1 5 kW to 100 1 10 kW, consistent with the requirement of American National Standards Institute Standard ANSI N510-1980, which allows that the heater dissipation rating can be 110% of the heater output. The staff finds this request also acceptable.
(3) Technical Specification Table 3.3.7.1-1. " Radiation Monitorina Instrumentation" The current Table 3.3.7.1-1, Action 74 states that with the offgas pretreat-mentradiationmonitorinoperable,therelease(s)Intothepathwaymaycon-tinue for up to 30 days provided:
- a. The offgas system is not bypassed, and
- b. The offgas post-treatment radiation monitor is OPERABLE, and
- c. Grab samples are taken at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and analyzed within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
In a letter dated June 13, 1986, the licensee requested deletion of Provi-sion (b) above, that is, deletion of the requirement that the offgas post-treatment radiation monitor be operable. The Hope Creek offgas system is designed to provide 35 days and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of delay time for xenon and krypton, respectively, with a total air inleakage of 75 standard cubic feet per minute.
The post-treatment monitor indication is, therefore, not for real-time measure-ment inlication. Instead, grab samples taken every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (analyzed within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) should be adequate to meet the limiting condition for operatian in Technical Specification 3.11.2.7. In addition, everything that passes through the offgas post-treatment monitor passes through the downstream north path vent monitor, which is required to be operable by Technical Specift-cation Table 4.11.2.1.2-1. The staff finds the requested change acceptable.
(4) Technical Specification Table 3.6.3-1. " Primary Containment Isolation Valves" The current Table 3.6.3-1 requires operability testing of the reactor vessel head seal leak detection line excess flow check valve. In a letter dated Hope Creek SSER 6 2 Appendix T
June 13, 1986, the licensee proposed Technical Specification change that would add a notation deleting the requirement to operability test excess flow check valve BB-XV-3649.
The reactor vessel head seal leak detection line excess flow check valve is not normally subjected to primary system pressure. If the reactor vessel head seal should fall, this valve could be pressurized to reactor pressure. However, any leakage is restricted at the source (head seal), by the instrument line orifice, and at the pressure instrument itself. This excess flow check valve has a 10 CFR 50, Appendix J, exemption for Type A leakage testing. Therefore, this valve need be not subject to operability testing by Technical Specifica-tion. The staff finds the requested change acceptable.
(5) Technical Specification 4.4.6.1.4 In a letter dated June 13, 1986, the licensee requested a change to Technical Specification 4.4.6.1.4 that would reduce the minimum temperature for closure stud tensioning from 79'F to 70*F.
The limiting RTNOT f material in the flange region is 19'F. The revision is permitted under Paragraph G-2222 (c) of the American Society of Mechanical En-gineers Boller and Pressure Vessel Code (ASME Code),Section III. The current Code states that when the flange and adjacent shell region are stressed by the full intended bolt preload and by pressure not exceeding 20% of system hydro-static test pressure, the minimum metal temperature in the stressed region should not be lower than RTNDT plus the effect of irradiation damage.
The proposed revision of the Technical Specification is in compilance with the ASME Code and, therefore, acceptable.
References American National Standards Institute, ANSI N510-1980, " Testing of Nuclear Air-Cleaning Systems."
American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Sec-tion Ill, Paragraph G-2222, 1971 Edition, Winter 1973 Addenda.
Hope Creek SSER 6 3 Appendix T
APPENDIX 0 PROBABILITY OF MISSILE GENERATION IN GENERAL ELECTRIC NUCLEAR TURBINES 1
SUMMARY
The objective of the NRC staff's review of the General Electric Company (GE) report, " Probability of Missile Generation in General Electric Nuclear Turbines" (submitted by the licensee in a letter dated July 11,1986), was to evaluate and, if appropriate, approve the methods and procedures utilized by the General Electric Company, Large Steam Turbine-Generator Department, to determine spe-cific turbine system inspection and testing intervals for its utility customers.
