ML20153A966

NRC Staff'S Motion for Summary Disposition for Contentions of Intervenors in Proceedings Re Subj Facils.Asserts That Contentions Lack Factual Basis;Thus Summary Disposition Is the Appropiate Remedy Per 3987 U.S.144,157.Cert of Svc Encl

ML20153A966
Person / Time
Site:	Black Fox
Issue date:	07/14/1978
From:	Desiree Davis, Paton W, Woodhead C NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
To:
References
NUDOCS 7811290026
Download: ML20153A966 (200)

Text

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eg 80 N og NRC PUBLIC DOCUMENT ROOM /f e

,mo S UNITED STATES OF AMERICA k; NUCLEAR REGULATORY COMMISSION l Ng g i 197B Y ,3 '

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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD g.

N e In the Matter of )

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PUBLIC SERVICE COMPANY OF OKLAHOMA, ASSOCIATED ELECTRIC Docket Nos. STN 50-556 r COOPERATIVE, INC. AND WESTERN STN50-55[y COOPERATIVE (Black Fox Station, Units 1 and 2) )

NRC STAFF'S MOTION FOR SUMMAR'? OISPOSITION

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THIS DOCUMENT CONTAINS P00R QUALITY PAGES L. Dow Davis Counsel for NRC Staff Colleen P. Woodhead Counsel for NRC Staff William D. Paton July 14, 1978 Counsel for NRC Staff 78112000m4

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' TABLE '0F CONTENTS Page 1

Introduction ...................................................

2 I. General Points of Law ...........>..........................

I I . C on ten ti on s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-1 A. Contention 1 ( Flow Induced Vibrations) . . . . . . . . . . . . . . . . .

Contenti on 2 (ECCS Testing) . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 B.

Contention 3 (Suppression Pool) ....................... 3-1 C.

Supports)................ 5-1

0. Contention 5 (Pressure Vessel 7-1 E. Contenti on 7 (Fire Protection ) . . . . . . . . . . . . . . . . . . . . . . . . .

F. Contention 8 (Cable Trays ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1 t

9-1 G. Contention 9 (Fire Protection) ............. ..........

Contention 10 (Quality Assurance) . . . . . . . . . . . . . . . . . . . . . 10-1 H.

Contention 12 (Spent Fuel Fool ) . . . . . . . . . . . . . . . . . . . . . . . 12-1 I.

Contenti on 13 (Emergency Pl an) . . . . . . . . . . . . . . . . . . . . . . . . 13-1 J.

Contention 15 (Stress Corroston Cracking) . . . . . . . . . . . . . 15-1 K.

18-1 L. Contention 18 (Financial Qualification) . . . . . . . . . . . . . . .

19-1 M. Cordention 19 (Turbine Missiles) . . . . . . . . . . . . . . . . . . . . . .

Contenti on 66 (Sabotage ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66-1 N.

0. ContentionA-1(OffgasExplosions)..................... Al-1 l

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UNITED STATES OF AMERICA NUCLEAR REGULATORY C0ft1ISSION .

'BEFORE THE ATOMIC SA'FETY AND LICENSING BOARG In the Matter of )

)

PUBLIC SERVICE COMPANY OF-OKLAHOMA AND ASSOCIATED l Occket Nos. STN 50-556

. ELECTRIC COOPERATIVE, INC. $ STN 50-557 (Black Fox Statiert, Units I ana 2) l

.NRC STAFF's. MOTION FOR SUWtARY DISPOSITION i

j The United States Nuclear Regulatory Commission Staff (Staff) moves that the contentions' listed below be dismissed pursuant to- 10 C.F.R. 52.,'49 for want of a genuine material issue of fact to- be adjudicated '

. j at the upcoming environmental hearings. The Staff is of the opinion j

that the atk: ached afffdavits, together with the Intervenors' various responses to discovery requests, demonstrate that they have failed to produce a sufficient factual basis for these contentions and that there are no issues of fact worthy of adjudication at the hearing. Accordingly, .

this Atomic Safety and Licensing Board (Board) should dismiss these-contentions a,s a matter of law.

Section I of this pleading wi11 discuss, in general tenns, the law appTicable to summary disposition motions. By means of the attached affidavits of NRC Staff members,Section II will show, contention by contention, that there are no material issues of fact raised by certain of the Intervenors' contentions. Legal arguments and statements of material facts as to which there are no genuine . issues will be listed infra by contention along with the. supporting Staff affidavit relating to that contention.

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I. General Points of Law l

0 l The requirement that there be a factual basis for each contention in l

issue in a. Nuclear Regulatory Casaission proceeding' derives from two i

sourcas: 1) the contention requirement of 10 C.F.R.12,714 and (2) .

the summary disposition provisions of 10 C.F.R.12.749. As will be shown below, a motion to dismiss will lie on the basis of either rule.

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l l A. Factual Bases for Contentions Under 10 C.F.R. Section 2.7%

The new.10 C.F.R. Section 2.714(h)(24(b) requires that there be a factual 3

basis for each contention set forth by each petitioner to intervene. E i

That section states that the Intervenor must file a supplemental to his petition to intervene that includes. ..." a list of the contentions

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whica petitioner seeks to have litigated in the matter and the bases for each_coAtention.. set _factit.d*h **=aaable. specificityv"- 43 Fed. Rg.

