ML20040A939

Testimony of DG Bridenbaugh & Gc Minor Re Contention 12. Power Operated Relief Valves & Block Valves & Valve Instruments,Controls & Structures Should Be Classified as safety-related.Certificate of Svc Encl

ML20040A939
Person / Time
Site:	Diablo Canyon
Issue date:	01/11/1982
From:	Bridenbaugh D, George Minor CALIFORNIA, STATE OF, MHB TECHNICAL ASSOCIATES
To:
Shared Package
ML20040A930	List: ML20040A929 ML20040A932 ML20040A935 ML20040A939
References
ISSUANCES-OL, NUDOCS 8201220325
Download: ML20040A939 (12)

Text

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4 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

)

In the Matter of )

)

PACIFIC GAS AND ELECTRIC COMPANY ) Docket Nos. 50-275 0.L.

) 50-323 0.L.

(Diablo Canyon Nuclear Power )

Plant, Units 1 and 2) )

)

PREPARED DIRECT TESTIMONY OF DALE G. BRIDENBAUGH AND GREGORY C. MINOR ON BEHALF OF GOVERNOR EDMUND G. BROWN JR.

REGARDING CONTENTION 12 January ll, 1982 820122 > 3pf

PREPARED DIRECT TESTIMONY OF DALE G. BRIDENBAUGH AND GREGORY C. MINOR l REGARDING CONTENTION 12

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I. INTRODUCTION i

! 1. My name is Dale G. Bridenbaugh. A statement of my qualifications and experience has previously been provided to this Board as part of my testimony on Contention 10 and in Attachment A to that testimony.

2. My name is Gregory C. Minor. A statement of my 1

) qualifications and experience has previously been provided to i

l this Board as part of my testimony on Contention 1 and in

Attachment B to that testimony.

l j II. STATEMENT OF CONTENTION -

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! 3. The purpose of our testimony is to respond to

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Contention 12 as admitted by the Board as follows:

I i Proper operation of power operated relief i valves, associated block valves and the I instruments and controls for these valves

) is essential to mitigate the consequences j of accidents. In addition, their failure

can cause or aggravate a LOCA. Therefore, 1

r i 1/ ASLB Memorandum and Order, September 30, 1981. On September 21, the Commission directed the Licensing Board to include in the full power proceeding Joint.

Intervenors' low power Contention 12.

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these valves must be classified as com-ponents important to safety and required to meat all safety,-grade design ,

criteria.

Further, the Appeal Board's order of December 11, 1981, expands Contention 12 to include "the testing and verification of these same cor.ponents" since " testing and verification of these com-ponents is an integral part of the qualification process."-2/

Thus, the adequacy of the qualification process, including the

's adequacy of the EPRI testing program, is included in the expanded scope of Contention 12. The results of our review of some of the important matters encompassed by this Contention are summarized in the following paragraphs.

III. DISCUSSION OF ISSUES III.A.: The NRC's Criteria for Equipment Classification are Confused

4. There is confusion as to the meaning of terms used to describe the safety significance of structures, systems, and components in nuclear power plants. The NRC issued a memorandum-3 2/ ASLAB Order, December 11, ' 1, 'p . 3.

-3/ Memorandum from H. R. Dentun to All NRC Personnel, Novem-  ;

ber 20, 1981,

Subject:

" Standard Definitions for Commonly-used Safety Classification Terms."

which provided definitions of the most often used safety classi-fication terms as follows (see Attachment A for the full text) :

Important to Safety:

Those structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public. .

Safety-Related:  !

Those structures, systems, or components designed to remain functional for the SSE (also termed

'saf ety fea tures ' ) necessary to assure required safety f unctions, i.e.,:

(1) the integrity of the reactor coolant pressure boundary; (2) the capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential off-site exposures comparable to the guideline exposures of this part.

Safety-Grade:

Term not used explicity in regulations but widely used/ applied by staff and industry in safety review process.