During the past few years, the staff has recommended a probabilistic approach to determine turbine rotor inspection intervals and turbine control system maintenance and testing frequencies so as to maintain the as-built turbine sys-tem integrity. The GE report describes such an approach generically and, to the extent possible, supports it with test and turbine system operating experi-ence data. The staff recognizes that probabilistic analyaes based on limited statistical data, especially for a complex system, will include inherent un-certainties. Nevertheless, when the overall approach includes conservative assumptions that overcome the uncertainties, then the ultimate results can be meaningful.
The staff concludes that the methodology dpscribed in the GE report is state of the art and is acceptable for use in establishing maintenance and inspection schedules for specific turbine systems.
Applicants or licensees who accept GE's recommendations, based on this report shouldconfirmtheircommitmenttothestaffandprovideadescriptionoftheIr specific maintenance and inspection program including a curve (or curves) of misslie probability (Pi ) versus service time for their specific turbine rotors.
2 BACKGROUNO Although large steam turbines and their auxiliaries are not safety-related systems as defined by NRC regulations, failures that occur in these turbines can produce large, high-energy missiles. If such missiles were to strike and damageplantsafety-relatedstructureslrsafetyfunction.
render them unavailable to perform the systems, and components, Consequently, General they could Design
" GeneralCriterion Design4Criteria
" Environmental for Nuclearand Missile Power Design Plants," Bases,"
to 10 of Ap')endix CFR 50, omesticA, Licensing of Production and Utilization Facilities," requires, in part, that structures systems, and components important to safety be appropriately pro-tectedagalnsttheaffectsofmissilesthatmightresultfromsuchfailures, in the past, with regard to construction permit and operating license applica-tions, evaluation of the effects of turbine failure on the public health and safety followed Regulatory Guide (RG) 1.115 " Protection Against Low-Trajectory Hope Creek SSER G 1 Appendix U
Turbine Missiles," and three essentially independent Standard Review Plan (SRP, NUREG-0800) sections: Sections 10.2, " Turbine Generator," 10.2.3, " Turbine Disk Integrity," and 3.5.1.3, " Turbine Missiles."
According to the NRC guidelines in SRP Section 2.2.3 and RG 1.115, the proba-bility of unacceptable damage from turbine missiles (P4 ) should be less than or equal to about 1 chance in 10 million per year for an individual plant, that is, P4< 10 7 per year. The probability of unacceptable dama turbine _ missiles is generally expressed as the product of the (1)ge resulting from probability of turbine failure resulting in the ejection of turbine disc (or internal struc-ture) fragments through the turbine casing (P ); (2) the probability of ejected t
missiles perforating intervening barriers and striking safety-related structures, systems, or components2(P ); and (3) the probability of struck structures, sys-tems, or components failing to perform their safety function (P3 ).
In the past, analyses assumed the probability of missile generation (P ) to be approximately 10 4 per turbine year based on the historical failure rate l (Bush,1973,1978). ThestrikeprobabIlity(P)wasestimatedonthebasisof 2
postulated missile sizes shapes, and energies and on available plant-specific informationsuchasturbIneplacementandorientation,numberandtypeof intervening barriers, target geometry, and potential missile trajectories.
(See SRP Section 3.5.1.3 for a description of the evaluation procedures pre-viously recommended by the staff.) The damage probability (P3 ) was generally assumed to be 1.0. The overall probability of unacceptable damage to safety- i related systems (P 4 ), which is the sum over all targets of the product of these probabilities, was then evaluated for compliance with the NRC safety objective. t This logic places the regulatory emphasis on the strike probability; that is, i it necessitates that Pg be made less than or equal to 10 8 and disregards all t the plant-specific factors that determine the actual Pi and its unique time 4 dependency.
Although for the mostthepart calculation being not of more strike probability than is not difficult a straightforward ballisticsinanalysis, principle,it ,
presents a problem in practice. The problem stems from the fact that numerous ["
modeling approximations and simplifying assumptions are required to make tract-able the jncorporation into acceptable models of available data on the (1) pro- !
perties 6f missiles, (2) interactions of missiles with barriers and obstacles, !