17802 (April 26,.1978).

Thus, prior to any- hearing, the Atomic Safety and Licensina Board must assure itself that each contention presents a genuine issue

- appropriate for resolution in the proceeding. Caselaw pmvides that See Duouesne Licht Co.,

M This concept is supported by prior casalaw.

et al (Beaver Valley Power Station, Unit No.1), ALAB-109, 6 AEC 243 NE TApril 2,1973); Vircinia Electric and Power Co. (North Anna Power 14,1973);

Station, Units 1 and 2), ALA8-146, 6 AEC 631, 533 (Saptember Wisconsin Electric Power Co. , et al. (Point Beach Nuclear Plant, Unit 2), ALAB-137, 6 AEC 491, 505 (July 17, 1973).

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l "Before canumancing an evidentiary hearing, a licensing board must, of course, pass upon the i sufficiency of every contention contained in an i intervention petition which has previously been granted. And ... the board is to exclude from j

consideration at that hearing any contention which does. not present a: genuine issue appro-pMate for resolution in the proceeding. Stated othemise, the hearin is not to embrace a. con-tantion which either 1) as presented, fails to

.i satisfy the requirements of 2.714; or (2) can t be susunaM1y rejected on the merits under the

! provisions of Section 2.749 of the rules of l practice.y i

As an illustration of 'this principle, in the Beaver Valley case, the Appeal Board stated that a Licensing Board e

" ... must be satisfied,. with respect to each

! contention which the- petitioner seeks to liti-

, gate, that a genuine issue in fact . exists. Any contention which on preliminary examination does i

not survive the application of that standard is to be excluded from consideration at the evi-dentiary heaMng." Duquesne Light Co., et al.

(Beaver Valley Power Station, Unit No.1), AL.AB-109, 6 AEC 243, 24S (ApH 1 2, 1973) (Emphasis - .

added). '

As will be shown later by the attached Staff affidavits, many of the

' intervenors' contentions lack a sufficient factual basis to be allowed to gar to hearing in this proceeding. The inadequacy of the Intervenors'

- position will be further spot 11 pitted by their responses to Staff inter-rogatoMes and cross examination of prospective witnesses when they were j '

asked for the factual bases of their contentions.

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E Mississipoi Power and Light Co. (Grand Gulf Nuclear Station, Units 1

. ano 2), ALA8 130, 5 AEC 423, 424-25 (June 19,1973),

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B. Summary Disposition Under 10 C.F.R. Section 2.749

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In addition to the factual basis requirement of Section 2.714, the )

Commission's rules provide that a moving party is entitled to summary  ;

l disposition if it can be shown that there are no material issues of fact to be adjudicated at the hearing. 10C.F.R.12.749. That Section states:

Summary Disposition on Pleadings 52.749 Authority of presiding officer to dispose j of certain issues on the pleadings.

(a) Any party to a proceeding may, at least i forty five (45) days before the time fixed for

. the hearing, move, with or without supportino affidavits, for a decision by the pre:. ding officer

( - in that party's favor as to all or any part of the matters involved in the proceeding.

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C. 8urden of Proof The Supreme Court and NRC.have clearly held that it is the party seeking sununary judgment, not the party opposing it, which has "the burden of showing,the absence of a genuine issue as to any material fact, . ." . Adickes v. Kress & Co., 398 U.S. 144, 157 (1970); Cleveland Electric Illuminating Co. (Perry Nuclear Power Plant),

ALAB-443, 6 NRC 741 (November 8, 1977).

The rules governing summary disposition are analogous to the law of Summary Judgment in the Federal Courts under the Federal Rules of

( Civil Procedure.E ni that the moving party mpst demonstrate that there is no genuine issue of fact remaining to be decided and that the uncontroverted facts entitle him to judgment as a matter of law.O )

Affidavits ",etting forth the material facts about which there are no l l

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genuine issues to be heard may accompany the motion to dispose of i issues in the pleadings, and the affidavits may be supplemented or l opposed by depositions, answers to interrogatories or further affi, davits.

i l U Alabama' Power Company (Joseph M. Farley Plant, Units 1 and 2), ALAB- '

182, 7 AEC 210, 217 (March 7, 1974); Public Service Comoany of New  !

Hampshire (Seabrook Station, Units 1 and 2), LSP-74-36, 7 AEC 877, 78

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(May 17, 1974); Gulf States Utilities Comoany (River Bend Station, '

Units 1 and 2), LBP-75-10, 1 NRCI 246, 247 (March 20, 1975).

b Adikes v. Kress, 398 U.S.144,158-161 (1970).

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While it is n'ot necessary to present evidence in order to defeat a motion for Sumary Disposition since the motion itself and accompanying affidsvits must discharge the movant's burden (and no defense to an 4

insufficient showing by movant is required) it is said to be perilous for an opposing party not to proffer any countering evidentiary materials or affidavits since the rule clearly states that a party t

opposing the sumary disposition motion may not rest upon the mere allegations or denials in his answer but rather must provide by affidavit, deposition or answers to interrogatories, specific facts 1

showing there is a genuine issue of fact in controversy. 10 C.F.R. l j

52.749(b); Perry, supra,at754..

