ML20072H700

Response Opposing Util & NRC Motions for Summary Disposition of Contentions I.A.2,I.B.1,I.D.2,I.G,II.B.3 & II.B.4. Some Contentions Withdrawn Since Discovery Satisfied Concerns.W/Certificate of Svc.Related Correspondence

ML20072H700
Person / Time
Site:	Seabrook
Issue date:	03/24/1983
From:	Curran D HARMON & WEISS, NEW ENGLAND COALITION ON NUCLEAR POLLUTION
To:	Atomic Safety and Licensing Board Panel
References
ISSUANCES-OL, NUDOCS 8303290536
Download: ML20072H700 (45)

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s ~ ,.,0' RELATED CORRESPONDENCR March'24', 1983. DOCKETED-USNac p

UNITED STATES OF AMERICA 13 NiiR 28 . Pl2'd5 NUCLEAR REGULATORY-COMMISSION BEFORE THE ATOMIC SAFETY-AND LICENSING 30AkD I In the Matter of .)

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PUBLIC SERVICE COMPANY OF ) Docket Nos. 50-443 OL-NEW H AMPS!! IRE, et al .

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) 50-444.0L.

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(Seabrook Station, Units l and 2) )

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NECNP OPPOSITION-TO MOTIONS FOR

SUMMARY

-DISPOSITION

+ AND NOTIFICATION OF WITHDRAWN CONTENTIONS Applicants and the NRC Staff have filed motions for summary disposition pursuant to 10 CFR S 2.749 on 21 of NECNP's contentions in this proceeding. #ith the exception of the motions on emergency planning issues, they are answered below.

NECNP has dropped a number of contentions because the discovery process has satisfied our safety concerns. NECNP opposes, however, summary disposition of Contentions.I.A.2, I.B.1, I . D .2, I .G , and II.B. 3 and 4. Neither Applicants nor Staff have met the burden of proof that they must carry for summary

, disposition. 10 CFR S 2.7 32. For each of these issues, there remain unresolved issues of fact or law, and therefore summary judgment may not be granted. We note that where "significant

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health and safety issues" are involved, motions for summary disposition should be granted only if the licensing board "is convinced from the material filed that the public health and

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F303290536 830324 gDRADOCK 05000443 A

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safety'...-will=beisatisfactorily-protected." 1 Cincinnati ~ Gas:&

Electric (William H. Zimmer" Nuclear Station)l LBP-81-2,.13 INRC

-36,140f(1981). .

I'.A.2; Qualification of Electric Valve' Operators Applic'anbs have moved for summary disposition of NECNP Contention, I.A.2, which asserts that:"the Applicant has not complied with Commission' standards regarding qualification tests of electric valve operators installed inside the containment." NECNP-Supplemental' Petition-to Intervene at 8' (filed April 21~/ 1982).

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In support of their motion, Applicants assert that " All Class lE electric valve operators -installed inside containment will be qualified in -conformity with Reg. Guide 1.73_(Rev. O, 1/74); Reg.. Guide 1.891(Rev.-.0, ll/74)Tand'NUREG-0588. .NECNP-

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. does not disput'e- this- f actual assertion. . ' The Applicants' error

.is legal,~not factual. -Applicants are required to environmentally qualify not only;those electric valve operators which are " safety related" or " Class lE,' but all clectric valve'operat' ors which are "important to safety." 10 CFR S 50.49(a). This includes non-safety related valve operators "whose f aily:e under postulated environmental conditions could prevent satisfactory accomplishment of safety functions...by-the . sa fety-r elated equipment ." 10 CPR S 50.49(b)(2).

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As shown by Applicants' Table I.A.2-3 (attached)'the only classifications considered by Applicants in qualifying electric valve operators are " safety related" and " Class lE", a subset of safety related. There are a number of electric valve operators inside'the containment which are not qualified because they are not '" safety related" or " Class lE" .

Applicants have not mado the required determination, however, of whether they are "important to safety." The analysis performed by Applicants in determining which electric valve operators must be qualified consists only of a determination regarding which valve operetors are required to operate during an accident. . Applicants have not considered whether failure of any of the other. electric. valve operators could prevent satisfactory completion of safety functions. Applicants have not met the legal requirements for environmental qualification of equipment important to safety, and therefore summary disposition must be denied.

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I STATEMENT OF FACTS AS TO WHICII .TIIERE IS A MATERIAL ISSUE TO BE IIEARD

1. NRC' regulations at 10 CFR S 50.49 require the

- envit onmental qualification of 'all electric equipment important to safety;

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2. The. electric valve operators inside;the containment-include valve operators which-are neither classified as " safety -

related" nor " Class lE." See Table -I . A.2-3

3. Applicants have.only environmentally' qualified electric valve operators which. they consider' to be " safety related" or " Class 1E".

4. With regard to the other electric valve operators, Applicants have made no determination as to whether they are nonsafety'related components which are nevertheless important to safety because their failure could prevent the operation of safety functions.

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a-I.B.1 Qualification of Residual Heat ~ Removal Equipment-Applicants and the NRC Staff have moved for summary.

disposition on NECNP Contention I.B.1, which asserts that Applicants have violated GDC 4 and GDC.34 in that they have not environmentally qualified all systems that may be required to remove heat from the steam generators during an accident, including steam dump valves, turbine valves, and the steam dump control' system. Applicants state, in support of their motion for summary disposition, that the Seabrook RHR system does not require the use of the steam dump valves, turbine valves, and the steam dump control system to meet the RHR requirements of' GDC 34; and that the systems at Seabrook'that a'e r essential to perform or support the function of RHR are safety grade and environmentally qualified. .The NRC Staff also asserts that these systems need not be qualified because they are "not ,

r equir ed" to remove residual heat from the core during an accident.

NECNP does not dispute the facts asserted by Applicants and Staff. Those facts support a conclusion that Applicants and Staf f have misapplied the NRC regulations.

The qualification of " essential" or " required" systems alone does not satisfy the Commission's requirements for i

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environmental qualification, and therefore.does not meet GDC 4.1! Under 10 CPR S 50.49, Applicants must qualify:all electric equipment "important to safety," which includes not only required safety related equipment, but "non safety-related

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electric equipment whose' failure under postulated environmental conditions could prevent satisfactory accomplishment of. safety.

f unctions . . . by safety-related equipment." ~ 10 CFR S 50.49(b)(2). The Applicants have not addressed the issue of whether heat. removal systems such as the turbine valves, steam dump valves, or steam dump ~ control system, are "important to safety." They are therefore not entitled to st nmary judgment.

The NRC Staf f, in a footnote, states that the components in question do not constitute nonsafety-related equipment which is "important to safety," but provide absolutely no support for that assertion. The Staff's conclusion is neither repeated nor explained in the " Statement of Facts as to Which There is No Genuine Issue to be Heard" or the affidavit supporting the 1/ In the absence of compliance with GDC 4, Applicants cannot provide a. reliable heat removal system, because unqualified components which are important to safety must be assumed to fail. Therefore Applicants violate GDC 34.

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lmohion. b ,'WithoutUany; fait lual) support dor %his'; assertion ( w

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the~NRCfStaf f, hdsin'ot 'sustainedlihsliliarden"ofl proof / "an,d :

' summa'ryidispos'itibn - must be ' denied.~-: -

Furthermore,the. Staff'ssafety(EvaIuation~Reportstrongly

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supports a conclusion'that1 systems'whichtremove,decayfh' eat?

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during normal operation, such'as?the' turbine valves,; steam 1 dump valves, steam dump control system,' main-feedwater and s

i condensate systems, condenser steam dump valves, condenser, and f

circulating water system, are important to. safety and shouldL_be

' environmentally qualified. In asserting that these' systems-are

-not required for decay heat removal during-an accident, the.