(3) trajectories of missiles as they interact with and perforate (or are de-flected by)' barriers, and (4) identification and location of safety-related targets. The particular approximations and assumptions made tend to have a ,
significant effect on the resulting value of P2 . Siellarly, a reasonably accurate specification of the damage probability (P3 ) is not a simple matter ,
because of the difficulty in defining the missile impact energy required to render given safety-related systems unavailable to perform their safety func- !
tions and the difficulty in postulating sequences of events that would follow '
a misslie producing turbine failure.
Operatingexperienceshowsthatnuclearturbinediscscrack(NorthernStates that turbine stop and control valves fail '
(Burns,1977;SouthernCaliforn PowerCo.,1981;NUREG/CR-1884)IaEdisonandSanDiegoGa 1982),
and that disc ruptures could result in the generation of h(gh energy miss es (Kalderon,1972). Analyses (Burns, 1977; Clark, Seth, and Shaffer, 1981) show ,
that missile generation can be modeled and the probability can be strongly in- .
fluenced by inservice testing and inspection frequencies. !
Hope Creek SSER 6 2 Appendix U r
- -__ - -,_ _ _ _ -,.. _ _ ... , _ _ _ ,. , - - . _ _ _ . _v-r --- - , _ _ - _ M
During the past few years, the results of turbine inspections at operating nuclear facilities indicate that cracking to various degrees has occurred at the inner radius of turbine discs of Westinghouse design. Within this period, a Westinghouse turbine disc failure occurred at one facility owned by the Yankee Atomic Electric Company (NUREG/CR-1884). More recent inspections of GE turbines have also discovered disc keyway cracking (Northern States Power Co.,
1981). Stress corrosion has been identifled by both manufacturers as the operative cracking mechanism.
In view of operating experience and NRC safety objectives, the NRC staff has shifted emphasis in the reviews of the turbine missile issue from the strike and damage probability (P 2 xPa) to the missile generation probability (P t) and, in the process, has attempted to integrate the various aspects of the issue into a single, coherent evaluation.
Through the experience of reviewing various licensing applications, the staff has concluded that P2 xPa analyses provide only " ball park" or " order-of-magnitude" values. On the basis of simple estimates for a variety of plant layouts, the staff also concludes that the strike and damage probability product (P xPa) 2 can be reasonably taken to fall in the characteristic narrow range that is depen-dent on the gross features of plant layout with respect to turbine generator orientation;thatis,(1)forfavorablyorientedturbinegenerators,PxPatends 2 to lie in the range of 10 4 to 10 3yr- and (2) for unfavorably oriented turbine generators, P2 xPa tends to lie in the range of 10 3 to 10 2yp.1 In addition, detailed analyses such as those discussed in this evaluation show that, depend-o ing maintenance and on the specific combination practices, P of material propertiest to 10can 1 perhave values turbine year from 10 0 pera depending on the turbine test and inspection intervals. For these reasons, in the evaluation of P 4 (P ixP2 xPa), the probability of unacceptable damage to safety-related systems from potential turbine missiles the staff is giving creditfortheproductofthestrikeanddamageprobabIlitiesof10.syr 1 for a favorably oriented turbine and 10 2yr 1 for an unfavorably oriented turbine, and is discouraging the elaborate calculation of these values.
The staff believes that maintaining an initial small value of P t through tur-bine testing and inspection is a reliable means of ensuring that the objectives precluding turbine missiles and unacceptable damage to safety-related structures, systems and components can be met. It simpilfies and improves procedures for evaluatIngturbinemissilerisksandensuresthatthepublichealthandsafety is maintained.
To implement this shift of emphasis, the staff recently has proposed guidelines for total turbine missile generation probabilities (Table U.1) to be used for determining (1) frequencies of turbine disc ultrasonic inservice inspections ar.d (2) maintenance and testing schedules for turbine control and overspeed protection systems. It should be noted that no change in safety criteria is associated with this change in emphasis.
3 SCOPE OF REVIEW There are essentially two modes of turbine disc failure that can result in tur-bine failure one resulting from rotor material failure at approximately the ratedoperatingspeed,oroneresultingfromfailureoftheoverspeedprotection systems resulting in excessive rotor speeds.
