ML20153A962

Informs Comm of Results of Staff Response to IE Bulletins 77-05 & 77-05A & Related QA Matters.Concludes Internal NRC Communications Could Be Improved.Found No Evidence of Utils Intentionally Misleading NRC

ML20153A962
Person / Time
Issue date:	06/26/1978
From:	Volgenau E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
To:
References
SECY-78-348, NUDOCS 7811290022
Download: ML20153A962 (86)

Text

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k dufS 26 i978 NUCLEAR REGULATORY COMMisslOf6

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SECY-78-348 WASHINGTON O C. 20555 4 'J,.y .

INFORMATION REPORT 0

For: The Commissioners From: Ernst Volgenau, Director DCJ m Office of Inspection and Enforcement Qh }>

Thru: Executive Director for Operations U

Subject:

EVALUATION OF RESPONSES TO IE BULLETINS 77-05 AND 77-05A, STAFF ,z:3 ACTIONS RELATED TO BULLETINS, AND RELATED QUALITY ASSURANCE MATTERS

Purpose:

To inform the Comnission of the results of the staff's evaluation and action.

Discussion: On April 13, 1978, the Commission issued a memorandum and order '

to the staff that included ten directives resulting from the Union of Concerned Scientists petition dated November 4,1977.

This paper addresses Directive No. 6 which states:

" Conduct a comprehensive " lessons learned" evaluation to include the following: (a) review all licensee responses (with partic-ular attention to the Pilgrim case), to determine conformance to applicable quality assurance documentation requirements, as well as the accuracy andetimeliness of information provided.

(Where justified, appropriate enforcement actions should be taken.); (b) review how electrica'l equipment, not fully quali-fied, came to be installed in those plants where found; (c) review staff actions in the Pilgrim case so that similar delays may be avoided in the future; and (d) review the need for further

. regulatory actions to include a possible NRC policy statement to re-emphasize the important safety responsibilities of licensees.

Evaluation to be completed within two months." .

_ Eval ua tion: IE Bulletins 77-05 and 77-05A were sent to 41 utilities operating an aggregate of 65 reactors. The responses to the bulletins indicated that 13 utilities had installed electrical connectors in the safety systems of 19 reactors.

All' responses for these reactors were evalucted to determine each utility's conformance to regulatory requirements for elec-trical connectors. The following seven utilities with an aggre-gate of 10 reactors had electrical connectors for which docu-mentation of qualification to regulatory requirements could not be demonstrated:

Conta c t:

THIS DOCUMiiNT CONTAINS N. C. Moseley, IE 492-3150 P0OR QUALITY PAGES 7811290477 -

c.

The' Commissioners 2 Evaluation: Indiana & Michiaan Electric Co. D. C. Cook 1 (continued) D. C. Cook 2 Connecticut Yankee Atomic Power Co. Haddam Neck Jersey Central Power & Light Co. Oyster Creek Niagara Mohawk Power Corp. hine Mile Point 1 Tennessee Valley Authority Browns Ferry 1 Browns Ferry 2 'd Browns Ferry 3 4

Boston Edison Co. Pilgrim 1 -

Maine Yankee Atomic Power Co. Maine Yankee With the exception of D. C. Cook 1 and 2, Haddam Neck, Browns Ferry 3, and Pilgrim 1, subsequent testing demonstrated that the electrical connectors for these reactors were qualified for use in safety systems. For D. C. Cook 1 and 2, the electrical connectors were qualified for the LOCA environment but not for the steam line break environment. The electrical connectors were replaced. The licensees for Haddam Neck and Pilgrim replaced the electrical connectors. instead of performing all the required qualification tests. The licensee for Browns Ferry 3 had quali-L.

fied electrical connectors but additional potting was required. '

Complete tests were not performed on the unpotted electrical

, connectors because the licensee elected to pot them. ,

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One of these utilities, Indiana & Michigan Electric Company, was cited for being in noncompliance with their quality assur-ance program for D. C. Cook 1; however, a response to the cita-tion was not required, as shown in Enclosure I, QA during con-struction was also investigated at Pilgrim, with no nonco;apliances identified. Conformance of the other utilities at the time these

. reactors were being constructed was not pursued because of the generally undeveloped nature of the regulations in.this area.

The lack of development of regulatory requirements at that time is discussed later. .

All of the responses from the utilities were evaluated for accuracy-and timeliness. The following utilities submitted information which was inaccurate with regard to the following rea,ctors :

Connecticut Yankee Atomic Power Co. Haddam Neck Boston Edison Co. Pilgrim 1

e l

'. .. l The Commisstoners- 3 Evaluation: Enforcement action against these utilities was not taken because -l' (continued) the information provided was preliminary and was subsequently corrected. All utilities, including Boston Edison Company, '

met the modified reporting date as given in Bulletin 77-05A, and thus their responses were timely. 4 l At the time of the investigation, D. C. Cook 2 had not yet been operated.

Nevertheless, enforcement action was initiated against l Indiana & Michigan Electric Company on the basis of material false statements in their application for license. p".. .l '

Installation of electrical equipment, not fully qualified, was due in part to lack of development of regulatory requirements. h l

Regulations and industrial standards related to nuclear power l plant construction and operation are continuously changing.

In this regard, it should be noted that the ten reactors men-tioned above were originally docketed'between 1963 and 1967. '

Construction permits were issued between May,1964 and March, r:

1969, and Operating licenses were issued between June, 1967 and December, 1977. [

While there were general requirements that F:

equipment should be qualified for intended service, the first "

i regulations to specifically include the General Design Criteria were established February 20, 1971. Note that this is two years '

after the latest construction permit was issued for the ten E reactors noted. , , g,

• Likewise, the formal quality ass 0rance regulation was established r

.with the issuance of Appendix B to 10 CFR 50 in June 1970.

L Since the construction permits for all of the ten reactor'; were issued prior to 1970, and there were no formalized backfit pro-visions, none of them had or were expected to have fully imple-mented Quality Assurance programs during their construction  :.

phases, although QA programs were implemented to some degree [

for a few of the reactors prior to completion of construction.

[I-As with the evolving regulations, the review process for nuclear

• power plant applications is also under continuous revision.

The Standard Review Plan was first published in final form in November 1975. While it had been published for comment and was being used in varying degrees as early as 1974, even the final (1975) version was not explicit regarding small electrical

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..l '

The Commissioners 4 l

l Evaluation: components or connections such as connectors, terminal boards *

(Continued) or splices. The inspection program also did not specifically address the types of compo' its listed above, but rather focused on motors, instruments, and cable qualifications.

There appear to be some cases in which individual reviewers or inspectors did .look at some of the types of. components mentioned above, but there was no consistent effort to do .,.

~'n so until more recently. Ed The staff has recognized that, as a result of evolving regu-latory requirements, issues like the environmental qualifica-tion of connectors may need attention on reactor facilities already licensed for operation. As a result, the staff has a l

ongoing programs to address safety levels in licensed plants q in areas where our regulatory requirements have changed. =

As discussed in SECY-77-561, the staff developed a specific program, the Systematic Evaluation Program (SEP), to address  %

the evolution of licensing criteria on older operating facilities.

The underlying need for this program was recognized sometime ago and in fact one of the topics included in the SEP was the environmental qualification of electrical equipment. j l

While the resources and time required to comprehensively address the changes in licensing cr.iteria under the SEP is significant, the staff also has many ac't'inties ongoing which address this cor.cern in a more a.d_ hoc manner: For example, in addition to licensing reviews and inspection activities that respond =

to chan, as in NRC regulations, the staff expends significant resourr when experience, such as research results, licensing review r operating incidents, indicates significant safety concer-- Much of the NRR's Technical Activities Program is aime at such issues. Also, current reviews of license amendments on operating reactors frequently address issues where our licensing criteria have evolved. In these cases, the IE and NRR staff assures adequate safety levels are main-tained. The significant increase in inspection activities that will result from the Resident Inspection Program will enhance this effort.

For the ten reactors identified as having unqualified con-nectors, detailed reviews of their applications were made.

3 7".e Ocmmissioners 5 Evaluation: As a result of this review, one reactor, D. C. Cook 2, was (continued) found not to have been constructed in full accordance with commitments made in their application. Enforcement action ..

~1 ,

2 has been initiated in this case, and is discussed in Enclosure I.

For the nine other reactors, it was found that license '

application statements regarding qualification of. electrical components were vague or nonexistent. These findings are consistent with the status of industry standards and NRC 4

requirements as published at the time these reactors were authorized for construction. A further discussion of the details of the review of the applications for these facilities is contained in Enclosure II.

In addition to the review of applications discussed above, two utilities responded to the IE bulletins (77-05 and 77-05A) with information that was later found to be incorrect.

Separate investigations of the responses from the Haddam Neck and Pilgrim reactors were conducted. The results of these investigations indicate that the original responses were based on the best information available, and that the licensees '

in'both cases had reason to believe that the information was correct.

'It should be noted that earlier information which indicated that connectors.wgre not used at Pilgrim, had been obtained~by telephone from NSSS'and AE representatives and was only the' belief of the indiv'iduals contacted.

it should be noted that the provisions of Bulletin 77-05AAdditionally, modified the required reporting due date for Bulletin 77-05.

Pilgrim' met the modified date.

4 The Commissioners 6 Evaluation: Based on the results of the investigations, no additional (continued) enforcement actions are planned. The investigation reports are included in Enclosures III and IV.

2

Conclusions:

Based on our review of these matters including the results of the Pilgrim Investigation, Enclosure IV, it appears that G coordination and communication between NRR and IE and among IE organizations could be improved. It is fair to say, however, that a large volume of information at a high level of detail was obtained from a large number of diverse organiza-tions within a relatively short time frame. In terms of the quality and accuracy of utility responses, we reviewed a great .d deal of information, performed three special investigations H and have given considerable thought to the matter. We found one case, D. C. Cook, where significant noncompliance findings q were made and high threshhold enforcement action has been j initiated. We did not find evidence that utilities had l intentionally misled the Commission.

The following actions will be taken based on those conclusions:

1.

' The IE/NRR interface group will implement practical improve-ments in interoffice communications. If specific additional l procedures are needed;they.will be developed in this forum, and implemented through line supervision.

2.

- IE, with input from NRR, will assess the need for a priority Bulletin procedure. Such a procedure could be used for those situations where specific information from licensees is required under short turnaround' circumstances. It could provide an identified communication and responsibility chain on a case-by-case basis, thus minimizing the chances for misunderstandings within NRC and by licensees.

3. IE is drafting a letter to be sent to all reactor licensees and permit holders. This letter will stress the importance of the accuracy of information submitted to NRC, both in applications and in response to Bulletins. As a related matter, IE is considering the desirability of improving our enforcement capability related to the accuracy of information submitted in Bulletin responses.

6

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.* E.

The Commissioners 7 s

Conclusions:

An IE Circular -(78-08) was issued on May 31, 1978 to all (continued) _ reactor licensees and permit holders discussing the variety 9"f..

of deficiencies related to environmental qualification of :tu .. . .

electrical equipment. Included in this Circular was a reitera- Zr~

, tion of the Commission's position regarding the need for accu- 2;;;.

rate and timely information. It also pointed out the need for Gai.

licensees to be knowledgeable and vigilant and that they must i.

take more initiative in ferreting out details of potential plant E ;;v weaknesses. =-

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[ E E Vol en ~ ..m; l Director . J ;

Office of Inspection and Enforcement "~l:i

Enclosures:

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1. Letter from E. Volgenau to ";;E l Indiana and Michigan Electric l

. Company dated May 30, 1978 l

2. Review of Applications for-l Reactors Having Unqualified '

Connectors .

3. Letter from W. G. Martin, ~

, Region I, to Connecticut Yankee Atomic Power Company dated May 3, 1978 =-

4. Investigation Report No.

50-293/78-13 dated May 30, 1978 DISTRIBUTION: "

Commissioners '

Commission Staff Offices Exec. Dir. for Opers ~-

Regional Offices Secretariat 1

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ENCLOSURE , - .

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Letter from E. Volgenau.to. Indiana..and Michigan Electric z.

Company dated'May~ 30,1978' 1

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tnciosure i WAY 3 01579 Indir.na and Michigan Electric ~

Cor.nany Docke't No: 50-315 '

Indiana and Michigan Power . Docket No. 50-316 Ccepany .

ATT!i: !<r. John Tillinghast

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Yice President -

P. ?.. T:x 10

• Gowling Green Station New York, KY 10004 l

Gantlecen: i This letter refers to the results of an investigation ennducted during December 1977 and January 1973, into the testing cf electrical cables connectors and penetrations at D.C. Cook Unit "os.1 and 2. At your meeting in my office on Aprilrequest, 12, 1973. the findings of this investication were dis  !

actions which ynu have already taken to prevent recur The . investigation determined that certain statements made by the licensec curing October 1977, to the NRC regardingstesting of electrical penetrations, and instrurent cable for D. C. Cook Unit Ho. 2, were false in that such tests had not been conducted as described.

thit the information presented regarding the testing of electricalFurtherm on the staff's acceptance of the qualificationThese of these ite s. pen

~'

civil penalties Appendix S.

• totaling $10,000 for these items, as set forth i In addition, there are three areas of concern. First, information and stater.ents relative to cable qualification testing were reviewed and incorrect information was found.

While the incorrect information sub itted regarding testing duraticn of power cable vas a catter of concern to the NRC.because the information furnished was false, it was not nw.terial acceptable because it did not affect the stcff's conclusion qualification. ,

on CERTIFIED MAIL sETu.t RECEIFT RE0' JESTED

l

. Indiana & Fiichigan Elec. Co. Zndiana. & liichigan Power Co.

d In November 1977, during a special revie'w by NRC of electrical connector qualifications for Unit No.1 it was determined that qualification could not be demonstrated. During this investigation a detailed review was made

. of your FSAR commitments concerning Unit No.1 connector qualifications, d Although the FSAR commitments' submitted in 1971 did not specifically refer to electricc1 connectors, these commitments did state that all electrical components were properly qualified, and from this statement u it could be inferred that electrical connectors were included. However, in view of the lack of specificity of this commitment made several years ago, and recognizing that the NRC issued an Order confirming the licensee's shutdown of Unit No. 1 on November 18, 1977, to assure the proper qualification of these electrical connectors, we do not plan to pursue this matter further.

During our review of testing of electrical components for Unit No.1, it was determined that six items of noncompliance relating to the QA program occurred during the 1970-1974 time period, and these are identified in Appendix C. We note that had the QA program functioned properly, the failure to test electri. cal connectors could have been identified at that time, and therefore would not have become an issue later. At this time, construction of Unit Nos.1 and 2 have been completed and operations are'under a different QA program. In addition, the licensee has'no nuclear plants,under. construction. Therefore, corrective action pertaining to a construction quality assurance. program is no longer necessary. No response to these items is required.

The NRC regulatory program is, of necessity, based on the premise that information provided by the licensees is factual, complete, and tech-nically supported by data, records, calculations and judgments of technically qualified individuals. The review, evaluation and inspec-tion processes involved in the regulatory program are therefore designed to function on that premise; that is, a program based on sampling and l auditing techniques. '

fnaccurate information could result in decisions which adversely' affect the health and safety of the public. It is, therefore, kperative for j licensees to exercise the utmost care in verifying information furnished i to the NRC. This burden of accuracy must be stressed throughout l licensee organizations.

We view material false statements as serious matters. In,all cases where substantive material false statements are identified, strong i

. enforcement action can be anticipated.  ;

i 1

l l

. - 3 diana & l'ichi an 5 Elec. Co .- -

Indiana -& Michigan Pc,ser Co.

l Your response to this notice and the results.of future inspections will determine if further escalated enforcement action is required to assure future compliance.

Sincerely, Ernst Volgenau Director Office of Inspection and Enforcement

Enclosures:

1. Appendix A - Notice of Violation Occket No.

50-316

2. Appendix B - Notice of l Proposed Imposition of Civil Penalties 50-316

3. Appendix C - Notice of

• l Violation Occket No.

, 50-315 cc: American Electric Power Service t i -

Corporation .