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L Applicants and Staff implicitly assume the integrity of the steam generators. The_SER, however,; recognizes the strong possibility of tube-leaks in the. steam generators, a. recurrent-

, problem with Westinghouse ~ equipment. With regard to the L

[ integrity of'the-steam generator. tubes, Unresolved Safety Issue e

ls  : 2/ In= fact, in Lits a'nswers ito NECNP's finterrogat'or'ies,: th'e ~' ~~

NRC-Staffcstateduthat ' systems that: perform the1 function:of?

i residual-heat removal during'all plantf operating conditions ares

.~ ...important~to safety." (emphasis added)

<NRC Staff Response.

.to NECNP First ' Set of Interrogat'ories at [ll1( filed , December J28,.

1982). Those systemscinclude the' turbine valves,imain i Efeedwater and condensate systems, condenser steamLdump valves, condenser, and circulating water system, which are normally used~during orderly plant shutdown to remove decay heat. ~See i NRCEStaff Response to NECNP~First Set of Interrogatories at 8.

Ati the' time, the NRC Staf f ~ defined "important to safety" as

" structures, systems, and components that provide reasonable l

assurance that the facility can be operated without undue risk

to the health and safety of the public." NRC Staff Response to j~ NECNP Second Set of Interrogatories at 2 (filed January 21, 1983). TheLNRC Staff has not explained why it has apparantly

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R A-3, the Staf f 'was unable to find that "this f acility can be -

operated before the' ultimate resolution of these generic issues without endangering the health and safety of the_public."

Seabrook SER a t C-7. Such leaks could allow radioactive primary coolant to escape into the secondary coolant system'.

Because the steam generators cannot be relied upon as a

' barrier to the escape of radiation, other barriers which might ordinarily be considered secondary and unnecessary gain importance. Those barriers include the valves and systems used to remove heat from the steam generators during normal operation. In the presence of tube leaks in.the_ steam-generators, failure of these valves and systems could result in releases of radiation to the environment. For example, with leaking steam generators, Applicants could not rely on the atmospheric dump valves to remove heat from the steam generators during an accident, because the valves would release radiation into the environment. The Applicants would be forced to rely on other systems such- as the condenser steam dump valves, which would channel the steam to the condensers instead i of the atmosphere. If those valves failed, there would be no safe outlet for the decay heat from the steam generators.

NECNP has given an example of a likely situation in which equipment currently classified as nonsafety related, and currently unqualified,.could be relied upon during an accident. It is not NECNP's burden to prove that this equipment must be environmentally qualified. Rather, it is the J

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Applicants' and Staff's burden to prove that qualification is not required because the equipment is not "important to safety." Applicants' and staff's motions give no indication that the ef f ect of these components' failure on the ability of safety equipment to function was ever considered. Because they have misapplied the legal standard, they have not satisfied their burden of proof, and summary disposition of Contention I .B.1 must be denied to Applicants and Staf f. NECNP has attached the affidavit of Gregory C. Minor in support of this opposition.

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STATEMENT OF MATERIAL FACT ~AS TO WHICH THERE IS A GENUINE ISSUE TO BE HEARD

' 1. Applicants have only environmentally qualified those parts of the steam generator decay heat removal system which

.they consider necessary to_ remove heat'from the steam generators during an accident.

2. . Applicants have not performed.any analysis to determine which of the nonsafety-related decay heat removal systems ae "important to safety" in that their failure would prevent operation of safety related functions.

3. In the Safety Evaluation Report for Seabrook, the NRC Staff has been unable to make a finding of. safe operation pending resolution of Task A-3, Steam Generator Tube Degradation. Therefore, tube degradation, which has historically-plagued Westinghouse steam generators, can be reasonbly expected to occur at Seabrook.

4. Where steam generator integrity is not assured, it can reasonably be expected that radioactivity will escape from the primary coolant into the secondary coolant. Therefore, all valves and systems which serve as a barrier between the steam generators and the environment must be classified as "important to safety" and enviromentally qualified because their failure to operate properly could result in the escape of radioactivity

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to the environment. 10 CFR 50.49(a). These valves and systems include the steam dump valves, turbine valves, condenser steam dump' valves, and other valves on lines leading from te main steam line; control systems for valves; the condensate steam system; and the coolant system related to the condensate system. See Minor affidavit, paragraphs 4 and 5.

5. During an accident, in the event'that the atmospheric dump valves could not be used to vent steam to the atmosphere because of radioactivity in the steam, Applicants would be forced to channel steam to the condensers via the turbine valves and condensate system and related cooling systems.

Failure of these components could prevent the safe removel of heat from the steam generators. Therefore, they are important to safety under the definition of 10 CPR S 50.49(b)(2) and must be environmentally qualified. See Minor affidavit, paragraph 6.

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' AFFIDAVIT OF GREGO'RY? C. MINOR.

Gregory C. Minor, being dulyLsworn,Meposes~and says:-

I -am a1 consulting ; engineer :with'MHB Techn'ical

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. Associates in San Jose, California.

2.- I have worked- for over 20. years in the nuclear industry in a variety of positions-including the design, manufacturing, construction, maintenance and analysis of nuclear plants, systems and components. A list of my professional qualifications is attached to this affidavit.

3. In the presence of steam generator tube leaks, there is an increased risk of radiation release to the environment through the' secondary cooling system of a nuclear plant.

4. Where there is uncertainty regarding the integrity of the steam generators, all valves which constitute a barrier between the steam generators and the' environment may be relied upon to prevent release of radioactivity to the environment.

These valves include the steam dump valves, turbine valves, l condenser steam dump valves, and other valves on lines leading l from the main steam line..

5. The other systems which may be required to function to prevent release of radiation include control systems for valves, the condensate system where steam is directed, and the' coolant system related to the condensate. system. ,

6. In the event of an accident'where there were tube ruptures in the steam' generators,-the atmospheric dump valves could not be used because they would release radioactivity to the environment. .In_such a case,-Applicants would be' forced.to r ely on the turbine . valves and . condensate system and r elated cooling systems to divert the steam to the condenser whereJits heat could be removed without release of radioactivity to. the enviroment. Failure of these components could prevent the safe removal of he,at from the steam generators.

/ @0%fY Gregorf C' Minor Subscribed and sworn to before me this,21 b day of /k/A1d ,

1983.-

-b42 hw W$uV NOTARY PUBLIC

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' PROFESSIONAL QUALIFICATIONS 0F 'GRECORY C. -MINOR GREGORY C.1 MINOR ~

MHB Technical Associates.

172 3 Hamilton Avenue Suite K

San Jose, California 95125 (408)-266-2716 E XPE RIEN CE

1976 - P RESENT Vice -P re s iden t - MH B Technical Associates, San Jose, California.

Engineering and energy consultant to state, federal, and private organizations and individusals. Major activities include studies.

of safety and risk involved in energy generation, providing tech-nical consulting to legislative, regulatory,.public and private-groups and expert witness in behalf of state organizations and citizens' groups. Was co-editor of a critique of-the Reactor Safety Study (W ASH- 1400) for the Union of Concerned S cientis ts and co-author of a risk analysis of Swedish reactors for the Swedish Energy Commission. Served on the Peer Review Group of the NRC/THI Special Inquiry Group-(Rogovin Committee). Actively involved in the Nuclear . Power Plant standards Committee work for:

the Instrument Society of America (ISA).'. .

1972 - 1976 Manager, Advan ced Con trol an'd Ins tru' men t'ation Engineering ; j General Electric Company, Nuclear' Energy" Division, San Jose,'

' California.

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Managed a design and"developm'entigroupsof thirty-four~ engineers-and support personnelide'signingf sys'tems ~ f or 'use uin the measurement,

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control and operation' of. nuclear reactors. . Involved coordination 1 with other reactor . design organizations, thel Nuclear Regula tory .