In this regard, a Licensing Board has said:

'In order to defeat a. motion for sumary disposition the Intervenor must establish (or the Board perceive from the record) that there does exist a genuine -

issue of material fact with respect to each contention so attacked. At this stage, mere allegations in the pleadings are not sufficient to establish the existence of an issue of material fact. 10 CFR 52.749(b); See Orvis

v. Brickman, 95 F.Supp. 605 (USDC, D.C.1951), afFd.

196 F.2d 762 (D.C. Cir.1952); qag, also 6 Moore T!IC15[3].

To defeat sumary disposition an opposing party must present facts in the proper fann; conclusions of law will not suffice. The opposing party's facts must be material, substantial, not fanciful, or merely suspicious.

One cannot avoid suninary disposition "on the mere hope that i at trial he will be able to discredit movant's evidence; he must, at the hearing, be able to point out to the court

! something indicating the existence of a triable issue of l'

material fact" 6 Moore's Federal Practice 50.15[4]. One l

( l cannot "go to trial on the vague supposition that something See may turn up." 6 Moore's Federal Practice 56.15[3].

l Radio City Music Hall v. U.S.136 F.2d 715 (2nd Cir.1943).

In Orvis v. Brickman, 95 E!Iiipp. 605 (0.0.C.1951), the Court, l in granting the defendant's motion for summary judgment under the Federal Rules said: l I .

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All the plaintiff has in this case is the hope that on cross-examination ...the defendants . . .

will contradict their respective affidavits.

This is purely speculative and to permit trial on such basis would nulify the purpose of Rule 56....

Gulf States Utilities Company, (River Bend Station, Units 1 and 2), LBP-75-10, 1 NRC 246, 248 (March 20, 1975) i (Fot '+esomitted) i Suninary disposition is appropriate in administrative hearings because it makes possible the prompt disposition ofa. case on its merits without a fonnal hearing by permitting a party to pierce his oppo-nents pleadings by presenting material evidence in affidavit fann l whichestablishesthatnofactualdisputeexists.E The Staff sub- f mits that such a procedure for saving hearing time by culling out , l baseless allegations is particularly appropriate in the instant case since, as will be shown below by affidavits and the parties' own answers to discovery and depositions, there is no factual basis for any of 'the Intervenors' contentions.

5/ ellaorn G and Robinson, Sumary Judament in Administrative Adjudication, 84 Harvard L.Rev. 612 (1971).

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Contention 1 I. Contention Intervenors# Contention number 1 deals with the adequacy of the Staff assessment of the possibjlity of flow induced vibrations on six . components.

For the reasons listed below, the NRC Staff believes that no litigible i

issue dealing with the metter of flow induced vibrations of components remains to be decided in this case.

I II. Argument ,.

The Staff Affiant, Mr. -Bill Kane (famer BFS Safety Lf censing l

Project Manager), analyzed the BFS Nuclear Steam Supply System (designed by General Electric and reviewed in the GESSAR-238 NSSS Standard Safety Analysis Report, docket No. STN 50-550). l

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A. Nuclear Steam Supply System The BFS NSSS is subdivided into three major categories;

1. reactor internals (includes jet pumps, spargers, control rodsandincoreinstrumentation),

2. piping (includes recirculation valves), and

3. fuel (includes fuel pins).

1 Reac'cor internals Information in the PSAR with respect to the ability of reactor internals to withstand vibration was reviewed by the Staff and found to contain no infomation wh'ich would indicate vibration would pose a safety threat to BFS. Kane Affidavit at 3-4.

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General Electric has designated the Perry reactor as being a prototype of the BFS station and preoperational tests of that reactor will be done prior to startup of the BFS units. Kane affidavit at 4-5. .

In addition. to preoperation tasts, visual inspections of both the BFS and Perry prototype reactors will be conducted.. Kane affidavit at. 3.

Should the Perry not become operational prior.to BFS, another reactor will become the prototype reactor for preoperational generic vibration analysis. Thus the' tests now scheduled for Perry are ensured of com-plation for BFS. Kane affidavit at 7. .

Applicant has comitted to install a loose parts monitoring system to detect any part which might vibrate icose. Kane affidavit at 7.

r. Pipino l As to vibration in NSSS piping, the Applicant has committed to perfom, i l

tests to evaluate the vibration dynamic effects tests for piping on Class 1 piping systems prior to operation of the Plant. The pre-op tests will be in accordance with Reg Guide 1.20 " Vibration Measurements on Reactor Internals". These tests are designed to provide additional verification that the SFS reactor coolant pressure boundary will withstand vibration dynamic effects. Kane affidavit at 2.

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' 3. . Fuel 8x8 fuel was covered in the GESSAR-238 r , clear Island SER and found to be safe from a vibration standpoint. Kane affidavit at 6. In addition, actual red vibration experiments and. operating information from fuel arrays similar in design, material and frequency to Black Fox have shown no vibration safety problem applicable to BFS. Kane affidavit at 5-7.