John A. Tillinghast, Vice

, Chairman, Engineering & ,

Construction 2 Broadway -

New York, NY 10004 w

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1 Appendix A

. NOTICE OF VIOLATION ~l i

i Indiana and Michigan Docket No. 50-316 Electric Company Indiana and Michigan Power Company ~g

. J This refers to the investigation conducted by representatives from our. ,

Region III (Chicago) Office and the Office of Inspection and Enforcemer'.  ;

(Bethesda) at the D. C. Cook Plant, Unit No. 2, Bridgman, Michigan, at "1 your offices in New York, and at Conax Corporation, Buffalo, New York, R of activities authorized by NRC License No. CPPR-61 and No. DPR-74. d During this . investigation conducted on various dates between December 9, h 1977 and January 19, 1978, the following apparent items of noncompliance were identified.

.l

1. As a part of the NRC staff review of information submitted by the licensee in support of a license application, the staff transmitted several questions in a letter dated . August 30, 1977. One NRC staff 1 question in part stated: .... l Ouestion 022.16 i "Your response to 022.8 is _ cot complete. We require the following additional informa'tibni ....

2. Identify the safety related equipment that will be qualified by test for service in the environmental conditions expected in the containment icwer compart-

. ment following a main steam line break containment lower compartment following a main steam line break (MSLB) (sic]. Show that the environmental conditions a.3 within the bounds of the environmental conditions simulated in the tests."

Resoonse The licensee's response dated October 19, 1977, which the staff relied upon in reaching its conclusions regarding the acceptability '

of the penetrations qualifications, in aoplicable parts stated: ....

"Part 2 Environmental qualification of equipment re;Jired to fun ~ction following a steam line break is discussed beic .

L -

l .

Appendix.A -

. . ~ . .

Electrical Penetrations Prototypes 'of the penetrations employing identical construction and materials to the penetrations used in the containment of. Unit 2 have been successfully tested by their manufacturer to'the requirements of =

IEEE 3 R Appendix A, Combined requirements of Tables .

~1 Al and 82."

The statement 'is false, in that prototypes of the penetrations ~

employing identical construction and materials to the penetrations ,

used in the containment of Unit 2-were not successfully tested to the requiraments of IEEE 323, . Appendix A, ccmbined re~ quire-ments of Table Al and A2, prior to October 19, 1977, nor had they been terted to this requirement at the time of this investigation.

Contrary to Section 186 of the Atomic Energy Act of 1954 as amended, this false statement was made under oath in careless disregarc of the true circumstance.

. i (Civil Penalt'. - $5,000) .

.t ,

2. As a part of the NRC staff review of infqrmation submitted by the licensee in support of a license application, the staff

, transmitted several questions in- a letter dated August 30, 1977.

One NRC staff question in part stated: ....

Question 022.16 "Your response to 022.8 is not complete'. We require the following additional information: .....

2. Identify the ~ safety related equipmant that will be qualified by test for service in the environmental conditions expected in the containment. lower compart-ment following a main steam-line break containment icwer compartment following a main steam ~line break (MSLB) (sic). Show that the environmental conditions are within the bounds of the environmental conditions simulated in the tests."

Recense

~% 'li:ersee's resconse dated.0ctober 19, 1977, which the staff cd fs:..;cq .ir si:,ir.;.i:  :: :i :n.d c: ge:i g Me .::-: g:4:i; i .y 1

of Lthe qualification testing, in applicable parts stated: ....

f 6 '

' c. s _ .

i

, ,l  :

Appendix A j "Part 2 i Environmental qualification of equipment required to function following a steam line break is discussed belcw.

Instrument Cables Twisted shielded pairs and quads used for instrumentati6n g were supplied by 3 different manufacturers. Two of the ;q manufacturers tested to the requirements of IEEE 323, 0 Appendix A, Figure Al with a maximum temperature of0340 F.

The third supplier tested at 1510 F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, 340 0F for -

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> followed by a decrease in temperature to 160 F H

in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, followed by 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> at 1600F.

Cable Documentation i

The documented results of containment environment tests conducted on prototype samples of cable with insulating q materials identical 'to those supplied to the D. C. Cook i Plant for use in the' service identified above for power, instrument and control cables are available for audit."

The statement is false, in that it.was determined that the two manufacturers did not test instruirlent cable (twisted shielded pairs and quads) but rather tested chble materials ar.d/or control cables.

Contrary to Section 186 of the Atomic Energy Act of 1954 as amended, this false statement was made under oath in careless disregard of

. the true circumstance.

(Civil Penalty - 55,000) -

This Notice of Violation is .sent to you pursuant to the provisions of Section 2.201 of.the NRC's Rules of Practice, Part 2, Title 10, Code of Federal Regulations. You are hereby required to submit to this office, within twenty (20) days of your receipt of this Notice, a written statement e avPlanatina in reply, including for each item of noncompliance; (1) admission or denial of the alleged items of noncompliance; (2) the' reasons for the items of noncompliance, if admitted; (3) the corrective steps which have been taken by you and the results' achieved; and -(4) corrective steps which will be taken to avoid .further noncompliance.

Your respense should focus on the broad issue of steps being taken to assure the accuracy of information submitted to the NRC.

, .j A;ocendix B'

. NOTICE'0F PROPOSED' IMPOSITION OF CIVIL PENALTIES Indiana and. Michigan Electric Company Docket No. 50-316 Indiana and Michigan power Company This. office.has considered the enforcement options available to the NRC including administrative actions in the form of Written notices of violation, civil monetary penalties, and orders pertaining to the _

=-

modification, suspension or revocation of a license. Based.on these considerations we propose to impose civil penalties pursuant to ..

~

Section 234 of the- Atomic Energy- Act of 1954, as amended, (42 USC 2282), ,

and .to 10 CFR.2.205 in the cumulative amount of Ten Thousand Dollars (510,000) .for the specific items of noncompliance set forth in ~

Appendix A to the cover letter. In proposing to impose civil penalties pursuant to- this section of the Act and in fixing the proposed amount of the penalties, the factors identified in the Statements of Consider-ation published in- the Federal Register with the rule making action which adopted 10 CFR 2.205 (36 FR 16894) August 26, 1971, and the

'" Criteria for Determining Enforcement Action," which was sent to NRC licensees on December 31, 1974, have been taken into account.

i i

Indiana and Michigan Electric Company and Indiana and Michigan Power Company may, within twenty (20) days of receipt of this notice pay the civil penalties in the cumulative amount of Ten Thou:;and Dollars ($10,000) 1 or'may protest the imposition of the civ.i.1 penalties in whole or in part by a written answer.

Should the licensed' fail to answer within the time specified, this office will issue an order imposing the civil penalties in the amount proposed above.

Should the licensee elect to file an

, answer protesting the civil penalties, such answer may: (a) deny the items of noncompliance listed in:the Notice of Violation in whole or in part, (b) demonstrate extenuating circumstances (c) show error in the Notice of not be imposed.

Violation, or (d) show other reasons w,hy the penalties should In addition to protesting the civil penalties in whole or in part, such answer may request remission or mitigation of the penalties.

Any written answer in accordance with 10 CFR 2.205 should be set forth separately from'your state:ent or explanation in reply pursuant

'to 10 CFR 2.201, but you may incorporate by specific reference (e.g.,

giving page and paragraph numbers)-to avoid repetition.

'* The licensee's attention is directed to the other provisions of 10 CFR '

2.205 regarding, in particular, We tc answer and ensuing orders; answer, consideration by this office, and ensuing orders; requests for hearings, hearings and ensuing orders; compromise; and collection.

9 8 q a  !

' Appendix B ,

2-lipon failure to pay any civil penalty due which has been subsequently determined in accordance with the applicable' provisions of 10 CFR 2.205, the matter may be referred to the Attorney General, and the penalty, unless compromised, remitted, or mitigated, may be collected by civil action ~ pursuant te Section 234c of the Atomic Energy Act of 1954, as amended, (42 USC 2282).

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Appendix C NOTICE OF VIOLATION _

Zndiana and Michigan Electric Company Docket No. 50-315 fndiana and Michigan Power Company This refers to the investigation conducted by representatives from our Region III (Chicago) Office and the Office of Inspection and Enforcement (Bethesda) at the D.C. Cook. Plant, Unit No.1, Bridgman, Michigan, at your offices in New York, and at Conax Corporation, Buffalo, New York, of activities authorized by NRC License No. C?PR-60 and No. DPR-58.

1 Based on the results of the investigation conducted during the period l December 9, 1977 through January 19, 1978, it appears that certain of l your activities relative to the design, fabrication, receipt and testing of D. C. Cook, Unit 1, electrical' penetrations were in noncompliance with NRC requirements as noted below. These are infractions.

1. 10 CFR Part 50, Appendix B, Criterien III states in part:

" Measures shall also be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems, and components."

l Section 5.1 of the D. C. Cook Nuclear Plant Quality Control Manual states in part: "For safety-related equipment in the AEPSC scope of supply, the cognizan%ngineer will prepare the design and specificatiens, integrating these with the nuclear steam supply system at each interface . . . . He will have the specifications checked for quality provisions by the Quality Assurance Engineer who, may obtain the recommendations of Southwest Research Institute . . . . Design changes will be processed through the same channels' of engineering that handled the initial design." .

Contrary to _the above: It was determined that during the period between November 23, 1970 and April 30, 1971, that design require-ments were changed to add multi-pin connectors to certain electrical penetrations for D. C. Cook, Unit 1, without assurances that appro-priate quality standards wer'e specified and included in the design documents. The penetrations involved included Nos. lIS, 2I8, 318,

- ~

415,1C4, 2C4, and 4CI (Type EP-11 and EP-12).

2. 10 CFR Part 50, Aopendix B, Criterion XI states in part: "A test prcgram shall be established to assure that all testing required to demcnstrate that structures, systems, and ccmponents will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the

' 1 : r s ' ad :::c:t:-ce liHts contained in apolicable design doc' :,so;5. "

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7 Appendix C Paragraph VII.B.4 of AEPSC Specifications No. DCCEE-ll5-QCN ~

Revision 1, Addendum A dated April 15, 1971, states in part:

'" Prototype testing compliance may be demonstrated by the seller, by the submittal of test data from tests. conducted on penetra-tions of equivalent. type and design as those being furnished for Donald C. Cook Nuclear Plant. Tests must comply with all require-ments of this specification."

Paragraph 1.7.11 of the D., C. Cook FSAR states in part: " Systems, subsystems, and components of. systems related to safety require proof or acceptance testing. Test procedures included check lists of prerequisites that had to be completed prior to testing, and the controlled ' conditions such as environments, test equipment, test fluid, and safety precautions under which the tests were run."

Contrary to the above: The test program conducted during May-July, 1971, end used to verify the adequacy of certain electrical penetra-tions for D. C. Cook, Unit 1, was not adequate in that the prototype penetration tested did not include multi-pin connectors. The D. C.

Cook penetra:: ions involved included Nos.115, 2I8, 318, 4I5, IC4, 2C4, and 4CI (Type EP-ll and EP-12).

3. 10 CFR Part 50,' Appendix B, Criterion IV . states in part: " Measures shall be established to assure the 'aiplicable reg;1atory require-ments, design bases, and other requirements which are necessary to insure adequate quality are suitably included or referenced in the documents for procurement of material, equipment, and services, whether purchased by the applicant or by its contractors or s ubcontractors. "

Paragraph VII.C of the AEPSC Specification N'o, DCCEE-ll5-QCM Revision 1, dated October 23, 1970, states in part: "A sample of each material prcposed to be used in the penetration shall, be tested for suitability for service under the radiation dosages listed in Section VA.1 and results submitted to purchaser."

Contrary to .the above: Applicable regulatory or other requirements were not included or referenced in the procurement documents for connectors purchased on various dates between August 20, 1971 and October 2,1974, for certain electrical penetration or as spares nor were the connector materials tested for suitability for service.

4 10 CFR 'Part 50, Appendix B, Criterion VII states in part:

" Documentary evidence that material and equipment conform to the prccurement recuirements shall be available at the nuclear oower piu: si:s :rier c installa:ior or use :f sucn riterial anc equipment."

A:pendix C - 3-Paragraph 1.7.7 of the D. C. Cook. FSAR states in part: "Ma te ri al ,

equipment, and services purchased' for systems important to safety -

were inspected to see that they conformed to the purchese documents.

This inspection was made in the vendor's shop and/or on delivery of the product by the cognizant engineer or his representative or the field organization. . . ..."

~~

" Material certifications, test reports, certificates of code inspection, review of deviation reports and other documentation _~ ~ . ."

were conducted in the inspection evaluation."

Contrary to the above: It was determined that between September and November 1972, certain electrical penetration's for D. C. Cook, Unit I were received and subsequently installed without. documentary evidence that they met requirements. The penetrations involved included IP6, 1P7, 2PS, 2P6, 3P7, 3P8, 4P7, and 4P8 (Type EP-1).

5. 10 CFR Part 50, Appendix B, Criterion XI states in part:

. . . . . Test results shall be documented and evaluated to assure that test requirements have been satisified."

1 The FSAR states that Section 1.7.11, entitled Test Control addressestthe following. " Planned test procedures that verify functional acceptance of materials, equipment and systems are prescribed." 74 Paragraph 1.7.11 of the D. C. Cook FSAR stated in part:

".... Test results were reviewed by the cognizant engineer and the results certified as satisfactory by him or referred back for modifications and retest. This certification was in accordance with the acceptance limits specified in the design, code cn purchase specification or the requirement of the functional system."

Contrary to.the above: It was determined that even though on or about May 10, 1972, the engineering review and evaluation of the test report relative to. accident operating environment tests was completed for the D. C. Cook Unit 1 electrical penetrations, that review and evaluation was inadequate in that it did not identify the lack of' test data. for some penetrations or that other penetra-tions.were not. listed in Table 1 of the report. Table 1 of the report compared D. C. Cook penetrations to the prototype penetra-tions which had been tested. The types of penetrations omitted from Table 1 include EP-1, EP-ll and EP-12.

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Appendix C

" Measures

6. 10 CFR Part 50, Appendix B, Criterion XV, states in part: ,

shall be established to control materials, parts, or componentsThese measures sh which do not conform to requirements . . . . m_

include, as appropriate, procedures for identification, documenta-tion, segregation. . ."

Section 8.3, QCM 8.3-1 of the D. C. Cook Nuclear Plant Quality Control Manual states in part: "If a required reference document has not been received, the cognizant enginee'r must be contacted q.

to obtain the document."

"The documents required with the' shipment should first be checked for existence. As prescribed by the cognizant engineer, they should also be verified for correctness of information by compar-2 ison with the reference documents."

"Upon completion of the inspection, the equipment or material shall be tagged for acceptance or for hold or rejection and processed outlined in QCM, Section 14.4.

be fnrwarded to the cognizant engineer."

"For Paragraph 1.7.G of the D. C. Cook FSAR states in part:

systems and components important to safety, the cognizant engineer was responsible for and directs the applicable control and identifi-cation of material during storagg.and. erection. Procedures were used to prevent the use of incorrect or defective materials, and material which had not been qualified by the required inspections

- and tests."

Contrary to the above: It was determined that at least eighteen (18)' O. C. Cook, Unit 1 electrical penetrations were received onsite 14, 1971, but were not identified or tagged as prior to Decembernonconforming for lack of documentation to establish that th The penetra-environmental and certain other test requirements.

tions involved included Nos. 4P6, 2P3, 4P3, 3P3, 2P7, 3C2, 3P2, 4P4, 4C3, 2C1, 1P2, 1P4, SP6, 2P4, 1P3, 4PS, 2P8, and IC3 (Types EP-2, EP-3, EP-4, EP-5, EP-7 and EP-13).

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ENCLOSURE 2 4 *

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i Review of Applications for Reactors Having Unqualified Connectors l S

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APPLICATION STATEMENTS FOR REVIEM OF OPERATING PLAfflS WITH ELECTRICAL CONNECTORS

Background

During the review of the responses to IE Bulletins 77-05 and;77-05A submitted by the licensees of operating nuclear power plants, the NRC i staff learned that seven sites _had insufficient documentation to assure l that electrical connectors used in safety systems were environmentally .. j qualified.

Scope As a result of the identification of these seven sites, a review was J undertaken to determine if false' statements existed in documents submitted -  ;

as part of the application addressing the' subject of electrical connectors.

For one plant, D.C. Cook, Unit No.1, which was required to be shut down on an innediate basis, there was an_ extensive investigation conducted which '

subsequently overinpped into a review of information related to D.C.