Commission, and; customers, both overseas and domestic. Responsi-bilities included coordinating and managing the design and development of control systems, safety systems, and new control' concepts for use on the next generation of reactors. The position included responsibility for standards ap plic ab le to control and instrumentation, as well as the design of short-term solutions to field problems. The disciplines involved included electrical and mechanical engineering, seismic design and process computer control /

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t-1970 -'1972 Manager, Reactor Control Systems Design, Cencral Electric Company, Nuclear Energy Division, San Jose, California.

Managed a' group of seven engineers and two support personnel in' the design'and. preparation.of the. detailed-system = drawings and

. control documents relating to safety ~and emergency systems for nuclear--reactors. Responsibility' required coordination' with

.cther design organizations and interaction withfthe customer's, engineering personnel, as well as regula tory personnel.

1963 - 1970 Design Engineer, General Electric Company, Nuclear Energy Division, San Jose, California.

Responsible for the design of specific control and instrumentation systems for nuclear reactors. Lead design responsibility for various subsystems of instrumentation used to measure.ncutron flux in the reactor during startup and intermediate power operation. Performed lead system design function in the design o f a maj or - sys tem for measuring the power generated in nuclear reactors. Other responsi-bilities included on-site checkout and testing of a complete re' actor control' system at an experimental' reactor in the Southwes t. Received patent for. Nuclear Power Monitoring System.

1960 - 1963 Advanced Enginee rin g P ro gram, General Electric Company; Assignments.

in Washington, California, and Arizona.

Rotating assignments in a variety of disciplines:

Engineer, reactor maintenance and instrument design, KE and D reactors, H an f ord , Washington , circuit design and equipment maintenance coordination.

Design engineer, Microwave Department, Palo Alto, Cali-fornia. Worked on design of cavity couplers f o r TWT's .

Design en g i*n e e r , Computer Department, Phoenix, Arizona.

Design of core driving circuitry.

- Design engineer, Atomic Power Equipment Department, San Jose, California. Circuit design and analysis.

Design engineer, Space Sys tems Department, Santa Barbara, California. Prepared control portion of satellite proposal.

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- TechnicaliStaff -- -T,echnical Milita ry -Planning Operation.

L(TEMPO) , -'S an ta- B arbara, ; Calif ornia. . Prepare analysis of missile e x c h a n g e s'. ;

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During'this period,~completedLthree-year General: Electric program.

of. extensive education.in advanced engineering 1 principles _of high-er mathematics, probability.and. analysis. :Also' completed courses' in Kep ner-Tregoe , .E f f ec cive Pres entation, Management Training Pro-gram,.and.various' technical' seminars.

EDUCATION

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University of' California at Berkeley, BSEE, 1960.

Advanced Course in Engineering - three-year curriculum, General Electric Company, 1963. 1 S tanf ord Univers ity, MSEE, 1966.

HONORS AND ASSOCI ATIONS

- Tau Beta Pi Engineering Honorary Society.

- Co-holder of U.S. Patent:No. 3,565-760,." Nuclear Reactor Power Monitoring Sys tem," February,.1971.

- Member

American Association for Advance of Science.

- Member: Nuclear Power Plant . S tandards Committee, I r c._ t r u -

ment Society of America.

i PERSONAL DATA l.

! Born: June 7, 1937 Married, three children Residence: San Jose, California l'

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O PUBLICATIONS AND TESTIMONY

1. G.C. Minor, S.E. Moore, " Control Rod Signal Multiplexing,"

IEEE Transactions on Nuclear Science, Vol. NS-19, February, 1972.

2. G.C. Minor, W.G. Milam, "An Integrated Control Room System for a Nuclear Power Plant," NE D0-10 6 5 8, presented at In-ternational Nuclear Industries Fair and Te chn ical Mee t in gs ,

October, 1972, Basle, Switzerland.

3. The above article was also published in the German Technical Magazine, NT, March, 1973.

4. Testimony of G.C. Minor, D.G. Bridenbaugh, and R.B. Hubbard before the Joint Committee on Atomic Energy, Hearings held February 18, 1976, and published by the U n io n of Concerned Scientists, Cambridge, Massachusetts.

5. Testimony of G.C. Minor, D.G. Bridenbaugh, and R .B . Hubbard before the California State Assembly Committee on Resources, Land Use, and Energy, March 8, 1976.

6. Testimony of G.C. Minor and R.B. Hubbard before the Cali-fornia State Senate Committee on Public Utilities, Transit, and Energy, March 23, 1976.

7. Testimony of G.C. Minor regarding the Grafenrheinfeld Nu-clear Plant, March 16-17, 1977, Wurzburg, Germany.

8. Testimony of G.C. Minor before the Cluff Lake Board of In-quiry, Regina, Saskatchewan, Canada, September 21, 1977.

9. The Risks of Nuclear Power Reactors: A Review of the NRC Reactor Safety Study WASH-1400 (NUREG-75/0140), H. Kendnll, et al, edited by G.C. Minor and R.B. Hubbard for the Union of Concerned Scientists, August, 1977.

10 . Swedish Reac tor S af e ty Study: Barseb'ack Risk Assessment, MHB Technical Associates, J an ua ry , 1978. (Published by Swedish Department of Industry as Document Sd1 1978:1)

11. Testimony by G.C. Minor before the Wis consin Public Service Co mmis s ion , February 13, 1978, Loss of Coolant Accidents:

Their P rob ab il i t y and Consequence.

12. Testimony by G.C. Minor before the California Legislature Assembly Committee on Resources, Land Use, and Energy, AB 3108, April 26, 1978, Sacramento, Califo rnia .

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- y-PUB LICATIONS ' AND TESTIMONY-

13. P'r e s en ta t io n by'G.C. Minor before the Federal' Ministry for Research and Technology (BMFT) , - Mee ting on Reactor S af e ty~ Research, Man / Machine Interface in Nuclear Reactors, August 21, and September.1, 1978, Bonn, Germany.

14. Testimony by G.C. Minor, D.G. Bridenbaugh, and R.B. Hubbard, before the Atomic Safety and Licensing Board, September 25, 1978, inythe matter of the Black -Fox Nuclear Power S tation -

Construction-PermitLHearings, Tulsa,: Oklahoma.-

15. .Te s timony o f HG .C . ' Min o r , ' ASLB ' He a rings; Rela t ed t o TMI-2 A c c id en t ,- Rancho Seco Power Plant, on behalf of Friends-of the Earth, September 13,'1979.

16.. Tes timony - of G. C . Minor ~before the Michigan' State Legisla-ture, Special Joint Committee on' Nuclear Energy, Implications of Three Mi_1_e__I s l a n d A c c i d e n t for Nuclear Power Plants in Michigan, 10/15/79

17. A Critical View of Reactor Safety, by'G.C. Minor, paper presented to the American As s o cia t ion for the Advancement of S cience, Symposium on: Nuclear. Reactor Safety, January 7, 1980, San Francisco, California.

18. The Effects of Agidg on Safety of Nuclear Power Plants, paper presented at Forum on Swedish Nuclear Referendum, Stockholm, Sweden, March 1, 1980.

19. Minnesota Nuclear Plants Gaseous Emissions Study, MHB Technical Associates , September, 1980, prepared for the Minnesota Pollution Control Agency, Roseville, MN.

20. Testimony of G.C. Minor and D.G. Bridenbaugh before the New York State Public Service Commission, Shoreham Nuclear Plant Construction Schedule, in the matter of Long Island Lighting Company Temporary Rate Case, September 22, 1980..

21. Testimony of G.C.' Minor and D.G. Bridenbaugh before the New Jersey Board of Public Utilities, Oyster Creek 1980 Refueling Outage Investigation, in the matter of Jersey Central Power and Light Rate Case, February 19, 1981.

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I.C.' NECNP has withdrawn this~ contention'.

I.D.l. -NECNP has' withdrawn'this contention.-

I.D.'2. Testing of Protection System Actuation Functions LApplicants have moved for summary disposition-of this contention on the ground that Reg. Guide l.22 has been fully.

complied with, and that therefore General Design Criterien 21 has been satisfied.d/ NECNP contends that, with respect'to testing of the manual system actuation functions, Applicants have complied with neither GDC 21 nor Reg. Guide 1.22, and that there is a factual dispute on the reliability-of the manual'

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trip function which cannot be resolved in f avor of Applicant's Eby summary. judgment.

