ML20054F912

Testimony of Jj Boseman,Rm Crawford,F Hayes,Jj Kreps & Jl Smith Re Implementation of NUREG-0737,Item II.K.3.16 Concerning Reduction of Challenges & Failures of Relief Valves.Schematic of three-stage Pilot Operated Valve Encl

ML20054F912
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	06/14/1982
From:	Boseman J, Crawford R, Hayes F, Kreps J, James Smith LONG ISLAND LIGHTING CO.
To:
Shared Package
ML20054F907	List: ML20054F906 ML20054F912
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.16, TASK-TM ISSUANCES-OL, NUDOCS 8206180193
Download: ML20054F912 (101)

Text

I l e- l UNITED STATES OF AMERICA .'. '"yg3 NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter ) l

) l LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322 (OL)

)

(Shoreham Nuclear Power Station, ) 1 Unit 1) )

TESTIMONY OF JOHN J. BOSEMAN, RAYMOND M. CRAWFORD, FRED HAYES, JOHN J. KREPS, AND JEFFREY L. SMITH FOR THE LONG ISLAND LIGHTING COMPANY ON SUFFOLK COUNTY CONTENTION 28a(vi)

AND SOC CONTENTION 7A(6) --

REDUCTION OF CHALLENGES TO SRV'S PURPOSE In response to NUREG-0737, " Clarification of TMI Action Plan Requirements", Item II.K.3.16, " Reduction of Challenges and Failures of Relief Valves -- Feasibility Study and System Modification," LILCO will reduce the probability of a stuck-open relief value (SORV) event by an order of magnitude. LILCO will not accomplish this reduction merely through procedural techniques, as the contention implies, but will use a modified valve design as well as procedural changes to reduce SORV events. The modified valve design of the Target Rock two-stage valve used at Shoreham provides improved SRV performance and reliability, as distinguished from the historically poor reliability of the Target Rock three-stage valve.

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ATTACHMENTS

1. NUREG-0737 Item II.K.3.16. -

II. Letter SNRC-557 of April 15, 1981 from J. P. Novarro to Harold R. Denton, with attachments.

III. Letter SNRC-579 of May 29, 1981 from J. P. Novarro to Harold R. Denton, with attachments.

IV. BWR Owners' Group Evaluation of NUREG-0737 Item II.K.3.16, Reduction of Challenges and Failures of Relief Valves.

V. Letter of March 31, 1981 from D. B. Walters to Darrell G.

Eisenhut, "BWR Owners' Group Evaluations of NUREG-0737 Requirements II.K.3.16 and II.K.3.18."

VI. Drawings of Target Rock Two-Stage Pilot Actuated Safety / Relief Valve.

Figure A: Schematic of Target Rock Two-Stage, Pilot actuated Safety / Relief Valve (Valve Closeo).

Figure B: Schematic of Internal Details, Target Rock Safety / Relief Valve Two-Stage Horizontal Design (Valve Closed).

Figure C: Target Rock Two-Stage Pilot Operated SRV (Horizontal Design) Valve Schematic (Open Position).

VII. Schematic of Target Rock Three-Stage Pilot Operated Satety/ Relief Valve (Models 7367F/74674) .

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LILCO, June 8, 19o2 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter )

)

LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322 (OL)

)

2, (Shoreham Nuclear Power Station, )

Unit 1) )

TESTIMONY OF JOHN J. BOSEMAN, RAYMOND M. CRAWFORD, FRED HAYES, JOHN J. KREPS, AND JEFFREY L. SMITH FOR THE LONG ISLAND LIGHTING COMPANY ON SUFFOLK COUNTY CONTENTION 28a(vi)

AND SOC CONTENTION 7A(6) --

REDUCTION OF CHALLENGE 5 TO SRV'S t

1. Q. Please state your names and busines. addresses, s

i A. My name is John J. Boseman; my business address is the General Electric Company, 175 Curtner Avenue, San Jose, California.

My name is Raymond M. Crawford; my business address is Science Applications Incorporatea, 1211 West 22nd Street, Oak Brook, Illinois.

My name is Fred Hayes; my business address is the General Electric Company, 175 Curtner Avenue, San 4

Jose, California.

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• My name is John Kreps; my business address is Post Office Box 618, Wading River, New York.

My name is Jeffrey L. Smith; my business address is Post Office Box 618, Wading River, New York.

2. Q. By whom and in what capacity are you employed?

A. (Boseman) I am a Senior Engineer and Technical Leader for General Electric.

(Crawford) I am Vice President or Science Applications Incorporated (SAI), and serve as Operations Manager of SAI's Oak Brook office.

(Hayes) I am a licensing engineer in the BWR eval-uation program, Nuclear Energy Business Operations, for General Electric.

(Kreps) I am a startup and test engineer with NUS Corporation and have been assigned to the Shoreham Nuclear Power Station as an operations assistant since April 1981.

(Smith) I am employed by Long Island Lighting Company as the Manager of Special Projects, Construction and Engineering Department. ,

3. Q. Please state your professional qualifications.

(Boseman) The attached resume summarizes my protes-sional qualifications. My familiarity with the issue of safety / relief valve (ShV) challenges stems

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evaluations, and programs to im; prove-the' reliability of safety / relief valve.' designs for BWR applications.

i (Crawford/ Smith) The attached resumes summarize our '

professional qualifications. Our familiarity with the issue of SRV challenges stems from our involve- - <

ment with the BWR Owners' Group SRV Performance N

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Testing Subgroup.

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(Hayes) The attached resume summarizes my profes-sional qualifications. My familiarity with the ,

issue of SRV challenges stems from my ' support of thk ' ,

BWR Owners' Group Program to respond to NUREG-0737 -

Item II.K.3.16. 1, j (Kreps) The attached resume summarizes my profes- ,. .

sional qualifications. My familiarity with the issue of SRV challenges stems from my coordinaIlon and preparation of-p1'nt a operating. procedures. 4

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4. Q. What are your responsibilities for this testimony?

A. (Boseman/Kreps) That part of the testimony. attribu-table to each of us is labeled.

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(Crawford/ Hayes / Smith) We developed the remainaer or ,

this testimony jointly, and are>2qually responsible for producing each part. . l c

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5. Q. Are you familiar with Suffolk County Contention 28a(vi) and Shoreham Opponents Coalition (SOC)-

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, Contention 7A(6)?

A. Yes.

6. ~Q. What issue is presented in those contentions?

A.' Suffolk County and SOC contend that, regarding NUREG-0737 Item II.K.3.16, LILCO hopes to reduce challenges to SRV's at Shoreham by procedural tech-niques rather than system modifications. The County and SOC also question the reliability of the SkV's chosen for Shoreham.

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7. Q. What is the function of SRV's at Shoreham?

i A. LILCO bas, installed eleven Tacget Rock two-stage t

s 6R10 type SRV'E, Model 7567F, at Shoreham. The pri-mary design function of these SRV's is to relieve excess pressure by releasing steam from the reactor vessel to the suppression pool. The SRV's perform this design function in two ways: (1) by providing i

automatic overpressure protection and (2), as part

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i of the automatic depressurization system (ADS), by depressurizip.g the vessel sufficiently to allow low-pressure emergency core cooling systems to oper-I- ate. In the unlikely event of a small-break loss of

cooling accident (LOCA) coupled with the a

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unavagla'bility of the 'hlgh' pressure cooling systems, the.6RV's would be actuated in an ADS mode. If a large-break LOCA were to occur, the reactor vessel would depressurize due.to/the break flow and the .

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SRV's would not be required to actuate.

8. Q. What issue regarding SRV's is identified in NUREG-0737 Item II.K.3.16?

A. NUREG-0737, " Clarification of TMI Action Plan Requirements,"-Item II.K.3.16, " Reduction of Challenges and Failures of Relief Valves --

Feasibility Study.and System Modification,"

(Attachment I) identifies _"the record of relief-valve failures to close for all boiling-water i

reactors (BWR's) in the past 3 years of plant opera-i tion [as] app'roy.imately 30 in 73 reactor-years (0.41 i

failures per reactor-year). This has demonstrated I

tha't the failure of a relief valve to close is the most likely cause of a small-break loss-of-coolant-accident (LOCA)." The intent of II.K.3.16

.is to. reduce the frequency of stuck-open

_ relief-valve (SORV) events.

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9. Q. Have the consequences of SORV events been evaluated for Shoreham?

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l A. Yes. The BWR Owners' Group evaluation of NUREG-0737 Item II.K.3.44, " Adequate Core Cooling for Transients with a Single Failure", establishes that adequate core cooling exists for the limiting core transient of a loss of feedwater flow coupleo with the worst single failure and a stuck-open relief valve. The letter of April 15, 1981 from J. P.

Novarro to Harold R. Denton, SNRC-557 (Attachment II), transmitted this evaluation to the NRC. The conclusions of the evaluation are applicable to Shoreham, as noted in the May 29, 1981 letter from J. P. Novarro to Harold R. Denton, SNRC-579 (Attachment III). Additionally, as is stated in the Shoreham FSAR at II.K.3.16-2, "it should be noted that adequate core cooling is maintained in a BWR following an SORV event even under degraded conoi-tions. Reduction of the frequency of SORV events is not, therefore, of great concern from the standpoint of assuring adequate core cooling."

10. Q. What does NUREG-0737 Item II.K.3.16 state with respect to reducing SORV events?

A. Item II.K.3.16 states that "[t]he high f ailure rate

[of SRV's] is the result of a high relief-valve challenge rate and a relatively high f ailure rate per challenge (0.16 f ailures per challenge) ." The

II.K.3.16 item suggests several potential design and operational modifications for reducing both the fre-quency of challenges to and the rate of failure or SRV's. It further requires that an investigation be conducted to examine the feasibility and contraindi-cations of implementing these and other methods for reducing challenges ano f ailures of the SRV's.

Although II.K.3.16 states that it is "[c]hallenges to the relief valves [that] should be reduced sub-stantially (by an order of magnitude)," it is clearly intended that utilities should reduce tail-ures of relief valves to close. Reduction of chal-1enges is merely one method that can be useo to reduce SORV events.

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11. Q. How has LILCO responded to NUREG-0737 Item II.K.3.16?

A. LILCO has responded to II.K.3.16 in three ways: (1) by participating in the BWR Owners' Group evaluation of the methods available to reduce challenges and failures of relief valves, (2) by using the improved two-stage design of the Target Rock safety /reliet valve, and (3) by implementing both procedural changes and design improvements that reauce the num-ber of challenges to and failures of Shoreham safety / relief valves.

12. Q. Taking the first item, how did the BWR Owners' Group evaluate the provisions in Item II.K.3.16 to reduce failures of relief valves to close?

A. The BWR Owners' Group performed a study in response to Item II.K.3.16. The results of that study were compiled in "BWR Owners' Group Evaluation ot NUREG-0737 Item II.K.3.16, Reduction of Challenges and Failures of Relief Valves" (Attachment IV), sub-mitted to the NRC by a letter dated March 31, 1981 from D. B. Waters, Chairman of the BWR Owners' Group, to Darrell G. Eisenhut, titled " bha Owners' Group Evaluations of NUREG-0737 Requirements II.K.3.16 and II.K.3.18" (Attachment V).

13. Q. What is the scope of this evaluation?

The evaluation noted that the intent of NUREG-0737

, Item II.K.3.16 was to reduce the number of stuck-open relief valve (SORV) events, not just to reduce the number of SRV challenges (Attachment IV at 1). Using a BWh 4 plant with Target Rock three-stage valves as a benchmark plant, the BWR Owners' Group evaluated the suggested modifications in II.K.3.16 to reouce failures and challenges, plus

' additional modifications suggested by the Owners' Group, to determine the relative benefits each of those methods offered in reducing challenges and failures. -

To estimate the impact of various design improve-ments and proceoural changes on the frequency of SORV events, data from operating BWk plants spanning 120 reactor-years of operation and approximately 1400 reactor scram events (including 720 SRV chal-lenge events) were investigated by the Owners' Group. By analyzing these data an estimate was made of the percentage by which various transient event frequencies would be reduced if a particular modifi-cation were to be implemented. The total number of valve openings for a given transient was determined analytically. The assumed frequency of transient events is consistent with the transient frequencies observed in operating plants. With these assump-t*ans, the relative contribution of each candidate madification for reauction of SRV challenges could be determined.

To determine relative SRV reliability, the SORV events in operating plants were reviewed, and the failure modes associated with the Target Rock three-stage valve were tabulated. Then, based on a study of the two-stage valve design, the Owners' Group assessed the failure modes eliminated by the two-stage design. Consideration was also given to any new f ailure modes that might develop in going from the three-stage to the two-stage design.

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14. Q. What conclusions were drawn from the Owners' Group study?

A. The study concluded that a reduction in SORV events can be achieved by a combination of (1) design changes to SRV's that improve their performance under transient relief conditions, (2) procedural changes to reduce challenges, and (3) design changes to reduce challenges. Consideration of feasible methods to reduce SRV challenges, by themselves, does not accomplish a significant reduction in SORV events, but a significant reduction is obtainea through design changes to SRV's to improve perform-ance in the event of a challenge. Consequently, both design and procedural moaifications were pre-sented in the Owners' Group study for the partici-pating utilities' consideration. The study conclu-ded that "the use of selected modifications from a list of candidates can produce a factor-of-ten reduction in stuck-open relief-valve events for BWh 4 plants" (Attachment IV at 31).

15. Q. What methods has LILCO applied at Shoreham to reduce SORV events?

A. Three methods for reducing SORV events have been applied at Shoreham: (1) the use of the Target Rock two-stage safety / relief valve, (2) the use of the

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Emergency Procedure Guidelines that provide for manual implementation of a low-low set relief, and (3) a lowering of reclosure setpoint on Shoreham SRV's. The first two of these methods were eval-uated in the BWR Owners' Group S.tudy; the combina-tion of the two was shown to reduce the failure of SORV events by an order of magnitude (Attachment IV at Table 5.1, Line E.)

16. Q. Were other modifications identified in the owners' Group study for reducing SRV challenges or SORV events?

A. Yes. However, many of the identified modifications provide little benefit (typically less than 5%) in reducing SRV challenges or SORV events. Some modi-fications, such as changing the pressure and water level setpoints for MSIV closure, would have an adverse impact on plant safety. Those methoos selected for Shoreham provide significant reouctions of SRV challenges and SORV events without compro-mising plant safety.