Hope Creek SSER 6 3 Appandix U
Failures of turbine discs at or below the design speed, nominally 120% of normal operating speed, can be caused by small flaws or cracks left during fabrication or those that initiate during operation and grow to critical size either by fatigue crack growth, by stress corrosion crack growth, or by a combination of both of those mechanisms. Cracks in the bore or hub region of turbine discs could eventually lead to disc failure.
Failures of turbine discs at the destructive overspeed can result from a fail-ure of the governor and overspeed protection systems consisting of speed sensing and tripping systems and steam valves. If the turbine is out of control, its speed can increase until failure occurs. For unflawed discs, destructive over-speed is reached at about 180 to 190% of the normal operating speed. In general, failures that occur at destructive overspeed are caused by stresses that exceed the materials tensile strength.
If a turbine disc should burst, high-velocity, missile-like fragments may break through the turbine casing, possibly generating secondary missiles. These mis-siles have a potential for damaging reactor safety systems. Alternately, the disc fragments could be arrested and contained by the turbine itself. Hence, in evaluating the risk associated with turbine disc rupture, it is necessary to determine whether or not missiles external to the casing can be generated by postulated disc ruptures.
This appendix considers the above possibilities and summarizes the review and evaluation of the GE report, which describes GE procedures for estimating (1) the design speed missile generation probability, (2) the destructive overspeed misslie generation probability, and (3) the perforation of the turbine casing by turbine disc burst fragstents.
4 DISCUSSION / EVALUATION This appendix presents an overview of the methodology in Section 2 of the GE report where three major components of the methodology are considered:
(1) probability of turbine overspeed (2) wheel burst probability (3) probability of casing penetration The probability of a wheel burst and the probability that a wheel fragment will penetrate the casing will depend on the speed at which a wheel bursts. Turbine conditions; however, when an speed abnormal event occurs, such as load rejection and /or failure of the control is close to 1800 rpm under normal operating system to function properly, turbine speed may reach 180 to 190% of the rated speed. The probability of attaining these various turbine overspeed levels, therefore, is a major component of the methodology.
Another major component of the methodology is the probability of a wheel burst at various operating conditions, which are defined by two important parameters:
speed and wheel temperature. The primary failure mode of the turbine wheel is assumed to be brittle fracture caused by the presence of a stress corrosion crack in the keyway near the bore of the shrunk-on wheel. The fracture mechan-les calculations include variations in the toughness of the wheci material in the depth of the crack, in the likelihood of crack initiation, in the ablIIty to detect crack sizes during inservice inspections, and in the rate of crack l growth during subsequent service.
Hope Creek SSER 6 4 Appendix U ,
The third major component of the methodology is the probability of a wheel fragment penetrating the turbine casing, given the wheel burst at a particular speed. The missile penetration probabilities are based on energy methods (Gonea, 1973) and laboratory tests. The variations involved in these calcula-tions lead to a probabilistic estimate of casing penetration as a function of burst speed.
Section 3 of the GE report describes overspeed protection systems. GE nuclear steam turbines are equipped with three speed-sensing devices for defense against turbine overspeed.
(1) Normal overspeed protection is achieved through the control valves, inter-cept valves, and check valves.
(2) An emergency overspeed protection device is set to close all steam valves if the speed reaches 110 to 111% of the operating level.
(3) A backup overspeed protection device is set to close all steam valves if the speed exceeds the emergency trip setpoint (112%).
Both mechanical hydraulic control (MHC) and electrohydraulic control (EHC) sys-tems are employed. Failure models for MHC and EHC systems are analyzed by a fault tree method, and the probability of attaining a given speed is calculated.
Section 4 considers wheel burst in both brittle and ductile modes. Operating experience shows that the primary failure mode of the turbine wheels is assumed to be brittle fracture resulting from the presence of stress corrosion cracks in the keyway near the bore of the shrunk-on wheel. After ascertaining the fracture toughness property at various depths, calculations are made to deter-mine crack length at a particular time from the initial service. Considerations in the probability analysis are gi in to variations in the likelihood of crack initiation, in the ability to detect and size cracks during inservice inspec-tions, and in the rate of crack growth during subsequent service.