General Design Criterion 20 -requires that the protection system be designed to-initiate the operation of' systems and' components important to safety. General Design Criterion 21 requires that the protection system be designed to permit periodic ' testing of its f unctioning when the . reactor is in operation. Reg. Guide 1.22, which implements these design criteria,. contains three ' narrow criteria for waiving the requirement for testing of power. Regulatory Position 4, page 22.2.

2/ Applicants also argue that NUREG-0737 does not apply to this contention. NECNP does not dispute this assertion.

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a The nuclear industry's history of-numerous-reactor' trip' failures, culminating;in-the near-disatrous Salem accident ~of February 25, demonstrates that?Applicaants:cannot-providefthe assurance , required by Regulatory.. Position 4.b. :that: *the pr'obabilitye tha't fthe protecti~on'.. system %ildf a'il 4 tofinitiat'e:

the--operation =of thel actuated > equipment is,fandican be maintained,Lacceptablytlow-without testing.the equipment-during.

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reactor operation." The Salem:accidentLdemonstrated the critical ~ role of the: manual trip function'i,nfavertingca serious accident. When both channes of the automatic trip function failed, the operator was required to immediately activate'the manual trip in order to sut the reactor down. The February 25 incident was preceded by a similar event on February,22.- The two Salem accidents were the first events in which both circuit' breakers failed; however, there have been 35 reported circuit breaker f ailures since 1973. Philadelphis Enquirer, March: 3, 1983 at 1-B. The NRC has also reported reactor trip breaker f ailure in two tests at San Onofre during the month of March.

See IE Bulletin No. 83-04.

! As a result of the Salem incident, the NRC.is requiring the i

' licensee to test the manual trip (or shunt trip) every month.

Because normal operational outages typically occur at much longer intervals, this means the manual trip functions must be tested at power.b!

$/ The fact that testing can be done at monthly intervals (the licensee actually proposed weekly intervals) indicates that it is possible to test the manual reactor trip without damaging the reactor, contrary to Applicants' unsupported assertion in the FSAR. at S 7.1.2.5.

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llECNP contends that- in light' of the -Salemiaccident and 'the.

.other .r eactor trip--f ailures which show a clow 'r'eliability of that function,-the' frequent test'ing.of:the manual. trip function at power is required for conformance with GDC_20 and 21,: and Reg. Guide-1.22.- Therefore, Applicants'"liotion.for summary disposition should be denied.

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CONTENTION I.D.2 STATEMENT OF MATERIAL FACTS AS TO WHICH THERE

.IS A GENUINE-ISSUE TO BE HEARD

1. General Design Criterion 21 requires that'"The protection system shall be designed for_high functional reliability and inservice testability commensurate with_the safety function to be performed." GDC_also requires that "The Protection system shall be designed to permit periodic testing. ,

of its functioning when the reactor is in operation,_ including

a capability to test.channelsiindependently.to determine' failures and losses of; redundance that may have occured." '

2. Applicants do not provide _for[ testing'at power offthe manual reactor" trip function..

3. The1unreliability offautomatic.reactorctrip has been

' demonstrated by a-large number of trip failuresLin recent

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years. See IE Bulletin 83-04 and Philadelphia Enquirer of March 3, 1983, attached, The most significant of these events was the failure of both atomatic reactor trip channels to initiate shutdown at the Salem reactor on February 22 and February 25, 1983. In those instances, the' Salem operator relied on the manual trip switch to bring the~ plant to safe shutdown. Had the plant been operating at full power, failure of the manual trip could have caused a major accident.

4. In order to insure reliability of reactor trip, the NRC has required the institution of monthly tests of.the manual

_ trip at Salem. SECY 83-98A, Attachment 1 at 27. Since normal outages occur at much greater intervals than monthly, it_may be assumed that the manual trip must be tested at power under these circumstances.

5. Aside from an unsupported assertion in the FSAR that testing the manual reactor trip at power would damage the reactor, Applicants have never given any reason for f ailing- to test the manual trip function at power. Applicants have not l stated that they are unable to test each channel of the manual l trip separately, which would avoid tripping the reactor.

! 6. Commensurate with the requirements set forth by the Staff for the Salem plant, the Applicants should be requied to perform periodic testing of the manual trip switch at power in l order to provide a reasonable-assurance of its reliability in i satisfaction of GDC 21.

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c. g March 14, 1983 \, ...../ ,

SECY-83-98A' POLICY ISSUE (Information)

' For: The Commissioners From: William-J. Dircks

' Executive Director for Operations

Subject:

SALEM RESTART

Purpose:

To provide the Commissioners with a report on the current status of the staff evaluation.of the .failurecto automatically scram events Lof February 22 andi25,1983_ at the. Salem ' Nuclear-Generating Station and the' staff action planFfor. authorizing restart of Units 1 'and.2.

-Discussion:

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During:a~ briefing on March 2,1983'concerning-the Salem reactor trip system -failure events, the Commissioners

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requested that the staff provide _its plan of action to resolve the . issues identified from the NRC evaluation of the Salem. events.

Enclosed is -the' Salem Restart status report which identifies the issues related to the recent Salem events and the'short- -

and long-term actions needed 'to resolve those issues.

'For the short-term actions, the staff has or intends to obtain specific commitments from the licensee ~to complete.

those actions and the staff will assure their satisfactory-completion prior to permitting restart of either Salem unit.

For satisfactory resolution of the long-term actions, the. m.

staff intends to develop'with the licensee an acceptable . -c is schedule for completion of those actions, obtain necessaiy _ jQ written commitments, and follow up their completion on the E -

agreed upon schedule. ] _

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Contact:

Gus Lainas X-27817 R. Starostecki FTS .488-1230 h  % , E- j A E iA @ N 5"lA M 2 f 5 M O 3 872*l W U i k 7 d A

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Theicommissioners "In addition to theLshort- and long-term actions identified ~'

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.in.the report,ithe staffchasf also:concludsd that a show

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cause ' order-shauld benissued to the licensee (seel enclosure 2).-

Theistaff. believes that' the;particular. circumstances at.

this facility, :as. further detailed in the=' start-up report,<

justify requiring th'at4these threelsepa. ate but interrelated sets of actions be implemented by;the licensee in a t.imely-fashion.

Subject [to satisfactory' implementation of these actions,- the staff has concluded that the Salem.ftcilities can oe ' restarted and operated .without undue' risk to thej health and safety.

of the public. Enforcement actions areLunder active consideration by the staff and will be discussed separately

with the~ Commission at'a later date.

.L LJL- L r Willia v.) Dircks Executive Director for Operations

Enclosures:

1. ' Salem Restart Status Report

2. Oft Show Cause Order DISTRIBUTION:

CoxJnissioners OGC.

OPE-OIA SECY-1 I

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l technician. At this time, the CRC-2-26 lubricant cleaner was sprayed on all four UV trip attachments associated with the Unit-1 circuit breaker. This

-lubricant is being procured by FRC for testing purposes.

List o'f Investigations To Be Performed by NRC Contractor (FRC)

1. The first test will be to perform various deenergirations and energiza-tions of the UV trip attachment and monitor the device under various conditions. ,

2. The second test'will be to disassemble the latch mechanism to observe the surfaces of the various parts of.the latch and to photograph these sur-ta:es through a microscope to determine the various levels of' wear on these surfaces. .

~3. The third test is to determine the effects of CRC-2-26 spray on the various types of metals used in this devices., An attempt will be made to use metals other than those in the actual attachment. If possible, the chemi:al consistency of this spray will be determined from the manu-

-facturer.