Equivalent to " Safety-Related," i.e., both terms apply to the same subset of the broad class "Important to Safety."

The writing of Contention 12 preceded the issuance of the clarifica tion document. If Contention 12 had been written using the definitions of the Denton memo, the term

" safety-related" would have been used instead of "important to safety." From our review of the Applicant's submissions, it is

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j unclear to us how the Applicant is using the safety classification j .

terms and how it defines "important to safety." .

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j III.B.: The Safety Significance or the PORVs' and Block Valves' Functions Justifies i

Safety-Related Classification i

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5. The design of Diablo Canyon includes 3 PORV's and 3 i

associated block valves. Two of the relief valves are described by PG&E as "important to safety" and, the third is not, having i

been added to provide capability for 100% load rejection without 4/

reactor trip.~ The three block valves are also described as j

{ "important to safety."

I i 6. The PORV's and/or Block Valves perform several functions l which have safety significance along the lines of one or more of f

j the definitions in paragraph 4. These functions are:

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! a. Maintain integrity of the primary pressure

{ boundary.

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b. Provide pressure relief for Low Temperature

Overpressurization conditions.

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c. Reduce the number of challenges to the 5

safety valves.

j d. Reduce the number of challenges to the ECCS.

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e. Provide a bleed capability during the feed-and-i bleed mode of operation to remove decay heat from the core. 5/

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~4/ PG&E-response to Interrogatory No. 46. (PG&E's response

) includes the term " local rejection" which is interpreted as a typographical error for " load rejection".)

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~5/ As used in the TMI-2 accident and as referred to in NUREG-l 0578, Sec. 2.2.1 and page A-1.

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Each of these functions is consistant with the definitions of i "important-to-safety," and the first two functions are also l

consistent with the definitions of " safety-related."

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7. The FSAR is vague as to the safety' classification of j4 the PORV's, Block velves, and their circuits and controls. The Applicant has stated that the qualification. level of the three POnV's and their circuits are not all identical. -

However, documents which the operator relies on for guidance in operating the plant during emergency conditions (Emergency Operating Pro-

, cedures) and deciding on an acceptable plant configuration r (Diablo Canyon Technical Specifications) provide no evidence

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of differentiation between the greater or less "q5alified'~~

valves or associated equipment."

8. The Block valves and/or PORV's are called upon to be operated or checked for misoperation in several of the Emergency

. Operating Procedures. For example, EOP-20 calls for checking the PORV's as a possible source of excessive leakage from the coolant i system (i.e., a small LOCA). EOP-38 (ATWT) describes.the need for automatic opening of the PORV's and checking later to see that they are not stuck open in the event of a pressure decay and coolant loss. EOP-2 describes the actions to prevent challenges j

to the pressurizer safety valves in the case of loss of secondary l 1 coolant. It too mentions that the transient may cause the PORV's I 3

to open and' requires that their resetting be checked, thus insuring

! against a small LOCA in the primary coolant.

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9. The emergency operating procedure for Emergency Shutdown (OP-22, describes copditions where the use of a PORV/BV combination may be needed to depressurize the primary loop so the safety injection pumps may be used for boration.

The PORV would be opened and the block valve used for throttling the flow. The procedure does not restrict the operator to any particular PORV nor does it identify a safety-grade alternative component to accomplish the task. Thus any of the PORV/BV combinations should be able to accomplish this safety-related task.

10. Emergency Procedure OP-13 on Malfunction of Reactor Pressure Control System calls for use of a PORV/BV combination j as a back-up pressure control technique. The same procedure I identifies techniques for finding stuck open PORV's which j

1 may be leaking coolant and exacerbating normal pressure control

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methods.

11. Section 3.4.9.3 of the Diablo Canyon Technical Specifica-l tions requires that two PORV's be operable during Hot Shutdown .