The statistical distribution is applied to crack initiation and growth behavior data obtained from inservice inspections performed on the majority of wheels of operating GE nuclear low pressure turbines. The relevant information can be extracted from these statistical distributions to arrive at a given value for any assigned probability or vice versa. Because of various parameters in-volved, the time to crack initiation varies significantly from wheel to wheel.
Tests on wheels with laboratory produced stress corrosion cracks and on those retired from service were used to define the ability of ultrasonic testing (UT) methods to detect and size wheel cracks. The GE analysis of the data shows that the crack depth is 0.07 in, larger on the average than the measured value for the wheel hub. The crack initiation distribution is influenced by the oxygen concentration in the steam and the type of locking ring that covers
, the keyways. The Weilbull distribution was fitted to the observed field data, l and the characteristic life, a Weilbull parameter, is approximated. Because of the limitations of UT equipment, undetected cracks might have initiated, and when these undetected cracks are taken into account and combined with the average crack growth rate from these initiation times, the actual distribution l
of crack depths can be estimated.
Hope Creek SSER 6 5 Appendix U
The report synthesizes stress corrosion crack growth with fracture appearance transition temperature (FATT), excess temperature versus KIC, and the calculated Kg , the stress intensity factor at various operating conditions.
FATT is determined from the test results on retired wheels and other laboratory-generated test data. The FATT value increases with distance from the surface to the interior of the wheel. The prediction of deep-seated FATT values is based on the regression analysis, which takes into account the range of three nickel alloys. The distributions of points about the median lines are normal with a standard deviation of 28 F. The overall standard deviation is 35 F when the cooling rate, ultimate tensile strength, percent carbon, and percent nickel error-distributions are taken into account.
The toughness of the wheel material can be ascertained from toughness curves ;
based on excess temperature (material temperature minus FATT) and the data I generated from valid American Society for Testing and Materials specimens. A semilog relation fits the data below 100 F excess temperature. The data are l more widely scattered at a lower excess temperature than at the higher values. 1 Here, the natural logarithm of standard deviation is a linear function of '
excess temperature. A lognormal relation is used for all upper-shelf values utilizing the Rolf-Novak relation for all the GE shrunk-wheel service data.
The stress intensity factor Kyis determined from a relationship involving a crack shape factor, the stress, the crack depth, and the geometry of the part near the crack. The general shape of stress corrosion cracks is assumed to be elliptical. They are quarter elliptical at corners and semielliptical in the interior. An average aspect ratio of death of crack (half the minor axis of the ellipse) to the half-length alcno the surface (half the major axis) is assumed to be about 0.4, on the basis of a study of three wheels that had several stress corrosion cracks. The average crack shape factor for corner cracks in the keyway under the hub was calculated to be 1.85. The average shape factor of 1.71 for semielliptical cracks under the web was calculated.
The log standard deviation for both of these factors is taken to be 0.01. The corrosion crack branching factor distribution is derived from test data on retired wheels and other data reported in the literature. The nominal bore stress is assumed to be lognormal with a standard deviation of 0.02. The
- keyway geometric function is based on the weight function method applied to I the results of finite element analyses.
After obtaining the probability of a crack initiating at time t and knowing crack depth "a" at inspection time t , and using the Weilbull distribution for i
i growth rate, the probability of having a crack depth "a" at time ti regardless of when it initiates is obtained by multiplying these two probabilities and integrating over the range from zero to t . Multiplying the probability of i
having a crack depth "a" at time ti by the probability of not detecting a crack depth "a" and integrating the product from zero to infinity for all possible crack depths, the probability of missing a crack of any depth at time t iis obtained. Thus, dividing the probability of missing a crack of depth "a" by the probability of missing a crack of any depth at ti will give the
! density function of any undetected cracks at time ti . Various combinations of temperatures and locking devices result in a median value of 0.03 in. with a lognormal standard deviation of 0.24. This shows that there is a 50% proba-bility of the undetected crack size being less than 0.03 in.