Cook, Unit No. 2, a duplicate plant, which at the time of issuance of the bulletins (11/77), was nearing an operating license. Because the Unit No. 2 effort was related to a facility which at the time did not have an operating license the results are-not contained herein. The 1 details of the investigation for both D.C. Cook units is documented in .1 lE Investigation Report No. 050-315/77-28; 050-310/77-41. The investi-gation consisted of extensive examinati.o'nt of procedures and records, and interviews with personnel at the plant site, ,the licensee's headquarters and the vendor's offices.

The review ofeffort.

a two-stage the other six sites with connectors was scheduled to be First, a review of licensee commitments and statements which are part of the. application records was to be conducted for each plant. If definitive statements were found regarding the qualification of electrical connectors, the second stage would proceed. The second -

stage would include the examination of procedures and records along with interviews of personnel of the licensee, the licensee's agents, and vendors for additional. information. The information gathered from the i

inspection would then be used to identify any necessary enforcement

, actions. l The discussion that follows presents for each of the original seven sites the details of the documents reviewed and the-conclusion reached. These-reviews were' conducted to determine whether material false statements were made by the licensee to the NRC relative to qualification of electrical connectors ,in safety (systems and instrur.entation.

l~

l

4 Inc'seven specific sites' involve'd in the review were:

Docket No. Plant. Date of CP Date of OL  !

1. 50-315 D. C. Cook, 3/69 10/25/74 ]

Unit No. 1

2. 50-213. Connecticut Yankee 5/64 6/30/67-POL j 12/27/74 .z .

3. 50-219 Oyster Creek- 12/64 4/9/69

4. 50-220 Nine itile Pt. 4/65 R 8/22/69-POL Unit No. 1 12/26/74 -l

5. 50-259 Browns Ferry 5/67 6/26/73 -. ; ;_

Unit No. 1 "

50-260 Browns Ferry 5/67 2/28/74 h

Unit No. 2 50-296 Browns Ferry 7/68 7/2/76 Unit No. 3 ._

6. 50-293 Pilgrim, 8/68 6/8/72 Unit No. 1 7.. 50-309 liaine Yankee 10/68 9/15/72

+s Plant Scecifics -

3

. D. C. Cook, Unit No. 1 (50-315)

An examination of the application records revealed that no specific references o'r commitments were made by the licensee as part of the application with regard to environmental qualification of electrical -

connectors. Neither were specific references to any existing industr codes or standards made part'of the docket for D.C. Cook. Unit No. 1 which. related to this subject. The references to environmental quali-fication which were made all were found to be general in nature.

.' .' The general topic of " compatibility of safety related equipment and post-accident environments" emerged in 1968 during the staff review of  ::-n the PSAR prior to the issuance of a CP. A meeting was held with representatives of the applicant and the various agents of the applicant '

and members of the then DRL staff. After the meeting the applicant "

documented the position found acceptable by the staff and identified those positions requiring.further consideration. The letter, dated . .; .

September 30, 1968, stated in part, the following:

.- 4;E!

"VI. Compatibility of Safety Related Equipment and Post-Accident =

Environments

" Safety related equipment located in the containment is considered subject to materials compatibility requirements if it falls in one or more of the following categories:

A. It facilitates emergency core cooling.

3. It facilitates emergency coeling of'the containment atmosphere.

C. It facilitates the removal of fission products from the containment atmosphere.

D. It insures against leakage from the containment.

E. It provides data to the operator which informs him of the status of safety equipment and which serves as a guide to his further. action to control or initiate recovery operations. ,,

"Experinantal evidence is considered necessary if no prior information is available covering the effects of environmental l conditions, or if the nature of the materials and configuration of the component does not preclude, beyond reasonable douot, ability to withstand the environment.

" Conditions must be equivalent to the ressure, temperature [and significant chemical environment of. the accident corresponding to the period of dependency of that component to perform its safety function. Environments to be cons;dered will include contaminants which may be leached from materials exposed to the ,

accident environment."

While this statement was very general and would not have constituted a false statement, no evidence of even a similar statement was found in the material which constituted the. application for a license.

The staff in the-safety evaluation report, dated January 14, 1969 stated the followir)g concerning the program proposed by the applicant.

1 -

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"The applicant has stated that he will attempt to verify from existing test data that the equipment, instruments and cables

-located in the primary containment that are required to function _._

under accident and seismic conditions will be capable of per-forming-their function.. If such data do not exist or are not adequate, the applicant has stated that tests will be performed 1

to verify equipment performance."

The current revies then focused on the statements made by the licensee in the FSAR regarding testing of electrical penetrations since the connectors were considered by the NRC to be logically part of the pene-trations as opposed to the cable. In section 5.2.4.1 of the FSAR the following statements were made.

" Inspection and Testing

" Electrical Penetrations - Prototypes Tests

" Prior to commencement of full production, a production prototype of the electrical penetrations listed in Table 5.2-3 must success-fully pass, in sequence, those tests indicated.by X.

"Upon completion of the above tests, each prototype must successfully pass, a second time, the High Potential and Leakage

, test prescribed.

"In addition to the prototype test listed in Table 5.2-3 all materials used in the penetrations will be quality control inspected, tested, and approved for service under operating and accident radiation

, dosages.

" Electrical Penetrations Production Tests "Each completed electrical penetration must successfully pass thle following tests prior to shipment:

1. Leaking

2. Conductor Continuity Test

3. High Potent'1a1 Test

4. Insulation Resistance l

l l

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" TABLE 5.2-3 ---

ELECTRICAL PENETRATION - PROTOTYPE TESTS Penetrations Test & Se auence 5 Kv Power 600 Volt Power and Instrumentation Penetra tion Control Penetration Penetration High Potential X X X Insulation Resistance X X Impulse X X

Corona X Leakage X X X Ampacity X X Accident Operating Environment 0 0 0 Short Circuit 0 0 0 Conductor Continuity X X X Note: 0 = Tests Conducted Simultaneously."

Section 7.5 of the FSAR addressed the environmental capability of electrical equipment of time neededrequired to function and prpvided a list of equipment and length to operate.

6elow.

The pertinent portions of Section 7.5 are quoted t i.

" Environmental Capability .

"The engineered safety features instrumentation equipment inside the containment is designed to operate under the accident environment of a steam-air mixture and radiation.

" Electrical equipment for the engineered safety features is located inside the containment and in the auxiliary building. Table 7:5-2 is a listing of the equipment inside the containment which is required for post-LOCA operation and indicates how long the equipment is required to function as well as specifying which components require qualification testing. F "An environmental testing program has been completed on components of the type to be used in the containment. Results are available and are summarized below."

The components listpd did not include electrical connectors.

i The l'censee was asked a series of questions which were answered and became part of the FSAR. The questions as part of Section 7.20 were as follows:

l 7  :

" QUESTION 7.21- l Provide the following additional information with regard to the electrical and mechanical equipment of the reactor protection containment.1/ system and the engineered safety features locate i

" QUESTION 7.21.1 "

Identify all safety related equipment and components (e.g. d motors, i?'

o detectors, instruments, controls, cables, filters, pump sea,ls, etc.) '"

located in the primary co'ntainment and required to oe operable during accident. and subsequent to a loss-of-coolant or a steamline break ^~~

this equipment.It is noted that Table 7.5-2 in the FSAR lists some'of complete. Complete the listing or confirm that it is ,

" QUESTION 7.21.2 State the testing that has been performed to assure the operability of those items in the combined temperature, pressure, humidity, radiation, accident and chemical environment associated with the postulated conditions.

. " QUESTION 7.21.3 \

have been performed on each ofIdentify thesethe items. Describe th applicable qualification test documentation for each item and provide such documentation for each item and provide such docu-mentation not previously submitted.

i

" 1

. QUESTION 7.21.4 i Your response should include the bases for the te and procedures and a discussion of how the testing is equivalent

. deterioration and wear due to normal long-term ope -

" 1/. Please note that our previous questions (July 13, 1971) requested such' in Chapter information 5,.6, andregarding 14 o specific functions and systems described certain of your response,s;f the FSAR. You may cncese to consolidate be made,regarding the location of the requested informationhcwe .

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l The licensee provided an answer which referred to Table. 7.5-2 of the FSAR as a complete list of those items required for operating during or after a LOCA or' 115L3.  ;

7.5-2. It was indicatedElectrical connectors were not mentioned in Table that the required environmental tests were done  !

on the items in the table. j No instances of material false statements were found with regard to electrical connectors since these components were not specifically mentioned in the FSAR or the answers to staff questions which became 1 '

FSAR~ame ients. The electrical connectors were supplied by the vendor with the penetrations or in some cases as loose spare parts. Because +q "

of this, one could reasonably conclude that the electrical connectors -

j

- should have been considered part of the penetration assembly.

{

In the review of the facts surrounding the use of electrical connectors at 0.C. Cook Unit No.1, IE identified six items of apparent non-compliance which were related to errors in the implementation of the QA H program.

Had the QA program functioned better in the case of the connec- j tors, been

.have the lack of proper qualifications for the Unit No.1 connectors could prevented. .]

incluced in IE Investigation ReportThe specific050-316/77-41.

050-315/77-28; details of these QA deficiencies are i Connecticut Yankee (50-213) 1 The review of the application records indicated no mention of electrical I connectors and no specific listing of components and systems required to meet specific environmental conditions.

Information was found which indicated the concern environmental qualification. of the staff andut])e..ACRS in the general area of

. During were heldthe on CP review12ofand December the13,1963 application, documents indicate that meetings Yankee, ~ the staff of the Division of Licensing and Regulation, and an AC Subcommittee.

As a result, the need for additional technical information was identified and a set of questions was transmitted by E. letter frob E. G. j Case to the licensee dated December 27, 1963. '

Question 9.b of the question set is quoted below.

"9.

Submitto:the design specifications which reflect the consideration

.given i

"b.-

Ability of all components and instrumentation within the containment to' withstand damage under both test pressure and accident conditions."

The a licensee's letter response dated January 15,~1974was submitted as Amendment #2 to the application The reply as stated in the amendment was:

1 i

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"All equipment necessary for safe depressurization of the containment after an accident, and all instrumentation necessary for monitoring tge accident will be specifically designed to operate in a 260 F, 40 psi gage steam-air atmosphere for indef-inite periods of time."

No further detailed information which would expand on this information was found during a review of the records for Connecticut Yankee. .!o '

follow-up questions were found, no indication that the design specifications were submitted was found, and no further infcrmation was found in the FSAR and subsequent OL review documentation related to connectors and their environ-mental qualification. The staff safety evaluation dated May 11, 1967 did not specifically address this subject either. -

Upon completion of this current review of the licensee's commitments during the licensing process it was determined that there could be no finding of material false statements made relative to the qualification of electrical connectors the CP review.due to the generality of the response to Question 9.b. during Additionally, the response was stated in the future tense.

No material false statements were found and it was concluded that further effort related to the subject of connectors would not be productive of instances requiring enforcement action.

Oyster Creek (50-219)

The only related information found during; a -review of the application records indicated that the staff had asked one question during the OL review which addressed environmental testing. The referdnce was contained in a letter to tha licensee dated June 21, 1977. The licensee was asked to describe the testing and qualification program used to determine tne behavior of sensors in the containment during a LOCA. Ho similar request for detail,s in the area of electrical connectors was found during a review of the files. .

The Safety Evaluation Report dated December 23, 1968 did not address the subject of electrical connectors or their qualifications for the accident environment.

In a letter dated August 6,1970, the NRC staff asked a question which dealt in general with the subject of the adequacy of safety systems in adverse environments. The question is quoted below and the pertinent parts of the response are summarized.

-9 "3.a. In light of the recent incident at Dresden Unit No. 2 caused by the inadvertent opening of the bypass valve as described in the enclosed letter from Commonwealth Edison dated July 6, 1970, provide the following information:

"An evaluation of the adequacy of the primary contain-ment and all safety system components therein for eE.

service at the maximum temperature and other environ-mental conditions which might be reached as a r~esult of steam releases inside the drywell. Safety and protective system instrumentation cables and standby gas treatment filters should be included in the evaluation."

The answer submitted as Amendment 62 on September 8,1970, indicated that the design environment inside the containment would be 62 psig :t and 2810 F. It was indicated that fully assembled penetrations were subjected to various environmental tests up to 124 psig and 3250F (saturated at 62 psig at steam) 3090F. and held for 30 minutes then held for 23-1/2 hours Drywell cables were also addressed. The con-clusion was that the cable installation will perform satisfactorily during a loss of coolant accident. Equipment was discussed in terms of tests parformed on electro-magnetic relief valves, Limitorque v' alve operators, and solenoid control valves, and the conditions of tes ti ng. _ The licensee ended the responsecto the question by stating the following: ,

"It is concluded that the design of the Oyster Creek Containment System and the drywell systems required during a loss of coolant accident is adequate to ;j

)

protect the public, to provide sufficient indication  !

of the status of the reactor, and to allow for the proper operation of safety systems in the event of .

a loss of coolant accident in the drywell."

As the review for a full-term license was begun, the licensee provided additional information in the subject area of environmental testing of electrical equipment and associated components. In a submittal dated March 7,1972 the. licensee addressed the Class IE electrical systems as defined in IEEE Standard 308, " Criteria for Class IE Electric Systems

.for Nuclear Power Generating Stations." The licensee stated the following in two sections submitted to the Commission.

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"Desian Basis Event Effects ,

" Design basis events were established for the individual plant and the severity and magnitude of each event defined. The Class IE electric system is capable of performing its function when sub-jected to 7.he effects of any design basis event at its location save for the possible exception noted previously in Section 4.4.1 General. ,

" Equipment Qualification i

"Each type of Class IE electric equipment was qualified to perform l its function under normal conditions by preoperational tests and l start-up conditions. d d

l

" Class IE electric equipment was specified to perform its function under the anticipated design basis event (s); vendors were advised of seismic requirements."

i The exception noted in Section 4.4.1_ General was as follows.

"The 4160 V switchgear units supplying power to ECCS equipment are both located in the same room of the Turbine Building, a seismic 2 structure. Also, the control leads emanating from these switchgear units are routed in exposed trays in the  ;

Turbine Buildings; however, the tr.ays follow separated paths. I Power cables are installed in conduit routed in the fill under the Turbine Building." '

[

The review of the information provided by the licensee as part of the license review process revealed no mention of, or statements made regarding.the qualification of electrical connectors. No specific questions were found which had been asked by the staff on this subject.

The licensee made no statement which would indicate that environmental qualification tests had been performed on any specific electrical related parts such as connectors. In fact, the licensee indicated that electrical equipment qualification consisted of their specifying that the items should perform under anticipated design basis events. No material false statements were found and it was concluded that further effort related, to the subject of connectors would not be productive of instances requiring enforcement action.

- [.

Nine Mile Point, Unit No. 1 (50-220)

A review of the information contained in the application records revealed considerably more in the general subject area of the performance of electrical equipment in the anticipated environments, including LOCA, than in the other sites reviewed.

During the review of the FSAR, the staff asked the following question as Question #5.

" Identify components or equipment which must function during and follwing the design basis accident (DBA). Specify the design rating Jif the components and equipment pertinent to their opera-tion while coping with the DBA.

" Describe the qualification or preoperational testing that has or i will be accomplished to determine that equipment will function in the accident environment."

The licensee's response was submitted as the 1st Supplement to the FSAR and was filed May 20, 1968. The response in part ' stated that:

i "The electrical ecuipment within the primary containment that must ,

function in an accident environment consists of a-c electric motor-operated isolation and core spray system valves (FSAR, Table VI-3),

with their associated limit switefies"and electrical cabling, and the solenoid-ar.tuated relief valves on the, main-steam lines.

"Except for the electrical solenoids on the reactor power relief valves, this equipment is designed to withstand continuous exposure to a drywell environment to 150 F and 100 percent relative humidity, and short-term accident conditions of 62 psig saturated steam (310 F).

"The solenoids on the reactor power relief va'1ves are designed for environmental conditions of 150 F maximum temperature and 60 percent humidity."

In a letter from the NRC staff there was an additional inquiry into the general area of performance in a post-LOCA envi,ronment. The question and answers were provided in an addendum to the licensee's petitions to increase power, dated October 1970. The pertinent portions of that question (Question 10) and reply are quoted below.