(Mode 4) conditions for overpressure protection. There is no guidance, however, to the operator to choose the more qualified'

-6/ It also notes that a stuck open PORV is designated as an

" UNUSUAL EVENT" and requires notification of offsite personnel per the emergency procedures (EOP General Appen-dix 2 - Notification of Offsite Personnel in the Event of j an Emergency). H l

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s PORV's. Thus, all PORV's should be qualified to the same level .

or the operators' EOP's shou.l'd restrict which two valves are to be used.

12. During operating modes 1, 2 and 3 (Power Operation, Startup, and Hot Standby), Tech Spec Section 3.4.4 requires that each PORV must be operable or isolated.by an operable block valve which is then deenergized. For these cases either the PORV's or their associated block valves are relied upon to protect the integrity of the primary pressure boundacy. However, accord-ing to the Technical Specification, it is possible to block the two higher-qualified valves and rely only on the lesser-qualified valve and its associated controls. We feel there should be instructions to the cperator to prevent this situation if the difference in valve classifica tion continues to exist.

III.C.: The Fact that Diablo Canyon Has More PORV's than Other Plants is Not Necessarily an Advantage

13. Since it is permissible to operate with one or more ,

of the PORV's isolated by their block valves, and there are no restrictions on which valves are isolated, it is possible that the PORV with the lesser classified components would be the only valve operable at a time when PORV operation was called upon by a transient or accident.

14. The fact that the Diablo Canyon design has more valves than most plants is commendable but it is not always a virtue.

The addition of the third valve may help the reactor ride through a load rejection transient, thus preventing a challenge to the protection system, but it also creates additional failure points which could result in a small LOCA, additional common mode failure mechanisms, and the possibility of systems interaction which could impact other safety-related functions.

III.D.: Qualification of PORV's and BV's is Incomplete,

15. Proper safety classification of the PORV/BV and their controls and instruments should insure proper design and quali-fication for worst case conditions and plant-specific evalua-tions.-7/

16. However, the qualification of the Diablo Canyon PORV's and Block valves is incomplete. The BV's have not been fully tested, and there apparently are no plans for further testing.-8/ The full range of conditions, including ATUS, has not

-7/ See June 16, 1981 PG&E Memorandum relating to EPRI safety valve testing, raising a potential issue regarding Diablo Canyon safety valves. Such testing designed to ensure qualification of valves will increase reliability of and confidence in Diablo Canyon systems.

-8/ See NRC Staff Response to Joint Intervenors' Second Set of Interrogatories, No. 61 (e) .

been tested and the plant-specific analysis has not been prepared to cover Diablo Canyon's design of PORV/BV's and their components, systems, and structures. Thus there can be no assurance that the configuration meets GDC 2 and 14. Also, the scheduled completion of the valve tests and the plant-specific analyses have been delayed until July 1, 1982.~9/This may not be soon enough to satisfy the terms of the Low Power Testing License, Section 1,

p. 6.

IV. CONCLUSIONS

17. Based on the functions and required operations of the PORV's and Block valves, as described above, and according to the NRC definitions of safety terms, the PORV's/BV's and their instruments, conex)ls and structures, should all be classed as

" safety-related." -10/

18. There are insufficient test data and plant-specific analyses to show that the Diablo Canyon PORV/BV's and associated equipment and structures have been properly qualified.

_9/ Ibid, No. 61 (d) .

10/ " Safety-grade" is also appropriate since it is defined as equivalent to " safety-related" by the NRC.