17. Q. What is the merit in the first modification mentioned--using Target Rock two-stage relief valves instead of three-stage?

~12-A. The Target Rock two-stage SRV design used at Shoreham was developed to provide improved SRV per-formance anc reliability as a result of the his-torically poor reliability of the Target Rock three-stage SRV design.

18. Q. Please describe the Target Rock two-stage SRV.

A. (Boseman) The Target Rock two,-stage pilot-operated SRV, Model #7567, (Attachment VI, Figures A, B, anc C) consists of two principal assemblies: (1) a single-stage pilot valve with an externally attached electro-pneumatic operator assembly, and (2) the main stage assembly. These two assemblies are directly coupled to provice a unitized dual-function SRV assembly that permits the valve to be incepen-dently opened either as a sarety valve or as a relief valve. When operated as a safety valve, tne valve actuates when the static inlet-system pressure reaches the setpoint spring permitting the pilot valve to open and, in turn, opens the mainstage disc. When operating as a relief valve, it actuates when an external electrical power signal is applied to the solenoid valve, permitting pneumatic pressurization of the operator, opening of the pilot valve, and in turn, opening of the mainstage disc.

19. Q. How does the two-stage design differ f rom the three-stage?

A. (Boseman) The Target Rock two-stage SRV design (Attacnment VI, Figure A) eliminated the second stage used in the three stage SRV design ( Attachment VII). This improvement (1) eliminates the major cause for spurious plant blowdowns and stuck-open relief valve events, and (2) simplifies the sequence of events required to open the main disc located in i

i the main stage assembly (Attachment VI, Figure A) .

20. Q. What other modifications have been made to the two-stage design to increase its reliability and improve its performance?

A. (Boseman) Design improvements made and incorporated into the Target Rock two-stage SRV design used at Shoreham are as follows:

(1) To preclude the SRV from cycling, an abnormal condition where the valve continuously opens and closes following blowdown, properly-sized bellows, pilot seat, stabilizer seat, and orifice diameters were used to assure that the SRV will properly blow down (re-set) with Shoreham-specitic discharge-line backpressures.

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(2) To minimize the potential for pilot valve leakage and sticking, thus extending the useful life of the valve betore maintenance is required, a 72 degree involuted pilot valve seat angle was used.

(3) To improve the pressure integrity of the valve body, forged material in lieu or cast material was used to eliminate the potential for hot tears, unacceptable material surface discontinuities, and cracks in a pressure-retaining member.

(4) To preclude accelerateo degradation of the seal, Viton 0-ring seal material was used in the electro-pneumatic valve separater in lieu of the previously used silicone rubber mate-1 rial.

J (5) To eliminate leakage between the body and 4

pilot valve, a new seal design has been incor-1 porated.

21. Q. Is the Target Rock two-stage design presently used in operating BWR's?

A. (Boseman) Yes. The two-stage design has been in use at the Browns Ferry 1, 2 and 3, Brunswick 1 and 2, Cooper, Hatch 1 and 2, Fitzpatrick, Millstone, and Pilgrim units.

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4 Shoreham uses the same generic SRV design as these plants. The only design dif ferences are the blow-down requirements for compatibility with plant-specific discharge-line backpressures. The in-service operating history of the Target Rock two-stage design indicates improved valve reliabi-lity over the Target Rock three-stage design.

22. Q. Turning to the second method for reducing the tail-ure rate of SRV's to close, what benefit is deriveo 4

from the use of low-low set relief at Shoreham?

I A. (Kreps) A significant reduction in the number of challenges to the SRV's is possible through the implementation of manual low-low set relief. For l

this mode of operation, a selectea valve is manually held open beyond the reclosure setpoint. The tech-nique enables the reactor to sufficiently depressurize through the removal of storea heat such that subsequent valve actuations are limitea to one valve for the removal of decay heat. Without this operational mode, insufficient stored heat would be

, removed from the vessel initially to prevent subse-j quent actuations of multiple SRV's. Consequently, the implementation of this operating mode will reduce the total number of SRV challenges for most transient events.

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The Shoreham procedures, such as emergency shutdown, level control, and cooldown, implement this mode of operation, as specified in the emergency procedures guidelines by reducing reactor pressure sufficiently to control the pressure on the turbine bypass valves, thus minimizing SRV cycling.

23. Q. How does the lower reclosure setpoint for the valves at Shoreham reduce challenges?

A. A recent design improvement in the Target Rock two-stage valve provides for a significantly lower reclosure setpoint for each valve. The reclosure setpoint will now be about 70 psi below the opening setpoint. This design change provides an automatic means for removing more stored heat with the initial SRV actuation to prevent subsequent actuation of multiple SRV's. This change has been applied to Shoreham.

24. Q. What is the resulting benefit to shoreham from the use of these three modifications?

A. The combination of these items will achieve a ten-fold reduction in SORV event frequency when com-pared to the benchmark BWR 4. This is documentea in Table 5.1, Item E of the Owners' Group stuoy (Attachment IV).

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25. Q. Please summarize your testimony.

A. LILCO has responaed to NUREG-0737 (1) by partici-pating in the BWR Owner's Group evaluation of the methods available to reduce challenges ano failures, (2) by using an improved valve design, ano (3) by implementing proceoural changes to reduce challenges to and failures of SRV's at Shoreham. LILCO will not accomplish a reduction in SORV events by proce-dural techniques alone, as the comtention implies.

The combined valve design improvements and proce-dural changes implemented at Shoreham will reduce SORV events by a factor of ten.

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PROFESSIONAL QUALIFICATIONS John J. Boseman Senior Engineer -- Technical Leader General Electric Company My name is John J. Boseman. My business address is 175 Curtner Avenue, San Jose, California. I am a Senior Engineer and Technical Leader for the General Electric Company (GE). In that position I am responsible for the design, development, qualification tests, programs, and related technological growth of equipment including all necessary technical assistance to

. support installation, testing, inspection and maintenance of the equ'ipment. I formulated, directed, and participated in the de s ig n , tests, evaluations and programs to improve the reliability of safety / relief valve designs for BWR J

applications.

I received the Bachelor of Science in Marine Engineering in 1964 from the United States Merchant Marine

Academy.

From 196'4 to 1965, I was Assistant Engineer with Delta Steamship Lines, Inc. in New Orleans, Louisiana. In that I capacity I tested, operated, maintained and repaired steam and l diesel power plant systems and equipment, including reviewing design proposals and major shipyard overhauls and repairs.

i During this period, I conceived and demonstrated an emergency l

technique to repack the stern gland while underway.

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During 1966 I was Assistant Engineer with The Boeing Company, New Orleans, Louisiana. In that capacity, I performed environmental and simulated testing of fluid power components m

and sub-systems applicable to the Saturn IV & V Booster System.

During this period, I established the cause for LOX and liftcheck valve failures and recommended corrective action to preclude recurrence.

From 1966 to 1968 I was Associate Engineer with Lockheed Missile & Space Co., Sunnyvale, California. There I analyzed, designed, developed, and tested the manuf acture of advanced electro-mechanical microwave antenna and antenna systems for Satellite and Polarts/ Poseidon Missile applications.

In 1968, I joined GE as a Project Engineer. In this position I designed, developed, tested, evaluated, and provided the field support for various types of Naval nuclear power plant fluid components. I then became Product Engineer. In that capacity I performed and provided technical direction for the design, applications, qualifications, development, installation and maintenance of certain equipment, including valves. I also participated in the Navy's 1970 Valve Design Review Task Force. I became a Senior Engineer in 1977.

I have authored the following articles:

(1) ASME 80-02/PVP Operability Assurance Testing of ASME Code, Class 1, Safety / Relief Valves.

(2) Evaluation of 3 Thermally Shock Tested 1/2-Inch Globe valves (MEDF # 54) - U.S.

Navy Document (Restricted).

(3) U. S. Navy Nuclear Valve Design Manual (VDM-71) (Classified).

(4) Plant Equipment Design Memorandum No.

126 Sealing hechanism Factors.

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PROFESSIONAL QUALIFICATIONS l DR. RAYMOND R. CRAWFORD Vice President and Operations Manager Science Applications Incorporated My name is Raymond Crawford. My business address is 1211 West 22nd Street, Oak Brook, Illinois. I am Vice President and Operations Manager of Science Applications Incorporated (SAI),

Oak Brook office. Sina:e joining SAI in 1981 I have been serving two major electric generating utilities and a BWR

! Owners Group as a technical consultant for test programs to evaluate safety and relief valve performance under severe fluid conditions. This includes development of an analytical model of valve performance including RELAP-5 analysis and also including discrete wave analysis. In addition, I have provided an independent assessment of the BWR scram system piping.

I received the Bachelor of Science in Chemical Engineering in 1958 and the Master of Science in Chemical Engineering in

, 1960, both from Wayne State University. I received the Ph.D.

in Engineering from the University of California in 1969.

From 1960 to 1963 I taught undergraduate and graduate courses in chemical engineering, nuclear engineering, process control and high vacuum technology at Wayne State University.

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. . 1 In 1962, I worked as a Technical Consultant for Atomic Power Development Associates, Inc. There I directed an experimental program on the reactions of hydrogen with sodium, designed a high vacuum system for the study, and performed the experiment and analysis. I also cerved as a Summer Fellow in 1962 and 1963 with the NASA-Lewis Research Center in Cleveland Ohio. I was an Assistant Professor at San Fernando Valley State College from 1963 to 1966, where I was responsible for the development of the Nuclear Engineering Laboratory and curriculum. I then became an NSF Science Faculty Fellow from 1966 to 1968.

After receiving my Ph.D., I joined the technical staff of Atomics International, where I was responsible for various aspects of Systems Analysis and Safety for the liquid metal fast breeder reactor (LMFBR), including analysis of various failure modes, event trees, and fault trees in the reactor core for various core design features. I continued my work on the LMFBR at the Argonne National Laboratory from 1971 to 1974.

In 1974 I joined Sargent & Lundy Engineers as Associate and Assistant Head Engineer. There I was responsible for and directed the work of a group of engineers engaged in direct thermodynamic and nuclear analyses in support of reactor safety systems. I worked with the clients, manufacturers' representatives and appropriate regulatory bodies to resolve potential safety problems and develop new and improved methods and techniques of analysis. I also coordinated the safety

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analysis work of the Nuclear Safeguards & Licensing Division l with the work of the other departments.

I am a Registered Professional Engineer in the State of i Illinois. In addition, I am a member of the American Chemical Society, the American Institute of Chemical Engineers, the (merican Nuclear Society, the Illinois Society of Professional Engineers, the National Society of Professional Engineers, the New York Academy of Sciences, and the Western Society of Eng ineer s . I served as Chairman of the American Nuclear Society, Chicago Section from 1978 to 1979, and am currently Chairman of the American Nuclear Society National Standards Writing Group. I have held numerous committee assignments in the National and Chicago Sections of the American Nuclear Society. I am a recipient of the Carr Award for Teaching Excellence, the ACEUG Achievement Award, the Engineering Society Faculty Advisor Award, and the Tau Beta Pi National Essay Award, and am listed in Who's Who in American Education, Who's Who in the Midwest, and Who's Who in Technology.

I have authored the following publications:

j Crawford, R. M., " Computer Simulation of a Batch Chemical

Reactor." ASEE Pacific Coast Meeting, Monterey, California.

(1965).

Crawford, R. M., and J. F. Paul. " Application of the Analog Computer in Process Simulation." ACEUG Applications 1 (1):

1-33 (1965).

l Crawford, R. M., " Computers in the Classroom." ASEE Annual Meeting. Lansing, Michigan. (1967).

Paulson, W. A., and R. M. Crawford. " Post Shutdown Cooling Requirements of the Tungsten-Water-Moderated Nuclear Rocket."

NASS TM X-1568 (1967).

1 Crawford, R. M., and W. E. Kastenberg. "On the Application of Semigroup Theory to Problems in Space Dependent Reactor Dynamics." Trans. Amer. Nucl. Soc. 12 (1): 241-242 (1969).

Crawford, R. M., and W. E. Kastenberg. " Analysis of Space-Time Reactor Systems Using the Method of Semigroups." Trans. Amer.

Nucl. Soc. 12 (2): 707-708 (1969).

Kastenberg, W. E., and R. M. Crawford. " Determination of Stability Domains in Space Dependent Reactor Dynamics." Nuc.

Sci. Engr. 37 (2): 311-316 (1969).

Crawford, R. M., and W. E. Kastenberg. " Stability Analysis of Distributed Parameter Systems in a Banach Space."

International Journal of Control 12 (6): 929-943 (1970).

Crawford, R. M., and W. E. Kastenberg. "On Semig roup Theory and Its Application to Space-Time Nuclear Reactor Dynamics."

Nucl. Sci. Engr. 47: 238-241 (1972).

Crawford, R. M., W. W. Marr, and A. Padilla Jr., " Coolant Temperature Downstream of a Subassembly Blockage." Trans.

Amer. Nucl. Soc. 15 (1): 352-353 (1972).

Marr, W. W., and R. M. Crawford. " Porous, Heat-Generating Blockag e in a Fuel Subassembly Blockage." Trans. Amer. Nucl.

Soc. 15 (1):350-351 (1972).

Marr, W. W., and R. M. Crawford. " Thermal-Hydraulic Analysis of Multipin In-Pile Experiments." Trans. Amer. Nucl. Soc. 15 (2): 880-881.

Padilla Jr., A., and R. F. Crawford. " Molten Fuel Relocation in an Over-Enriched Fuel Pin." Trans. Amer. Nucl. Soc. 15 (1):

342-343 (1972).

Padilla Jr., A., T. J. Marcinak, and R. M. Crawford. " Pressure Pulse on Subassembly Wall Due to a Local Molten Fuel-Coolant Interaction." Trans. Amer. Nucl. Soc. 15 (2): 809 (1972).

Wang, P. Y., and R. M. Crawford. " Thermal Stress Analysis of a

! Subassembly Duct Wall During Several Thermal Loading." Trans.

j Amer. Nucl. Soc. 15 (2): 853-854 (1972).