' . l

'"10. In light of the recent. incident at Dresden-2 caused by inadvertent opening of the bypass valves, as described in the enclosed letter from Comuonwealth Edison dated July 6, 1970, pruvide the following. information as applicable to Nine Mile Point fluclear Station: 9 "a. Evaluate the adequacy of the Nine Mile Point primary i containment design temperature of. 310 F for the maxi- H mum temperatures and other environmental conditions -

which might be reached as a result of steam release 1

inside the drywell. Include the basis of qualification of all safety and protective system components and instrument cables required to function following .a LOCA =

to operate at the maximum temperatures achievable for-the requisite period of time."

_[

Reply: _

"A Limitorque valve operator with the same design rating as those installed at Nine flile Point was performance-tested f continuously for six days. The valve operator was exposed i to saturated steam at pressures up to 90 psig. Temperatures ranged from 250 F to 335 F during the test period. 'l "An at electromatic 62.psig and 300relief-valve F. was tested 10 consecutive hours Results of this test show that the j

electromatic relief-valve performed satisfactorily and was j capable of producing the requ.

period." #'

ired lift for at least this time As the first IEEE standards became available for use, the licensee began to i address .the environmental qualification of electrical equipment to those standards. In a technical supplement to the petition for a full-term license '

the licensee submitted'the following information.

, was July 5,'1972. The date of the submittal "8. guplifying' Class 1 Electric E0uiement for Nuclear Power Gyeratino Stations (IEEE 323-April 1971)

"The qualification program for Class I elactrical equipment is discussed in the FSAR.101.The protection and engineered safe-guards systems'-instrumentation has been selected on the basis of demonstrated performance in similar applications and environ-ments on othed nuclear installations. Therefore, this instrumentation is considered to have passed qualification testing.

No. 1 The instrument cables used at Nine title Point Unit 4

cable c are high-temperature cross-linked polyethylene (Vulkene) month. gble of withstanding 3a0 F for a period of one a

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- 17. -

"The equipment that must withstand the most severe environment is that which is in the primary containment. "

Class I equipment was orginally designed to withstand short All of this term accident conditions of saturated steam at 62 psig and 310 F except'the solenoids on the electromatic relief valves.

humidity.were originally designed for 150 F and 60 percent relative These General Electric Company conducted a special test program on an electromatic relief valve actuator under severe During this qualification test the relief valve perform satisfactorily. =~i In addition, a Limitorque valve operator, typical of those used on the motor-operated valves inside the primary containment was also performance tested under severe accident conditions. The actuator was exposed to saturated

~ 1.

250 F to 335'F for six days. steam at pressures up to 90 psig

- an It, too, performed satisfactorily "

There was one instance during the review of the seven plants where a specific question on electrical ' connections or connectors was found and _;

it appeared in the records for Nine liile Point, Unit No. -1 as part of the full-term license review.

licensee on September 28, 1973 The question was transmitted to the and was as follows:

"11. The following additional information regarding electrical and instrumentation systems is required.

"d.'ArecentincidentoccuiNnginaB'.lRresultedintemperature _

inside the primary containmen't exceeding those temperatures located inside the containment.specified in the design of equi Provide the results of environmental qualification type tests for all electrical Class IE equipment, including electrical penetration assemblies andrequired are connections, located inside the primary containment that for safety.

Section llI.C.3 and 8 of the Application." Include this information in The reply to part d. is quoted below.The licensee's response , 1973. was pro "d.

Environmental- Quali fications The following Class IE environmental type tests were performed on electrical equipment at Nine !!ile Point No. 1.

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"(1) Electrical penetrations - I These penetrations were terted for leak tightness under the following accident environmental conditions:

Temperature Containment 310 F Reactor building 50 - 150 F Pressure 62 psig . . .

Relative Humidity 100 percent

"(2) Limotorque valve operators These operators were exposed to saturated steam at pressures up to 90 psig and temperatures ranging from 250 F to 335 F.

The operators performed satisfactorily. ~',

"(3) Electromatic relief valves These 300 F. valves were tested for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> at 62 psig and The vaivt function during this time.vas capable of performing its required

"(4) Control Cable

. High temperature cross-linked polyethylene (Vulkene) cable was used, period of onewhich is capabl,e of withstanding 340 F for a month." D' The review of the facts then focused on whether the omission of could be considered a material false statement.about " connection question had in fact clearly requested the results of all environmentalIE decided qualification type tests with some emphasis placed on the electrical penetra-tion assemblies and connections. It was also the opinion of the rece'nt reviewers of these documents that the licensee answered the question clea The answer indicated, by the fact that no results were provided on electrical .

connections, connections. that no tests were performed to environmentally qualify the The review of the facts then focused on what staff follow-up to this reply was completed si.nce there was apparently a need for the information.The only follow-up question related to this subject was transmitted to the licensee inquiring about the capability of main steam line isolation valve position switches to perform in and after an accident.

on connections was fcund in the record. No follow-up

As a consequence of this recent review of the record contained in the files, the conclusion was reached that the licensee did not make any specific statements related to connectors. It is also clear that by the licensee not responding to the direct question regarding the test results on connections, the tests had not been performed. No material false statements were found and it was concluded that further effort related to the subject of connectors would not be productive _ of instances requiring enforcement action.

Browns Ferry, Units No. 1, 2 and 3 (50-259/260/296)

The review of the application records indicated a reference to electrical connectors as follows. This quotation is from Section 7.2.3.10 of the FSAR which addressed wiring.

" Wiring and cables for Reactor Protection System instrumentation

~ are selected to avoid excessive deterioration due to temperature and humidity during the design life of the plan. Cables and connectors used inside the primary containment are designed for continuous operation at in ambient temperature of 105o F and a relative humidity of 99%." l In Appendix I of the FSAR the licensee stated that testing would be i performed, but the commitment to perform environmental tests was only for

' cable and valve operators. The pertinent parts of Appendix I are quoted beloW. 4 I

" Electrical Equipment Inside Containmen't -

Electrical equipment that must operate inside primary containment in an accident environ-ment is limited to cables and operators for isolation valves. . . .

"The tests will demonstrate that the material and eouipment will survive the accident conditions of simultaneous pressure, temperature, and humidity for a period of time essential for their operation.

Successful demonstration by these tests will satisfy our requirements.

" Type tests of typical valves have been performed and are reported in Millstone FSAR Amendment 18, Question A-9, Docket No. 50-245.

The quality control plan will ensure that components identical to those which successfully passed the tests will be installed at Browns Ferry."

The staff followed up on this reply with a question list which included Question 7.7. The list was datec May 22, 1971.

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A' .

. .' H 7.7.

In Appendix I of the FSAR.you state that the containment environmental type test data presented in the Millstone Unit -

No.1 FSAR, Amendment 18, Docket ilo. 50-245 are directly applicable.to Browns Ferry since the.same safety-related equipment and components will be installed. In this regard, confi rm that: . . . .

(3)thesafety-relatedcomponentsandequipmentareidentical 1

to those used in Millstone or submit qualification test  ;.

results for those which differ." ~"'

3 The licensee's response stated:

"The safety-related components and equipment within the Browns Ferry containments are identical to those used in fiillstone except that the.d-c solenoid operators on the' relief valves and main 125 volts. steam isolation valves are rateJ at 250 volts instead of The construction of the solenoids is the same and the change in voltage should have no effect on their environ-mental capabilities; therefore, no additional testing is considered necessary."

I J

There was no indication in any of the records reviewed that the specific subject of environmental qualification of electrical connectors was {

discussed by the licensee or the staff. It appears that the concern j focused during equipment this time frame on the ctble and the specific pieces of or ini.truments.

f!o material false statements were found and it was concluded that further effort related to the subject of connectors

. would not be productive of instances requiring enforcement action. "

, Pilorim, Unit No. 1 (50-293) n A review of the application records for statements and information in the subject area of environmental qualification of' electrical connectors at Pilgrim, Unit No. I shows the FSAR was submitted on December 31, 1969 i i

and states the fol.. lowing in Section 7.2.3.10 which addresses wiring.

" Wiring and cables for Reactor Protection System instrumentation are' selected to avoid excessive deterioration due to temperature ~l and humidity during the design life of tne plant. Cables and  ;

connectors used inside the primary containment are designed for continuous operation at an ambient temperature of 1500 F and a re'lative humidity of 99%."

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- . ~ . . --. - -- - - . - . . - - . . - . . -

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- With regard to the components of the reactor protection system, the

.FSAR stated the following in Section 7.2.4.

d "The environmental conditions in which the instruments and equipment of..the Reactor Protection System must operate were considered in setting the environmental specification given in Table 7.2.2.

For the instruments located in the reactor or turbine buildings, the specifications are based on the worst opertte.

must expected ambient- conditions in which the instruments -

The Reactor Protection System components which are-located inside the primary containment and which must function in the environment resulting from a break of the nuclear system process barrier inside the primary ' con -

tainment are the temperature equalizing columns and condensing. chambers.. Special precautions are taken to ensure satisfactory operability after the accident. The temperature equalizing columns and condensing chambers are similar to those:that have successfully undergone qualifica-tion testing in connection with other projects.

l

" Table 7.2.2 REACTOR PROTECTION.SYSTEit INSTRUMENTATION ENVIRONMENTAL CONDITI0iIS

[.' 4.

3 Ins trumant Temperature Pre,ssure Location Range Relative Range Humidity Range Primary Containment 40-1500F -2 to +62 psig 20-99%

4 Reactor Building- 40-1400F Atmospheric 20-99%

Turbine Building 40-1200F Atmospheric 20-99%

• Main Control Room 60-800F Atmospheric 4

20-60%"

During the review of the FSAR, the staff asked several questions in the broadequ'ipment.

electrical area of environmental analysis and testing with respect to June 16, August 24, October Questions 20, andwere Decembertransmitted to the licensee on-1 of 1970 .

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The licensee responded to the staff's questions in several amendments to the.FSAR. In Amendment 13, which was submitted on January 14, 1971, the licensee responded to' Question 7.1.5 which requested the design criteria for ~the reactor protection system and engineered safety fea-

'tures for related electrical and mechanical equipment located anywhere i

in the plant with respect to all radiation effects. There was also a request for a description of the analysis and testing performed to verify compliance with the criteria. The response in part stated:

" Safety. system equipment shall be capable of withstanding without B

' loss of function the potential effects of. the total integrated radiation dose from normal operation during the expected life-  ::]

time of the station plus a one-time dose from an accident chpo s u re .

In addition, the lict d ce referenced GE Topical Report, GE-APED-5756, i.)

dated March 1969 and testing which had been performed on cabling with  :.J regard to radiation. No mention was made of electrical connectors.

]

Question 7.1.6 which was asked by the staff was the most direct question

' in the subject of environmental qualification of safety-related items.

The question is quoted below.

"7.1.6 Identify all safety-related equipment and components  !

. (e.g. , motor, cable, filters, pump seals) located within the primary containment which are required to be operable .;

during and subsequent.to,.any. loss-of-coolant or a steam-

~

line break accident. Des,cribe the qualification tests which have been ur will be performed on each of these items to assure their availability in a combined high temperature, pressure, and humidity environment. Evaluate these qualification tests for the maximum temperature and

, pressure combinations assuming a complete spectrum of-break sizes. Of special concern are those items that must remain operable following incidents not in the major accident category; e.g., small loss-of-coolant accidents which do not cause a complete loss'of the facility and therefore plant restart is desirable."

The response, which is quoted below, does not address the electrical connectors which were described in the FSAR in conjunction with the cabl e'. -

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" Safety-related equipuent inside the primary containment which must operate in an accident environment is limited to isolation valves, recirculation line valves, . auto-relief valves and their associated electrical penetrations and cabling. These cables, .;

operators, and penetrations are designed for in excess of the required period of usage- under their design basis' accident condi-I tions. In addition,; indication of drywell . pressure, and torus water temperature are required and provided for the full range of _.

postulated accident environmental conditions. The containment __

1 1

icolation, and recirculation line valves are required to operate scr only short time durations. This is true for the full spectrum  !

i of break sizes.

" Electrical cables will utilize insulation of a type that has been

~

environmentally tested to demonstrate suitability for this service for periods. of time greatly in excess of the required operating time.

" Specifications for the above equipment included the requirement to remain operable as required during and/or subsequent to a design basis loss-of-cnolant accident. Manufacturer's proposals, manuals, and drawings are carefully reviewed for selection of materials and l conformance requirements. to the specified temperature, pressure, and humidity i

" Qualification tests have not been: required for the temperature 1 sensors because their design range of operation and calibration i

encompasses the' environmental conditions expected within the

' primary containment as the result of loss-of-coolant or steamline break accidents.

"Qualif.ication tests (prototype tests) have been performed on the the Limitorque operators, power and control cabling run to .

engineered safety ' feature equipment inside the primary containment and on electrical peretrations to ensure requirtl op ration and/or leakage integrity under LOCA conditions.

"The Limitorque valve operators and solenoids associated with the relief and main steamline isolation valve have been tested and'the results of these tests have been submitted in Amendment 18 of Millstone Power Plant FSAR.

4 4

s

" Power-and control cabling samples were tested in a standard steam environment significantly above. the peak design basis LOCA conditions. Sat:ifactory operation was verified at peak test conditions of 32/ F,0 85 psig, and 100% RH Subsequent to these_ tests, . dielectric tests were performed to verify that insulation properties were naintained."

"Drywell electrical penetrations are presently being manufactured and as a result the qualification tests have not yet been accom-plished.

The design spec.fications of these components require a '

'that a prototype of each type be tested under design basis LOCA conditions. The ' tests must . satisfactorily demonstrate the penetra-tions electrical and leak tightness characteristics at peak conditions of 3000F, 56 psig and.100%:RH.

..q "For equipment inside- the drywell that is' not safaty related, l purchase specifications include the requirement cemain operable under-normal operating environmental conditions e nonsafety-related component's ability to function for reste'. following a lesser incident has not been' evaluated. Following any loss-of-cccler,t :::ident, equipment and components would be checked for f aictional operability before station res' tart. " -i in the review of the foregoing quoted stat ments one section in the fifth pcf0 graph was focused upon since it was considered to be'a potential faise statement statement is, depending upon the interpretation of the meaning. The

~

'lification tests (prototy'pe tests) have been performed on. . . . . power and cuntrol ca,blino run to' en equipment...tc ensure requireo operation..gineers safety feature

..under u0CA conditions."

The words " cabling run" could be interpreted to mean a specific section of cable run _from point to point and as such should have included the terr mations which would have been in some cases electrical connectors.

IE reached the conclusion that in the context of the entire sentence the intent was that the " cabling run" indicated that c'able used between-safety equipment had been tested and did not mean a " cable run."

No material false staten.ents wert found and it was concluded that further effort slated to the subject of connectors would not be productive of i ns ta: '. 3 ^ requiring enforcement action. 1 s tiaine Yankee (50-309)-

During the review of the application-records there were statements found in the .FSAR whi:h directly addressed _ electrical connectors. Pertinent portions of Section B.4 of the FSAR are. quoted below.

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" Prior 'to installation, the major components of the electricci systems are' inspected and tested according to industry-accepted procedures. Insulation values for cables and motors are measured...

"In general, industry standards and test data are used to insure that electrical equipment will operate properly in the environment in which it may be called.upon to function...

" Environmental' tests will be performed on the reactor containment .i

-electrical connectors to prove their suitability for post-incident U"'

operation." ~!

/

By' the time the' staff had written the safety evaluation, February 25, 1972, '-

the testing was not yet completed. The staff highlighted the fact that the ""-

licensee had not originally intended to complete environmental te;.ing. The pertinent portion of the staff safety evaluation is quoted below.

"3.2.4.5 Seismic and Environmental Testing ..

The-vital cables and connectors within containment have not been subjected to accident (LOCA) environment tests. Fur-ther, no seismic testing for any vital instrumentation or electrical equipment has been performed. In both areas 1

. the applicant proposed to rely on purchase specifications to assure that the equipment, has. been designed to perform properly under accident uor seismic conditions. In our judgment,' the satisfactory performance of this vital equipm^nt should be demonstrated by qualification testing of categories of equipment or "6o e testing may not be practical by analysis.

"The. applicant has agreed to perform this testing prior to operation. The program for' qualification testing a.hd 14bsequently the results from the testing will be submitted

- for evaluation. We conclude that this arrangement

.ceptable."

the licensee submitted a test report entitled, " Environ-

. ion of Vital Instrumentation and Electrical Cables, and '

• ttors." The pertinent sections of that report related

'ectors are quoted below. ~

..c ihe assembly was electrically tested, subjected to the environmental conditions of a Design Batis Accident for'48' hours _and:again electrically checked.