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In. addition to the showing and review made by the Staff above to sustain l their burden for sumary disposition, the Staff notes that as of June 1978, _ )

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Intervenors had no facts with which to controvert the Staff affidavit, Bridenbaugh Deposition at 113-117, nor had they even reviewed the PSAR or SER on the subject. Tr. 116. ]

III. Statement of Material Facts About Which flo Genuine Issue Remains To Be Litigated 1

1. The BFS NSS5 system has been reviewed and found to contain l i

no vibrational problems.

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2. In order to confinn the Staff on Applicants' NSSS analysis, prototype and pre-op vibration testing will be accomplished, furnishing I

a reasonable assurance that no vibrational safety items will exist when BFS comences operation.

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UNITED SS TES OF AMERICA tWCLEAR RiiGULATORY CDMMISSIN .

BEFORE 'IEE ATmIC SAFETY AND LICENSING BOARD In the Matter of )

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PUBLIC SERVICE COWANY OF OKLAIOMA )'

ASSOCIATED ELECTRIC COOPERATIVE, INC. ) Oocket Nos. S m 50-556 AND WESTERN FARMERS COOPERATIVE, INC. ) STN 50-557

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(Black Fox Station, Units 1 and 2) ) 1 AFFIDAVIT OF W. F. KANE IlffERVENORS CCN1ElffIN 1 My name is W. F. Kane. I an employed by the Nuclear angulatory Commission as a Licensing Project Manager in the Division of Project Management, Office of Nuclear Reactor Regulation. I have been enployed in this position since June 1973. My professional qualifications are contained in an attachment to this affidavit. This affidavit was prepared by me or under my supervision.

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\ The purpose of this affidavit is to address Contention 1 which reads as follows:

The Intervenors contend that the applicant has not adequately assessed for Black Fox Station, Units 1 and 2, flow induced vibrations on the following couponents:

(1) Jet Pumps (2) Spargers (3) Fuel Pins (4) ancirculation Line Valves (5) Control Bods (6) Incore Instrumentation All of these % Ants are within the nuclear steam supply system as described in the GESSAR-238 NSSS Safety Analysis Report, Docket No. STN 50-550, which has been referenced in the Black Fox Station, Units 1 and 2 Preliminary

. Safety Analysis Report. Each of these components is included and under of three.

l . areas which the staff has previously evaluated and found acceptacle in its l

review of the GESSAR-238 NSSS design, namely, piping systems, fuel assemoly,

. and reactor internals.

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Piping Systems me staff has evaluated the information provided by the applicants in the Black Fox Station, Units 1 and 2 Preliminary Safety Analysis Report and with respect to vibration of piping systems has concluded that it is accept-aole for :.he construction permit stage. Our evaluation of the vibration of piping systems which are part of the nuclear steam supply system is provided in Section 3.7.1.1 of Appendix A to the Black Fox Station, Units 1 and 2 Safety Evaluation Report and is provided below.

Preoperational piping vibration tests will be conducted on the main steam lines and the . recirculation system. In response to our concern, the General Electric Company has comitted to conduct preoperational vibration

! tests on all American Society of Mechanical Engineers Boiler and Pressure vessel Code Class 1 piping in the GESSAR-238 NSSS scope. With this require-ment met, the preoperational vibration test program which will be conducted ~

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for the nuclear steam supply system during startup and initial operating conditions on all safety-related American Society of Mechanical Engineers Boiler and Pressure vessel Code Class 1 systems, restraints, components, and supports is an acceptable program for issuance of the Preliminary Design Approval. me tests will pr' ovide additional verification that the piping and piping restraints of the system are designed to withstand vibrational dynamic effects due to valve closures, pu _ rips, and other operating modes associated with the design operational transients. The formal detailed description of the planned tests will be included at the final design stage of review. The planned tests will develop loads similar to those excerienced during reactor operation. .

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Ccnipliance with this test program constitutes an acceptable casis for verifying the existence of adequate design margins as specified in Criterion 15 of the General Design Criteria and is acceptable for the. Preliminary Design Approval.

The above evaluation is applicable to the recirculation line valves, item 4 of the contention, as well as the remainder of the components of the piping system that form the reactor coolant pressure boundary.

Reactor Internals The staff has evaluated the information provided by the applicants in the Black Fox Station, Units 1 and 2 Preliminary Safety Analysis Report with-respect to vibration of reactor internals and has concluded that it is

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I acceptable for the construction permit stage. Our evaluation of the vibration of the reactor internals is contained in Section.3.9.1.3 of the Black Fox Station, Units'1 and 2 Safety Evaluation Report and is provided l

below.

Dynamic systems analysis methods for reactor internals design and for correlation of analytical and test results are oiscussed in Sections 3.7.1.3 and 3.7.1.4 of the GESSAR-238 NSSS Safety Evaluation Report ( Appendix A to this report). The applicants will conduct confirmatory vibration testing and subsequent visual inspection as part of the Black Fox Station preoperational tests to provide added confirmation of the capability of the structural elements of the reactor internals to sustain flow-induced vibrations. The ,

proposed program is consistent with angulatory Guide 1.20, " Vibration s

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7 Measurements on anactor Internals"- for "similar to prototype" testing provisions stated in Items D.1 and D.2 of the guide. The applicants have further agreed that, in the event' that neither Perry Unit 1 (Docket No. 50-440),

the designated 238-inch vessel prototype, nor any other boiling water reactor plant becomes a valid prototype before the Black Fox Station operating license

- stage review is completed, Black Fox Station will be considered a " prototype" plant.