Marr, W. W., and R. M. Crawford. " Effects of Local Coolant

Flow Diversion Out of a 217-Pin FTR-Type Subassembly." Trans.

uMner . Nucl . Soc. 18 (1)

207 (1974).

Krieg, R., and R. M. Crawford. " Deformations in a 19-Pin l Bundle Due to a Sudden Fluid Expansion." Amerc. Nucl. Soc.

Fast Reactor Safety Meeting. Beverly Hills, California, April 1974.

Marr, W. W., and R. M. Crawford. " Local Coolant Flow Diversion from a Failed LMFBR Subassembly." Trans. Amer. Nucl. Soc. 17 (1): (1974).

Marr, W. W., and R. M. Crawford. " Thermal Hydraulic Analysis of a Partially Crushed FTR-Type Fuel Subassembly." Trans.

Amer. Nucl. Soc. 18 (1): 204 (1974).

Misra, B., and R. M. Crawford. " Pressure Decay and Fission Gas Release Rates for LMFBR Fuel Pins." Trans. Amer. Nucl. Soc. 18 (1); 204 (1974).

Wang, P. Y., and R. M. Crawford. " Failure Prediction of Cladding Under Internal Pressures at Elevated Temperatures."

Trans. Amer. Nucl. Soc. 17 (1): 397 (1974).

Wang, P. Y., and R. M. Crawford. "Probabilistic Analysis of Reactor Duct Response." ASCE National Structural Engineering Meeting. Cincinnati, Ohio, April 1974.

Crawford, R. M., W. W. Marr, A Padilla Jr., and P. Y. Wang.

" Studies on LMFBR Subassembly Boundry Integrity." ANL-75-27 (May 1975).

Crawford, R. M., et. al. " Mark II Containment Dynamic Forcing Function Information Report." NEDO-21061 (September 1975).

Crawford, R. M., W. W. Marr, A. Padilla Jr., and P. W. Wang.

"The Safety Consequences of Local Initiating svents in an LMFBR." ANL-65-73 (December 1975).

Hammersley, R. J., and R. M. Crawford. " Rapid Depressurization of Drywell." Trans. Amer. Nucl. Soc. 21 (1): (1975).

Krieg, R., P. Y. Wang , and R. M. Crawford. " Maximum Uniform Elongation of Rods with Small Variations in Cross-Section Area and Material Properties." Trans. of the 3rd International Conference on Structural Mechanics in Reactor Technology, Vol. 1, Part C (September 1975).

Marr, W. W., P. Y. Wang, B. Misra, A. Padilla Jr., and R. M.

Crawford. " Analytical Investigation of Certain Aspects of LMFBR Subassembly-Failure Propagation. " ANS-76-19 (February 1976).

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a

• o PROFESSIONAL QUALIFICATIONS Johnny J. Kreps Startup and Test Engineer NUS Corporation My name is John Kreps. My business address is Shoreham Nuclear Pow 6r Station, Post Office Box 618, Wading River, New York. I am employed by NUS Corporation as a startup and test engineer and have been assigned to the Shoreham Nuclear Power Station as an operations assistant since April 1981. In this capacity I have been responsible for the coordination and preparation of plant operating procedures, emergency procedures, surveillance procedures and alarm responses and for.

coordinating modifications resulting from human factors evaluations, INPO audits, TMI action plan and NRC inspections.

I have recently accepted a position with EDS Nuclear in Texas but will continue to be available to Shoreham for consulting purposes.

Following six years' service in the United States Navy Nuclear Power Program (1968-1974), I joined Stone & Webster Engineering Corporation (1974-1975) as an engineering aide in the operations service division where I performed design review, operating and maintenance procedure development, ISI scheduling and feasibility studies. From May 1975 until July 1980 I was employed by Arkansas Power & Light Company in plant

. . operations at their two nuclear units near Russellville, Arkansas. In this capacity, I was responsible for the safe and efficient operation of the plant and supervision / coordination of an operating shift. During this time period I participated in all phases of startup for Unit #2 from preoperational testing through hot functional testing, power ascension testing, and commercial operation. I was licensed as a Reactor Operator. From July 1980 until April 1981, I was employed as a startup and test engineer by Gilbert Associates at the Calloway Nuclear Plant with responsibilities including operations support in the development of station operating procedures, including system procedures, emergency prvcedures, alarm responses and surveillance testing.

My educational background includes one year of en-gineering studies at the University of Oklahoma, Nuclear Power Trainina from the U.S. Navy and various courses associated with obtaining an NRC reactor operator's license, including plant design and simulator certification.

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PROFESSIONAL QUALIFICATIONS Jeffrey L. Smith Manager.of Special Projects Construction and Engineering Department Long Island Lighting Company My name is Jeffrey Smith. My business address is Post Office Box 618, Wading River, New York. I am employed by the Long Island Lighting Company (LILCO) as Manager of Special Projects, Construction and Engineering Department and have held this position since November 1981. In this capacity, I have overall responsibility for technical licensing matters associated with Shoreham, including coordination of the ASME N-5 certification program, the large-bore and small-bore piping "as-built" program, the Maintainability Task Force, the small-bore piping and instrumentation design groups, and the Station Modification Support Program.

I received a Bachelor of Science degree in Mechanical Engineering in 1967 from Clarkson College of Technology, and a Master of Science degree in Nuclear Engineering in 1978 from Polytechnic Institute of New York. I completed the General Electric Boiling Water Reactor Simulator Program in December, 1979 and am certified as a Senior Reactor Operator. I have also completed numerous industry seminars and training programs related to BWR technology, construction, and operation.

My professional experience began with my employment by LILCO in 1966 as Assistant Engineer at the Port Jefferson Power Station. In this position I was responsible for various operations and maintenance administrative activities, and for the design and installation of numerous modifications at the station.

After two years of service in the United States Army as Mechanical Engineering Assistant at Munitions Command Headquarters, Picatinny Arsenal, Dover, New Jersey, I returned to LILCO in 1969 as Associate Engineer and Plant Engineer at the Northport Power Station. In 1972 I became Operations / Controls Engineer at Northport. In that capacity I was responsible for the direction of all operations, instrumentation, and controls testing and water chemistry functions at the three-400 MWe units comprising Northport Station. My duties included the startup and initial operation of Northport Unit No. 3.

2 In 1974 I was assigned to Hicksville Operations Center in the Electric Production Department as the Staff Engineer responsible for coordination and liaison with the Jamesport Nuclear Project on all matters dealing with operations staffing, training, Service Building layout and operational reviews. In that capacity, I was also responsible for the direction of turbine and boiler capability and equipment-performance testing of cl1 electric generating stations on the LILCO system.

In 1975 I was promoted to Manager of Operational Quality Assurance. After training at Rochester Gas and Electric Corporation's-Quality Assurance Department where I worked on the Station Quality Control staff, I returned to LILCO as Manager. In that capacity, I was responsible for establishing and assuring the overall implementation of the Operational Quality Assurance Program, defining the content and changes to the Operational Quality Assurance Manual, and evaluating the manner in which quality reviews affecting activities both onsite and offsite are conducted. During this period, I directed engineering personnel in the development of an Operational Quality Assurance Program and Procedures Manual and participated in audits of Shoreham. I also conceptualized and developed the Nuclear Operations Corporate Policy Manual that is presently used to define corporate organizational interfaces and responsibilities for the support and operation of Shoreham.

Prior to assuming my current position, I was assigned in 1979 as Regulatory Supervisor reporting to the Manager, Nuclear Operations Support Division, with overall responsibility for the management and coordination of nuclear regulatory matters that are under the jurisdiction of the Vice President-Nuclear.

These regulatory matters included licensing and compliance activities associated with maintaining a full-power operating license, Nuclear Review Board affairs, special compliance projects and programs, and company commitments to Federal, state and local agencies. I directed and participated in the

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I am a member of the American Nuclear Society"and;American i-  ;.

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g ATTACHMENT I to LILCO Testimony on SC 28a(vi) and SOC 7A(6) :

NUREG-0737 Item II.K.3.16

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) II.K.3.16 REDUCTION OF CHALLENGES AND FAILURES OF RELIEF VALVES--FEASIB

STUDY AND SYSTEM MODIFICATION

- 1 Position l The record of relief-valve failures to close for all boiling-water reactors l

(BWRs) in the past 3 years of plant operation is approximately 30 in 73 reactor-years (0.41 failures per reactor year). This has demonstrated that the failure of a relief valve to close would be the most likely cause of a small-break j loss-of-coolant accident (LOCA). The high failure rate is the result of a high relief-valve challenge rate and a relatively high failure rate per challenge (0.16 failures per challenge). Typically, five valves are challenged in each event. This results in an equivalent failure rate per challenge of j

O.03. The challenge and failure rates can be reduced in the following ways:

(1) Additional anticipatory scram on loss of feedwater, 1

(2) Revised relief-valve actuation setpoints, (3) Increased emergency core cooling (ECC) flow, .

! (4) Lower cperating pressures, (5) Earlier initiation of ECC systems

]

t (6) Heat removal through emergency condensers, l

i (7) Offset valve setpoints to open fewer valves per challenge, (8) Installation of additional relief vales with a block- or isolation-valve feature to eliminate opening of the safety / relief valves (SRVs), consistent with the ASME Code,

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(9) ~ Increasing the high's' team line flow setpoint for main steam line isolation

' valve (MSIV) closure, (10) Lowering the pressure setpoint for MSIV closure, (11) Reducing the testing frequency of the MSIVs, (12) More-stringent valve leakage criteria, and i

(13) Early removal of leakinc valves.

An investigation of the feasibility and contraindications of reducing challenges to the relief valves by use of the aforementioned methods should be Thoseconducted.

changes i

Other methods should also be included in the feasibility study.

which are shown to reduce relief-valve challenges without compromising the performance of the relief valves or other systems should be implemented.

Challenges to the relief valves should be reduced substantially (by an order of magnitude). ,

i II.K.3.16 t _3_ 156 - - _ - - _ .

Changes to Previous Requirements and Guidance The schedule for plant modifications has been changed to allow time for staff review of evaluation and purchase of required hardware.

Clarification Failure of the power-operated relief valve (PORV) to reclose during the TMI-2 accident resulted in damage to the reactor core. As a consequence, relief valves in all plants, including BWRs, are being examined with a view toward their possible role in a small-break LOCA.

The safety / elief valves (SRV) are dual-function pilot-operated relief valves that use a , 'ing-actuated pilot for the safety function and an external air-diaphragr. ctuated pilot for the relief function.

The operating history of the SRV has been poor. A new design is used in some plants but the operational history is too brief to evaluate the effectiveness of the new design. Another way of improving the performance of the valves is to reduce the number of challenges to the valves. This may be done by the -

methods described above or by other means. The feasibility and contraindica-tions of reducing the number of challenges to the valves by the various methods should be studied. Those changes which are shown to decrease the number of challenges without compromising the performance of the valves or other systems

- should be implemented.

The failure of an SRV to reclose will be the most probable cause of a small-break LOCA. Based on the above guidance and clarification, results of a d2 tailed evaluation should be submitted to the staff. The licensee shall document the proposed system changes for staff approval before implementation.

Applicability This requirement applies to all operating BWRs and BWR operating license cpplicants.

i Implementation Results of the evaluation shall be submitted by April 1, 1981 for staff review.

The actual modification shall be accomplished during the next scheduled refueling l

outage folltwing staff approval or no later than 1 year following staff approval.

Modification to be implemented should be documented at the time of implementation.

Type of Review _

A preimplementation review will be performed.

Documentation Required l By April 1, 1981, licensees must submit the results of the feasibility study for reducing SRV challenges and F W ase any necessary modif.ications for reducing SRV challenges.

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LILCO, June 14, 1982 j

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i' ATTACHMENT II to LILCO Testimony on SC 28a(vi) i and SOC 7 A(6): -

j Letter SNRC-557, of April 15, 1981 from J. P. Novarro to Harold R. Denton,

! with attachments A -

l LONG ISLAND LIGH~s ING COM PANY fyu:.py SHOREHAM NUCLEAR POWER STATION

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• WAOING RIVER. N.Y.11792 April 15, 1901 SNRC-557 Mr. Harold R. Senton, Directer Office of Nuci3ar Reactor Regulation U. S. Nuclear tegulatory Commission .

Washington, D.C. 20555 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322

Dear Mr. Denton:

Forwarded herein are fifteen (15) copies of our positiens related to Post-TMI Requirements outlined in NUREG-0737. These require-monts were transmitted via le:ter dated October 31, 1980 frca Mr.*Darrell G. Eisenhut. Where applicable, our responses to certain NUREG-0578 items (SNRC-503, dated August 29, 1980) are superceded by the respective responses contained herein.

The positions included in t.his submittal are listed in Isttachment

1. The present schedule for submittal of the remaining items is as shown in Attachment 2. This information will be contained in a future hmendment to the FSAR.

Please note that regarding our response to NUREG-0737, item II.K.3.27, Ccanon Water Level Reference, we are in receipt of and presently assessing your letter from Mr. D. Eisenhut co Mr. D. B.

Waters, Chairman, BWR Owners Group, dated April 6, 1981.

Very.truly yours, e

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J{ L"? ~tv5 YO J. P. Novarro ,

Q Project Manager

,Shoreham Nuclear Power Station RWG:mp cc: J. Higgins L FG-0080

1 SNPS-1 FSAR II.K.3.44 Evaluation of Anticipated Transients with Sinole Failure to Verifv No Puel Failure NRC Position For anticipated transients combined with the worst single failure and assuming proper operator actions, licensees should demonstrate that the core remains covered or provide analysis to show that no significant fuel damage results from core uncovery.

Transients which result in a stuck-open relief valve should be included in this category.

BWR Owners' Group Discussion Analyses have been performed for the worst anticipated transient of those identified in R.G. 1.70 (loss of feedwater event) with the worst single active failure (loss of HPCI) which demonstrate that the reactor core remains covered with water until stable conditions are achieved. Analyses have also been performed for further degraded conditions involving a stuck-open relief valve in addition to the worst transient and single failure. The results of these analyses show that, with proper operator action, the core remains covered. For a complete discussion of the Owners

• Group response, refer to Attachment 1, "BWR Owners' Group Evaluation of NUREG-0737 Item II.K.3.44."