To prove:that the. sample UV trip attachment is Lidentical to all 'such Salem-devices, a visual i 9spectionfof all ixisting Salem-Unit'l and:2 UV trip:

attachments will be performed. Thisican_take~ place'at Salem, with 'no disas-sembly needed. The inspection can be'made with_the devices mounted on the' circuit breakers or loore. 'These -inspections should be' done as soon as

.p'ossible, and Tuesday, March 8, 1983.is. recommended.

If further tests are required they-will be based'on the results of these initial tests. All cests'will be nondestructive such that the device can be used for further testing and returned to the utility. -

Additional Test To Be Conducted by the Licenseh, as Revised by NRC Staff This test will require the ese of a spare circuit breaker. The UV trio and shunt trip attachments will be mounted on the breaker, and the breaker will be operated repeate'dly to determine the effect on the shunt and UV trip attach-

• ments. It is surmised that wh'le the attachments are energized,and the breabr trips and closes a number of times, additional friction or the trip latch may occur from the vibration. This test is described in detail in the following section..

'II. REVISED SURVEILLANCE OF REACTOR-TRIP CIRCUIT BREAKER OPERATION AND VERIFICA110N TESTING The licensee proposed the following increased surveillance of reactor-trip circuit breaker operation:

1. Main and bypass breakers will be shunt-tripped weekly.

2. Main breakers will be UV-tripped monthly.

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3 The acceptability of _ this revised surveillance of reactor-trip circuit- breaker .

' operation has been evaluated by NRC staff. Based on an analysis conducted by NRC staff, which ccnsidered reactor-trip system unavailability, reactor-trip circuit breaker failure rates,' and test intervals, the following conclusions were drawn. First, the proposed test of'each reactor-trip circuit breaker UV trip attachment once every 30-days is acceptable. Second, the proposed test of the shunt trip attachment once every seven days is considered to be exces-sive and may impact on the reliability of the reactor trip system by increas-ing _the potential for a single failure.. During: testing, a single failure in -

.the logic portion of the reactor trip system could prevent an automatic SCRAM.

Thus, it is recommended that the shunt trip attachment be tested on the same schedule as the UV trip attachment; that is, once every 30 days. It is also recommended that the UV trip of the bypass breakers be teste'd prior to restart and every refueling thereafter.

Discussion The acceptability of the proposed test intervals for the reactor-trip circuit

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breakers was based on NRC staff review of reactor-trip circuit breaker failure rate data ootained from Licensee Event Reports-(LERs). The generic RPS unavailability of 3 x 10 3 (used in both NUREG-0460, " Anticipated Transients '

Without Scram for Light Water Reactors," and by the A WS Task Force and

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Steering Group in the dselopment of the proposed ATUS Rule) was used in

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evaluating the licensee's proposed test in_tervals. In addition, the following considerations were incorporated into the NRC' staff recommendation:

1. The shunt trip attachment provides a diverse means of tripping the reactor-trip circuit breaker, which is electrically independert of the UV trip attachment. The UV trip attachment is surplied by a 48-V dc source and is deenergized to trip. The shunt trip attachment is supplied by a-125-V de source and is energized to trip.

2. The shunt trip attachment is an energize-to-actuate device and is'not.

" fail safe" in that a loss of power will not cause a trip. However, the

. shunt' trip is powered from a reliable Class 1E battery-backed source.

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'Since the shunt trip attachment is an energize-to-a_ctuate device, it is 3.

not subject to the constant heating effects that the ' continuously en-ergized UV trip attachment experiences. .The heating effects may contrib-ute to the higher failure rate of thc UV trip attachment.

4. The mechanical construction of the shunt trip attachment is less complex than that of the UV trip attachment. The shunt trip attachment does not rely on the successful operation of the complex latching mechanism that has been determined to be the source of the majority of the failures of the UV trip attachment. '

5. The majority of the electrical circuit breakers used in the high-voltage electrical distribution system have dc powered energize-to-actuate shunt trip attachments. These circuit breakers are used for manual, as well as automatic, trip functions for load shedding and power switching. Relia-bility of energize-to actuate shunt trips in similar applications through-J 26

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out the nuclear power industry has been shown to be significantly higher

,than for devices that are constantly-energized.

Over 70% of the-kncwn reactor-trip circuit breaker failures were caused

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'by UV trip' attachment failures.

7. Most of the concerns relating to the events at Salem on February. 22 and 25, 1983'are related to the operation ~of the UV' trip attachment. During the events at Salem -the shunt trip attachment functioned properly.

8. The bypass breakers are required to trip in response to a UV trip demand signal should this occur when the main breakers are being tested. Since the test frequency of the main breakers has been increased, the bypass breakers should be tested to verify the capability to perform their backup safecy function. ,

Verification Testing It is recommended that a bench test be perform 5d on one 08-50 reactor-trip circuit breaker. The purpose of the test Gill ~ be to cycle the DB-50 with the UV trip and shunt trip attachments in place for a total of 2000 cycles to determine if any adverse effects can be-identified and, if there are no adverse effects, show thet a properly maintained breaker and its subcomponents can operate for an extended number of cycles.. The breaker will be tripped, ips. The ambient with each cycle temperature shouldbeing be 100 alternated withthe F to simulate the UV'andserv expected shunt'tr, ice environment,~and-the circuit breaker should be cycled no more often than once every 30 minutes

- to allow for return' to steady-state. conditions. _ The. results of each circuit

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breaker operation willibe documented and.a visual. check:made. . AdditionalL details for this type- te'st'will'be provided at a.later time.

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% Mtilabelpftia 3fnquirer section city suburbs region Thursday. Wrch 3,1983 ~

B NRG Salem plant's saMWs%Tl rm.W .n 2S,Wes%uM

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sibd some time to sort this out."

ce Ily Matt Yancey' The disclosures created concern .

A.mpd Nm yc$!crday Pmong the [)ve NRC com. *8s Operating at only 12 percent $f

't>ASillNGTON J Federal officials yesterday blamed r maintenance me.y.

g. its capacity last Friday BecausG;'

., that. they said, the plant was ncver

• for the unprece ented failure - * '#8 d8 E" twice within a that happened here," said Commis-Jersey n'ucleah@cek - of a Southplant's automatic But had the p'lant becc operating a 5{n(r Vict t t r Gilinsky. who added full power. hiattson said, the pres.

safety system.an'd expressed ! car that e wanted a report on correc.

the problert could be widespread 1

ns that 'hc utlhty was making sures inside the reactor and its cool.

among the nation's reactors.. the ing system would have created many Pye allowmg it to restart more problems during the 24%ecc.

The Nuclear:Regu% tory Commis. orid ' gap. including possible- core -

ston (NRC) disclosed that the system ..jI fcy labeled these incorrectly, damage. lie said he haddnot yet con.

that automatically shuts down a reac. isn t Ihere a chance they labeled ther icces of equipment incorrect. ducted studies to' determine just how tt!r when thereire indications of tauch. damage could have occurred.

unsafe conditiotisihad failed for a issi nu khn Ahearne Eiid'tlin'e~last week at Public Scr-sked . Officials said a similar total bicak.

vice Electric'& Cas Cols plant in NRC officials ackn.owled8-

~. ed'that the Salem incidents raised new ques- at the plant occurred three days car.