Additional evaluation of vibration of the reactor internals is contained in Section 3.7.1.3 of Appendix A to the Black Fox Station, Units 1 and 2 Safety Evaluation Report and is provided below.

With regard to flow-induced vibration testing of reactor internals, the

(' General F.lectric Ccapany has stated in the GESSAR-238 NSSS Safety Analysis aeport that the first 238-inch size BWR/6 system will be considered a proto-type design and will be instrumented and subjected to both cold and hot two-phase flow testing to demonstrate that flow-induced vibrations similar to those expected during operation will not cause damage. The 238-inch BWR/6 system currently scheduled for prototype testing is that for the Perry Nuclear Power Plants Units 1 and 2, Docket Nu:eers 50-440 and 50-441.

Specific predictions and acceptance criteria will be supplied at the operating license review stage for the Perry Nuclear Power Plants, Units 1 and 2.

The preoperational vibration assurance program for the reactor internals provides an acceptacle basis for verifying the design adequacy of these internals under test loading conditions that will be comparacle to chose experienced during operation. The comoination of tests, predictive analysis,

'and post-test inspection provide' adequate assurance that the reactor internals may _

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be expected during their service lifetime to withstand the flow-induced vibrations of reactor operations without loss of structural integrity. The continued integrity of.the reactor internals in service is essential to assure the retention of all fuel ammamh1les in their place as well as to permit unimpaired operation of the control rod assemblies to permit safe reactor operation and shutdown.

The conduct of the preoperational vibration tests constitutes an acceptable basis for demonstrating design , adequacy of the reactor internals in fulfilling the applicable requirements of Criteria 1 and 4 of the General Design Criteria and in conforming with the provisions of Regulatory Guide 1.20, " Comprehensive Vibration Assessment Program for Beactor Internals i During Preoperational and Initial Startup Testing." ,  ;

The above evaluation is applicable to the jet pumps, spargers, control  !

rods, and incere instrumentatien, Items 1, 2, 5, and 6 of the contention as well as the remainder of the retator internals.

Fuel Assembly Our evaluation of vibration of the fuel pins, Item 3 of the Contention, is provided by reference in Section 4.2.1 of Appendix A to the Black Fox i

Station, Units 1 and 2, safety Evaluation Report which states:

Details of our evaluation of the 8x8 fuel assembly design are included in Appendix D to the GESSAR-238 Nuclear Island Safety Evaluation Report, which

. deals with 8x8 fuel assembly reloads. The only difference between the GESSAR-238 NSSS 8x8 fuel assencly and the reload 8x8 fuel assembly is that the total active fuel length is 4 inches greater in the GESSAR-238 NSSS fuel r -- . _ __ ,

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1 rods _ and the fission gas plenum length is 0.75 inches greater in the GESSAR-238 NSSS fuel rods. Rose changes do not affect our conclusions regarding the 8x8 ruel evaluation since:

1. Se design limits for both types of 8x8 fuel are the same. j i

2. Se service life and peak linear heat generation rate are the same for both designs.

3. Se same analytical methodology applies and the accident consequences are the same for both designs. l Se pertinent material in-Appendix b to the GESSAR-238 Muclear Island Safety Evaluation aeport, dealing with fuel pin vibration, is provided f below.

Confidence that the vibration and fretting characteristics of the 8x8 assemblies are known, is based on rod vibration experiments (Footnote 7) and the operating exterience with other types of fuel assenolies in general and the 7x7 design in particular. Se 7x7 and 8x8 assemblies are very similar in this regard. Se fuel rods in both are of similar design, are made of the same material, and have nearly the same natural frequency. Se fuel rod spacer grids in both types of assembly also are of similar design, are made of the same materials, and exert the same spring force. Both j operate at the same pressure and temperature with nearly identical fluid (7) GEAP-4059, V1 oration of fuel aods in Parallel Flow," E. P. Quinn, July 1962 l

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I velocities and quality. Further verification of the adequacy of the design has been provided by the testing of an assembly of similar design  :

for 7000 hours0.081 days <br />1.944 hours <br />0.0116 weeks <br />0.00266 months <br /> in high pressure, two-phase flow loop (Footnote 8). This test was performed by ASFA-Atom, a ,Swedish Boiling Water Reactor and.a General Electric licensee as part of a fuel development program.

Icose Parts Monitoring The staff has evaluated the.information provided by the applicants in the Bisck Fox Station, Units 1 and 2 Preliminary. Safety Analysis Report with respect to the monitoring of load parts. Our evaluation of the montoring of loose parts as contained in Section 5.2.1.3 of the Black Fox Station, Units 1 and 2 Safety Evaluation Report and is provided below.

The applicants have agreed to install a loose parts acnitoring system on the Black Fox Station plants to satisfy one of the staff interface requirements <

t listed in the Safety Evaluation Report for GESSAR-238 NSSS (see Section 5.2.1.'3 of AEpendix A to this report). We find this commitment acceptable for this stage of review. During our review of the final design we will verify that this conmitment has been appropriately inglemented. l During the course of the discovery phase, the intervenors addressed ,

l a number of interrogatories to the staff that dealt with Contention 1. The interrogatories and the responses thereto supplement the staff's evaluation regarding flow-induced vibration provided in the Black Fox Station, Units 1 and 2 Safety Evaluation Report and_ repeated earlier in this affidavit.