BWR Owners' Grouc Imolementation Criteria No implementation criteria are applicable for this item.

LILCO Position

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LILCO endorses the' result of analyses described in Attachment 1 demonstrating that the core remains covered for the worst anticipated transient (lose of feedwater) with the worst single active failure (loss of HPCI) . LILCO also endorses the results of additional analyses demonstrating the capability to keep the core covered for the conditions described above in combination '

with a stuck-open relief valve. For supporting documentation, refer to Attachment 1.

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j II.K.3.44-1

ATTACHMENT 1 i

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l BWR OWNERS' GROUP EVALUATION OF l

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! NUREG-0737 ITEM II.K.3.44 i

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] ADEQUATE CORE COOLING FOR TRANSIENTS WITH

} A SINGLE FAILURE 1

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1 Table of Contents i

fag Summary -

I. Introduction 1 II. Criteria, Scope and Assumptions 2 .

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! III. Discussion 3 IV. Conclusion 1LO Y. References 11 l Appendix A - Participating' Utilities 12 i

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Summary Analyses of the worst anticipated transient (loss of feedwater event) with the worst single failure (loss of a high pressure inventory makeup or hea,t removal system) were performed to demonstrate adequate core cooling capability. It is shown that, for the BWR/2 through BWR/6 I plants, adequate core cooling is maintained for these worst-case condi-tions. Analyses of further degraded conditions involving a stuck-open relief valve in addition to the worst transient and single failure were also performed. The results show that, with' proper operator action, the core remains covered and therefore adequate core cooling is achieved.

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ADEQUATE CORE COOLING FOR TRANSIENTS WITH A SINGLE FAILURE I. Introduction This report has been prepared as the BWR Owners' Group generic response to NUREG-0737 Task Item II.K.3.44 which addresses the issue of adequate core cooling for transients with a single failure for those plants identified in Appendix A. The text of Item II.K.3.44 is as follows:

For anticipated transients combined with the worsts. sir.gle failure and assuming proper operator actions, licensees should demonstrate that the core remains covered or provide analysis to show that no significant fuel damage results from core uncovery. Transients which result in a stuck-open relief valve should be included in this category."

At the outset it should be noted that the conditions described in II.K.3.44 (i.e., transients plus single failures) go beyond the current BWR design basis and that the item's reference to transients with multiple failures goes beyond the regulatory requirements as specified in Regulatory Gui'de 1.70, Rev. 3. The multiple failures specified involve consideration of a stuck-open relief valve (SORV) combined with the worst single failure. GE and the Owners Group continues to support the current BWR design basis approac.h. This report is intended to provide information to address Item II.K.3.44, but it does not reflect our intention to change the current BWR design basis approach.

It is shown that, for the GE BWR/2 through BWR/6 plants, the core remains covered for any transient with the worst single failure.

This is achieved without any operator action to manually initiate emergency core cooling system (ECCS) or other inventory makeup systems. The worst transient with the worst single failuro is shown to be the loss of feedwater (LOF) event with a failure of l

l the high pressure ECCS or one isolation condenser (IC) loop, whichever is applicable.

For the bounding LOF event, studies which included even more degraded j conditions have been documented in Reference 1. The degraded

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conditions cover the, failure of HPCS (or HPCI or FWCI or IC) and one 50RV. Reference 1 shows that the core will remain covered and therefore, that no fuel failure would occur.  ;

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! II. Criteria, Scoce and Assumptions i NUREG-0737 Item II.K.3.44 requires that the licensees demonstrate adequate core cooling to prevent the fuel from incurring significant l damage for the anticipated transients combined with the worst

single failure. In order to meet this requirement, either one of

the following two criteria should be satisfied

1. The reactor core remains covered with water until stable

! conditions are achieved; or

2. No significant fuel damage results from core uncovery.

For BWR plants, this report will show that criterion 1 is met. The report makes the following assumptions:

a. A representative plant of each BWR product line, BWR/2 through BWR/6, is used to represent all of the plants of that product line.'

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b. The anticipated transients as identified in NRC Regulatory Guide 1.70, Revision 3 were considered.

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c. The single failure is interpreted as an active failure. -

I d. All plant systems and components are assumed to function j normally, unless identified as being failed.

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III. Discussion Table 1 lists all of the transients which were considered in this study. The event sequence of each transient was examined for each product line to determine the impact on core cooling. The following three factors were used to determine the worst transient and the worst single failure:

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a. Reduction or loss of main feedwater or coolant makeup or heat removal systems, especially high pressure systems, e.g., HPCI, FWCI, HPCS, RCIC or IC.

b. Steam release paths causing rapid reactor coolant inventory loss, e.g., S/RV's, turbine, or turbine bypass valves.

c. Power level, especially the timing of scram.

Based on these considerations, a comparison was made among the transients in Table 1.

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TABLE 1

SUMMARY

OF INITIATING TRANSIENTS

(

Reference:

NRC Regulatory Guide 1.70, Revision 3)

1. Loss of Feedwater Heating

2. Feedwater Controller Failure - Maximum Demand l 3. Pressure Regulator Failure - Open l 4. Inadvertent Safety / Relief Valve Opening l S. Inadvertent Residual Heat Removal (RHR) Shutdown Cooling Operation

6. Pressure Regulator Failure - Closed GeneratorLoadRejection

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7.

8. Turbine Trip l 9. Main Steam Isolation Valve (MSIV) Closure

10. Loss of Condenser Vacuum

11. Loss of Normal AC Power

12. Loss of Feedwater Flow

13. Failure of RHR Shutdown Cooling l

14. Recirculation Pump Trip

15. Recirculation Flow Control Failure - Decreasing Flow

16. Rod Withdrawal Error I

17. Abnormal Startup of Idle Recirculation Pump

18. Recirculation Flow Control Failure - Increasing Flow l

19. Fuel Loading Error

20. Inadvertent Startup of High Pressure Core Spray (HPCS) or High Pressure Coolant Injection (HPCI) or Feedwater Coolant Injection (FWCI) or Isolation Condenser (IC), whichever is applicable.

In Reference 2, the events of Table 1 are compared in detail for e typical BWR/4 plant. In particular the impact on core cooling for each transient is evaluated by comparison to the analysis results for the LOF event in the section titled " Applicability of Analyses."

It is found that the LOF event is the most severe transient from thecorecoolingvie@ointduetoitsrapiddepletionofreactor

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coolant inventory. This conclusion has generic applicability to all BWR product lines covered by this study.

The same approach was also used to select the single failures which would pose the greatest challenge to core cooling. Among all of the possible failures considered (Table 2), the following failures are identified as the most important ones:

1. Failure of HPCI or HPCS or FWCI or one IC loop, whichever ~is applicable.

I l 2. Failure of RCIC.

j 3. One of the S/RV's, which has opened as a result of the transient, fails to close.

I Items 1 and 2 are the poss'ible limiting failures because they represent loss of high pressure inventory makeup or hett removal

systems which would be relied on following a loss of feedwater event. Item 3 is a possible limiting failure, because it results

in the largest steam release rate from the vessel compared to other possible release paths (e.g., a stuck-open turbine bypass valve).

No other failares identified in Table 2 result in a direct challenge to core cooling capability.

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TABLE 2 LIST OF SINGLE FAILURES WHICH CAN POTENTIALLY DEGRADE THE COURSE OF A BWR TRANSIENT

1. One or all of the bypass valves fail to modulate open when required.

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2. One of the bypass valves, which has opened as a result of the transient, fails to close.

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3. Failure to trip the turbine or feedwater pumps on high water level. .

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4. One main steam isolation valve (MSIV) fails to close when required.

1 5. One of the safety / relief valves fails to open when required.

6. One of the safety / relief valves, which has opened as a result of 4

the transient, fails to close.

7. Failure to trip one recirculation pump.

8. Failure to run back the recirculation pumps.

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9. Failure of high pressure coolant injection (HPCI) or high pressure l core spray (HPCS) or feedwater coolart injection (FWCI) or one -

isolation condenser (IC) loop, whichever is applicable.

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10. Failure of reactor core isolation cooling (RCIC) or one IC loop, whichever is applicable.

11. Failure of one low pressure coolant injection (LPCI) loop or the low pressure core spray (LPCS) system.

i TABLE 2 (CONT'0) l

12. Loss of one residual heat removal (RHR) system heat exchanger.

13. A, single control red stuck while the remainder of the centrol rods are moving. - .

14. Failure to achieve the rod block function (i.e., a single control rod will withdraw upon erroneous withdrawa.1 demand).

15. Loss of one diesel generator if loss of AC power was the initiating event.

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Because of the relatively low steam loss capacity through one 50RV (Failure 3, Page 5) ccmpared to the makeup water capacity of the highest capacity makeup water system, the failure of the highest capacity high pressure makeup system (Failure 1, Page 5) would be worse than a stuck open relief valve (Failure 3, Page 5). For

- example, for a typical BWR/4, representative values of HPCI makeup and S/RV flow are 18% and 6% of rated feedwater flow, respectively.

Because of the higher makeup rate of HPCI/HPCS relative to RCIC (3%

of rated feedwater flow), Failure 1 would be worse than Failure 2.

Table 3 lists the worst combination of transient and single failure for the GE BWR product lines covered by this study.

Even with the worst single failure in combination with the LOF event, the RCIC or at least one IC loop will function to provide makeup and/or to remove decay heat while the vessel pressure remains high. The design basis for the RCIC or the IC is such that they are capable of removing decay heat with the vessel being isolated.

Analyses of the LOF ivent with the worst single failure have been performed to support this conclusion. For example, for BWR/2 plants, such analyses are documented in Reference 1 Table 3.2.1.1.5-5.

These analyses show that the isolation condenser heat removal i capacity is greater than the decay heat generation rate and will

! lead to a safe and stable condition. Similar analyses have been I performed for representative plants with the RCIC system. These analyses show that for the worst transient with the worst single '

failure, the minimum whter level for different BWR product lines ranges from 6 ft to 11 ft above the top of the active fuel.

With even more degraded conditions, i.e., one SORY in addition to the worst case transient with the worst single failure, reference plant analyses in Reference 1 Tables 3.2.1.1.5-9 and 3.2.1.1.5-10 show that for the plants analyzed the RCIC system can automatically provide sufficient inventory to keep the core covered even with a ,. l single failure plus a SORV. This capability is not a design basis l for the RCIC system, and not all plants have been analyzcd to demonstrate this capability. If a plant should not have this l

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TABLE 3

.THE WORST CASE OF TRANSIENT WITH A SINGLE FAILURE FOR DIFFERENT BWR PRODUCT LINES Product Line Transient with a Single Failure (The Worst Case)

BWR/2 LOF + Failure of one IC Loop (0yster Creek only)

LOF + Failure of FWCI (Nine Mile Point only)

BWR/3 ,

LOF + Failure of PJCI (Mi11stene only)

LOF + Failure of HPCI (others)

BWR/4 LOF + Failure of HPCI BWR/5 LOF + Failure of HPCS BWR/6 LOF + Failure of HPCS capability, manual depressurization will avoid core uncovery for the case of LOF plus worst single failure plus 50RV. It should be noted that manual depressurization is the proper operator a'ction l for all plants during loss of inventory conditions when the high p,ressure cooling system (s),are unable to restore and maintain RPV level. These proper operator actions are allowed for in the l NUREG-0737 requirement.  !

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For plants without RCIC, manual depressurization will avoid core uncovery for the case of LOF plus worst single failure plus 50RV.

IV. Conclusion The anticipated transients in NRC Regulatory Guide 1.70, Revision 3 i were reviewed for all BWR product lines BWR/2 through BWR/6 from a core cooling viewpoint. The LOF event was identified to be the j most limiting transient which would challenge core cooling. The i BWR is designed so triat the high pressure makeup or inventory maintenance systems or heat removal systems (HPCI, HPCS, FWCI, RCIC or IC) are independently capable of maintaining the water level above the top of the active fuel given a loss of feedwater. The

detailed analyses show that even with the worst single failure in combination with the LOF event, the core remains covered.

Furthermore, even with more degraded conditions involving one SORY f in addition to the worst transient with the worst single failure, l studies show that the core remains covered during the whole course of the transient either due to RCIC operation or due to manual .

depressurization.

It is concluded that for anticipated transients combined with the

, " worst single failure the core remains covered. Additionally, it is concluded that for severely degraded transients beyond the design i

j I basis where it is. assumed that a S/RV sticks open and an additional failure occurs the core remains covered with proper operator action.

V. References

l. Section 3.2.1 (prepublication form) cf " Additional Information l Required for NRC Staff Generic Report on Boiling Water Reactors,"

NE00-24708, March 31, 1980

2. Section 3.2.2 (prepublication form) of " Additional Information l Required for NRC Staff Generic Report on Boiling Water Reactors,"

NEDD-24708, June 30, 1980

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f 3. Section 3.5.2.1 (prepublication form) of " Additional Information l

Required for NRC Staff Generic Report on Boiling Water Reactors,"

NEDO-24708, August 31, 1979 i

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APPENDIX A PARTICIFATING UTILITZES NUREG-0737, II.K.3.44 i

This report applies to the following plants, whose owners participated in the report's development.