ATTACHMENT A

/ga ttog'o, UNITED STATES jT .. , y j NUCLE AR REGULATORY COMMISSION

~

'. **S g WASHINGTON. D. C. 20555 sh%, wE f J

..... f NOV 2 01951 J:

MEMORANDUM FOR: All NRR Personnel  ; ,

FROM: Haroid R. Denton, Director Office of Nuclear Reactor Regulation-

SUBJECT:

STANDARD DEFINITIONS FOR COMMONLY-USED SAFETY CLASSIFICATION

. TERMS' l

Litignion of one of the principal issues in the TMI-1 Restart Hearing brought l to light the fact that 'there is not complete consistency among all elements of 1 the HRR staff in the application of safety classification terms used frequently in the conduct of NRR's safety review and licensing activities. More specifi-cally, it appears that terms "important to safety," " safety grade," and " safety-related" have been used at times interchangeably, or in ways not completely consistent with the definitions and usage of such terms in the regulations, and which d'o not fully reflect the intent of the regulations o.r current licensing practice.

Efforts have been underway for some months now to develop guidance for the

_ consistent usage of these terms. These efforts have included: (a) review of ,

a large number of Reg Guides and SRP's, in conjunction with parts of the regula-tions upon which they are based, for consistency in the application of safety classification terminology, (2) extensive discussions among cognizant NRR, RES (Stds. Devel.) and ELD representatives regarding proper interpretation and application of such terms, including consideration of alternative " standard" definitions and (3) consultation with the cognizant ACRS Subcommittee regarding these matters, and consideration by the full ACRS as well.

As a result of these efforts, I am endorsing and prescribing for use by all' NRR personnel the standard definitions set forth in the enclosure to this letter.

It should be noted that in connection with long-term efforts to develop means for ranking reactor plant systems with respect to degree of importance to safety, and in. connection with related efforts to develop a graded Q.A. approach in reactor licensing, the general question of safety classifications and safety classification terminologies will be reexamined; and this could result in changes to the defini-tions set forth in the enclosure or perhaps in development of a completely new scheme in this regard. For the time being, however, the definitions in the en-closure should be considered " standard" and should be applied consistently by all NRR personnel in all aspects of our safety review and licensing activities and should be appropriately reflected in our regulato.y guidance documents.

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U1 tiRR Personnel .

pen -

It is expected that minor editorial revisions will have to be made to some

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axisting Reg Guides and SRP's in order to make their wording corsistent with these definitionsf You should review the regulatory guidance decuments within your purview in this regard and recommend the necessary changes; it is not expected that this will involve extensive revision efforts. I want to make clear that my interest here is only in establishing consistency in the langtiage used by all cognizant groups within NRR in expressing our technical require.nents.

It is not my intention by this action to dictate new technical requirements, to modify existing technical requirements, or to broaden the exist.ing scope of

lRR licensing review. .

de N n Harold R. Denton, Director ' '

Office of fluclear Reactor Regulation

Enclosure:

Definition of Terms l

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DEFINITIOfi OF TERMS Important to Sa'f ky a

Definition - From 10 CFR 50, Appendix A (General Design Criteria) - see first paragraph of " Introduction."

"Those structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the. health and safety of the public"." ,

) Encompasses the bf cad class of plant features, covered (not necessarily exolicitly) in the General Design Criteria, that contribute in important way to safe operation and protection of the pLblic in all phases and aspects of facility operation-(i.e., normal operation and transient control as well as accident mitigation).

• Includes Safety-Grade (or Safety-Rela ted) as a subset.

Sa fe ty- iela ted o

Definition - From 10 CFR 100, Appendix A - see sections III.(c), VI.a.(1), and VI.b.(3).

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Those structure, systems, or components designed to remain functional for the SSE (also termed ' safety features') necessary to assure recuired safety functions, i .e. :

(1) the integrity of the reactor coolant pressure boundary; (2) the capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential off-site exposures comparable to the guideline exposures or this part.

3 Subset of "Important to Safety" Regulatory Guide 1.29 providesan LWR-oeneric, function-oriented listing of

" safety-related" structures, systems, anc components neeced to provide or perform' required safety functions. Additional information (e.g., NSSS type, BOP design A-E, etc.) is needed to generate the complete listing of safety-related SSC's for any specific facility.