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' I "1.3 The electrical connectors showed no direct electrical or mechanical degradation due to; exposure.

q "4.5 Run the cable assembly for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> at the following conditions:

g P - 55 psig T - 2800F RH 100%

~

"6.1 The receptacle, which is the primary barrier to radiation leakage is unaffected physically or electrically by expo-sure'to the DBA environment. '

1 "6.0 Conclusions ~"

1 "6.2 The cable-plug system will still function electrically for most -

. :a circuits.in a nuc1 car power station. Even the degradation which took place during this test is not to the extent that circuit performance would be degraded except in the most critical of j signal. circuits or unless severe overloading occurred which would accelerate further the deterioration of the primary H insulation."

With the. submittal of this report and the absence in the files of additional.

staff inquiries the licensea had apparently met the design criteria and the FSAR commitment. No material false statements were found and it wcs concluded that further effort related to the subject of connectors would.not be productive of instances requir..ing enforcement action.

l Summarv and Conclusions '

.l

- As a result of' the IE review of the units at these seven sites, the staff '

has reached some general conclusions.

These conclusions are listed below.

1.

No material false statements were found, but.there were two cases which were looked at in great detail since they presented potentici for false statements. These include the statements made on Pilgrim, Unit Ho, 1 which focused on the term " cable run" and the lack of a state-ment on Nine Mile Point, Unit >Ho.1 following a direct staff question.

2.

In generai :Se licensee only addressed the environmental qualifications in general terms and in cases where specific environmental parameters were listed' they applied only to specific pieces of electrical equipment and were silent on hardware such as the electrical connectors.

3.

Where. connections or connectors were addressed'by the licensee the statements.were in terms'of, " environmental conditions will be speci-fied or. design for" or "the equipment is selected to meet the, conditions."

. . _ ,, w-*

• 23 -

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- 4.. One direct question was asked by the staff'of a licensee in order to have the qualification test results for connectors provided to the staff.' :The question was not answered for connectors, indicating, = -

we believe,. that.the tests had not been made since the licensee in the remainder of. the answer listed results of tests on other components.which the staff had also addressed.

5. As a result of the. time span into the past for the issuance of a CP, plant construction and licensing of these units, it should.be noted that there.was not definitive guidance available for the industry within the Commission's General Design Criteria of 10 CFR 50, Appendix A or the Quality Assurance of 10 CFR 50, Appendix .B; regulatory guides or various ..----

. national codes and standards.

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ENCLOSdRE3 . m o

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Letter from W. G. Martin, Region'I, to Connecticut Yankee Atomic Power Company dated May 3, 1978 e

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UNITED STATES NUCLEAR R EGULATORY CO~.. sslON 5 . [ ~ $2 -

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I' R EGION I '

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A is+os Docket No. 50-213 U .? O .1 ic78  :

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, Connecticut Yankee Atomic Power Company -

' ATTN: Mr. D.'C. Switzer - - :

President P. O. Box 270 Hartford, Connecticut . 06101

-Gentlemen: ~ ' E

Subject:

Investigation 50-213/78-10 d This refers to the investigation conducted by Mr. R. Smith of this j

. office on. Apri1~ 10 andl ll,197S, at Berlin, Connecticut, of activities authorized by URC License No. DPR-61, and to the discussion of our findings hel4 by Mr. R. Smith of this office' with Mr. - A. Roby of your staff at the conclusion of the investigation.

^ 1

..reas examined during this investigatien are described in the Office of Inspection and Enforcement:Investigatton Report which is enclosed with  !

this-letter. Within these areas, th'ehivestigation consisted cf selective examinction of representative records, interviews'with personnel, and-o bservations by.the. investigator.

Within the scope of th'is investigation no items of noncompliance were g observed.,

In accatdance'with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy'of this letter'and the enclosed investigation report will.be placed in the NRC's Public Occument Rcom.

If this report contains any information that you (or your contractor) believe to be proprietary, it is necessary that you make a written application within 20 da --

from pubiic. disclosure. ys to this- office to withhold such information Any such application.must be accompanied by an affidavit executed bysthe owner of the 'information, which identifies the document or, part scught to.be withheld, and which contains a statement .

of reasons which addresses with specificity the items which will be considered by the Connission as listed in subparagraph (b) (4) of Section 2.790.: The infcrmation2scught to be withheld shall be incorporated as-  :

fari as. cossible into a"s'ecarate part of the affidavit. If we do not '

near from ycu in this rerard withir, the scecified period, the report will be piace:. in the Puulic Cccument Room.

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Connecticut Yankee Atomic Power 2

. Company

• l' O 1 1978

. .t: . . . . .

No reply to this letter is required; however, should you have any cues- "~

tici.. concerning this investigation, we will be pleased to discuss them wi th you, Sincerely. -

g' . ::

Walter G. Martin, Chief ---

Safeguard,s Branch

Enclosure:

Office of Inspection and Enforcement . .- . .

Investigation Report Number 50-213/78-10 cc w/ enc 1:

R. Graves, Plant Superintendent O. G. Diedrick, Manager of Quality Assurance

.bec w/ encl:

IE Mail & Files (For Appropriate Distribution) '

Central Files .r4 Public Document Room (POR) .

Local Public Document Room (LPOR)

Nuclear Safety Information Center Technical REG:I Reading Room Information Center (TIC)(NSIC)

State of Connecticut .

6 6

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,. nuLLtM Kr.bdLA)URY.LUMMIS$1 0FFICE OF INSPEC7IDH AND EHFORCEMENT Region I Report No. 50-213/78-10 , ~.

Docket No. 50-213 ..

Licen::c Ho, OPR-61 Friority --

Category C Licensee: Connecticut Yankee Atomic Power Company P. O. Box 270 -

Hartford, Connecticut 06101 Facility Name: Haddam Neck Plant, Haddam, Connecticut Inspection at: Berlin, Connecticut ;l Inspection conducted- April 10 and 11,1978 Inspcetors: wm I .u c. [ 7 Ra o H. Smith, Infestigation Speci list ca t( si gned ll date signed

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74 da,te s'gned Approyed fl /  ?~s k '

Jf C DevlTni Chief. Security and /.sete signed -

/ Investigation Section, Safeguards

/ Branch ]

insocction Summary: -

'I Investicatien on Acril 10 and 11, 1978 (Recort No. 50-213/78-10) .l Areas investigated: The possiole misrepresentation regarding the qualification of terminal blocks as stated in the ifcensee's response dated December 8,1977, to IE Bulletins 77-05 and 77 05A. This investigation involved 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> onsite by -

one NRC investigator. 'i '

Results: The investigation found no indication of intent by the licensee to msrepresent a material fact and no ite.ms of noncompliance were identified.

9 i

Region ! cenn 12 (9ev. W il 77) i

f

../..'...

TABLE OF CONTENTS 7:

A. Introduction .

B. Possible Misrepre,;entation .

C. Sumary of Findings D. Details ,

1. Persons Contacted "" '

2.. IE Eu11etins 77-05 and 77-05A '

3. Licensee's Bulletin Response -

4 .Insestigation e.nd Findings

5. Management Meeting

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A. Introduction The_ purpose of this investigation was to gather information con- *?

cerning the possible misrepresentation regarding the environmental

. qualification of terminal; blocks as stated in the licensee's Bulletin ' response. * ~

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B. Possible Misrepresentation

=

The licensee's ~ response dated December 8,1977, 'to IE Sulletins 77-05 and 77-05A contained the following two paragraphs:

Paragraoh No. 3

" Documentation is available indicating that these connectors

- are of high quality, however no documentation is available indicating they are specifically qualified for a LCCA environ-ment. On this basis the connectors.were removed and replacad.

with' qualified terminal blocks enclosed in sealed junction

~

boxes."

.Paracrauh No. 4 ' '

~

"These connectors'were orfeinally installed as a convenience item to facilitate refuelihg"oberations and, in order. to reestaclish this convenience, it'is intended to replace the  !

terminal blocks and junction boxes with cualified cennectors during the next refueling outage."

The NRC questioned the validity of the statement in the third paragraph .i.e.,

".'.. replaced with qualified

• terminal blocks en- .,

closed.in sealed junction boxes.", which was interpreted by the NRC to mean that the terminal blocks were qualified for a LOCA environ-ment.

This statement also appeared to conflict with the statement in the fourth paragraph,.i.e., "...it is intended to replace tne terminal blocks and junction boxes with qualified connectors...."  !

j C. Sumary of Findinos '

The.NRC to misrepresent investigation a material found no indication of intent by the licensee fact.

1 e

t D. Details

1. Persons Contacted The following individuals are with Northeast Utilities Service Company (NUSCO) which provides engineering support- a services to the Connecticut Yankee Atomic Power Company (Conn Yankee). ,

L. Davison, Generation Electrical Engineer ~ l J. Kufel, Superintendent of Nuclear Produccion l A. Roby, Chief of Generation Electrical Engineering

(

2. IE Bulletins 77-05 and 77-05A These Bulletins were entitled " Electrical Connector Assem-blies" and were issued on November 8 and 15,1977, for licensee's and permit holders to determine whether such described elec-trical connector / cable ass'emblies had been properly qualified H for their intended service or whether these types of assemblies were utilized in applications that must continue to operate reliably in a LOCA environment. j

3. Licensee's Bulletin Resoonse

. . A.

The licensee response to IE Bulletins 77-05 and 77-05A dated  !

December 8,'1977 described four unqualified connectors which

, were identified and subsequently removed. The response also identified the replacement for the removed connectors as being "... qualified terminal blocks enclosed in sealeo junction boxes which would be replaced with qualified connectors during the next refueling outage."

Based on a telephone request to the licensee from NRC Region I personnel on January 4,1978, the licensee respondec by telephone (on January 13,1978) that they had been unable to obtain sufficient documentation to support qualification of the terminal blocks, but that tests would be conducted to confirm this assumption. This telephone report was followed up by receipt in the Region I office of a written report concerninn the environmental cualification of the terminal blocks /bexes. This response contained the following statement:

e G

N a 5

3

" Initially, our decision to use terminal blocks instead of connectors was based on using an approved terminal-block, which' in conjunction with a suitable terminal box would provide a combination assuring that the electrical and mechanical characteri'stics of the terminal block would design not degrade basis LOCA."to an unacceptable level during a

~

This response also contained an evaluation of the terminal block / box-combination to indicate performance under postulated LOCA environment. In addition, the response described . plans - _;

the terminal block / box. combination.for qualification testing to d

4. Investication and Findinos The investigator contacted the NUSCO Generat~ ion Electrical- -

Engineering and based on disr.us.sinns, the following informa-tion was developed. t The use of the word " qualified", as used 'in Paragraph 3 of the licensee's. response dated December 8, 1977, was intended to l establish that the terminal blocks in their enclosure would satisfactorily design perform basis LOCA theiV fun ~ction in the containment under conditions. .

Although the terminal blocks were not considered as electrical connectors (of the type addressed by the Bulletins), the con-cept of using terminal blecks in place of connectors was dis-cussed 21, 1977 by NUSCO Generation Enginearing Persennel en November alternative.and was agreed in principle to be an acceptable on November 22.The 1978.terminal block / box combinations were installe engineering discussion. There was no record maintained of the The basis for the judgement of acceptability for the terminal blocks cotained was that they were seismic Category 1 devices The investigator examined purchase order number .

12179-04998 No. E350, Category I, Hon-Encineered.and noted that the te The licensee representatives stated that the .

described in their response was obtainec from the purchase order.

m-_

8 e Ri

. 6 )

l i

On November 22, 1977, the connectors were replaced with terminal block / box combinations. Two terminal blocks were installed in '

existing aluminum boxes which had contained the removed con-nectors and two terminal bloc'ss were installed in steel boxes obtained from the Millstone site. -

1 The planned testing described in the licensee's response dated January 13, 1978, was initiated on March 16, 1978. The test duration was five and one half days. The NRC was informed of the test results on March 21, 1978, which showed failure of the terminal' blocks in aluminum boxes.

steel boxes performed satisfactorily. The terminal blocks in The Haddam Neck Plant was shutdown on March 21, 1978, f'or replacement of the aluminum boxes with steel boxes.I The licensee representative also, stated that although there was no immediate certainty that environmental qualification had been proven by test, the generic use of terminal blocks' in nuclear in junction stations boxesled them would beto conclude fully that qualified for the theirterminal intended blocks use, where they could be exposed to a LOCA environment.

The licensee representative further stated that confirming this judgement, full serie's environmental tests (a ation, temperature, pressure, and chemical spray) recently ginc, radi-completed on a terminal block / box combination as installad at Haddam Neck have confirmed the suitability of this config-uration conditions. to satisfactorily perform its function during LOCA The licensee representative stated that it is still their intent to replace the terminal block / box cambinations with more able. convenient qualified connectors when they become avail-ations. This change will be made to facilitate refueling oper-i Office of Inspection and Enforcement Inspection Report Number 50-213/78-09

;  ;+ -

. _ _ _ _ _ _ -.___ __ ___m

1

. 7  ;

.l Bared on the above developed information, the NRC investi-gation found no indication of intent by the licensee to mis-represent a material fact.

5. Manaoement Meeting- l

. On April 11, 1978, the investigator informed Mr. A. Rcby, Chief of Generation Electrical Engineering of the investiga-tion findings and that no items of noncompliance had been identified.

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ENCLOSURE.. 4 ..

Investigation Report No. 50-293/78-13 dated May 30, 1978

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,.f *'.* % , UNITED STATES I"C """ '

.,.,,,' 'h NUC LE A A R EG ULlJO R Y COMMISSION

. ggi 2 Y REClON I g {UQ/' Q._9 j s31 PARK AVENUE MING OF PRUS$l A, PENNSYLVANI A '9404

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Gr =r j

• ..e JUN 0 71976 Docket No. 50-293 . _..

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'I Boston Edison Company M/C Nuclear ATTN: Mr. G. Carl Andognini, Manager Nuclear Operations Department 800 Boylston Street Boston, Massachusetts 02199 ..

Gentlemen:

Subject:

Investigation 50-293/78-13 1 This refers to the investigation conducted by Messrs. S. Ebneter and J. l Devlin of this office during the period April 26 through May 18, 1978, ,.

WR at your Boston Office and at your Pilgrim Nuclear Power Station, Unit 1, $h Plymouth, Massachusetts, of. activities authorized by NRC License No.

DPR-35.

This also refers to the discussion of the preliminary findings S$

held by Messrs.. Ebneter and Devlin with members of your staff at the conclusion of the investigation at both locations and to a subsequent  :

te.lephone cn discussion between Mr. Ebneter and Mr. Ziemanski of your staff May 12, 1978. j g j, t

Areas examined during this investigation are' described in the Office of j Inspection and Enforcement Investigation Report which is enclosed with this letter. Within these areas, the investigation consisted of selec- i i

tive examinations of procedures and representative records, interviews with personnel, and observations by the investigator.

Within observed.

the scope of this investigation, no items of noncompliance were

] i In~accordance with Section 2 790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the i

enclosed investigation report will be placed in the NRC's Public Document Room. {

If this report contains any information that you (or your contractor) believe to be proprietary, it is necessary that you make a written i application within 20 days to this office to withhold such information from public disclosure.

Any such application must be acccmpanied by an affidavit executed by the. owner of the infonnation, which identifies the i i

document or part sought to be withheld, and which centains a statemen-  !

cf reascns w,hich addresses with specificity the items which will be i

I

.  :=

%::tc,n Edison. Company 2

'iC , .Nuci ear JUN 0 71978 considered by the Commission as--listed in subparagraph (b) (4) of Section 2.790. The information sought to be withheld shall be incorporated as .

far as possible into a separate part of the affidavit. If we do not hear from you in this regard within the specified period, the report will be placed -in the Public Document Room.