Boston Edison .e Pilgrim 1 1 Carolina Power & Light -

Brunswick 1 & 2 i

Commonwealth Edison LaSalle 1 & 2, Dresden 1-3, 4

Quad Cities 1 & 2 l

Georgia Power -

Hatch 1 & 2'

Iowa Electric Light & Power Duane Arnold Jersey Central Power & Light Oyster Creek 1 Niagara Mohawk Power Nine Mile Point 1 & 2 1

i Nebraska Public Power District Cooper Northeast Utilities Millstone 1 Philadelphia E1cetric ,

Peach Bottom 2 & 3; Limerick 1 & 2 Power Authority of the State of New York Fitzpatrick Tennessee Valley Authority Browns Ferry 1-3; Hartsville 1-4, Phipps Bend 1 & 2 Vermont Yankee Nuclear Power Vermont Yankee 2

Detroit Edison Enrico Fermi 2 Mississippi Power & Light Grand Gulf 1 & 2 Pennsylvania Power & Light Susquehanna 1 & 2 Washington Public Power Hanford 2 Supply System Cleveland Electric Illuminating Perry 1 & 2 Houston Lighting & Power Allens Creek ,

Illinois Power ,

Clinton Station 1 & 2 Public Service of Oklahoma Black Fox 1 & 2 Long Island Lighting Shoreham Northern States Power Monticello i

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1 ATTACHMENT III i to LILCO Testimony o'n SC 28a(vi) l and SOC 7A(6):

i i Letter of SNRC-579 of May 29, 1931 from J. P. Novarro to Harold R. Denton, I with attachments i

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LONG ! GLAND LIGHT!NG COM PANY

[~,J;.;[. .I.3,. SHCREMAM NUCLEA 4 POWER STATION P.O. DOX 613. NORTH COUNTRY RC AO . WADING RIVER. N.Y.11792 May 29, 1981 SNRC-579 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Shoreham Muclear Power Station - Unit 1 Oc:k:t N0. 5C-222 D ar Mr. Denton:

Forwarded herewith are sixty (60) copies of our positions related to Post-TMI Recuirements outlined in NUREG-0737,

" Clarification of TMI Action Plan Requirements". The soventeca specific items adcressed are listed in Attachment 1 to this letter. Where applicable, cur responses to certcin NUREG-0578 items (SMRC-503, dated S/29/80) are superceded by the respective responses contained herein.

Two previous submittals have been made regarding NUREG-0737 1.e., letters SNRC-557 dated 4/15/82 and SNRC-563 dated 5/15/81. As a result of discussions with your staff on these items, the following clarifications are being made:

1. II.K.3.44 " Evaluation of Anticipated Transients with Single Failure to 'larify No Fuel Failure" - The analyscs noted in the BWR Owners' Group Evaluation were done for BWR 2's through BWR 6's. The analyses done for BWR 4's .

and the conclusions reached as a result of these analyses, i.e., no core uncovery, are applicable to Shoreham.

2. II.K.3.18 " Modification of Automatic Depressurization '

System Logic" - The proposed modification will be implemented in accordance with the requirements of NUREG 0737, i.e., during the first refueling to occur six months after NRC staff approval of the proposed modi-fication.

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Harold R. Donton 11ay 29, 19S1 Page 2 This submittal ecmpletes our racponses on SUPIG-0737. All of those submittal

and the clarifications noted above will

• b3 incorporated into an FSAR Amendment.

Vary trul'y yours, Origir.al cig.ne:i by

\ lb. P. IIOVrd0 s.

J. P. Novarro Project Itanager Shoreham Nuclear Power Station W/'RWG:mp

-)) Enclosures cc: J. Iliggins bec: Dist. List #14 Eng. File /SR2...A21.010 w/ attachment M.H. Milligan w/ attachment i "

R.A. Loper A. Pedersen ' w/o httachment O

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ATTAC:iM .:;I 1 NUREG-0737 ITEMS INCL *.:DED IN'T1:IS SU3MITTAL I.A.2.1 Immediate Upgrading of Reactor Operator and Senior Reactor Cperator Training and Qualifications I.B.1.2 Evaluation of Organization and Management I.C.1 Guidance for the Evaluation and Development of Procedures for Transients and Accidents I.C.6 Procedures for Verfication of Correct Performance of Operating Activities I.D.2 Plant Safety Parameter Display Console I.G.1 Training Requirements During Low Power. Testing 21.B.1 Reactor Coolant System Vents II.B.4 Training for Mitigating Core Damage II.D.1 Performance Testing of Boiling-Water Reactor and Pressurized-Water Reactor Relief and Safety Valves II.E.4.2 Containment Isolation Dependability II.F.2 Id.entification of and Recovery from Conditions Leading to Inadequate Core Cooling II.K.3.30 Revised Small Break Loss of Coolant Accident Methods to Show Compliance with 10 CFR 50, Appendix K i .

II.K.3.31 Plant-Specific Calculations to Show Compliance with 10 CFR 50.46 ,.

1 III.A.l.1 Upgrade Emergen'cy Preparedness III.A.l.2 Upgrade Licensee Emergency Support Facilities III.A.2 Improving Licensee Emergency Preparedness -

Long Term ,

III.D.l.1 Primary Coolant Sources Outside the Containment Structure l

LILCO, June 14, 1982 1

ATTACHMENT IV to LILCO Testimony on SC 28 a(vi) and SOC 7A(6):

BWR Owners' Group Evaluation of NUREG-0737 Item II.K.3.16, Reduction of Challenges and Failures of Relief Valves

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BWR OWNERS GROUP EVALUATION

'0F NUREG-0737 ITEM II.K.3.16 REDUCTION OF CHALLENGES AND FAILURES OF RELIEF VALVES e

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TABLE OF CONTENTS J

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• l ABSTRAI.T . . . . . . . . . . . . . . . . . . . . . . . . . . i i.

1. Introduction . . . . . . . . . . . . . . . . . . . . . . . . 1

' 2. BWR Response to a Transient with a Stuck Open Relief Valve . 4

3. Candidate Modifications Evaluated for Reducing SORV Event l

. Frequency. . . . . . . . . . . . . . . . . . . . . . . . . . 6 i

4. Methodology . . . . . ................... 23

5. Results . . . . . . . ................... 27

6. Conclusion . . . . . . . . . . . . . . . . . . . . . . . . . 31 l , .

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I Appendix A - Participating Utilities '. . . . . . . . . . . . . .- 33 9

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A55 TRACT

~ This report documents a study performed in response'to NUREG-0737 item II.K.3.16 which requires an evaluation of the feasibility and contrain-dications of reducing cha11enges to the relief valves by various methods -

in BWRs by April 1, 1981. The report reviews potential methods of ....

l reducing the likelihood of stuck open relief valve (50RV) events in BWRs .

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- - -and estimates the reduction in such events that can be achieved by .

.inplementing these methods. The reduction was estimated by cocputing -

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the reduction in Safety Relief Valve (SRV) actuations achievabic by various design and operating modifications and by estimating the relative-

' probability of various types of SRVs to stick open. Using the BWR/4 plant as a measure of operating experience, it was concluded that BWR/2, BWR/3 with isolation condenser, BWR/5 and BWR/6 plants already include design features which yield a significant reduction in the occurrence of

-50RV events. The remaining plants can reduce the 50RV event frequency by methods evaluated herein. The report applies to the plants listed in

. Appendix A. -

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. 1. INTRODUCTION This report documents a study performed in response to NUREG-0737 item ,,

II.K.3.16 which requires an evaluation of the feasibility and contrain- .

' dictations of reducing challenges to the relief valves by various methods in BWRs by April 1, 1981. The report reviews potential methods of _ , , ,

reducing the likelihood of stuck open relief valve (SORV) events in BWRs and estimates the reduction in such events that can be achieved by ,,,

icplementing these methods. -.

Reducing the likelihood of S/RV challenges will directly reduce the likelihood of a 50RV. In addition, attention is also given to codific:-

tierts-which could reduce spurious SRV blowdowns and to modifications which could reduce the probability of SRVs to stick open when challenged.

The , report applies to the plants listed in Appendix A.

1.1 NRC Recuirement HUREG-0737 item II.K.3.16 requires that a feasibility study be performed to identify modifications to reduce S/RV challenges. The NUREG-0737

. position stater, "An investigation of the feasibility and contraindica-tions of reducing the challenges to the relief valves...should be conducted.

Those changes which are shown to reduce relief-valve challenges without compromising the performance of the relief valves or other systems should be implemented. Challenges to the relief valves should be reduced substantially (by.an order of magnitude)." .

); 1.2 Obfective. Standard and Goal Although the NUREG 0737 position geah pimarily with reduction of challenges to S/RVs, its clear intent is to reduce the incidence of 50RV events. Reducing challenges is only one of three approaches to reducing 50RV events. The others are reducing the causes of spurious blowdowns and reducing the probability of S/RVs to : tick open when challenged.

All three of these approaches present feasible and effective opportunities

. for reducing the incidence of uncontrolled blowdowns via S/RVs. Further,

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i as discussed in the study, the feasible approaches to reducing S/RV challenges do not by themselves accomplish the desired " order of magni-tude" (factor of ten) reduction in SORY events. Consideration of the  !

other two approaches, however, shows a factor of ten reduction in the incidence of 50RVs to be feasible. Such a reduction is the objective of .

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a this study. ,

j The NUREG-0737 position does not state a standard by which. the desired .

factor of ten reduction should be judged. Using the number of 5/RV years I

in plant operation as a criterion, it can be concluded from Table 1.1

. that operating BWR/4 units provide the most representative basis for

!- suchjudgements. The methods employed in this study make.it desirable i to take a single BWR product line as a benchmark, due to the differences in transient response, valve types, and other reactor systems (e.g.,

4 isolation condensers) among the product lines. Thus, the 50RV experience of currently operating BWR/4s without the design improvements described in this study was selected as the criterion by which all plants are to be judged.

! Considering all of the above, the goal *of this study is to identify feasible modifications to BWR design and operation which reduce the i f's:pancy of uncontrolled S/RV blowdowns for each product line to a rautor of ten below the frequency experienced in BWR/4 units.

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TABLE 1.1 .

S/RV OPERATING EXPERIENCE **

SRV-Years of Operatien*

. Product Line End of 1977 End of 1(X)

BWR/1 -

N/A N/A S!!R/2 93 126

. BWR/3 260 361 BWR/4 342 715 BWit/5 None None ,,

BWR/6 None Nona .

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• Number of power-operated relief or safety / relief' valves per plant, ...

times number of years of operation of plant, totalled for all plants in product line through end of year. ,

• • Includes US experience only.

N/A - Not applicable since these plants were not considered in the ,

study. .

2. BWR RESPONSE TO A TRANSIENT WITH A STUCK OPEN RELIEF VALVE _ _ _ ,

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The Boiling Water' Reactor design assures that core integrity will be maintained following a stuck open relief valve (50RV) transient. The response of a typical BWR to a 50RV transient is discussed in this section in order to provide a perspective on typical sequenc2s of events which' lead to and follow SORY events and the consequence of various manual or automatic actions. .

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(. The safety / relief valves (S/RVs) of a BUR are designed to protect tha reactor coolant pressure boundary from overpressurization. Transients resulting in pressurization frequently raise the reactor vessel pressure to the S/RV setpoint~ causing the S/RVs to open so that safety limits are not reached. If any relief valve fails to close after the pressure peaks and decreases, further steam release will deplete the water inventory in the reactor vessel and challenge the numerous water delivery systems l .

which assure adequate core cooling. In a few instances, S/RVs have

spuriously opened. E,ither event is termed a " Stuck Open Relief Valve l (SORV)" event.

l For any anticipated transient, if a 50RV is the only additional failure, the vessel inventory lost through the 50RV can be easily made up by

{ various high pressure and/or low pressure water delivery systems. The consequences of a 50RV are not a safety concern and reactor shutdown is uncomplicated, as proven by numerous field occurrences.

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l Following transie.nts which result in loss of feedwater flow, a SORV could challenge the emergency core cooling systems. There are several

!~~ such transients which result in the loss of the feedwater system, e.g. ,

" loss of feedsater flow," " loss of AC power," "MSIV Closure" cr "Feedwater controller failure maximum demand." The loss of feedwater flow event

is a typical and bou'nding transient from the core cooling viewpoint.

The BWR response to this transient with a SORV and with more severo l 1

degradations is discussed in detail in NEDO-24708, Sections 3.1, 3.2 and '

3.5 (Refs. 1-4). These evaluations demonstrate that adequate core

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cooling is assured in the BWR following an SORV event.

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. i It is concluded that adaq0 ate core cooling is maintained in a B'JR following an 50RV event even under degraded conditions. It follows, then, that

' reduction of the frequency of SORV events is not of great concern from the standpoint of assuring adequate core cooling. .

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3. CAftDIDATE MODIFICATI0fts FOR REDUCING SORV EVE!!T FREOUF!!CY Three different approaches can be taken to reduce the frequency of SORV

events

1. Reduction ( f challenges to the S/RVs; l

i 2. Reduction of the probability of the S/RVs to stick open when challenged;

3. Redu: tion of spurious blowdown of S/RVs.

I Each rf these approaches leads to the identification of feasible and j effective opportunities for reducing the incidence,of uncontrolled blowdowns via S/RVs. Based on the recommendations in NUREG-0737 and on

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l the judgment and experience of GE and utility personnel, a number of candidate acdifications have been selected for consideration in this

! study. A description of these candidate modifications is provided in this section. The benefit associated with the implementation of each candidate modification, as estimated in this study, is also presented.

Other candidate modifications may exist which are not addressed in this study. ,

The effectiveness of many of the candidate modifications will vary

amongst the BWR product lines, due to design variations. Thus, a range

! of potential benefits are presented. For instance, lower water level I isolation is expected to result in a 24-36% reduction in S/RV challenges.

l The 24% reduction is applicable for BWR/S plants and 35% for BWR/3

! plants without Isolation Condenser. Further, the values cited ara

! maximum achievable benefits evaluated based on the' ass'umption that the candidate modification will completely eliminate all of the challenges associated with the modification.

j 3.1 Candidate Modifications Which Reduce S/RV Challences i Most of the candidate modifications to reduce S/RV challenges reeuce the frequency of transient events which cause S/RVs to open. The remairing 6

candidates reduce the number of relief valves which open during a given transient event. ,

3.1.1 Main Steam Line Isolation Two' candidate modifications will reduce the frequency of main steam line

' solation during transients. One involves lowering the water level isolation setpoint; and the other, lowering the pressure isolation setpoint.

3.1.1.1 Lower Water , Level Isolation Satpoint ,

Definition - Lower the RPV water level isolation setpoint for MSIV closure from Level 2 to Level 1.

Discussion - This candide.te modification would reduce the number of times the reactor is isolated from the main condenseE This results in reduced S/RV challenges by eliminating isolation cycling of the S/RV's resulting from transients such as feedwater controller failure, trip of both recirculation pumps, and loss of feedwater flow. This modification is expected to result in a reduction of 24 to 36% in S/RV challenges for plants without isolation condensers.