Salem One of the breakdowns had lier, on Feb. 22. It went uitnoticed come io light previously. e a ut " ~

, n nce gas n they said because the operator, scE to I e si u tanc s la ure -

to clear plants around the country. Jng in;!Jqations o[ unsa(e..pondjtlong ,

circuit breakers that immediateg) in Thecommission ordered a detailed on his instruments, shut the plant sert controt rodsinto a rcactor.w n investigation of similar. circuit dowrt manually only four seconqr cither one 1s , activated, to stop t ic. breakers used in other plants and - after the failure. '~

how they are maintained. The own- ~ The only previous incident of ti R cr at spn r ctor of the ers of all Westinghouse reactors reactorjalling to n. cpats, NRCs systems inYegration division, were instructed to test their ciruit pretelypurreM . sferry {

said thcre' liad.ibcen 35 incidents breakers over the weekend, and t' 5 smcc 1973 in which one circuii none failed. NRC officials said.

g plaht;trr:I Alabama ,inrgds Thery,gK of thp.th$

nlybl'1980.:f breaker fared in an automatic sa ety g gg .

system. Ilut he said the Iwo incidents at the Salem plant last week marked been ordered to supply data en cir-cuit breakers of similar design and 'InOfauf06ta[ic'rcult;

.ibo M ci aty breaker. (affurcs r ci u t b ca er a i s bac u on hw mey a maWnd m3 g han & 'N[syyte'ms.:since' al N Sakm .

Ialting simuhaneously. . NRC ofucials satd the second fail. p'lant, including the four last week.

Because the circuit breakers are ure at the Salem plant last Friday and htatt' son said.~

e h phfore' There are five"other plants at u i ticsare uir d to o c the control room operator shut down which there has been more than one rate' detailed procedures in mam. the reacia manually could, hne occasion on which one of the two caused a severe accident with possi. safet il off ci is said that Pubhc Scr. ole damage to the reactor core if the~ f work.yThey system breakers involve failed seven failurcs at to>

vice Electric & Gas had never te. Pl ant hao been operating at full pow. o.thcLOconee plant in South Carolina, ceived maintenance bulletins that cr. cs at @ Lon plant in h "3 The last nuclear accident that %conshtlt f%wo failures at the Robinson <

c '9uipmc t.- utou n 19 caused core damage in the IJnited plant in South Carolina and three in addnian, wlien- ' the brFakers~

States wasin March 19N at the Thice failures each at the Point Beach were taken apart for maintenance in Mile Island nuclear plant near flar. t

) January, they were cassified incor.

risburg. There invbet bgcapte the ' f reactor ant inafWisconsin Russelville. and the Arkansa rectly as nonsafety. equi ment and worst acStdch,th th'e ipdustry's his-did not get the priority I ey should ~

have reccived, officials said.

'Obviously, we take this situation

. tory, workedtheh(authmatk.)afety systemsoperly,'f tntPtlWppgrator.

we're m very, very seriously. said believ ' .,#

cover' that nitMl.MthOcap),tQt WUhem vrpsafely off.

Richard Eckert Public Service's se-nior vice pr$ident, who blamed hu. NRC officials Wd th'e Salem linit i man crror for the incorrect clawf reactor, which was starting back up nuon er the equipment. %e nee t armr P '^wn for refueling.

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IEB 83-04' -

March 11,'1983.

Page 2 of 3 out-of-adjustment within'the linkage mechanism of the UV trip device installed in General. Electric (GE) type AK-2 (i.e. .~ AK2A-15, 25, 50, 75,100) circuit breakers. Failures have occurred at ANO-1, Crystal River-3, Oconee Units 1 and 3, TMI-1, and St. Lucie. As a result of these events, the NRC-issued IE Bulletin No. 79-09 dated April 17, 1979 and IE Circular 81-12 dated July 20, 1981. Subsequently, failures have been' reported at ANO-l'and Rancho .

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Required Actions for' Holders of Operating Licenses for Pressurized Water Reactors:

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PWR licensees with other than W DB type breakers in Reactor Protective System applications are requested to:

1. Perform surveillance tests of undervoltage trip function independent of the shunt trip function within 5 days of receipt of this Bulletin unless equivalent testing has been performed within 10 days. Those plants currently shutdown should complete this item before resuming operation or within 10 days, whichever is sooner. Those plants for which on-line testability is not provided should complete this item at the next plant.-

shutdown if currently operating.-

2. Review the maintenance program for conformance to the latest manufacturer's recommendation, including frequency and lubrication. Verify actual implementation of the program.* If maintenance does not conform, initiate such maintenance within 5 days of receipt of this bulletin or provide.an a alternate maintenance program. Repeat the testing required in item 1 prior to declaring the breaker OPERABLE.

3. Notify all licensed operators of the failure-to-trip event which occurred at Salem (see IE Bulletin 83-01) and the testing failures at San Onofre Units 2 and 3 described above. Review the appropriate emergency operating procedures for the event of failure-to-trip with each operator upon his arrival on-shift.

4. Provide a written reply within 10 days of receipt of this bulletini

a. Identify results of testing perfomed in response to item 1. - Pl_ ants without on-line testability should report the date and results of the most recent test.

b. Identify conformance of the maintenance program to manufacturer's recommendation and describe results of maintenance performed directly i as a result of this Bulletin in response to item 2. 6

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• IE Bulletin 79-09, dated April 17, 1979, had as an attachment an extract of General' Electric'(GE) Service Advice Letter'No.175(CPDD)9.3 which is applicable to GE type AK-2 breakers.

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March 11 1983 Page 3 of 3'

c. Provide a statement that provisions are in place to notify licensed operators of the Salem and San Onofre events and bring to their attention appropriate failure-to-trip emergency procedures upon their arrival on-shift.,  % .. s

d. Provide a description af fall RPS breaker malfunctions not previously reported to the~NRC. '

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.e._ Verify that procurement, testing and maintenanid activities treat the

.RPS breaker and UV devices as safety related. Report the results of this verification to the NRC.- ,

5. Any RPS breaker failure identified as a result"of testing requested by this bulletin should be promptly reported to the NRC via the emergency notification system, regardless of the operating mode of the plant at the time of the failure. -

• The written report required shall be telefaxed to Richard C. DeYoung Director, Office of Inspection and Enforcement within 10 days of receipt of this bulletin.**

At the same time, the report shall be submitted to the appropriate Regional Administrator under oath or affirmation under provisions of Section 182a, s Atomic Energy Act of 1954, as amended. The original copy of the cover letters and a copy of the reports shall be transmitted to the U. S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555 for reproduction and distribution. ,

This request for information was approved by the Office of Management and Budget under a blanket clearance number 3150-00012 which expires April 30, 1985. Coninents on burden and duplication may be directed to the Office of Management and Budget, Reports Management, Room 3208, New Executive Office Building, Washington, D.C. 20503.

If you have any questions regarding this matter, please contact the Regional Administrator of the NRC Regional Office or the technical contact listed below.

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!sq i chard C. DeYoung, Dir to /

fice of Inspection and Enforcement I

Technical

Contact:

I. Villalva IE . Thomas, IE j 301-492-9635 301-492-4755

Attachment:

1. List of Recently Issued IE Bulletins

• • Rapidfax (301) 492-8187 or (301) 492-7376 3M Remote Copies (301) 492-7285 j

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I.D.3. NECNP has withdrawn this contention.

I . D .4 . NECNP has' Withdrawn this contention.

I.F. NECNP has withdrawn this contention.

I.G. Pressure Instrument Reliability In this contention, NECNP asserts that the Seabrook wide range pressure instruments cannot be relied upon for ' accurate information, and thus may lead to inappropriate operator actions jeopardizing the cooling of the reactor. This contention was based on IE Notice no. 82-11 (April 9, 1982) which reported a significant margin of inaccuracy (1363 psig actuation and 1390 psig indication) in the instruments during qualification tests in a post-accident high energy line break environment. The NRC concluded that these inaccuracies could result in " inappropriate operator actions."

In their answers to NECNP's interrogatories, Applicants

' identified the wide range pressure transmitters as PCT 403 and PCT 405, and supplied NECNP with a drawing (No. 9763-M-506635) showing the tyo transmitters to be outside the containment.

By motion of February 11, 1983, the Applicants requested the Board to summarily dispose of NECNP Contention I.G. on the sole ground that "the seabrook wide range pressure transmitters are located outside the containment and thus are not subject to the high .,aergy line break environment which caused the inaccuracies addressed in IE Information Notice 82-11."

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. Applicants attached an affidavit stating that the wide range pressure. transmitted are located outside containment, but supplied no diagrams in support of this assertion.