H (a) _ Letter, J. Hinos to V.. Moore, February 4, 1974 _

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^K i- .'. WILLINI F. KANE .

'*f DIVISION IF PRa7ECT3tdbAG5ENT

U. 'S. NUCI. EAR REGULATCRY cot.IISSICN jj

',r PROFESSIONAL CUALIFICATIONS

. LI am a Licensing Project Manager on the technical staff of Light Water Reactors Branch 4, Division of Project Management, U. S. b.hciear

,lIih,l i*

l Regulatory Ccmnission. As a member of this staff, I have the function l;' lj af analyting and evaluating the radiological safety a.spects of nuclear Il {;i c facilities for which the Ccanission is responsible, including licenses , j, and authoritations for the design approvals, constmetion and operation >

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of nuclear resc,. ors and standard nuclear pcwer plant designs. I have the technical respensibility for the management and technical coordina- Ph3" tien of the safety reviews for the applicatiens assigned to me, , /,

I accepted an appoint:nent with the technical staff of the U. S. Atomic Q!q.,

Energy regulatory organi:ation in 1973. I have the primary respcnsibility f:r the management of the review of radiological safety matters on five

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..:llity applications for nuclear power plants which include Black Fox 4 c ation, Units 1 and 2, Ccmanche Peak Steam Electric Station, Units 1 .

l :nd 2, Hartsville hbclear Plant, Units A1, A2, B1 and B2, Nine Mile ~

Point Nuc'. ear Statien, Unit 2 and River Bend Staticn, Units 1 and 2. h.1

., ddition I have similar responsibility for three General Electric '

1 i any standard design applications, GF.SSAR-233 Nuclear Island and the .g

.cSAR-251 and GESSAR-238 Nuclear Steam Supply Systems. I am also the ;g* t aaff's technical coordinator for the review of the General Electric Ocpany's Mark III containment design. i

' graduated fr m Widener College in 1961 with a Bachelor of Science f Legree in Mechanical Engineering. I perfonned my post-graduate work j in Nuclear Engineering at Catholic University in the peried frem 1971 - .)

!973. I am a registered professional engineer in the state :t California.

I have a total of 12 years of professional experience, all in the nuclear '

. field. Six of the years I was 'emoloyed by the Atomic Energy Division cf ,4 the Allis Chalmers Manufacturing Ccmpany where I was affiliated with the i,%f mechanical design and field installation of the nuclear steam supply system- fer the Lacrosse Boiling Water Reactor. Fo11cwing this I was $pf

-mpicyed by hbclear Associates International Corporatien where I worked ,

as a censultant to the utility industry speciali:ing in the steas of [..4 mechancial design and analysis of nuclear steam supply system tomcenents if and quality assurance. Subsecuent to this -I was employed by the O Reactor Fuels Jivisien of Nuclear Fuel Services, Inc. , were [ speciali:2d ..7 in the area of -he mechanical design and analytical .ncdel develcpment for Ip 4 nuclear fuel re', cads fer light water reactors. ,

.4 ,

I ~am a meder :f the .american; Nuclear Society, American 5cciat/ cf (;,

• *anical Sng:neers, and the Henorary Society of Tau Beta Pi. I am tuthor or co-author 46 several technical publications. 4'[ '

8*-l 1

.l I have read the foregoing affidavit and swear that it is true and accurate to the best of my knowledge.

~~

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William F. Kane, Project Manager Light Water Reactors Branch No. 4 Division of Project Management Subscribed and sworn to before me 4

Gthis N day of July 1978.

- - L V fla L l ifh O'-m -'

Nottry Puolic 'q My Ccanission Expires MS.k / / f[l I/ 'I

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2-1

'ContEnti6n 2

. 1 1

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I. Contention l Intervenors' Contention 2 maintains that the BFS emergency core cooling system will not comply with the provisions of 10 CFR Part 50, Appendix K. However, as the enclosed affidavit of G. Mazetis and R. Frahm indicates, the NRC Staff has examined General Electric's Section 6.3 of the GESSAR-238 NSSS Safety Analysis Report and found that the document, including all references, shows that the BFS ECCS is in .

compliance with Appendix K.

II. Arcument The enclosed affidavit and SER report indicates that the NRC Staff has reviewed the Emergency Core Cooling System for Black Fox and that it complies in all respects to Appendix K to 10 C.F.R. Part 50. Mazetis and Frahm Affidavit ,at 1. However, in their November 30, 1976 Answer to Staff Interrogatories, Intervenors through their consultant, Mr. Grecory Minor, stated that it was the Intervenors' position that Appendix K compliance l for Black Fox ECCS had not been demonstrated because of inherent problems and inaccuracies in the computer codes used to show Appendix K compliance.

Intervenors' November 30, 1976 Resnonse to NRC Staff Interrogatories at C2-3. I l

l

', J

)

,' 2-2 However, at the June 2,1978 deposition of Mr. Minor, the Intervenors' prospective witness on ECCS repeated that although he doubted the l l

general ecuracy of ECCS evaluation codes, in order to find. any j non-conservatisms in an ECCS code, detailed analyses of the codes used must be conducted and that he had not done any such analyses.