This modification is feasibic for BWR/4-5 plants which do not a1 randy include the feature. It is not feasible for BWR/2-3 because of the need for additional reactor water level instrumentation.- .

3.1.1.2 Lcwer Reactor Pressure Isolation Setpoint ,

, Definition - If the reactor is in the "run" mode and the main steam line pressure drcps below 825 psig, the reactor is isolated in order to prevent a rapid cooldown resulting from a pressure control malfunction.

This candidate modification would reduce the pressure at which the reactor is isolated.under these conditions. The extent of reduction in pressure setpoint has not been established but is expected to be in the neighborhood of 50 psig.

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. Discussion - Prior to 1975, the main steam line isolation was initiated at 850 to 880 psig depending upon the product line. Operating experience at that time showed that this setpoint was too close to the normal operating pressure. As a result, the noise level of the pressure switch i hydraulic sensing line or small pressure transients in the main steam lines could initiate reactor isolation. After a careful review, General Electric determined that the isolation setpoint could be safely lowered j to 825 psig. This setpoint provides adequate protection against spurious j isolation events. A review of recent operating plant experience shows that the additional reduction in S/RV challenges resulting frem further

! reduction in isolation.setpoint would be less than 11 .-

! 3.1.2 Feedwater Control System Modifications 1

j A number of candidate modifications improve the feedwater control system i

as a means to reduce S/RV challenges by reducing the number of main

steam line isolation events. Feedwater control system failures have

! contributed to about 0.7 isolation events per plant year. The thrust of

the modifications is to control water level between the high water level

< trip and the ECCS initiation /MSIV isolation trip setting for various i transients. -

e Water level is controlled in a BWR by a Feedwater Control System that l- utilizes a single primary channel for control. The control system

! utilizes a water level sensor input (the primary element) and the dif-ference'between two secondary elements, namely feedwater ficw and steam flow. Water level must be under positive control by the Feedwater

Control System during plant operation, because feedwater flow must respond to changes in steam ficw while maintaining the water level in the vessel. Therefore, when a component in the control system or in the power supply to the control system fails, the water level in the l reactor vessel can drift out of limits, ultimately causing the reactor to scram. Such a water level excursion is usually so rapid that the i reactor operator is. unable to respond in time to prevent a scram from an t

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abnormal water level condition. Frequently these scrams are followed by reactor isolation with consequent vessel pressurization causing the i S/RV's to open. , ,

3.1.2.1 , T*iple Redundant or Single Failure Proof Control System 1

1 Definition - A triple redundant control system is a candidate modifica-l tion which could reduce isolations. Such a system would have three channels of control, with the highest and lowest values being ignored.

- Thus failure of the controlling element (either upscale or downscale) -

would result in another channel taking control. The single failure proof

! contro'1 system would have two channels of control, in which the second l channel acts as a backup for transfer of control when a failure of the

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controlling channel is detected.

Discussion - The improved control system would reduce _ transients resulting from failure of components in the existing single channel Feedwater

! . Control System. By eliminating these transients, the associated reactor scrams and isolation events are also eliminated, which reduces the f number of S/RV challenges. This results in a reduction of about 2 to 4%

i in S/RV challenges. The small reduction in S/RV challenges alone does not justify the high cost of implementing the modification.

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3.1.2.2 Uninterrupted & Redundant Control System Power Definition - Failures in Feedwater Control System power supplies have

. caused reactor isolations in operating plants. This candidate provides ,

for an uninterruptible and redundant source of pcwer such that the controller is not affected by failures in the power supply system.

Discussion - This change could eliminate S/RV challenges associated with isolation events arising out of failures in the power supply. A maximum of 0.07 isolation events per plant year can be eliminated by implementir.g this modification, resulting in about a 1% reduction in S/RV challenges.

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3.1.2.3 Condensate System Hodifications and Condensate /Feedwater Integration -

, Definition - The controls of the condensate system (including the demin-eralizers), on which the feedwater system depends for proper operation, -

could be in'egrated t with the Feedwater Control System so that failures in the condensate system would be detected in such a way that reactor scram and isolation could be avoided.  :: -. .

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This candidate modification calls for an integration of the cor.dansate -

and feedwater control systems in providing input signals for reactor operation. Examples of the possible integration and design modifications of the condensate system are as follows: .

a) Typically, three 50% or four 33% capacity condensate pumps are provided per plant. If one of the condensate pumps fails, the redundant pump could be automatically sta.-ted.

b) If there is, high differential pressure across a condensate deminer-alizer, a signal could be provided to cut back reactor power by running back recirculation flow.

c) The Feedwater Control System could be designed to assure that a loss of a single condensate or condensate booster pump or feed pump will not result in reactor scram or isolation.

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Discussion - Imp 1ementation of these candidate modifi,c,a_ti,ons could result in a reduction in reactor isolation events for BWR/3 plants j

without Isolation Condenser, BWR/4 and SWR /S plants, and consequently, relief valve challenges resulting from failures in the condensate and feedwater systems.

This could result in a reduction of 3-4% in S/RV challenges.

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l j The implementation of this modification would increase the complexity of l the feedwater and recirculation control systems, and thereby introduce additional failure modes. Therefore this modification could have an l adverse impact on the reliability of these systems.

3.1.2.4 Feedwater Runback i Definition - Feedwater runback is a method of controlling reactor water Icvel to avoid high vessel water level (LB) trip, following certain transicnts. This would prevent tripping the fcedwater pumps cnd subsequent reactor isolation on icw water icvel.

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i Discussion - The implementation of this candidate codiff. cation would

result in the elimination of 5/RV challenges associated with the trip of l both recirculation pumps and recirculation controller failure. A reduc-f tion of 6 to 12% in S/RV challenges can be expected by implementing this modification. ,

The implementation of this modification would increase the complexity of the feedwater control system, and thereby introduce additional failure modes. Therefore, this modification could have an adverse impact on the reliability of that system.

3.1.2.5 Additional Anticipatory Scram on Loss of Feedwater Definition - A description of this candidate modification is quoted from NUREG-0626: "

...The challenge rate could be reduced by providing anticipa-tory signals on the feedwater pump trip similar to the scram signal derived from turbine stop valve closure on a turbine trip. This mcdifi-cation will reduce the reactor power quickly and thereby reduce the severity or magnitude of the pressure spike."

DiscussionL - A review of the Loss of Feedwater Flow (LFWF) event in NEDO-24708 shows that LFWF causes a low water level scram approximately 7-15 seconds following initiation of the transient at full power, depending upon the product line. LFWF results in reactor isolation for BWR/1 thru

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i BWR/5,andanticipatoryscramonfeebraterpumptripdoesnotprevent

, reactor isolation at reactor icvel 2 or the associated cycling of relief valves. LFWF does not result in reactor isolation for BWR/6 due to isolation at reactor level 1. This candidate modification is of no

. benefit in reducing S/RV challenges for BWR 1-5 if the low water level isolation modification is also carried out, since the latter modifica- .

tion prevents reactor isolation following the LFWF event. A possible disadvantage of implementing this candidate c:odification is that it denies the operator the opportunity to prevent a scram by restarting

! -feedwater pumps. In summary, anticipatory scram or loss of feedwater is

- considered to be an insignificant contributor (less than 1%) to S/RV-l challenge reduction for all BWR product lines.

3.1. 3 SRV Control Logic /SRV Setooint Revision .

The following candidate modifications are expected to reduce S/RV

! challenges through changes to SRV control logic or through revision of I S/RV setpoints.

3.1.3.1 Low-Low Set Relief or Equivalent Manual Acticns

Definition - Some BWR plants are equipped with a ' Low-Low Set' design feature which changes the setpoints of selected SRVs following the initial opening of a number of S/RVs. This assures that folicwing the

initial pressurization the pressure will be relieved by the ' Low-Low I

Set' valve alone, and the remaining S/RVs will not experience any subse-quent actuation. This faature could be applied to plants which do not j currently include it. However, the BWR Emergency Procedure Guidelines l (Ref. 5) call for the equivalent manual action.

Discussion - The ' Low-Low Set' design or equivalent manual action will reduce the total number of S/RV challenges by limiting the second and subsequent opening of the S/RVs to the Low-Law Set valve. It is esticated i that a 23-62% reduction in S/RV challenges can be achieved by implementing this modification. This modification is practical for all B'nR product lines.

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3.1.3.2 Revised Relief Valve Setpoints Definition - A description of this candidate modification is quoted from NUREG-0626: "The relief valve setpoints could be revised upward to allow more margin to the relief valve opening setpoint. Another method to provide margin is to lower the operating pressue. A combination of the relief valve sotpoint and operating pressure will increasa the plant's ability to withstand a pressure increase transient without .

causing the relief valve to open." ,

Discussion - There are two setpoints associated with safety / relief valves, namely relief setpoint and. safety setpoint. The relief setpoint, i

which is the lower of the two setpoints, is used to provide pressure relief following an overpressure transient. The safety setpoint limits i the reactor pressure to the ASME code allowable limits.

! Both the relief and spring setpoint values are constrained by a number of factors. In the case of the Target Rock valve, the factors are:

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1. The ASME Code .

2. High pressure injection system (High Pressure Coolant Injection 1

(llPCI). High Pressure Core Spray (HPCS), Reactor Cere Isolation Cooling (RCIC)) design discharge pressure. If the spring setpoints are higher than the purr.p discharge pressure, high pressure coolant cannot be infected into the reactor under all design conditions. ,

3. A need to offset relief valve setpoints where applicabic.

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This is done to prevent all valves from opening simultaneously.

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j 4. Tol rance on the relief valve setpoints. Setpoint drifts in l one direction should not result in the valve openirg in the i

safety mode, nor should a drift in the opposite direction

! result in relief valve operation for minor overpressure transients.

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I In the case of Crosby and Dikkers valves there exists another factor which is a practical consideration of reouiring the valve to reclose in a relief mode rather than the spring mode, i.e., the spring mode reclosure setpoint should be always higher than the relief mode reclosure setpoint, even after allowing for setpoint drifts. - .

In' the case of Three Stage Target Rock valves, it has b'een determined -

r. hat the spring setpoint could be raised by about 15 to 50 psig depending upon'the plant. The impact of such a modification is a reduced incidence l' of spurious S/RV actuations, through increased simmer margin. Based on a review of failure data and engineering judgment, a 5% reduction in SORV events in plants with Three Stage Target Rock valves is expected through increasing the spring setpoint to the maximum value possible.

No reduction in S/RV challenges is likely because a 15-50 psi increase

! in setpoint is insignificant compared to the pressure rise experienced i in most overpressure transients.

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In the case of Crosby and Dikkers valves (SWR /5-6 plants) the setpoints are already near their maximum possible value and can be increased by no more than about 15 psig. Such an increase in the relief setpoint will l not cause any significant reduction in SRV challenges.

In the case of Two Stage Target Rock valves, pilot valve leakage does

! not lead to spurious opening. Therefore, the conclusion for Crosby and i Dikkers valves also applies for the Two Stage Target Rock valve.

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In conclusion, the plants with Three Stage Target Rock valves may be j able to achieve a 5% reduction in spurious SRV openings through an j increased spring setpoint. None of the plants can achieve any signifi-i cant reduction in SRV challenges through an increased relief setpoint.

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The NUREG-0737 suggestion also refers to lowering the operating pressure.

But as stated above for setpoint increases, modest changes are insignifi-cant compared to the pressure rise experienced in cost overprersure l

transients. ' Thus, lowering the operating pressure by a modest amount 14

would not result in any significant reduction of S/RV challenges. In

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addition such a change w'ould result in undetermined penalties in terms

! of plant thermal efficiency and fuel utilization.

3.1.3.3 Offset Relief Valve Setpoints , ,

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! Definition - A description of this candidate modification is quoted from i : HUREG-0626 as follows: "The valve pressure setpoint. could also be j'- modified or offset such that fewer valves are challenged.." -

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Discussion - Offsetting relief valve setpoints does not contribute to G'

S/RV challenge reduction during isolation cycling since only one or two j . valves participate in such cycling. During the initial blowdown there could be some reduction in S/RV openings for some transients. It is

! noted_that small but unavoidable setpoint drifts result in a ,d_e facto

! offsetting even when several valves are nominally set at the same value.

! 3.1.3.4 Increase Main Steam Line Flow Setpoint I

Definition - A description of this candidate modification is quoted from NUREG-0626 as follows: " Increasing the high steam line flow setpoint I for main steam line isolation valve ~(MSIV) closure (can reduce SRV challengo and failure rates)."

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Discussien - The MSIVs are designed to close when a break occurs in the ~

main steam lines. An abnormal increase in the main steam line flow i's taken as an indication of a main steam line break. High steam line flow setpoints are selected in a manner as to assure a high probability of isolation on a steam line break while keeping the probability of inadver-tent clonre resulting from operational transients small. A review of the BWR experience data has revealed no instance of spurious MSIV closure resulting from plant transient events. However, a number of inadvertent isolation events have occurred during MSIV closure surveillance testing.

These occurred when a second MSIV was closed without resetting the first MSIV that was tested. The sudden increase in steam flow in the remaining lines results in reactor isolation. The maximum reduction in SRV challenges 15

that can theoretically be ob2afned through an increased hfgh steam flow ,

I setpoint is about 0.5%. However, such a reduction may not be practical

,' since increasing the setpoints will reduce the reliability of isolation j following a main steam pipe break. A more practical approach to achieve  !

the same goal is through reduction in MSIV test frequency, discussed in i

3.1.4.4.

1 l 3.1.4 Other Candidate Modification -

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Candidate modifications pertaining to other systems.are discussed in

this section.

3.1.4.1 Analog Transmitter / Trip Unit System .

Definition - Most operating BWRs use direct acting pressure, differential j pressure and water level switches as input into the reactor protection,

l l main steamline isolation, and emergency core cooling systems. Technical specifications for this type of process sensor typically require surveil-lance testing once a month while*the plant is at power. In the past, j during the monthly surveillance tests, errors have caused scrz.ms and challenges to relief valves. If an improved system were installed which uses an analog transmitter and bi-stable trip unit instead of the pressure l switch, the number of unnecessary scrams (and associated SRV challenges) could be reduced. The transmitter-trip unit combination can also be designed to be highly stable and easily testable. Calibration requirements for this new syste'm are thus greatly reduced.