Applicants'1 statement regarding the location of .the wide range pressureLtransmitters.is" squarely contradicted by the PSAR. Figure 5.1-1, SH 2 and:Sil.5 of the FSt.R (attached),

entitled " Reactor Coolant System' Loop No. 1, P & 1D Diagram",

show the wide range pressure l transmitters.. (PCT 403 and 405) to

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be inside the containment. 1 Applicants. provide;no explanation for the discrepancy between the statement inithe~ summaryL judgment motion:and the-FSAR.. Nor have Applicants explained-why.the drawing supplied in response to NECNP's. interrogatories

-differs from the FSAR.

The location of the wide range pressure transmitters'is the only ground raised by Applicants in moving for summary judgment. The Applicants' own documents establish a clear conflict in the evidence-on this issue. Since material issue of f act remains to be resolved, Applicants are not entitled to ,

summary judgment.

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'l MATERIAL FACTS AS TO WilICil TilERE IS A GEtiUINE ISSUE TO BE IIEARD.

1. According-to Applicants' Final' Safety. Analysis Report,

' Figure 5.1-1,-Sil 2.and SHS, the wide range pressure. .

' transmitters (PCT 403 and PCT 405) are located inside the. t containment and are'therefore subject to.the type of accident environment which caused the' instrument. ambiguities noted in IE tio tice 8 2-11. -

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I.I. Environmental Qualification of a Path to Cold shutdown In Contention I.I. , NECNP asserted that Applicants mu'st

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identify and environmentally qualify a path to cold shutdown.

Applicants have moved for summary disposition on the ground that the NRC's latest rule on environmental qualification 10 CFF 50.49, excluded a requirement for environmental qualification of a path to cold shutdown. We note that in doing so, the Commission stated that shutdown decay heat removal remains an Unresolved Safety Issue which the Commission continues to study. 48 Fed. Reg. at 2731-(January 21, 1983).

Ther efore, the adequacy of residual heat removal, including consideration of cold shutdown capacity, must still be resolved I

on a plant-specific basis.

In any event, NECNP considers this issue to be resolved to our satisfaction. . Despite their protests in this proceeding that it is not required, Applicants have informed the NRC Staff by. letter that~they have qualified a path to cold shutdown, with the exception of the pressurizer. heaters. Letter from J.

DeVincentis, Yankee Atomic Electric Co., to George W. Knighton, U . S . N PC , re:. Clarification to RAI 440.133 (Letter No. SBN480)

(February 25, 1982). In light of this commitment, NECNP withdraws contention I.I.

II. B.l. NECNP has withdrawn this contention, f

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'II.B.3. Independence of QA program i NECNP Contention II.B.3. asserts that the Applicants' Cuality Assurance ("QA") organization for operations does not have the independence required by 10 CPR Appendix B, Criterion

1. As reflected in the basis for the contention and in NECNP's responses to Applicants' interrogatories on the point, the contention is based upon and limited to the fact that the Nuclear Quality Manager, who is responsible for the QA program, reports to the Vice President - Nuclear Production, who is responsible for the operation of the seabrook station, and therefore is responsible for assuring that Seabrook remains on line as much as possible to maximize the financial benefit to the company.

Criterion I of Appendix B requires, in portinent part,' that

-[t]he persons and organizations performing quality assurance functions shall have sufficient authority and organizational freedom to identify quality problems; to initiate, r ecommend, . or provide solutions; and to verify implementation of solutions.

Such persons and organizations performing quality assurance functions shall report to a management level such that this required authority and organizational freedom, including sufficient independence from cost and schedule when opposed to safety considerations are provided.

(Emphasis supplied).

As previously noted, at Seabrook, the Nuclear Quality Manager reports directly to the Vice President - Nuclear Production. FSAR 17.2.1.3.a.l. However, the FSAR also establishes that the Vice President - Nuclear Production "is responsible for the operation and operational support of s

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s Seabrook Station.' FSAR 13.1.1.1.,'FSAR 17.2.1.3.a. This description alone, particularly when read in light of the title of the position, establishes that the Vice President for Nuclear Production is responsible for maximizing the operation of Seabrook Station so that it produces electricity and contributes to' the Applicants' rate base to the maximum extent possible. Neither the Applicants nor the Staff has denied that this individual is responsible for maximizing electricity production. Moreover, a review of Chapters 13 and 17 of the FSAR reveals no other manager above the site-based staff who has this responsibility. The necessary conclusion is that the Vice President - Nuclear Production has it, and that this is his or her primary responsibility.

Neither the Applicants nor the Staff contest this. Rather, they rely upon (1) the fact that NUREG-0731 shows a structure in which the QA Department reports to a Vice President -

Nuclear Production (Applicants' Hotion, Material Facts #2, Affidavit #2(b)), (2) the fact that the Seabrook organization-arguably comports with the type of organization suggested in.

the Standard Review Plan, which is used by the Staff in reviewing license applications (Staff Motion, Material Facts

1. 5), and (3) the opinion of a Staff member that, because the SRP suggestion is met,;the QA program at Seabrook has sufficient independence under Appendix B.

These assertions avail the Applicants and Staff of nothing. Both NUREG-0731 and the SRP are merely Staf f

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t documents. They have no force of law. They reflect no review

- of Seabrook itself. More important, they do not refute the fact that at Sebrook, according to the PSAR, the QA program

-would be reporting directly to the ind vi idual with a major vested interest.in assuring that QA problems do not cause shutdown or delayed operation,' which would interfere with electricity production, this Vice President's primary

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responsibility.

Other portions' of the FSAR are revealing and disturbing in

-this regard. According to FSAR 17.2.1.3.a.1, the_ Nuclear Quality Group has the authority to "r equest work stoppages or remedial actions if conditions adverse to quality are encounter ed" (emphasis supplied). This conflicts with a reference at FSAR Page 17.2-6 that the Quality Assurance

) Department may exercise stop-Work authority.- At a minimum, the FSAR and the record now before the Board are unclear on the extent of the authority vested in the Nuclear Quality Group.

In addition, all disputes between QA and production personnel are to be mediated by the Vice President - Nuclear Production, t,he very individual responsibile for maximizing the operation of Seabrook. Although the dispute may be appealed to the Executive Vice President - Engineering and Production, the PSAR gives no indication that the QA department has direct access to that officer. Moreover, to appeal any dispute to this relatively impartial officer, the Nuclear Quality Manager would have to contest a decision by his immediate superior, who appears from the FSAR to be responsible for all aspects of the, n

-s QA function, including whether.the Nuclear Quality Manager' keeps his job. This has a-serious chilling effect'on any necessary appeals'to the first individual in the line of command who appears tua be sufficiently- removed from production pressures to ren' der an independent opinion.

tiothing presented by the~ Staff or Applicants has resolved or even approached the factual dispute here, which involves the s

independence of a QA program' under the direct control of a production-oriented official. Compliance with two non-r egulatory Staff documents- indicates nothing. In the absence of evidence refuting the foregoing discussion, which is based on the language of the ?SAR, these motions must be denied.

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II.B.4. QA of Replacement Parts NECNP Contention II.B.4. asserts that the Seabrook QA program _for operations as described in the FSAR does not demonstrate how the Applicants will assure that replacement parts are equivalent to original parts and installed in accordance with proper procedures. It also asserts that the QA program does not demonstrate how repaired or reworked items will be adeluately inspected and tested. It is based on the absence of any demonstration in.the FSAR itself.

Applicants move for summary disposition on the ground that these matters are covered in the FSAR at SS 17.2.4 and 17.2.15.- The supporting af fidavit similarly references those sections. In addition, it references a matrix of procedures being prepared to address various quality assurance i

requirements. According to the affidavit, "These procedures are presently being developed and previewed; will be available prior to fuel load; and will provide-the implementation of.the Quality Assurance Program."