No reply to this letter is required; however, should you have.any questions concerning this investigation, we will be pleased to discuss them with you. H Sincerely, l

GW CLY, Boyce H. Grier Dire,ctor

Enclosure:

Office of Inspection and Enforcement Investigation Report Number 50-203/78-13 cc w/ encl:

• P. J. McGuire, Pilgrim Station Manager A. Z. Roisman, Natural Resources Defense Council (Without Report) ~"

. . e.

e 3

GFFICEOFINSPECT10tiANDkb?OEi$kNT Region I Report No. 50-293/78-13 Docket No. 50-293 '

License No. OPR-35 Priority --

Category C y

Licensee: Boston Edison Comoany M/C Nuclear R

'l 800 Boylston Street -l

.I Boston, Massachusetts 01299 Facility Name: ,Pilcrim Nuclear Power 5tation, Unit 1

. Investigation at: Boston, Massachusetts; Plymouth, Massachusetts; Bethesda, Maryland .q q

Investigation conducted: April '26-28; May 4, 8-11,16-18,1978 -

0 4

Investigators: f4' /d

1. D. ione;er, Cnlar, engineering M/h4 cate siened Support e'etion o. 2, RC&ES Branch ~

C ll! L & ~

i W. Devlin, cnier, decurity e investigation f 7f j Section, Safeguards Branch cate signed .1 1

i cate signec

. .~ i. 1 cate signec

, -Approved by: -

W. G. Martin, Chief, Safeguards

1. Mhd date signed Branch .

Investication Sumary:

]

Investication Cenducted on Acril 26-28, May 8-11, and 16-18,1978 (Recort No.

50-293/78-13)

Areas In'resticated: Areas investigated included a review of th.e licensee's response and actions with respect to verifying the qualification test data for electrical connectors which are' required to operate in safety systems. Also reviewed was the timaliness and accuracy in obtaining and reviewing the Pilgrim, Unit 1 response documentation to IE Bulletins _77-05 and 77-05A by the NRC staff.

?.es ul ts : Within the scope of the investigation, no items of noncompliance were 1centified. The delay in the receipt and evaluation of the response data could not ce attributed to the licensee.-

9 7 ien I =c m 143

Fev. Octooer 1977) .

t'

,1 ,'. .

TABLE OF CONTENTS I. BACKGROUND

=

A. Reason for Investigation E. Identification of Involved Organizations

..=

C.

IE Bulletins 77-05 and 77-05A s

D. SECo Response II.

SUMMARY

OF FINDINGS A. Investigation Firidings '

B. Conclusion .

III. DETAILS Introduction B. Scope of Investigation C.

Personnel InvestigationDirectly Interviewedand/or Contacted During the D.

Chronology of BECos Response to Bulletins 77-05 and 77-05A E. .-

Pilgrim Connectors Information Gathering by NRR/IE F. Screening and Diagnostic Tests of Bendix Connectors G. .

Requirements and Commitments of the Licensee Related to Qualification of Connectors H. Quality Assurance - Pilgrim Construction I. Source 6f Qualification Records IV. MANAGEMENT EXIT INTERVIEW

l .

~

3 o

I. BACKGROUND A. Reason for Investication

.. +. ;

Sy memorandum dated April 20, 1978, from H. 'D. ' Thornburg, Director, Division of Reactor Operations Inspection (ROI),

D Inspection and Enforcement (IE), to Boyce H. -Grier, Director, s+

Region I (RI), it was directed 'that an investigation be conducted to review the response from the Boston Edison Company to IE Bulletins 77-05 and 77-05A. The' purpose of the' investigation was to examine primarily.the Boston Edison Company's actions and response with -appropriate inquiry to determine conformance .. _

to applicable quality assurance documentation requirements as -

well as the accuracy and timeliness of the information provided, m The investigation also reviewed how electrical equipment, not -

fully qualified, came to be installed in the Pilgrim Nuclear Power Station, Unit I. ,

.l R

The request for'this investigation ls based upon the directions of thethe Commission matter'of

" Petition:as forstated in the-Memorandum Emergency 'and Remedial Action."0 and Order, in )

An investigation to review and evaluate BEco's actions in 'this matter'was initiated by the NRC"*' 8egional Office on April 26, 1978.

B. Identification of Involved Orcanizations

1. BostonEdisonCompany-(BECo) 800 Boylston Street Boston, Massachusetts 02199 A utility company licensed by the NRC to operate Pilgrim Nuclear Power' Reactor, Plymouth, Massachusetts.

2 .- Bendix Corporation .

Electrical Components . Division Sidney..New: York 13938' A manufacturer and supplier of electrical connectors installed on safety' rela:ed equipment and required to perform in the LOCA environment. o

+

l ID Federal Register 90lume 43,'No. 76, Wednesday, April-19, i

_1973 (15561-16575) l l

I T -+ e- .i.- e-e--* y- - r--- ,- - _ _ -

• _.- - -__-__m__--u--___

l 4

l C. IE Bulletins 77-05 and 77-05A On November 8, 1977, IE Bulletin No. 77-05 on. Electrical Connector- Assemblies was issued by the Regional Office of the NRC Office of Inspection and Enforcement to all power reactor -

facilities with.an operating license or a construction permit.

The Bulletin required licensees to: =s

"1. Determine if in their facility electrical connector assemblies of the type tested by_ Sandia Laboratories or any other similar type are utilized or if they plan to utilize such electrical connector assemblies in systems' that are located inside conthinment that -

are subject to a LOCA (loss of coolant accident) environment and are required to be operable during a -

LOCA."

"2. If any such applications are identified, review the adequacy of qualification testing for the assemblies and submit the documentation for NRC review."

"3. If evidence is not available to support a c*onclusion l

' of adequacy, submit'your plans and programs toward

. qualifying existing equipment or your plans for

. replacing unqualified -a,ssemblies with qualified ec.uipmint." ,

The Bulletin required licensees of operating plants to provide a written response within thirty (30)-days from date of issue of IE Bulletin 77-05 and holders of construction permits to provide a similar response within sixty (,60) days. The written response to this Bulletin from plants'having operating licenses was due on'0ecember 8, 1977. The' licensees on November 23, 1977, had been requested to provide prompt notification to the i l

NRC of any problem areas discovered prior to that time. On this same date, Pilgrim was instructed by telephone.from Region I, that the responses to IE Bulletins 77-05 and 77-05A should.be telephoned to the region as soon as possible but no later than December 8,1977.

J On November 15, 1977, IE Bulletin 77-05A on Electrical Connector Assemblies was' issued as a supplement to IE Bulletin 77-05 with the instructions.that the two documents ~ should be censidered '

i.

1

_ . - - . _ . _ . . . _ - . - .~ - _ . _ . - , _ _ . . . . , _ , -

p ,

5 together and. the response tv Bulletin 77-05 should reflect the expanded scope stipulated in Bulletin 77-05A as follows:

" Electrical connectors should be qualified to perform their. intended. function af ter having been subjected to accident conditions if they are contained in a system whose function is to mitigate that-accident."

"The location of the connectors that must be qualified is '

not. limited to those inside containment."

" Action To Be Taken'By Licensees and permit Holders:

1. - Actions requested by Bulletin 77-05 should be expanded to include all connectors in safety systems which .

are required to function to mitigate an accident ~q where the accident itself could- adversely affect the ability of the system'to perform its safety function.

The examination is not to 'be limited to only LOCA's nor to. areas only within containment.

2. Responses should be provided within 30 or 60 days, as appropriate, of the date of this supplement." .

The response date to this supplemental Bulletin for plants -

with operating licenses was December.15, 1977.

D. BEco Rescense At the request of IE:HQ, Region I contacted SEco by telephone, on November 23, 1977, and requested they verbally respond to the Bulletins by December 8,1977. On December 8, BEco contacted Region I and provided an oral response to the subject Bulletins.

The licensee stated.that Bendix connectors were utilized at Pilgrim I, and that, based upon telephone comunications with Bendix, they met the requirements .of Military Specifications

MIL-C-5015D and MIL-C-26482. The licensee additionally stated that certification correspondence from ~the Sendix Corporation had not been received by SECo, that the written response would- .

be ~ delayed until this correspondence was received, and that the response'would be transmitted on December 12, 1977. A L copy of- the , regional telephone- notes was transmitted by facsimiie to IE:HQ on December 9, 1977. '

I t 9 --. , , s_.-- . . , _ , y + ,..,---,.% ,n. ,ev.--- w.-.+-.-, - - - , -m-w e -,-

e

, ,- 6 On December 13, 1977, K. Seyfrit, . Assistant Director, Technical Programs, Inspection and Enforcement, contacted E. Ziemianski, BECo, Licensing Engineer, and informed him that the response to Bulletins 77-05 and 77-05A was overdue. Although Mr.

Ziemianski j stating the position'of the licenses, contended the responses were not due until 30 days after the Supplement (77-05A) Mr. Ziemianski agreed to telecopy the response letter to NRC by noon of that date. It was further agreed that the '

supporting documentation would not be telecopied but would be mailed on the same date. ,

Region I received the response on December 19, 1977, including all documentation, and forwarded the correspondence to IE:HQ on' December 20, 1977. -

The investigators were unable to establish when the response was received at IE Headquarters mail unit. On December 27, 1977, the response from Boston Edison Company (Pilgrim, Unit I) was received by the organization responsible for evaluating the response in IE Headquarters. It also could not be established whether the response received was the one submitted by BECo or the package forwarded by Region I. .

Following a review by Headquarters staff of the qualification information provided by the Boston Edison Company for the Pilgrim Nuclear Power Station, Urtit.1, it was determined,by IE:HQj to be incomplete to confirin the adecuacy of environ-mental qualification af certain electHeal connectors in safety systems located inside containment and required to be operable in the event of a LOCA. The NRC staff met with the licensee on December 29, 1977, to discuss the safety aspects of the electrical connectors that are in use in the Pilgrim Unit i facility.

11.

SUMMARY

OF FINDINGS '

A.

Investigation Findings A review of the licensee's response to IE Bulletins 77-05 and 77-05A, and their actions with respect to verifying the qualifi-cation test data, for electrical connectors which are required to operate in safety systems indicates the sequence cf events was aggressively pursued, responsive, and the results reported accurately. The delay in obtaining and reviewing the Pilgrim documentation by the NRC staff could not be attributed to any

a _;

,' .' \

7 l

actions on the part of the licensee. As reported in the January 6, January 13, and February 3,1978 memoranda oy the staff to the Comission, the actions taken by the staff. to assure that the licensee would achieve prompt qualification of unqualified connectors or replacement with a qualified alternative S are adequately demonstrated. I B. Conclusions f

Within the scope of the investigation, no items of noncom- 1 pliance were identified. q u

111. DETAILS '

h A. Introduction :i i

This investigation was initiated at the direction of the 1

?

Comission as requested by memo', andum dated April 20,197o, 9 IE:ROI to IE:RI to ascertain the cause and extent of possible i inaccuracies and untimeliness of a response to IE Bulletins j 77-05 and 77-05A (Electrical Connector Assemblies) by Soston '

Edison Company (Pilgrim Nuclear Power Station, Unit 1).

Based upon the sequence of events as reported to the Comis- ...l sion in memoranda dated January;6410,- 13,1978 and February l

3,1978, and IE and NRR memoranda dated November 11 and 18, l 1977, December 13, 1977, and January 5 and 6, 1978, it was concluded by the Commission that the review of the electrical ^

connectors in use at the Pilgrim Nuclear Power Station was unsatis factory.

B. Scoce' of Investication This investigation included a review of actions and docu-mentation associated with the preparation of the SECo's (Pilgrim, Unit 1) response to IE Bulletins 77-05 and 77-05A and inter-

' views with licensee personnel involved in the response review and documentation. It also included a review of NRC documenta-tion on this subject and interviews with staff personnel involved in the review and evaluation of the qualification of electrical connectors used in plant safety systems, at the 4

Pilgrim Nuclear Power Station, Unit 1.

6 ,

C. Personnel Directly ' Interviewed and/or Contacted Durino the Investication

1. Boston Edison Comoany m G. C. Andognini, Nuclear Operations Manager W. M. Sides, QA Manager ~~

• A. -Morisi, Group Leader, Power and Control Systems

'*S. J: Little, Licensing Engineer .

E. Ziemianski, Licensing Engineer ==-

• M. X. Yessaillion, -QA Lead Engineer J. Sullivan, QA Engineer '

• R. Kennedy, QA Specialist -

~

• R. Machon, Plant Support Group Leader

• D. Stoodley, BEco Legal . Counsel '

• S. Stowe, BECo Legal Counsel "D. Horan, BECo Legal Counsel -

2. Nuclear Reculatory Commissio, ~

K. V. Seyfrit, Assistant Director, Reactor Operations Inspection, Inspection and Enforcement _.

l T. Ippolito, Chief, Instrumencation and. Control Systems  !

Branch, DSS I V. D. Thomas, Senior Instruqentation and Controls Engineer '

O. Mcdonald, Systematic Evaluatibn Program

• • P. O'Connor, Licensing Project Manager (, Pilgrim)

R. Haynes, Chief, Engineering Support Section No. 1, Reactor Construction & Engineering Support' Branch

.R. Architzel, Reactor Inspector, Reactor Operations

• & Nuclear Support Branch D. Caphton, Chief, Nuclear Support Sect' ion, RI B. Faulkenberry, Chief, Career Management Branch W. Rutherford, Senior Mechanical Engineers RCI W. Mullinix, Chief, IE Mail File Unit

• Representatives at corporate exit meeting.

• • Presently assigned to Systematic Evaluation Program D. Chronolocy of BECo's Resoonse to Bulletins 77-05 and 77-OSA The following chronology of events is the actions taken by the licensee folicwing the receipt of tne IE'5clietins. This-l

9 sequential history of activity was obtained from SEco person- e nel; S. J. Little, Licensing Engineer.and A. Morisi, Group Leader, Electrical and Control Systems. A significant number of these activities are supported by telephone logs and/or correspondence.

=..

History of Pilgrim Nuclear Power Station, Unit 1, electrical 3' H connector review, test and modification:

[ .

1: Notification of Recuest for Data by NRC 0 l

a. On November 8,1977, NRC issued IE Sulletin 77-05 requesting data on use of electrical connectors at .

' Pilgrim Station in containment that are subject to a d LOCA environment. ~!

1

b. On November 14, 1977, BEco received IE Sulletin 77- :1 05 in Boston Corporate Offices. I

c. On November 15, 1977, NRC issued IE Sulletin 77-05A (supplement to IE 77-05) requesting electrical connector study be expanded to include all electri-cal connectors in safety systems which are required to function to mitigate an accident where the accident ==

itself could adverse]y affect the ability of the system to perfonn its's'a fety function. The examina-tion of the use of electrical connectors should not l be 1.imited to only LOCA nor to areas only within containment,

d. On November 17, 1977, BECo received supplemental IE q Bulletin 77-05A (which expanded the scope of the review) .

2. Utility Action with Resoect to Recuest for Data

a. A review of QA documents germane to the' subject was instituted by Nuclear Operations Department -(NOD) and Nuclear Engineering Department (NED) (November 14,1977).

b.- In conjunction with the records search, the follow-ing concurrent' actions took place starting on November 14, 1977. i

.- 1 l

l l l'

10 (1) A review of -

y drawings was instituted for types .onnectors used at Pilgrim Station.

(2) A search er spare electrical connectors was implemen d to obtain information on types of connectc 5 used at Pilgrim Nuclear Power Station.

(3) Electrical penetrations outside containment-were inspected for the use of electrical connec-tors. On November 21, 1977, inspections revealed the use of Bendix Electrical Connectors on cables outside containment.

(4) A request.was processed for deinerting the drywell to inspect electrical penetrations inside containment. On December 1, 1977, after' deinerting the drywell, a physical inspection verified the use of Bendix Connectors on elec-trical cable inside containment.

(5) The Architectural Engineer (Bechtel) and the supplier of the nuclear steam supply (General Electric) were contacted and requested to review home office records for types of connectors used and for connector qualification test results or to ascertain if any other types were used. This occurred either in late November or early December 1977,'but continual contact was maintained with the General Electric Company, Wellesley, Massachusetts.

. c. On November 21, 22, and 30,1977, several telephone conversations were held with the NRC concerning the Bulletins. During this period, General Electric, Bechtel, and Bendix were notified to search their files for any information in addition to that in BECo's 2.a vault to respond to NRC. Raychem was contacted by BECo to supply qualified splice kits.

NOTE: The telephone contact on November 21 from NRC te BECo discussed shutdown of D. C.

Cook.

Telecon on November 22, NRR invited BECo to

, meeting to be held at NRC:HO :n ACS valves.

11 Telecon on November 30, NRR to BECo, discussed -"

Sandia. Pilgrim connector findings were not discussed.

'd. BEco QA/QC records and plant drawing: re-coafi rmed ,

the use of Bendix connectors on seven (7) electrical penetrations (three power and four control).