Discussion - The use of the analog transmitter / trip unit system would reduce the number of reactor scrams resulting from procedural and physical

.ctrors during surveillance tests. A 2 to 5% reduction in S/RV challenges

'could be expected due to the implementation of this candidate modification.

16

Discussion -The frequency of MSIV tests is contained in a picnt's technical specifications and generally conforms to ASME Code Section XI recommendations. The extent of frequency reduction that is possible without impacting reliability of isolation capability should be con-sidered in the detailed design of this modification. However, the maximum benefit that could be expected is about 2 to.3% reduction in S/RV challenges. .

3.1.4.5 Installation of New Relief Valve With Block Valve in Series Definition - A description of this candidate modification is quoted from HUREG-0626: " Plants could also be modified by installing new relief valves with normally open isolat.'on or block valves that would eliminate the opening of present S/RVs that may fail to close and cannot be isolated."

Discussion - The following factors would have to be considered in the implementation of this candidate modification:

1. The suggested modification could be in violation of the ASME

. Code, unless scme valves are dedicated for the safety function and others for the relief function. Any modification would need to be reviewed to assure compliance with the ASME Code.

2. The new~ relief valves, pipes and b'eck valves would have to be designed for the reactor design pressure (1250 psig). Currently, ,

the relief valve discharge flange and piping is designed for about 600 psig. .

3. In::dvertent closure of the block valves would cause the new relief valves to become unavailable for the relief function.

This candidate modification would not reduce the 50RV event frequency, but would reduce the consequence of such an event. Theoretically it is possible to design a system that will mitigate any future 50RV event; however, from a practical standpoint this may be an impossible modifica-tion to implement since there would not be sufficient room in the drywell 18

of most plants to accommodate the additional piping and equipment. The large expense in terms of cost and personnel exposure required to implement this concept is not justified when the low risk of core damage resulting from a 50RV event is considered.

3.1.4.6 Earlier Initiation and Increased Flow of Emergency Core Coolant .

Definition - A description of this candidate modification is quoted from NUREG-0625: "Another method that could be employed is to provide additional

~ccergency core coolant (ECC) flow to act as a heat sink (stear.: condenser) to accommodate the pressure increase due to swelling of-the coolant. .

j This could also be accomplished by modifying the plant instrumentation - --

! to provide earlier ECC system initiation. The combination of increased ECC flow at an earlier time in the transient could provide the necessary t

heat sink to absorb the power or pressure spike before the relief valve

setpoint is reached."

Discussion - ECC flow could not be initiated early enough by any practical means to result in any significant reduction of the number of valves ,

'that open during the initial blowdown, because of the steep rate of pressure rise following a transient. Further, earlier initiation of ECC flow could result in ECC initiation and L8 feedwater pump trip on transients such as turbine trip (following which ECCS is not expected to initiata),

causing simultaneous loss of the preferred coolant source (feedwater) and the preferred heat sink (the main condenser). Such a modification cannot be justified.

.. 3.2 Reduction of the Relative Probability of the Valve to Stick 0:en Hat.y operating B'n'R plants are equipped with Three Stage Target Rock valves. The Three Stage Target Rock valves have exhibited a higher probability to stick open in the past than other types of valves. A dnailed review of the 50RV events associated with the Three Stage

~

Target Rock valve was carried out by General Electric and Target Rock Company, and the results have been used to identify valve modifications which improve the valve performance. Design and operational modifications 19 _ --

~

have been identified for tha Three Stage Target Rock valve which reduce the probability of the Three Stage Target Rock valves in service to stick open.9 In addition, the valve topworks have been redesigned to minimize the probability of the valve to stick open: The new design is referred to as the "two stage" design. Suc.h valves-have been installed .

in some operating plants. The valves thus modified are referred to as Two Stage Target Rock valves in this study. -

Some operating B'4R plants are equipped with Dresser Electromatic relief .

valves. B'iR/5-6 plants are equipped with Crosby and Dikkers dual-functica safety /relicf valves. - - - - - -

Assigning a nonnalized SORV probability factor = 1.0 for the Three Stage Target Rock valve (which is taken as a benchmark valve), the relative SORV probability factors for other valves were determined as follows:

Two Stage Target Rock Valve = 0.50 Dresser Electromatic Valve = 0.25 Crosby Valve = 0.125 Dikkers Valve = 0.125 3.3 Reducing Causes of Sourious Blowdowns i The following candidate design modifications are expected to directly affect the number of SORVs by eliminating the causes of spuricus blowdcwns.

3.3.1 Eliminate Spurious Safety / Relief Valve Openings Resulting fica DC Power Supply Ground Faults l Definition - Inadvertent S/RV openings can be reduced by'previding doubic pole single throw switches, or other means of protection that disconnect both the positive and the negative sides of the DC power supply, for ener-l gizing and deenergizing the solenoids of safety relief valves.

4 j

• Plants with improved Three Stage Target Rock valves and plants employing operational modifications will address their valve reliability on a I

plant-unique basis. -

i 20

Discussion - The potential benefit of this candidate modification is ,

the avoidance of spurious depressurization of the reactor as the result l of grounding faults in the DC power supply. It is estimated that approxi-mately one spurious relief valve opening or failure to reclose after i

proper opening per 50 reactor years would be eliminated by this modifi-cation. -

Detailed design should assure that the new switching device will not be less reliable than the existing device in performing the functions of

[- energizing and deenergizing the solenoid coil on de'::and. -

1 i

l 3.3.2 Control of Pneumatic SuppTy Pressure to S/RV's.

!. Definition - High pneumatic supply pressure to the actuating solanoids

!- of Target Rock S/RVs caused one spurious blowdown in an operating plant.

1 Improved pneumatic supply pressure control would eliminate the cause.

Distussion - The implementation of this candidate modification would assure that this mode of spurious S/RV actuation will be' eliminated.

. This modification results in a maximum reduction of 2% in spurious l blowdowns.

3.3.3 Revised S/RV Spring Setpoint. See discussion in Section 3.1.3.3.

j .

3.3.4 More Stringent Leakage Criteria and Early Removal of Leaking l

Valves i

Definition - These candidate modifications were suggested in NUREG-0525.

"More stringent leakage criteria" is assumed here to refer to leaking l safety / relief valves while in operation. "Early removal of leaking I

valves" refers to a planned action of removing the valves which begin to leak.

l

}

l 21

Discussion - Leaking Three Stage Target Rock valves can result in spurfous i

blowdown. Analysis has shown that a maximum of 40 to 60% reduction in spurious operation of S/RVs could be obtained by identifying and replacing valves with high leakage. Since all valves leak to some extent, it is difficult to develop an absolute leakage criterion. Additional study is .

required to develop a leakage criterion which is practical and a system l to detect the leakage. With the use of the Two Stage Target Rock... -

Crosby or Dikkers valves, the leakage is not a concern because leakage l- does not significantly affect the spurious blowdown. probability. The --

- impiccentation of this candidate modification will not reduce spurious l

! bloudowns in the Two Stage Target Rock, Crosby and Dikkers valves.

3.3.5 Use of Two Stage Target Rock Valves i

Definition - The Three Stage Target Rock valves could be changed to Two Stage Target Rock Valves.

i i

Discussion - The Two Stage design eliminates most spurious blowdown

! modes associated with the Three Stage Valves. A 40-60% reduction in j spurious blowdcwns can be achieved by changing the Three Stage Target i Rock valves to Two Stage valves.

. 6 _ _

i 22

l

4. METHODOLOGY

! This section discusses the methodology used in this study.

! 4.1 Introduction .

l Although the NUREG-0737 position deals primarily with the reduction of challenges to S/RVs, its clear intent is to reduce the incidence of 50RV events. Reducing challenges is only one of three approaches to reduction of 50RV events. The others arc reducing the causes of spurious blowdowr.:;

I and reducing the probability of S/RVs to stick open when challenged.

All.three approaches are required to achieve an " order of magnitude" (factoroften)reductionin50RVevents. BWR/4 units equippad with Three Stage Target Rock valves were used as the basis for such judgments.

The BWR/4 with Three Stage Target Rock valves is referred to as the

" benchmark" plant in this discussion.

4.2 Approach -

l A comparison of the 50RV event frequency that can be expected over the l lifetime of each BWR product line is made by multiplying the expected l, number of S/RV openings during a plant's_ lifetime by the relative proba-

bility factor for the S/RV to stick open. The 50RV event frequencies j thus computed for each product line were normalized to that of the l benchmark plant taken as 100.

4 1

The reduction of spurious operation of relief valves was estimated based .

onoperatingexperienceandengineeringjudgment.

i 4.3 Probability of 5/RV's to Stick Ocen i

For comparing the various valves, the Three Stage Target Rock valve was taken as the benchmark valve with an assumed normalized factor of 1.0 l for probability to stick open when challenged. Similar factors for

'l other types of valves were obtained as described belcw.

1 23

4.3.1 Two Stage Target Rock .

A detailed review of all the 50RV events in operating plants was made, and the failure modes associated with the Three Stage Target Rock valve were tabulated. Then based on a study of the Two Stage valve design, an assessment was made of all the failure modes that are. climinated by the l Two Stage design. Consideration was also given to any new failure modes f- which might develop in going from the three stage to the two stage i . design. With this information, a relative probability factor of 0.50 l was assigned for the Two Stage valve to stick open, when challenged.

4.3.2 Dresser Electromatic l Based on a review of operating experience and engineering judgment, the l

1 Dresser valve was assigned a factor of 0.25 for its relative probability

! to stick open, when challenged.

l 4.3.3 Crosby and Dikkers l

The actual experience with the Crosby & Dikkers valves is too limited to l be used for estimating the relative 50RV probability factor. However, since those valves are direct acting (unlike the Three Stage Target Rock Valve which is pilot operated) seat leakage is not likely to be a signifi-cant concern as in the case of Target Rock valves. Based on valve qual'ification test data and limited operating experience, .a factor of

] 0.125 was assigned for their relative probability to stick open, when challenged.

! 4.4 Estimation of S/RV Challences . . . . .

j The total number of 5/RV challenges expected over the design life of a j plant was estimated as' described below for each BWR product line. The total S/RV challenges during a plant lifetice was taken to be the summation of the product of the frequency of various design transients and the estimated number of valve openings per occurrence of a transient. These value:; were then normalized to the benchmark plant whose'value was taken as 100.

l 4.4.1 Frequency of Transients It was assumed that each plant will experience the same number of various transients as were considered in its design basis. To estimate the l impact of various design improvements on the frequency of transients, --

data from operating BWR plants spanning 120 reactor years of operation and approximately 1400 reactor scram events (which include 720 S/RV challenge events) were investigated. By analyzing the data an estimate l was made of the percentage by which various transient event frequencies l would be modified if each of the candidate modifications discussed in Section 3 were impicmented. This estimate was used to modify tha frc-j quency of design basis transients. 1 4.4.2 Total Number of Valve Openings

The total number of valve ocenings for various transients was computed by using the General Ele S. ic long-term thermal hydraulics model (SAFE i and REDY codes).

There are many operational transients which can result in a pressure

! rise in the reactor vessel. The safety / relief valves will open if

! necessary to prevent the pressure from exceeding allowable limits. For l most of these events, the safety / relief valves will open only once.

l However, there are several types of transient evants which can result in a closure of the main steam isolation valves. Although a scraa occurs ,

immediately when the isolation valves close, the reactor continues to I generate steam due to decay heat. The safety / relief valves are then the primary means of reacter pressure control. One or more valves may cpen s with the initial pressurization following MSIV closure. These safety / relief j

valves will open when their pressure setpoints are reached and will j discharge steam to the suppression pool until the vessel pressure decreases to the closure setpoint of the valve. Reactor pressure will then increase

{

~again until the lowest safety / relief valve's opening setpoint is reached.

In most instances, only one S/RV will open on subsequent actuations. If j no operator action occurs, one valve will continue to cycle open and closed. The total number of safety / relief valve lifts is thus based on 25

l three factors--the number of transient events which result in cpening of )

the safety / relief valves, the number of valves which cpen in the initial l 1

pressurization, and the number of cycles which subsequently occur.  ;

4.4.3 Discussion of Assumotions .

Following are the key bases and assumptions used in the analysis.- .

i l-4.4.3.1 The frequencies of transient events are based ~upon the EUR/S

. design document for plant duty requirements. Overall the estimated number of relief valve openings based on the design transient frcq::ency differed only by 13 to 21% from the estimate based on transient frequency.

actually experienced by the plant. Since these numbers were ured cnly

! to determine the relative contribution of various modifications, this I difference of 13 to 21% is not significant.

4.4.3.2 The maximum specified relief valve reclosure setpoint is used since a smaller difference between the opening and closing setpoints results in a larger number of cycles. A 25 psi blowdown per relief valve opening is used for pilot operated valves such as Target Rock and I Dresser, and a 50 psi blowdown is used for direct-acting valves such as Crosby and Dikkers.

l l

4.4.3.3 Initial plant operating conditions cre at 105% nuclear boiler

?ated steam ficw.

i

l 4.4.3.4 BWR/2 and 3 plants equipped with isolation condensers are

' assumed to be capable of avoiding relief valve cycling after the initial relief valve discharge due to a transient.

) ,

4.4.3.5 For isolation transients, subsequent single S/RV discharges continue for 30 minutes.

I i

26

4

5. RES'JLTS . .

5.1 Expected SORV Frequency The expected SORY event frequency (normalized to th'e benchmark plant) for some of the most effective modifications are shown in T.able 5.1.

The following conclusions can be reached from this table.

I 5.1.1 BWR/2, BWR/3 with isolation condenser, BWR/5 and BWR/6 plants f

l :. are estimated to have a SORV frequency which is a factor of 10 less than .

the benchmark plant (BWR/4).

l 5.1.2 BWR/4 plants can reduce the SORV frequency by a factor of ten or more by implementing selected modifications from Section 3.

l 5.1.3 BWR/3 plants without isolation condensers can reduce the 50RV i frequency by a factor of ten or more by implementing selected modifi-l cations from Section 3. ,

! 5.1.4 The effect of isolation condensers on plants not so equipped is shown for comparison even though they are not practical for backfit application. The effect of high steam bypass capability (here, 110%) is i shown because some plants are so equipped.

l 5.1.5 The relative impact of each candidate modification en the S/RV challenge reduction alcng is shown in Table 5.2. It should be noted .

that more than one candidate modification could reduce S/RV challenges by addressing a common characteristic; therefore the percentage reductior.s in S/RV challenge rates attributable to the candidate modifications are not necessarily additive.