The Staff's motion cites further FSAR references, including a cryptic statment by Applicants that it complies with certain Regulatory Guides. These provisions involve general commitments to meet QA requirements.

There is no f actual-dispute as to the language of the FSAR or the various referenced documents. The issue may thus be decided as a matter of law based on that language, s

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The governing regulation is 10 C.F.R. S 50.34(b)(6)(ii),

which requires that the PSAR meet. the following standard:

The information on the controls to be used for a nuclear power. plant...shall include a discussion of' how the applicable requirements of Appendix 3 will be satisfied.

.(Emphasis supplied). The PSAR falls woefully short- of this standard, and the affidavit supplied by the Applicants, as well as Applicants' response to NECNP's interrogatories on this issue, demonstrate the Applicants have not yet even determined how the QA requirements will be met. We address in turn (1) the alleged QA program for replacement parts, and (2) the alleged QA program for rework and repair.

The FSAR discussion of QA for replacement parts appears at S 17.2.4. That section, a total of only three pages of text, is purported to demonstrate how QA requirements will be met for replacement parts. It does no such thing. To the contrary,.it includes only a very general discussion describing generally the framework of what the Applicants expect to do in this area.

The discussion indicates, for example, that all purchase requests will include, " as appropriate," " Quality Assurance Requirements." But there is no indication what those requirements will be or how Applicants will assure that they are met by the vendor. There follows a mention of the " quality review," but there is no indication of what that review might be. Indeed, for "off-the-shelf" items, Applicants appear to indicate that they have not decided how they will perform

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l quality control. Rather, they will develop special requirements for each item. (FSAR Page 17 . 2 -21 ) . 5;/ Not only is there not even the slightest. outline of what these requirements might be, there is no indication of how they will be developed. Who will be responsible for developing them?.

Who will review them when they are developed? Who has initial approval responsibility? Who has ultimate approval authority?

The FSAR provides nona of this information.

The same is true of the rest of the discussion. It is simply impossible to determine from the sketchy information in the FSAR how the QA program will be implemented for replacement parts. _The discussion constitutes nothing more than a general ~^

assertion that the Applicants intends to implement it. That'is not enough under 10 CFR S 50.34(b)(6)(ii) .

With respect to repair and rework, the Applicants rely on less than one page of text in FSAR S 17.2.15. Again, this is only the most general discussion of the point. There is no 5/ In fact, the FSAR does not identify who will develop these requirements. It says simply that they "will be established."

This leaves open the possibility that the Applicants will rely on the vendors to do the work, although the Applicants are responsible. The FSAR commits the Applicants to nothing more than that.

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indication of how or by whom the inspections will be performed. Beyond the cryptic mention of nonconformance reports, there is no indication of how the QA will be documented. Perhaps more important, there'is no indication of how or with what frequency the crucial trend analysis will be performed to prevent repetition of significant quality problems. The FSAR states only that some such system will be maintained, but does no explain how it will be implemented.

The Staff has added little to the Applicants' discussion.

The fact that the Applicants have commited to meeting various Regulatory Guides or other standards does not demonstrate how the Applicants will do so. That is what the regulation requires.

Finally, as previously noted, the Applicants' affidavit establishes that they have not yet determined, or at least nave not yet chosen to reveal, how the QA program will be implemented. Similarly, in response to NECNP's interrogatories,.

Applicants stated that The procedures which implement the QA program are not yet fully approved and are still under review and development. They will be made available when finally approved.

Applicants' Answers to "NECNP Second Set of Interrogatories and Requests for Documents to Applicants on Contentions I.D.l.,

I . D .4 . , I . F . , I.I., I.L., and II.B" and Motion for Protective Order, filed January 29, 1983, at 15. These procedures are precisely what are needed to determine how the QA requirements will be met for replacement parts. According to Applicants' ,,

affidavit, we can expect them only sometime " prior to fuel load."

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The very purpose of detailed procedures in any complex operation is to establish how the relevant task will be 4

accomplished. ' It may be that something less than the procedures themselveas would suffice, but the information does

, not appear in the FSAR, as required by the regulation. .Thus the Applicants are avoiding public scrutiny of' one of_ the most serious issues of reactor operation - the quality assurance.

program that is necessary to safety when actions are.taken to alter or repair the reactor.

For these reasons, we urge the Board to deny the motions for summary disposition filed by the Applicants and the Staff.

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6 CONTENTION II.B.4 STATEMENT OF MATERIAL FACT AS TO WHICH THERE EXISTS A GENUINE ISSUE TO BE HEARD

1. The portions of the FSAR cited by the applicant' and.

Staff establish that the required program for inspection and testing of repaired'or replaced. parts is not complete.

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FSAR-SS 17.2.4, 17.2.15; Applicants Answer to "NECNP Second Set of Interrogatories. . .on Contentions I .D.1, I .D.4. , I .F. , I.I.,

I.L., and II.B" and Motion for Protective Order, filed' January 2 9, 1983, a t 15.

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- - II.B.5.: NECNP'has withdrawn-this contention.

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i Respectfully submitted,

' Diane Curran.

William S. Jordan'III Lee L. Bishop

HARMON & WEISS >

1725 Eye.St. N.W. i Suite 506

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Washington, D.C. 20006 (202) 833-9070 t

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CERTIFICATE OF SERVICE I certify that on March 24, 1983, copies of NECNP OPPOSITION TO MOTIONS FOR

SUMMARY

DISPOSITION AND NOTIFICATION'OF WITHDRAWN CONTENTIONS were served by first class mail or as otherwise indicated on the following:

• Helen F. Hoyt, Esq. Robert A. Backus, Esq.

' Chairperson, Atomic Safety 116 Lowell Street and Licensing Board P.O. Box 516-U.S. Nuclear Regulatory Commission Manchester, NH 03105 Washinton, D.C. 20555 Phillip Ahrens, Esq.

• Dr. Jerry Harbour Assistant Atty. General Administrative Judge State-House,. Station #6 Atomic Safety and Licensing Board Augusta, ME 04333 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Jo Ann shotwell, Esq.

Assistant Atty. General

• Dr. Emmoth A. Luebke Office of the Atty.' Gen..

Administrative Judge One Ashburton Placei-Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Beverly Hollingworth Washington, D.C. 20555 Coastal Chamber of Co mmerce

• Roy P. Lessy,-Jr., Esq. 822 Lafayette Rd.

Robert Perlis, Esq. P.O. Box 596 Office of Exec. Legal Dir.

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Hampton,:NH 03842 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Anne Verge,-Chair Board of Selectmen Atomic Safety.and Licensing Board Panel Town Hall U.S. Nuclear Regulatory' Commission South Hampton, NH 03842 Washington, D.C. 20555

• • Robert K. Gad, Esq.

Atomic Safety and Licensing Board Thomas G. Dignan, Jr., Esq Appeal Panel Ropes and Gray U.S. Nuclear Regulatory Commission 225 Franklin Street Washington, D.C. 20555 Boston, MA 02110

• Ruthanne G. Miller, Esq. Sandra Gavutis Law Clerk to the Board RPD 1 Atomic Safety and Licensing Board East Kensington, NH 03827 U .S. Nuclear Regulatory Commission Washington, D.C. 20555 s'

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George Dana Disbee, Esq. c. John B'. Tanzer Edward Cross, Esq. 5 Morningside Dr'ive Asst. Atty. Generals . Ila mpton, - NII '03842- ,

Office of the Atty. General.

Dr. Mauray Tye, President- Letty Hett,-Selectman Sun Valley Association RFD Dalton ~ Road 98 Emmerson Street Dr entwood, till 03833 Haverhill, MA 01830 Calvin A.-Cannery Edward P. !!cany City Manager 155 washington Rd. City llall R y e , till 03870 Portsmouth, Nil 03801 Docketing and Service U.S. Nuclear Regulatory Commission Washington, D.C. 20555

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Diane Curran-

• By Hand

• • By Federal Express W

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