L

e. A review of QA/QC records produced certificates of 2 conformance (the ability of the connectors to remain 5 operable during and after a LOCA). ~

1

f. BEco received letter from Bendix on December 8, 2 1977. The letter restating that connectors would remain operable under LOCA conditions ~(if potted and '

mated properly). ]

3. Resoonse to NRC Recuest for Data

a. At the request of NRC, Region I, BECo telephonically responded on December 8, 1977, to IE Bulletins.

NRC, Region I was informed that Bendix connectors are in use in safety systems. The systems in which they are used: ADS solenoids; Isolation Valves for HPCI, RCIC, RWCU; Recirculation pump suction; and, discharge valves. ,

LNOTE: The telephone survey record at Region I also indicates, in addition to the above, that it was reported that the connectors were environ-mentally qualified and meet the following specs: MIL-C-50150 and MIL-C-26482)

Bulletin 77.-05A; not applicable, no connectors outside containment.

b. On December 13, 1977, BECo responded in writing to NRC Bulletins 77-05 and 77-05A based on Bendix assurances of satisfactory operation of connectors "if mated and potted properly." BECo tranrmitted, by telecopy, the cover letter without attachment by agreement with ROI.

' Following the teleccoying cf the cover letter, the encire package containing cer:ificata Of conformance

1

-l

.* e'

' and manufacturer's letter.was ~placed in the' mail for-

NP.C Headquarters and_ Region.I.

c. On December,19,1977, letter and attachments were received by NRC, Region I..

d. On December 27, 1977,. IE Headquarters received BECo letter and' notifies BECo.that a meeting in Bethesda, ~~

Maryland was necessary to discuss the situation in

. greater detail.

4. Actions Taken by Utility to Obtain Qualification Test Data- -

+:0

a. An order was placed with Bendix to supply connectors }

' (same as inservice) and was told *by manufacturer, that the shipment would require'18-22 weeks (no longer. a shelf item).

b. BECo requested Bendix to provide them with a random selection of duplicate inplant connectors and to expedite shipment (r,ecember. 15, 1977 - approximate time). Materials were expedited. Connectors fabri-cated and received. by BEco on January 6,1978..

c. In coniunction with;o'rdering spare cennectors for testin., BECo decided to' evaluate the removal of in-service connectors for qualification testing.

d. Meeting in Bethesda, Maryland, with NRC en December ,

29, 1977, to discuss BECo response to NRC Bulletins '

77-05; 05A. Agreement reached with NRC to perform

, qualification testing on existing connectors.

However, continued operation up to the planned outage in January was acceptable with the NRC.

4 e.

4 Letter to NRC dated December 30, 1978, from BECo committing to agreements reached at December 29, 1977 meeting. Continued contacts with GE, Bechtel, and Bendix. plant review was initiated to locate a connector assembly which could be removed for type testing. Raychem Corporation was contacted as a

. potential supplier of qualified splice kits. Bendix searched the country fer qualified reolacement connectors and f:r connectors of the type used at pilgrim for test ourposes.

4 I d

- - y v , M m-re

13

f. On January 7,1978, BECo removed cable / connector assembly from Pilgrim Nuclear Power Station Unit No.

1/nonsafety circuit.

g. On Janu' y 8,1978, connector / cable assembly was ,

trans rurted by BECo vehicle to Bendix Corocration at Sidney, New York.for high altitude water immersi'on test.

==

h. On January 9,1978, Bendix notified BECo of failure "

of cable / connector assembly.in high altitude immer-

, sian test.

~

i. On January 10, 1978, Pilcrim Station, Unit I was removed from service at 0215 AM. -

5. Utility Contincency Plans

a. On December 9, 1978, BECo instituted an engineering study to provide contingency courses-of-action for replacement of or modification to electrical connec-tors. - *

b. Engineering design modifications were evaluated on December 9 - 12/14, .J977, alternatives consisted of:

(1) Replacement of connec'tcrs with new qualified cable / connector assemblies. -

(2) Replace connectors with qualified splices.

(3) A combination of (1) and (2).

6. Contincency Modifications

a. .The AE was requested to evaluate the contingency

' fixes (January 11,1978).

b. The connector manufacturer was requested to evaluate contingency fixes - January 10, 1978.

c. Wyle Laboratories were contracted to perform tests on contingency fixes.(as required) and in addition

' to perform qualification tests on Sendix connectors.

(December 13, 1972).

i 14 d.. Portion of.new connectors received by SEco on January 6,Ji978, were returned to connector manufacturer to perform and evaluate contingency fixes.

7. Engineering Effort by BECn

a. Additional electrical . engineers were contracted and commenced = preparing design package for modifications. _

' Initiated December 9,1977 - Terminated February 4,  ;~

1978.

. ..a

b. NRC accepted BEco. proposed fix on January 16, 1978, with stipulation that materials be environmentally tested for compatability. -

c. Materials tested at Wyle Labs - January 30-31, 1978. -

Complied with test requirements,

d. .NRC briefed and appraised.cf test and test results February 1,1978.

E. Pilgrim Connectors Informa' on Gathering by NRR/IE

'l. Summary In light of the Commission ls i.nvolvement in the Union of Concerned Scientists (,UCS)- petit, ion and rec.uest for information in a limited time frame, several NRC offices became involved in attempting to acquire the information related to qualification of the electrical connectors at

, operating The Office of' Nuclear Reactor Regula-

. tion (NRR)surveyed plants. the NRR staff involved in safety evaluations of older plants and made telephone surveys of architect / engineers (A/E) and nuclear steam systems suppliers (NSSS). ' Concurrent with these surveys, the Office of' Inspection and. Enforcement (IE) issued Bulletin

' 77-05 on November 8, 1977 and-Supplement 77-05A dated November 15, 1977 requesting the licensees to provide written response.about the qualification of connectors installed at their facilities. Within IE,Bulletin 77-05 and. Supplement 77-05A were issued by the Division of Reactor Construction Inspection (RCI) with instructions for .the licensees to respond to the Regions and RCI.

Although IE:RCI issued the Bulletins, and the responses

' were to- be~submittedjtc RCI, ncrmal practice is that IE Division of Reactor Operations Inspection (R0!) has the responsibility for the evaluation of coerating plant 1

licensee responses' .

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13 l

2. . Discussion 1

Initial activity by NRR in' early November in ~9sponse to the UCS petition took the form-of a survey at .ne NRR 1 staff members in the Instrumentation and Control Systems =

, Branch. Staff members could not recall that any connec-tors-of the type.specified in the Sandia report and -

Bulletin 77-05 had been used in nuclear plant safety systems. This was based on their recall of interfaces -:

and~ reviews they had performed on the older plants. The result of this survey was the basis of the initial NRR y

. staff report to the Commission dated November 9, 1977 -;

that states "On the basis of presently available informa- -

tion, we have fuund that electrical connectors of the type -that failed in the Sandia tests. are not being used . .j in safety systems...." .

On November 9, 1977, NRR initiated a telechone survey of y A/Es and NSSSs to supplement ~the staff survey. With I regard to' Pilgrim the A/E was Bechtel Fewer Corporation ^

and' the NSSS was General. Electric Company. NRR personnel stated to the investigators during the interviews that the A/Es and NSSSs were instructed to perform a " quick and dirty" survey of their staffr to determine if any connectors of the type inNquestion were installed in the facilities / systems they had b,een responsible for. They were not to conduct a detailed drawing review; the response was required quickly. Initial responses were negative (connectors in nuclear instrumentation and rod position indicationexcepted). The licensee had not provided any i

information about' Pilgrim at this time.

On November 18, 1977, NRR submitted a staff report to the Commission which summari:ed -the results of the survey.

Pilgrim was not identified as a plant containing the

, connectors based on the above surveys. In interviews 4

t with the licensee, BECo personnel stated that Bechtel had prorided the NRC with information that connectors of the type in question were used at Pilgrim on neutron monitor-ing systems. However, NRR had previously specified that connectors used in nuclear instrumentation and rod position ind,ication were not required to operata during a LOCA.

a g -- -

er , . . ,

i

g The NSSS for Pilgrim, General Electric (GE), notified he  !

NRC on November- 22, 1977, that the Target Rock valves 1 used on the Automatic Depressurization. System (ADS) at Pilgrim utilized connectors of the type described in 1 Bulletins 77-05 and 77-05A. This was reported in a general ,;

sense in the staff report to the Commission dated November '

22,.1977, as being applicable to BWR but Pilgrim was ,not specifically identified. The licensee had not reported the use of. any connectors up to this point. All informa-tion had been obtained from other sources.

On November 22' the NRR licensing project manager invited U the licensee, by telephone, to attend a meeting with NRC :s~

and GE in Bethesda, Maryland to be held on November 23, to discuss Target Rock values. Also on November 22, 1977, the staff submitted a report to th'e Commission '

which defined the connectors used on the Target Rock valves as not being 3 quired to function during a LOCA and, therefore, of no concern to the issue being discussed. I

'The staff submitted to the Commission a report dated December 6,1977, whic.h reported further results.of the preliminary survey on the use of electrical connectors in safety systems. This report stated that the preliminary survey was comolete and thst4 responses to IE Bulletins 77-05 and 77-05A concerning electrical connectors were due to be received by December 8, 1977.

The due date of December 8, 1977, was based on the licensee reporting in writing within 30 days of the date of issue

. of Bulletin 77-05. However, the official due date for written response had been extended to December 15 by the wording of the Supplement 77-05A. IE Region I had previ-ously notified the licensees on November 23 that a tele-phonic response was r. eded on December 8,1977.

BECo notified Region I by telecon on December 8,1977, that connectors of the type described in Bulletin 77-05 and 77-05A were installed inside containment at Pilgrim.

The licensee had a record of the telecon and this was substantiated by telephone logs in Regien I'. T a licensee identified those systems c:ntaining the'conne . ors as ADS, RWCU, HPCI, RHR, Recirculation A&E Suctira and

. discharge valves. TheLlicensee stated nat tne connectors I 4

, y . .- - -- ,,,-4, - , -y , , , - - - ,,,vr, , - . , , , , - - - -I

,1 ," 17' meet applicable parts of MIL-C-50150 and MIL-C-25482 and that the written response would be transmitted on December 12, 1977. This was the first response directly from BECo to Bulletin 77-05 and Supplement 77-05A and the informa-tion supplied was based on drawing reviews, plant inspec-tions, and retrieval of data from the QA files.

4 This information was contradictory only to information supplied by other sources. '

=

The infoimation supplied by BECo telephonically on December. c 8, 1977 was-tabulated by RI and transmitted to IE:RCI by facsimile on December 9,1977.

On-December 13, 1977, IE:ROI-contacted BECo by telephone

-and requested the response to Bulletin 77-05. BECo was ~

instructed to supply the response by telecopy prisr to '

12:00 noon on December 13 and to include in the response a discussion of why the. response was late. SECo perso u nel stated that at the time .they did not think the response was late. BEco discussed the response with IE:ROI and related that it consisted of a cover letter and nine supporting attachments. In an interview by telephone J with BECo licensing engineer, it was stated emphatically that he asked the.IE:ROI representative if he should telecopy the cover letter.and all the attachments. The response was "No the cover'l'etter would be sufficient."

The response had been prepared and was ready for trans-mittal. 'BECo_ revised the letter to include the paragraph presenting their position that the response was not late.

The cover-letter was then telecopied to IE:ROI on December

. ' 13, 1977, without the attachments. The licensee stated that the original letter with all attachments was then placed in.the company mail system addressed to Region I.

A copy of the original and all attachments addressed to IE:RCI was also placed in the company mail system on December 13, 1977. This complied with the instruction in =

Bulletin 77-05 and 77-05A.

IE':ROI received the telecopy of the cover letter on December 13, 1977 and used the information in the response to provide an input to NRR; this was subsequently used in preparing 15,.1977.

the staff report to the Commission dated Decemoer Table B-1 of the December 15 staff recort icenti-fied pilgrim' connectors as being cualified by test-awaiting formal documentatien.

The formal documenta-icn was the attachments te the cover letter.

-- _- , , - . ,_,_.--m,.-7.,..,_o,-,,_ .,e3

P

1. s 18 The licensee could substantiate the telecopy transmitted by telephone charge records. However, no records are kept of outgoing mail that could verify the exact day the two letters were picked up from BECo. This pick up could  :

have been on a day subsequent to December 13 even if -

placed in the internal mail system on December 13, 1977.

NRC, Region I (RI) received the letter and attachments on December 19, 1977. Mail received in RI is date stamped upon receipt. A Xero:: copy of the letter and the attach-ments was mailed to IE:HQ from RI on December 20, 1977.

The mailed response from the licer'see was addressed to IE:RCI as specified ia the Bulletins. Mcwever, it was IE:ROI that was awaiting the responses frcm operating reactors (including Pilgrim).

The organization responsible for the evaluation' received '

a copy of the mailed response on December 27, 1977.

The l

copy received could not be positively identified as being frem the licensee or RI. There is no clearcut explana-tion for the length of time involved in receiving the mailed response at ROI and the time of receipt of corre-  !

spondence could not be estahl.ished since no record of '

receipt is maintair.ed. ""

F. Screeninc and Diacnostic Tests of the Bendix Connectors BECo had been informed by NRC:RI by telecon on December 27, 1977, that the information submitted to support the qualifi-cation of the Bendix connectors installed at Pilgrim was inadequate. BECo met with' representatives of NRC in Bethesda, Maryland on December 29, 1977 to discuss acceptable options related to the connectors. In a letter dated December 30, 1977, BECo committed to further evaluation of the connectors to resolve NRC concerns. BECo had defined several options which could be pursued including (a) replacement of the existing connectors with a type fully qualified; (b) verification that the existing connector configuration could be qualified; and (c) replacement of the connectors with a butt splice covered

. . l 39 l

l with an insulating sleeve. To detemine which option was the best to pursue SECo had to perform certain preliminary functions.

The existing connector configuration consisted of one half of d the connector that had been installed by the GE facility under controlled conditions and potted with a silicone material; the other half of the connector had been installed in the field under less than ideal conditions and potted with a white material suspected of being a GE silicone material. No actual test data was available to verify adequacy of the connector ..

o assembly and positive identification of the potting material was not available. The connector manufacturer, Bendix, had l L

stated their position in a . letter to BECo dated December 8, 1977, that the connectors, when mated and potted properly were -

capable of meeting requirements. SECo agreed to supply a '

typical connector assembly to Bendix for examination of the potting and mating characteristi.cs. These were not qualifi-cation tests. -

The licensee reviewed the location and function of all cennec-tors and selected a typical Bendix connector located outside containment in a non-safety circuit as the candidate for examination. The connector as'sembly was removed on January 7, 1978, and submitted to Bendix for diagnostic screening tests en January 8,1978; the licensee As"very specific in inter-views that -he tests were not intended .as cualification tests'.

The report frca NRC Office of Nuclear Reactor Regulation (NRR) to the Ccmmission dated January 13, 1978, also identifies these tests as " preliminary environmental screening anc diag-nostic examination." The intent of the tests was to aid the licensee in selecting the most viable approach to obtain fully qualified connector assemblies at pilgrim. Based on the Bendix evaluation, the licensee intended to select an approach and then fully qualify the selected configuration in accord-ance with IEEE-323.

Bendix performed a visual examination of the assembly and determined that the connector mating was satisfactory. The insulation resistance from pin-to-pin and pin-to-shell was measured prior to and after perfoming a high altitude imersion test (HAIT) per MIL-STD-13 M , Method 1004.1. Initial insulation resistance values were acceptable but some of the post-HAIT values were considerably lower and unacceptable. Sendix x-rayed the assembly and performed a tear dcwn analysis to

~

-l 8 4 20 I

determine the cause of low insulation resistance. They concluded that there was " lack of sufficient potting compound

~ .l at the rear of the plug assembly." The connector side referred .l to is the half of the connector that was field installed.

BEco concluded that the workmanship of all field potted connec- -

tors was in question and voluntarily arrived at the decision +

to shut Pilgrim down. BECo notified RI on January 9,1978, of t..

l the problems with the connectors and subsequent shutdown by telephone and submitted LER 78-001/01X-0 on January 10, 1978. '

G. Recuirements and Commitments of the Licensee Related to Quaiification of Connectors l

The investigators reviewed the Pilgrim Nuclear Power Station (PNPS) FSAR to determine the requirements imposed on the ,

licensee and/or commitments applicable to the connector a:sem-blies. The PNPS FSAR is dated January 1970 with subsequent .!

revisions made in response to AEC staff or ACRS questions. l Several sections of the FSAR discuss the ability of electrical equipment to withstand accident conditions.