Hope Creek SSER 6 6 Appendix U
The probability of cracks existing when no indication is found is computed by dividing the probability of missing any depth crack at t i by the sum of the probability of missing any depth crack and the probability of no crack existing at time t .i This probability is the same as the probability of missing a crack at inspection time t . The difference between the true crack initiation distri-i bution and the observed crack initiation distribution is considered as a per-centage difference for a given time. Thus, the percentage of cracked wheels is higher for the true crack initiation distribution than for the observed distribution. -
After adjusting the field data based on true crack initiation distribution, the Weilbull distribution as a function of the temperature parameter, reciprocal to temperature, reactor type, and type of locking ring, showed that the Weilbull slope was close to unity and the characteristic growth rate distribution for the third iteration remains indistinguishable from that of the first iteration.
The probability of wheel burst at any time is a function of speed and tempera-ture during a cycle between two refueling outages. The cumulative probability of burst increases in time since the last inspection. The probability of wheel burst at time t ,2 P OB ),2can occur either when the wheel bursts at normal operation PBN(t 2 ) or it bursts at abnormal operation PBA(t2 ).
H wever, burst at abnormal speed will occur only if there is no burst at normal speed. The annual rate of missile generation during normal operation is calculated by multiplying PBN(tp ) by the probability of a missile given a burst at normal speed. The probability of burst for abnormal events is derived from the assump-1 tion that a burst will not occur until the cumulative burst probability exceeds the level attained during normal operation. An abnormal event occurs at a given temperature and a given maximum speed. This probability difference is summed up for all temperature levels for this abnormal event. Further, summing up for all abnormal events gives the probability of external missile generation P,(t2 )'
which depends on the speed at which the wheel bursts. Hence, the event (missile) must be integrated over the speed ranges for a given temperature. This differ-ence must be multiplied by the probability of speed and temperature occurring, and summed for all temperatures that can occur for the abnormal event. This probability must be again multiplied by the annual probability of an abnormal event occurring and summed for all possible abnormal events. Thus, the proba-bility of a missile resulting from abnormal events is obtained. The final probability Pi is the sum of the probability of a missile resulting from normal and abnormal events.
The second mode of failure is ductile fracture of the wheel during an abnor-mally high overspeed occurrence. Failure occurs when the average tangential stress across the wheel section exceeds the tensile strength of the material.
Since both brittle and ductile modes are statistically independent, the com-bined probability of failure is expressed as a standardized normal distribution.
Section 5 of the GE report discusses the values of the casing escape probability
- of each shrunk-on wheel of nuclear turbines manufactured by GE. Earlier analyses assumed that the energy absorption was due to a gross deformation of many compo-nents of the low pressure turbine casing. However, present tests show that the absorption is a local " punching" mechanism. Electric Power Research Institute Hope Creek SSER 6 7 Appendix U i
full-scale casing penetration tests consisted of accelerating a 120* segment of an actual turbine wheel at 180% speed of the turbine. Test results show that empirical formulas are overly conservative (McHugh, Seaman, and Gupta, 1983).
The actual penetration of the missile is only halfway through the wall when a 8300-1b missile at 450 ft/sec strikes the wall. The range of final energy vari-ation (energy remaining after absorption) is based on normal distribution with two sigma limits.
Section 6 of the report gives an overall determination of a wheel burst proba-bility that is a function of time, temperature, and speed. During a typical ncrmal operating cycle, the temperature varies from 50*F at the start (0 speed) to 220*F at full loading (1800 rpm), then to 120 F after the coastdown. The probability of the annual failure rate is calculated for both the normal and abnormal operating conditions. By combining these two probabilities, the proba-bility P ,1 the generation of an external missile, is derived.