Note: The tem " safety-related" also appears in 10 CFR 50, Appendix B (0.A. Program Requirements); however, in that context it is framed in somewhat different language than its definition in 10 CFR 100, Appendix A. That difference in language between the two appendices has contributed to confusion and misunderstanding regarding the exact meaning of " safety-related" and its relationship to "important to I

safety" and " safety-grade." A revision to the language of Appendix 8 has been proposed 'to clarify this situation and remove any ambiquity in the meaning of these terms.

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! Sa fe ty-Gra de fe Term not used explicitly in regulations but widely used/ applied by staff (

and irdustry in safety review process. "

e Equivalent 'Io " Safety-Related," i.e., both terms apply to the same subset ,

of the broad class "Important to ,56fety." '

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3 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

! )

i In the Matter of )

)

{ PACIFIC GAS AND ELECTRIC COMPANY ) Docket Nos. 50-275 0.L. .

) 50-323 0.L.

]

(Diablo Canyon Nuclear Power )

< Plant, Units 1 and 2) ) ,

)  !

) i i

1 CERTIFICATE OF SERVICE l

, I hereby certify that on this 13th day of January, 1982, I have served copies of the foregoing JOINT INTERVENORS' RESPONSE I IN OPPOSITION TO NRC STAFF AND PACIFIC GAS AND ELECTRIC COMPANY MOTIONS FOR

SUMMARY

DISPOSITION, together with attached exhibits, mailing them through the U. S. mails, first class, postage l

i prepaid.

• Admin. Judge John F. Wolf,

• Docket & Service Branch Chairman. Office of the Secretary Atomic Safety & Licensing U.S. Nuclear Regulatory l Board Commission U.S. Nuclear Regulatory Washington, D.C. 20555 Commission Washington, D.C. 20555 William Olmstead, Esq.

Marc R. Staenberg, Esq.

• Glenn O. Bright Edward G. Ketchen, Esq.

Atomic Safety & Licensing Office of the Executive Legal j Board Director - BETH'042 U.S. Nuclear Regulatory U.S. Nuclear Regulatory i Commission Commission Washington, D.C. 20555 Washington, D.C. 20555

• Express Mail 1

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• Dr. Jerry R. Kline Nancy Culver Atomic Safety & Licensing 192 Luneta l Board San Luis Obispo, CA 93401 U.S. Nuclear Regulatory  !

4 Commission

! Washington, D.C. 20555 ,

l

!i Mr. Fredrick Eissler Malcolm H. Furbrush, Esq.

I Scenic Shoreline Preservation Vice President and General

! Conference, Inc. Counsel  !

3 4623 More Mesa Drive Philip A. Crane, Esq.

t Santa Barbara, CA 93105 Pacific Gas & Electric Company ,

2 P. O. Box 7442 i i Sandra A. Silver San Francisco, CA 94106  ;

Gordon Silver

]j 1760 Alisal Street Arthur C. Gehr, Esq.  ;

j San Luis Obispo, CA 93401 Snell & Wilmer i 4 3100 Valley Center  ;

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I I P. O. Box 1178 i Oklahoma City, OK 73101 Carl Neiburger  ;

) Telegram Tribune

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3216 N. Third Street San Luis Obispo, CA 93402 d

Suite 202 l i Phoenix, AZ 85012 Byron Georgiou, Esq.  ;

i Legal Affairs Secretary to Janice E. Kerr, Esq. the Governor r Lawrence Q. Garcia, Esq. State Capitol Building l

, J. Calvin Simpson, Esq. Sacramento, CA 95814  !-

{ California Public Utilities  !

Commission Lawrence Coe Lanpher, Esq.

J 5246 State Building Hill, Christopher & Phillips -

l 350 McAllister Street 1900 M. Street, N.W. f San Francisco, CA 94102 Washington, D.C. 20036  !

! l i MHB Technical Associates i 1723 Hamilton Avenue .

Suite K j San Jose, CA 95725 ,

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t CtA.bt.( ~

A. S. VARONA i

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