The AEC staff in FSAR question 7.1.6 requested the licensee to identify all safety related equipment and components (e.g.,

motors, cables, filters, pump seals) located within the primary containment which are required.to be coerable during and '

subsequent to any LOCA or a stsa.ilir.e break accident. It asked the licensee to describe the qualification tests which have been or will be performed in a combined high temperature, pressure, and humidity environment. The licensee's response in Amendment 26 stated that safety related ecuipment inside primary containment which must operate in an tecident environ-ment is limited to isolation valves, recircuistion valves, auto relief valves, and their associated electrical penetra-tions and cabling. Note the absence of connectors or termina-tions.

Section 7.1.4 of the FSAR discusses cetails of cable install-ation design criteria. It states that for the purpose of cable installation, the protection system circuits shall be interpreted in their broadest sense to include sensors, signal cables, control cables, power cables, and the actuated devices.

There is no specific reference to connectors or terminations.

The licensee's statement in 7.1.4 subpart a.3, " Installation Evaluation" states that an evaluatien of the installation sh'all be performed to ascertain that na installation has

'u i complied with the design intent. Field inspections verifies proper installation workmanship and compliance with design instructions; including cable type, identifications, routing, and connections.

FSAR section J.3.2.16, titled " Electrical Equipment Inside Containment Test Program" discusses the ACRS question of ability of electrical equipment to withstand an accident environment inside containment. ACRS statement S.2 states that electrical equipment which must operate inside primary containment in an accident. environment is limited to cables and operators for isolation valves...In addition to designing the equipment to withstand. the accident environment long enough to operate the applicant has agreed to test the perform-ance of the equipment. The resolution of the ACRS concern is -

stated as " Tests have been completed on cables and operators for isolation valves, recirculation line valves, and relief valves. The test results indicate the capability of-the equipment to meet or exceed the design requirements and to perform while under postulated acc.ident conditions." There is no reference to connectors or terminations.

FSAR section J.3.4. ll, "Drywel1 Acciden+ Condition - Electri-cal Component Testing," states that the applicant will provide test data to confirm that those . electrical compenents, instru-men s, and cable located in the ' primary containment will be capable of performing their function during and subsequent to a design basis accident for the length of time required. The resolution of this refers to FSAR section J.3.2.16. As noted above, the licensee did not commit to any qualification testing of connectors.

FSAR section J.3.4.12 states that the RPS and the instru-mentation which actuates the ESF are being designed to the proposed IEEE-Standard. (Reference IEEE-279) The early issues of IEEE-279 allowed equipment qualification by type test data or reasonable engineering extrapolation based on test data.

In summary, requirements for qualification of electrical equipment at the pNpS are basically referenced to IEEE-279 which was issued as proposed criteria in August 1968. This document was written for application to protection systems and states that qualification consisting of type test data er reasonable engineering extrapolaticn based on test data shall be availa:le. IEEE-323 was in preparaden during the time

period under consideration and was released as a trial-use guide in April 1971. The accepted practices to demonstrate "

qualification of electrical equipment as reflected in IEEE-323 ~

was by (a) type test; (b) partial type test augmented by analysis, extrapolation or operating experience; and, (c) ..

analysis or operating-experience. Type tests were the preferred "

qualification demonstration method, but clearly. other approaches, notably the use of analysis extrapolation, and operating experience were acceptable and in widespread use.

d In response to specific AEC/ACRS questions, the licensee R specifically defined those items which were subject to qualifi-cation testing and they'did not include connectors.

H. Quality Assurance - pilarim Construction -

The pilgrim FSAR-discusses quality assurance in Appendix 0.

• The QA plan description in Appendix D defines organizational  %

relationships and functional aspects of Pilgrim quality assurance. H The QA program does not follow the exact format of 10 CFR 50, 9 Appendix B, primarily:because Appendix B was not in effect when the pilgrim program was developed. There were several amendments to the FSAR and -QA program to bring the pilgrim .

program into reasonable con.formance with Appendix B.

The time period when the connectors in question were purchased and installed include the yearF1968 to 1972. This was a transiti:n period in the nuclear industry for quality assur-ance activities due to the AEC issuance of 10 CFR 50, Appendix

. B. On April 17, 1969, the AEC published in the Federal Register for public ccmment a procosed amendment to 10 CFR 50 which would add Appendix B. On April 27, 1970, Appendix B was officially published in the Federal Register, Volume 35, Number 125.

In implemeating a QA' program, BECo had prepared a QA program document dated October 1, 1968. This was revised and re- .

issued on November 11, 1969, and was more definitive. The h revised program included additional requirements such as '

noncenformance control which aligned it more with the require-ments of proposed Appendix B.

The QA program assured that overall responsibility for QA was retained by BECo. The functional aspects of CA/QC however,  !

- were delegated to the A/E - Bechtel power Corporation. The lic~ensee's responsibilities were ac::m:lished primarily by

< n _ s - - . - - - - - ,

23' performing audits to verify implementation of'the overall p rogram. :

Procurement activities and-control 'are of significance to the connector situation. The investigators reviewed procurement documentation to verify that qualification oflecuipment was required. Most of the documentation reviewed was related to

' the procurement of electrical penetrations.because 'the connec-tors in question were procured from the Physical Sciences ....'

Corporation'as an integral part of the penetrations. The

' -applicable procurement specification was Bechtel document 6498-E-28. Section 5.5 of this specification defined the qualification requirements for the penetration and included 100% relative humidity, peak pressure of 56 psig, ?qd peak -

temperature ' of 300* F. These requirements appear to be within l

, 'l the parameters specified in FSAR Section 14, Station Safety Analysis and Section 5, Containment. The specification also 1

contained qualification requirements for radiation dose and-seismic. The qualification requirements were imposed on the entire assembly which included'ene half of the connector assembly. Section 10 of 6498-E-28 defined the specific tests, test duration and sequence to be followed.

l Physical Sciences Corporation supplied' that part or the connec-tor assembly which would mate with the penetration connectors.

Therefore, one half of the conn.ee:or' assembly was installed by craft workers as a field operation. ,3echtel personnel installed the field connectors purportedly to the requirements specified in the Physical Sciences Corporation instruction manual. A microfilm copy of Bechtel Field Inspection Manual dated August 3,1970, was reviewed. Field Inspection Procedure E-2 speci-fied that cable connectors were to be inspected to verify assembly and test per vendor directions. There were no inspec-tion reports on file to verify this and the QC forms did not require this degree of detail of inspection data.

The Physical Sciences instruction manual covered installation of connectors and pottings. Section 1.6 of the manual recommended potting with GE RTV-112 and specified that the method of potting was covered in CE Catalog No. CDS 852, Attachment C.

Attachment C'was revised to include an option allowing use of  !

Dow Corning "Sylgard No.186 f f flame retarding was desired '

and st'ated that the resin' manufacturers should be contacted for processing details. .During review of QA records

-< x , + e- -

w--,, row,,y ,r,--,,-w-e -- -,, - ,

24 it was noted that Field Material Requisition E-7852 dated July 31, 1971 was for " Material for Plugs at Penetration." It specified GE RTV-11 Liquid Silicone Rubber, white. It could not be verified if this was an approved equivalent to GE-RTV-112 or an acceptable substitute. Again, lack of records at thic level and field procurement practices were not defined. m..

The penetrations supplied by Physical Sciences to the above criteria were found to be unacceptable due to inability to meet the stringent qualification criteria. All of the Physical Sciences' penetrations (except two) were subsequently removed and replaced by GE penetrations.

The GE penetrations were purchased by Bechtel frem GE to the same procurement specification with certain revisions. Revision-2 of 64g8-E-28 modified the environmental enveloped with time and increased the accumulated radiation dose to 3.3 X 107 rads. A specific requirement was included to provide connec-tors which would match existing plug's in the field. "

The qualification of GE penetrations was documented on summary sheets which did 'not contain complete test data or test pro-cedures. It could not be verified that the connector types in question were tested as an integral part of the penetration.

The objective evidence provided: to ' support qualification of the connector assembly consisted of letters of certification from Bendix and engineering extrapolation based on experience,

, tests, and/or judgement.

Vendor inspections were specified in the QA program and the procurement specifications as part of the overall program to assur'e a quality product. There was ample evidence that Bechtel had performed vendor shop inspections in conformance with QA program requirements during the manufacture of the penetrations at Physical Sciences and GE. QA records at Pilgrim contained Inspection Reports of periodic shop inspec-tions conducted by Bechtel personnel. Frequent references to tests and unsatisfactory conditions were documented and subse-quently corrected.

e ,- , e ,

7. -

' j e

. -c

. 25 l

Other evidence of quality assurance / quality control was avail-able which indicated the connectors and penetrations were receipt inspected. I The QA records also contained Bechtel Form QC-103, Final _ Inspection Report; Interim Field Reports; and nonconformance reports related to penetrations. However, no records were available to verify inspections were conducted of

-the connector installation. Licensee personnel interviewed 3 stated that they were not involved with the cable termina- .

l

~

tien/ connector installation during-construction, and, there-fore, could not provide any positive information related to the installation or inspection of the connectors and/or potting.

The licensee developed Field QA Auditing _ Procedures, October, '

1970, to define audit requirements as a means of determining '

Bechtel/GE conformance to overall requirements. _ The following procedures were reviewed: .

EQA-5, Clectrical System Audit EQA-6, Electrical Equipment' Audit Sheet (yendor)

EQA-7, Electrical System' Audit Sheet (. cable pulls, termina-tions)

.L The investigators reviewed numerous. licensee audit reports including audits 14, 40,11,177, 246, and 341, related to electrical systems installation.

The' investigators reviewed the AEC Compliance Office inspec-tion feports related to Pilgrim during 1970,1971, and 1972.

Numerous inspections were conducted with regard to electrical systems and primary emphasis appeared to be on cable routing and separation. The inspections were conducted in accordance with the Division of Compliance program at that time. In early 1972, Inspection Reports 72-03, 72-06, 72-10, and 72-13 all referenced inspections of electrical systems but there was no reference to connector installation inspections. The investi-gating inspectors interviewed two AEC inspectors actively involved. in the Pilgrim inspections; neither.could recall any specific problems' in inspections related to connectors. 'This is consistent with the. inspection' program requirements for the 4

_,.m -

_ , r ,, .- - . - m_-- . ., , , . . - . , - - -, ,-,.m-

. . .. . . .~ - . -

26 1969-1972' time period which was defined in TI-3800, Attach-ments H and I. Attachment H was related to Instrumentation =

and Attachment I contained electrical inspection requirements.

Neither required a specific inspection of connectors or environ- u mental qualification data. =

In sumary, it appears that the licensee had developed and implemented _a QA program consistent with AEC requirements for the time period in question. There was objective evidence in =

the form of procurement specifications, vendor inspector -

reports, and other inspection reports of QA/QC activity. The AEC Division of Cat . 'ance reports delineated many areas of noncompliance related to electrical systems but the ccmpliance program did not place any requirements on connectors or provide any guidance for inspection of qualification data. With -

regard to retention of inspection records / checklists, existing standard for records related to nuclear power plants ANSI N45. 2. 9-1974 does not require retention of field workmanship checklists for a period greater tha'n one year. Certainly the program' implemented by BEco in 1969-1970 would not meet present NRC requirements. However, in relation to the time frame in question, the inspectors' determined that the program was in conformance with requirements and no basis for enforcement action was. defined. ~

I. Source of Qualification Records ,

The licensee researched the QA records for the Pilgrim plant for evidence of qualification of the electrical connectors in question. To further substantiate the findings which were certificates of conformance, BECo contacted the A/E, NSSS, and the supplier of the connectors. The information obtained was suoplied as nine attachments to the BEco cover letter response dated December 13, 1977.

Eight of the attachments are dated October / November 1970, and contain material certifications furnished to Physical Sciences Corporation (pSC) by Bendix. The connectors were procured by PSC and supplied to Bechtel as part of the penetration procure-ment'6498-E-28-AC. The certification appears to be based on some test data and extrapolations. -

The material used for the connectors insert is reported to be an alastomeric chloroprene based synthetic rubber, i.e. ,

m--,

27

-Neoprene type W with a clay. filler. The certification states service temperature could be as 'high as 350*F and contains a t

heat aging curve with an extrapolated lower end.

The certification states that Dupont reports Neoprene samples tested under. reverse stress have been exposed to outdoors conditions for 30 years; encapsulation of the Neoprene would act to retard the aging process.

In October 1970, pSC in reference to MIL-C-50150 and MIL-C-26482 defined the environmental conditions and asked Bendix if the connectors'would remain operable under those conditions.

Bendix responded that assuming the connectors are mated and -

~

the receptacle containing a main joint sealing gasket, the Bendix connectors will remain operable under the conditions specified in the reference wire.

The certifications also contain'ed two tables related to the effects of gamma . radiation on physical properties of elas-tomers. The data presented were the resu.lts of tests and included information on dose, hardness, elongation, and tensile strength. The data was obtained from Rubber Age. The original source of the data was not identified.

BEco received a letter from Bendix.; dated December 8, 1977, in response to their request for i'nformation on the qualification of the connectors. Sendix stated tha't the "sucolied" connectors -

were not oualified but are of a type similar to tnose quali-fied under the applicable parts of MIL-C-50150 and MIL-C-25482 (emphasisadded). After a review by their (Bendix) engineering staff, it was their opinion that the supplied connectors, when mated and potted properly were still capable of meeting the a5cve mentioned requirements (as stated in the Bendix letter).

Bendix in closing the letter stated "The above conclusions are Bendix Engineering opinions. A qualification test should be -

set up and run to confirm."

In reviewing all of the above, the concepts of qualification by partial test and extrapolation apparently were used to certify the connector materials. In addition, qualification by similarity is implied in the 1977 letter frem Bendix in reference to MIL-C-52150 and MIL-C-25482. The two milita ry

. specifications contain rigorous qualification test require-ments. However, Bendix did not specify any of the tests performed nor_ did they provide any data relative to :ne actual connectors of the "similar" connectors.

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- The above techniques were commonly used Lin the early 1970s to ,

qualify equipment and the basis for the connector qualifi-cation was presented to Bechtel/BEco in the form of a certif-icate' of compliance which was an acceptable means of providing

. objective evidence'to support qualification. However, require-ments are more rigorous today and, as Bendix recommended, a qualification test should be run.

o Th~e licensee in- preparing his response relied on the certifi- q cation- submitted, qualification by similarity, and the use of B the Sandia test results. The licensee as part of the response H stated:

"A review of the Sandia test results further sustains our conclusions of the adequacy of the existing connector assemblies. In the simultaneous mode of synergistic testing the Bendix connector assemblies withstood the harsh environmental conditions for periods in excess of seventeen days. - As stated in, your bulletin, ' test results of the Bendix connector assemblies during the sequential mode of testing are inconclusive in determining the failing component."

. In the official submittal,'the cover letter dated December 13, licensee did not specifically 1977, state'that and thethe attachments, connectors atEP theilgrim were qualified by test.

Neither did.the licensee state in the cover letter that Bendix recomended a qualification test be run although the letter from Bendix was included as an attachment. Without the attach-ment, it is reasonable for one to imply from the response that qualification testing may have been performed since the Bulletin question being answered asked the. licensee to review the adequacy of qualification testing.

In summary, the licensee provided all the information avail-able to him at the time. No direct statement that qualification tests had been performed was made. The determination that the environmental qualification of the connector was substantiated by th5 available infonnation was an engineering judgement.

.The licensee's representatives stated in interviews that they felt the response was accurate and timely. They took exception 6 4 y -

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29 to the statements in the April 12, 1978 Commission's response to the UCS, particularly those on page 29. The licensee stated that internal discussions were in. progress related to the filing of a statement taking e ception the Commission

l s tatements. H l

!Y. HARAGEMENT EXIT INTERVIEW 0 The investigators met with those licensee representatives identi- q fied by asterisk in Section III.c of this report to discuss the  !

findings of that part of this investigation conducted at BECo 1 during the period from May 8 through May 11, 1978. The investi-ga : ors stated that the investigation was not complete and required additional inquiry at NRC facilities. BEco was informed that they 4 woJld be furnished a copy of the investigation report.

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