5. 2 Exceeted Sourious Blowdewn Frecuency Reduction The expected reduction in frequency of spurious blowdowns alone through implementation of various candidate modifications is summarized in Table 5.3.

27 i

N TABLE 5.1 SORV EVENT FREQUENCY J

~

T TOTAL SORV EVENT FREQUENCY (NORMALIZED)

BWR/4 CANDIDATE w/3 Stage BWR/3 BWR/2/3 H0DIFICATION Target Rock without with Valve (Bench- Isolation Isolation

_. ______ i. mark Plant) Condenser Condenser BWR/5 EWR/G (A): None 100.' 78 8 8 G (B): Low Water Level 69 - 50 8 6

• Isolation Setpoint (C): Low-Low Set Relief or 44 29 8 6

• Equivalent Manual Action (D): (B + C) 35 21 8 5 *

(E): Low-Low Set R'elief or 11 7 2 N/A N/A Equivalent Manual Action .

+ 2 Stage Target Rock Valve (F): (C) + Early Removal of 22 15 4 N/A N/A Leaky 3-Stage Target Rock Valve (G): Low Water Level Isolation 9- 5 2 N/A N/A

+ Low-Low Set Relief or -

Equivalent Manual Action .

+ 2 Stage Target Rock Valve (H): (B) + (C) + Early Removal of 18 10 4 N/A N/A

! 3 Stage Target Rock Valve ,

(I): 110% Steam Bypasst 88 73 4

~

6 4 (J): 110% Steam Bypasst 22 18 2 N/A N/A

+ 2 Stage Target Rock. Valve (K): Isolation Condensert 23 8

• 3 4 (L): Isolation Condensert 6 2 2 N/A N/A

+ 2 Stage Target .

Rock Yhlve

• Already implemented t for comparison only -

N/A Not Applicable 'since these plants are equipped with Crosby /Dikkers valves.

NOTES: 1. This table shows the 50RV' event frequency reduction due to some of the most effective contributors.

2. SORV event frequencies (normalized) shown above are obtained by multiplying total S/RV challenges (normalized) in Table 5.2 by relative 50RV probability factor. In the case of Two Stage Target Rock Valves the benefit in reduction of spurious blowdowns (from Table 5.3) has been included above.

~

28

.x ,

TABLE 5.2 t i S/RV CHALLENGES TOTAL S/RV CHALLENGES (NORMALIZED) 8WR/4 >

CANDIDATE 2/3 Stage BWR/3 BWR/2/3 MODIFICATION Target Rock without 'with Valve (Bench- Isolation Isolation mark Plant) Condenser Condenser ,BWR/5 BWR/6 H:ne -

m- - 100 78 -

8, 63 47 40

• Low Water Level Isolation <

69 50 8 -

Sctpoint 8 49

• Low-Lou Set Relief or 44 29 -

Equivalent Manual Action Feedwater Runback 91 69 7 58 44

^

Reduce Surveillance Test Error ** 95 74 7 60 45 R; duce MSIV Test Frequency"* 98 76 8 . 62 46 Feedwater Control System 89 68 7 56 45 Modification -

Feedwater System Improvement ** 97 75 8 62 47 Turbine System Improvement ** i BS 67 6 53 38

~

Analog Transmitter / Trip Unit ** 97 75 8 . 61 46 Improved Recirculation Flow 96 73 8 60 46 Control .

8-

• 27 34 Isolation Condensert ~

23 110% Steam Bypasst . 88 73 4 47 28 "Already implemented C*See Note 3 .

t for comparison only NOTES: 1. To obtain SORV event frequency (normalized) multiply the values in the table by relative SORV probability factor.

2. Relative 50RV probability factor for varicus valves are as follows:

3 Stage TR Valve: 1.0 Dresser Electromatic: 0.25 2 Stage TR Valve: 0.50 Dikkers & Crosby: 0.125

3. The benefits due to various candidate acdifications are not additive except where noted by ^*.

4. Values shown above may vary from plant to plant depending upon utility operating practice.

29

l TABLE ~5.3

REDUCTION IN SPURIOUS BLOWDOWN EVENT FREQUENCY D '

- Percentage ' '

l Reduction in

,,,IORV Events Applicability Candidate Modification Eliminate S/RY Ground Faults 1-2% All Plants i

Improved Pneumatic Supply 2-3% Plants with Target Control System Rock valves only Revise Spring Setpoint to . EE Plants with Target Increase Simmer Margin Rock valves only More Stringent Leakage Criteria 40-60% Plants with 3-Stage

& Early Replacement of Leaking Target Rock Valves

, Valves Only l

-Replace Valve Topworks with 40-60%* Plants with 3-Stage Two-Stage Design Target Rock Valves Only i'

• This modification also reduces the probability of the valve to stick open.

See Table 5.1 for the total impact on 50RV event frequency.

1 .

- n--- --

. _ _ _ _ _ _ _ _ .. . 3 0 _ _ _

6. CONCLUSIONS ,

i Adequate core cooling is maintained in a BWR following an 50RV event

. even under degraded conditions. It follows, then, that reduction of of frequency of SRV events is not of great concern from the standpoint of assuring adequate core cooling. ,

It is concluded that BWR/2, BWR/3 with isolation condenser, BWR/5 and BWR/G plants are expected to have a 50RV frequency which is a factor of I at least ten below that for the benchmark plant. The use of selected niodifications from a list of candidates can produce a-factor ef -t:n-reduction in stuck open relief valve event frequency for BWR/4 plants i and BWR/3 plants without isolation condenser. It should be noted l that additional candidate modifications may exist which could reduce SORV event frequencies but have not been addrcssed in this report.

e 1 .

J i

9 1

i 31 _ _ ___. _ _

7. REFEREllCES -

i

1. HEDO-24708A (December, 1980), " Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors," Sections 3.1.1,

3. 2.'1.

2. NE00-24708A, Section 3.2.1.

~

3. ' NEDO-24708A, Section 3.2.2.

4. HE00-24708A, Section 3.5.2.1.

5. BWR Emergency Procedure Guidelines, Rev. 1 (prepublication form),

submitted January 31, 1981.

e I

4 e

e O

e o

e S

O

APPENDIX A Participating Utilities .

~ '

, NUREG-0737 II.K.3.16 l

This report applies to the following plants, whose Owners participated in the report's development. -

Utility Plant .

Boston Edison Pilgrim 1 Carolina Pcwer & Light Brunswick 1&2 Commonwealth Edison . LaSalle 1&2, Dresden 2-3, Quad Citias 1&2 Georgia Power Hatch 1&2 l Iowa Electric Light & Power Duane Arnold l Jersey Central Power & Light Oyster Creek 1 l Niagara Mohawk Power Nine Mile Point 1&2 Nebraska Putlic Power District Cooper Northeast Utilities Millstone 1 -

Northern States Power

• Monticello l Philadelphia Electric Peach Bottom 2&3, Limerick 1&2 Power Authority of the State of FitzPatrick New York Detroit Edison Enrico Fermi 2 Long Island Lighting Shoreham Mississippi Power & Light Grand Gulf 1&2 Pennsylvania Power & Light Susquehanna 1&2 ,

Washington Public Power Supply Hanford 2 .

System ,

Cleveland Electric Illuminating Perry 1&2 -

Houston Lighting & Power Allens Creek

~

Illinois Power C1Tnton Station 1&2 Public Service of Oklahoma Black Fox 1&2 Vermont Yankee Nuclear Power Vermont Yankee -

Tennessee Valley Authority Brcuns Ferry 1-3; Hartsville 1-4; Phipps Bend 1-2 Gulf States Utilities River Bend 33

LILCO, June 14, 1982 ATTACHMENT V to LILCO Testimony on SC 28a(vi) and SOC 7A(6):

Letter of March 31, 1981 from D. B. Walters to Darrell G. Eisenhut, "BWR Owners' Group Evaluations of NUREG-0737 Requirements II.K.3.16 and II.K.3.18"

~'t f s (4

dC./ . fYUN:.h.. r} b% n#t.itI E N Dn...--,-,3 D2YtdU$ 0 E C.:: s O: * --

PO 80r 1551 e Relegn Nor: Ceae -e 27eC2 e (o19; S36.c584 BWROG-8134 March 31,1981 U.S. Nuclear Regulatory Commission Division of Licensing Office of Nuclear Reactor Regulation Washington, D.C. 20555 Attention: Darrell G. Eisenhut. Director

~

SUBJECT:

BWR Owners' Group Evaluations of NUREG-0737 Requirements II.K.3.16 and II.K.3.10 Gentlemen:

This letter transmits feasibility studies performed by the BWR Owners' Group for the following NUREG-0737 items:

I I . K. 3.16 : Reduction of Challenges and Failures of Relief Valves - Feasibility Study and System Modification II . K. 3.18 : Modification of Automatic Depressurization Systen Logic - Feasibility Study for Increased

Diversity for Some Event Sequences 1

The submittal of an Owners' Group position developed in response to an NRC l requirement does not indicate that the Owners' Group unanimously endorses that position; rather, it indicates that a substantial number of members believe the position is responsive to the NRC requirement and adequately satisfies the requirement. Each member must formally endorse a position so developed

• and submitted in order for the position to become the member's oosition.

General Electric will ' provide sixty (60) additional copies of the attachment in a separate mailing.

Please contact me at (919) 836-6584 if you have any questions concerning the enclosed information.

l l Sincerely, hw[ ,

hCLbj D. B. Waters, Chairnan 9W9 Cm ers' Gro x

, DBW:PWM:na l Erclosures cc: SA Cwners' Geer M. U. Hodges (NRC) c." ":-r' c " W J. A. 01shinksi r .,

(!RC) e c. . .. ...3

'l LILCO, June 14, 1982 ATTACHMENT VI to LILCO Testimony on SC 28a(vi) and SOC 7A(6):

Drawings of Target Rock Two-Stage Pilot-Actuated Safety / Relief Valve Figure A: Schematic of Targe.t Rock Two-Stage Pilot-Actuated Safety / Relief Valve (Valve Closed)

Figure B: Schematic of Internal Details, Target Rock Safety / Relief Valve Two-Stage Horizon tal Design (Valve Closed)

- Figure C: Target Rock Two-Stage Pilot Operated SRV (Horizonta l Design) Valve Schematic (Open Position).

ATTACHMENT VI , FIGURE A i

i SCHEMATIC OF

" TARGET ROCK" TWO STAGE,

PILOT ACTUATED, SAFETY / RELIEF VALVE l (VALVE CLOSED) l l ELECTRO-PNEUMATIC OPERATOR

~

! i p _ _- -

1 ELECTRO-PNEUMATIC < N i OPERATOR ASSEMBLY ,

9~g e .

I w Q 4>?;

p m q= l =< -

I - SOL ANL F e ""

! AIR CONTROL VALVE e qf), e ASSEMBLY l ' I *E PILOT VAINE VENT -MAIN DISC "

TO. DISCHARGE- - -

. .d*

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d * "

t g- L .

. *% w g;y ==

9 ,. _; r. - v 1. - __-: w.-=-<--

3l r e d .1 m . w m.. ms e =- A ~

PILOT P- VALVE -

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METER - _

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9

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l1' VALVE WEIGHT: 1100/1150 lbs.

VALVE SIZE:,

INW HEIGHT APPROX. 36 INCHES I 9 ' LENGTH APPROX. 30 INCHES 6 INCH DIAMETER WIDTH APPROX. 18 INCHES i

l l

l

i ATTACFEENT VI , FIGURE B l l

i l LECTRO-PNEUMATIC OPERATOR

- SOLENCID AIR CONTROL VALVE I _

d E

_ i 1

1 -[] Q , ,

g< SPRING LOAD e-_,y g .. 9. -- ADJUSTING RING l lb ~I ~

w "

5 l\/

L _ _.-

=

! SET POINT D

D D f.. SPRING J

O w

__bEj'D I u _

'1 p SPRING WASHER e.: ; l J  % &_

MAIN

~,,

I l- ,

PILOT L f VALVE STAGE r

LOT s PILOT DISC VE pnwnstgg l O i t ,

CHARGE lj l lg I bN s q l

.l y BLOWDOWN DISC f

4 3 BLOWDOWN ORIFICE MAIN PISTON SCHEMATIC OF GIASING SPRING INTERNAL DETAILS

" TARGET ROCK" SAFETY / RELIEF VALVE TWO-STAGE HORIZONTAL DESIGN (VALVE CLOSED) l

ATTACHMENT VI , FIGURE C SOLENDIO -

VALVE ASSE MB LY AIR OPER ATOR ASSEMBLY I i 1 1 I I 5

PILOT '

• BONNET /

OISCH AR GE

. . . . p SPRING PORTING

, , , [ e SUB ASSE MB LY e

MAIN PISTON 4 4 CHAMBER TO 4 4 PILOT VENT PORT l

i J

Q ,

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INLET

~,4[. J ! SYSTEM IHIGHI PRESSURE sm uNNNN OlSCH ARGE ILom PRESSURE i <suu l TARGET ROCK TWO-STAGE PILOT OPERATED SRV (HORIZONTAL' DESIGN)

Valve Schematic (Open Position) 1

i LILCO, June 14, 1982 ATTACHMENT VII to LILCO Testimony on SC 28a(vi) and SOC 7A(6) :

Schematic of Target Rock Three-Stage Pilot Operated Safety / Relief Valve (Models 7367F/746TF) l i

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6 ELECTRO-PNEUMATIC OPERATOR 4

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SONNE]T]/ j PtLOT STAGE 0iSC SELLOWS LEAMACE ICLOSEDI AL AAM PQRT

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l OUTLET I

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lNLET M@xA 5 FLUSD PREssunE i

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I Schamatic of Target Rock Three-Stage l Pilot Operated Safety / Relief Valve Olodels 7367F/74677) l I

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