Section 7 of the report discusses typical results of calculations. To provide further insight into the influence of various factors involved in the method,
- missile probability calculations have been made for a typical GE turbine used with a boiling-water reactor. Two tables summarize the information for each of the 32 wheels used on the turbine. The median value of calculated deep-seated FATT is given for each wheel together with the measured values of sur-face FATT and tensile strength. The type of locking ring used with the axial key is also noted. Tables also describe the wheel temperature under full-load l
conditions and the median value of crack growth rate, which is calculated using this design temperature. Another table describes the results of missile proba-bility calculations for each wheel of the low pressure rotor turbine at various times since the last inspection. On the basis of these calculations, the risk l
of missile generation for each rotor and the unit can be estimated.
l 5 CONCLUSIONS AND RECOMMENDATIONS l
The methodology used in the GE report for the calculation of disc rupture and turbine missile generation probabilities is a straightforward application of
! probabilistic concepts to variations in surface FATT and deep-seated FATT, overspeed due to load rejection, and/or failure of the control system to func-tion properly. The fracture mechanics calculations include the statistical l
variations in the toughness of the wheel material, in the depth of the crack,
- in the likelihood of crack initiation, in the ability to detect and to size cracks during inservice inspections, and in the rate of crack growth during l subsequent service. Because of the mixing of surface FATT and the deep-seated FATT values, the overall standard deviation is a larger value than the staff would anticipate. In this way conservatism is introduced at each step. The population of experimental tests and the actual data from the retired wheels are still small, and this results in a large standard deviation, thereby giving a conservative estimate. The staff finds that the GE crack growth equation gives a somewhat lower growth rate than that by another vendor; however, the allowable crack length is only one-half the critical crack length for the deter-mination of an inspection interval. Again it should be emphasized that the upper-shelf value of toughness is code allowable (200 Ksi-in.b), which is again a conservative value.
To arrive at the final probability of missile generation under normal and ab-normal operating conditions, a series of numerical integrations is required Hope Creek SSER 6 8 Appendix U
and this may introduce some uncertainty. However, the missile penetration formula used is conservative (Woodfin, 1983) so that disc fragments as heavy as 4600 lb at velocities as great as 300 mph penetrated less than half the thickness of walls at impact velocities that would have produced complete perforation according to other formulas.
The staff has completed its review of the GE report after it met with GE per-sonnel at Schenectady, New York, to resolve some questions. The staff believes that various safety factors or margins used in arriving at the final inspection interval are adequate and the report describes an acceptable method to determine such inspection intervals.
Therefore, the staff concludes that the report may be used in determining the inspection interval for turbine discs in operating and new reactor plants. The inspection interval will vary from plant to plant on the basis of the type of turbine in service and the previous inspection results. Applicants or licensees who wish to reference this report should commit to the turbine inspection intervals determined by GE and should submit a brief summary of how the GE method is used for their specific tur~ines.
o The summary should 1nclude a plot of missile probability versus inspection interval.
6 REFERENCES Burns, J. J. , Jr, " Reliability of Nuclear Power Plant Steam Turbine Overspeed Control Systems," 1977 ASME Failure Prevention and Reliability Conference, Chicago, Illinois, September 1977, p. 27.
Bush, S. H., " Probability of Damage to Nuclear Components," Nuclear Safety, 14(3): May-June 1973, p. 187.
-- , "A Reassessment of Turbine-Generator Failure Probability," Nuclear Safety, 19(6): November-December 1978, p. 681. -
Clark, W. G. , Jr. , B. B. Seth, and D. H. Shaffer, " Procedures for Estimating the Probability of Steam Turbine Disc Rupture From Stress Corrosion Cracking,"
ASME/IEEE Power Generation Conference, St. Louis, Missouri, October 4-8, 1981.
Code of Federal Regulations, Title 10, " Energy," U.S. Government Printing Office, W 4shington, D.C.
Gonea, D. C., "An Analysis of the Energy of Hypothetical Wheel Missiles Escap-ing From Turbine Casings," General Electric Company, Turbine Department Report, February 1973.
Kalderon, D. , " Steam Turbine Failure at Hinkley Point A," Proceedings of the Institution of Mechanical Engineers, 186(31/72): 1972, p. 341.
McHugh, S., L. Seaman, and Y. Gupta, " Scale Modeling of Turbine Missile Impact Into Concrete," Electric Power Research Institute, Final Report NP-2746, February 1983.