Tr. 113-114.

1 Intervenors' witness, in both his answers to interrogatories and deposition, maintained that full scale testing cP the ECCS was necessary to show compliance with Appendix K. Intervonors' Answers to Staff Discovery at page 3; deposition Tr. at 114. As will be shown below, the NRC

( Staff believes that there'is no legal requirement that fuTT scale ECCS testing is required by Appendix K. Appendix K to 10 CFR Part 50 provides that: .

- 4. To the extent practicable, predictions of the evaluation model, or portions thereof, shall be i

compcred with applicable experimental information.

Section II4 to 10 CFR 8 Part 50, Appendix K. J (emphasis added)

That full scale testing (as opposed to model analyses or predictions) are not required to be conducted is further emphasized by the following l statement contained in the Camnission's comments in the opinion accom-panying the Rulemaking Hearing on Acceptance Criteria for Emergency Core Cooling Systems for Light-Water-Cooled Nuclear Power Reactors:

2-3 4

i ..- . ..

The need for canparisons of the calculations of analytical models with experimental data is dis-cussed and the value is recognized in the written testimony of nearly all of the- participants, includ-ing the Regulatory Staff (Exhibits 1001 & 1113).

Westinghouse- has stated the existance of some probles in interpreting the requirment for such canparisons. It is reasonable to restrict these comparisons to unose enan cno proponenu or evo iua-tion models have made of the r own volition to check out certain features and to comparisons requested .

by the Regulatory Staff. Rulemaking Hearing on .

Acceptance Criteria for ECCS Systes for Light Water Reactors, CLI-73-39, 6 AEC 1085, 1127 (December- 28,1973).

(Emphasis added.)

As indicated in Section 4 of " Status Report by the Directorate. of Licensing In the Matter of General ElectHc ECCS Evaluation Model Conformance to 10 CFR S0, Appendix K" (submitted to ACRS in October 1974), comparisons to applicable experimntal information have been performed and accepted.

,Therefore, while the staff continues to be openly supportive of practicable

! confirmatory experimental programs, the staff concludes that Black Fox .- l Station has demonstrated compliance with 10 CFR 50, Appendix K.

Thus, the NRC Staff maintains that the Applicant has shown compliance .

1 withAppendixK,thatfullscaletestingoftheECCStoshowAppendixKcom-)

pliance is not required as a matter o'f law and that there are no further facts to be litigated in this matter. Consequently, Contention 2 should.

be dismissed for want of any disputed facts to be litigated.'

III. MateHal Facts As To Which There Are No Genuine Issues

1. A review of General Electric's Section 6.3 of GESSAR-238 NSSS Safety Analysis shows that the ECCS designed for BFS complies with Appendix K.

2. As a matter of law, no full ale ECCS testing is needed to sher compliance with Aopendix K.

. = -

UNITED STATES OF AMEPTCA NUCLEAR REGULATORY C0h .sSION BEFORE THE ATOMIC SAFETY AND LICENTING BOARD i

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In the Matter of )

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) I PUBLIC SERVICE COMPANY OF ) Docket Nos. STN 50-556 OKLAHOMA AND ASSOCIATED ) STN 50-557 i ELECTRIC COOPERATIVE, INC. l AND WESTERN FARMERS ELECTRIC COOPERATIVE

)

(Black Fox Station, Units 1 an3 2) )  ;

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AFFIDAVIT OF GERALD R. MAZETIS AND RONALD K. FRAHM ON CONTENTION 2 4

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AFFIDAVIT OF G' NAZETIS AND'R. FRAHM ON CONTENTION 2 The purpose of this discussion is to address intervenor Contention No. 2 regarding the acceptability of the Black Fox emergency core cooling systems. Contention No. 2 relates to compliance of Black Fox Units 1 and 2

- ~ ~ to the requirements of 10 CFR 50, Appendix K. Black Fox is a BWR/6 standard plant using the same emergency core cooling system design previously accepted on GESSAR-238 NSSS (see Section 6.3.2 of the GESSAR-238 NSSS Safety Evaluation Report copy enclosed). The staff has concluded that GESSAR-238 NSSS, and, therefore, Black Fox, has demonstrated compliance with 10 CFR 50, Appendix K. Staff Interrogatory No. 3 requested additional specifics upon which intervenors contend that 10 CF'R 50, Appendix K, has not been complied with by the applicant. The intervenor's response referred to Criterion II-4 of Appendix K to 10 CFR 50 (Exhibit A-2). This criterion states: ,

('

"To the extent practicable, predictions of the evaluation model, or portions thereof, shall be compared with applicable experimental information. " ,

Intervenors further contend that to meet the specific requirements of 10 CFR 50, Appent K, requires full-scale reactor plant loss-of-coolant tests, and a demon: : ration that a loss-of-coolant accident occurring as a result of transient has a negligible probability. Intervenors also request that the use and limitations of the Power Burst Facility for LOCA test'ing of BWR feel be provided. Other interrogatories related to the adequacy of experimental programs have also been submitted by the intervenor(25,27,28,29,30).