Northern States Power Co., Preliminary Notification of Event or Unusual Occur-rence, PN0-III-81-104, " Circle in the Hub of the Eleventh Stage Wheel in the Main Turbine," Monticello Nuclear Power Station, November 24, 1981.
Hope Creek SSER 6 9 Appendix U
Southern California Edison and San Diego Gas & Electric Co., Licensee Event Report No.82-132, Docket No. 50-361, " Failure of Turbine Stop Valve 2VV-2200E To Close Fully," San Onofre Nuclear Generating Station, Unit 2, November 29, 1982.
U.S. Nuclear Regulatory Commission, NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," July 1981.
-- , NUREG/CR-1884, " Observations and Comments on the Turbine Failure at Yankee Atomic Electric Company, Rowe, Massachusetts," March 1981.
Woodfin, R. L., " Full-Scale Turbine Missile Concrete Impact Experiments," pre-pared by Sandia National Laboratories under Electric Power Research Institute Research Project 399-1, Final Report NP-2745, February 1983.
Table U.1 Turbine system reliability criteria I
Probability, yr 1 Favorably Unfavorably oriented oriented
[
i turbine turbine Required licensee action (A) P < 10 4 Pi < 10 5 This is the general, minimum re-liability requirement for loading i
the turbine and bringing the system on line.
(B) 10 4 < P < 10 3 10 5 < Pi < 10 4 3
If this condition is reached during operation, the turbine may be kept in service until the next scheduled outage, at which time the licensee is to take action to reduce P1 to meet the appropriate A criterion (above) before returning the turbine to service. '
(C) 10 3 < Pi< 10 2 10 4 < P < 10 3 i
If this condition is reached during operation, the turbine is to be isolated from the steam supply within 60 days, at which time the licensee is to take action to reduce Pi to meet the appropriate A cri-terion (above) before returning the turbine to service.
10 3 < P i If this condition is reached at any (D) 10 2 < P t time during operation, the turbine is to be isolated from the steam supply within 6 days, at which time the licensee is to take action to reduce P1 to meet the appropriate A criterion (above) before returning the turbine to service.
Hope Creek SSER 6 10 Appendix U
Poam ass u s. NuCLEs.n tivLtioav Commission i > EPOar NumeE A (As.gnea py TtOC, ser var Ng,,r eny; E"252 ' BELIO2RAPHIC DATA SHEET NUREG-1048 ut iN TruCriONs Os.rs.E .Evtast Supplement No. 6
- 3. TITLE &ND SueTITLE 3 LE AVE 8 LANK Sa fet Evaluation Report related to the operation of j Hope C ek Generating Station /
[ 4 DATE aEPOa7 COMPLETED
-ONT VEAa l
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/ July 1986
[ 6 DATE EPORT i$$vED MONTH VEAR
, ,Ea Dauis.o Onam2 ATiON = Aq AND Aiu=o ADOatss t, .i.c ,
./ July 1986 f raO;ECTa As==Oax unit Nuuna Division of BWR L' nsing Office of Nuclear ctor Regulation 7 +iN Oa oaANT Nuesta U. S. Nuclear Regula y Commission Washington, D. C. 20 i$ SPONAiNG ORGANi2 ATION NAME AND MA.LIN DRES$flerko lpCo., iis TYPE OF aEPOaT Safety Evaluation Supplement Same as 7. above = *Eaim COv t a ED <,~~~ -i in sur,u ENT A < NOTE l
Pertains to Docket No. 50-354 i2A T-AC1,m . ,s ,
/
i Supplement No. 6 to the Safety Evaluati Report on the application filed by !
Public Service Electric and Gas Company o its own behalf as co-owner and as agent for the other co-owner, the Atla ic ity Electric Company, for a license to operate Hope Creek Generating Stati n ha aeen prepared by the Office of Nuclear Reactor Regulation of the U. . Nucle, Regulatory Commission. The facility is located in Lower Alloway Creek To hip in Salem County, New Jersey.
This supplement reports the status certain i s that had not been resolved at the time of the publication of t Safety Eval tion Report. This supplement supports the issuance of a full-pow license to o ate the Hope Creek Generating Station.
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