It has been the staff's judgment that the Black Fox Station is in compliance with the cited Criter_ ion II-4 of Appendix K. This criterion is clearly not intended to specify full-scale testing requirements, nor r-, -, --- -- -

2-

,n I is it intended to dictate any particular test program. :This interpretation is supported by the Opinion of'the Comissioners issued December 28, 1973, page 133:

"The need for comparisons of the calculations of analytical models with experimental data is discussed and the value is recognized in the written testimony of nearly all of the participants, including

~

the Regulatory Staff (Exhibits 1001 & 1113). Westinghouse has stated the existence of some problems in interpreting the requirement for such comparisons. It is reasonable to res*.rict these comparisons /

to those that the proponents of evaluation models have made of their own volition to check out certain features and to comparisons requested by the Regulatory Staff."

As indicated in Section 4 of " Status Report by the Directorate of Licensing In the Matter of General Electric ECCS Evaluation Model Conformance to 10 CFR 50, Appendix K" (submitted to ACRS in October 1974), comparisons to

, applicable experimental information have been performed and accepted.

Therefore, while the staff continues to be openly supportive of practicable confirmatory experimental programs, the staff concludes that Black Fox Station has demonstrated compliance with 10 CFR 30, Appendix K.

We, Gerald Mazetis and Ronald Frahm, having been duiy sworn,. state that:

we are employed as Section Leader and reactor engineer, rr ively, for the Reactor Systems Branch, Division of Systems Safety in the Office of Nuclear Reactor Regulation with the U.S. Nuclear Regulatory Comission.

A copy of our professional qualifications is attached to this affidavit.

We are the authors of the above affidavit and,believe it to be accurate to the best of'our knowledge and belief, d .A e -,4.

87~p75 o,p- //b74y/MO . *A'W " A o" Mbsr-60n#ey Ovey r y . Gerald R. Mazetis 7 ones hh,

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Ronald K. Frahm.

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GERALD R. MAZETIS .

PROFESSIONAL QUALIFICATIONS t

I an employed as a Section Leader in the Reactor Systems Branch, Division of Systems Safety, Office of Nuclear Reactor Regulation. My responsibtLities include supervising the safecy' reviews of the reaccor coolanc, emergency core cooling, and other auxiliary systems which are assigned to me during the review of nuclear power reactor license applicacions.

I graduated from the U.S. Naval Academy in 1963'wich a Bachelor of S:isnca degree._'In 1968 I received a Hascar's degree in .Nucicar Engineering fre-l Catholic University of America, i

From 1968 to 1972, I was an engineer with the General Elact ric Corapany l

where I was involved in .che licensing of boiling water reactors. My

( ductas included coordinating technical inputs to safety analysis reporta j and participac$ng in various safety reviews of General Ele.:cric reaccor j systems.

In January 1973, I accepted emplo,yment with the Acomic Energy Comnission (now the Nuclear Regulatory Commi,ssion) in the Reactor Systems Branch.

I have been responsible for the review of various safacy systems ci Davis Besse Unic No. 1, Clinton Station, and the B&W standird planc design.

In addition, I have reviewed the LOCA analyses for several pressuri:eu water reactors.

m 6

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,N _

RONALD K. FRAKM PROFESSIONAL QUALITI.'ATIONS

-I am a Reactor Engineer in the Reactor Systems Branch, Division of Systems Safecy, Of fice of Nuclear Reactor Regulacion. I am responsible for conducting saf ety reviews and evaluacions for light wacer ,re.sctor emergency core cocling, reactor coolanc, and various auxiliary rystems assigned to me during the review of nuclear power reaccer license applicacions. 1 I received a Mechanical Engineering degree in 1959 and a Pascer of Science degree in 1962 from Scevens Inscicuce of Technology. From 1964 to 1975 I I

have received 21 credics coward a D<,ccor of Philosophy degree ac cne Catholic University of America.

\

During 1961 to 1966 I v'sa an Assistant Professor in the Engineering

(' .

Department of the United Scaces Naval Academy coaching courses in mechanical engineering.

During 1966 to 1967 I was staff mechanical engineer in che Maintenance Engineering Branch at the Nacional Inscicutes of Healch. I was involved in the design and review of central heacing, vencilacing, and air conditioning systems and various equipment and systems associac4d with controlling laboratory and hospical environmencs.

From 1967 to 1972 I was a senior project engineer ac che U.S. Naval Ship Research and Development Laboratory. I was responsible for develap cis and coordinating rosearch and developmenc programs in the area Of propulsion systems for' naval vessels, o

in Jewe 1972 I accepted employment with the Atomic Enargy Commissi:n ' .sv che Nuclear Regulatory Commis 41cn) in the Reaccor Sys t s.as 3 ranch. I %2"c been responsthie for reviewing safety syscams on tid hc water ro..e.ars.

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tes on the BWR/6 must be suositted, for *1RC staf f review. The results in tne G

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al El ectri moanyTopical'R'eporft:E00108'46, "OuR Core Spray Sistribution," ' for ,

314R/4 and SWR onfigurations, with results for DUR/6 configurations - mised later.

A final copy of the rietary version of the spray cooling tas- as suomittea for NRC staff review. He will re on our evaluation of these it= at tne final design stage of review.

The low pressure ecolant injection system ne of the indepenannt operating si.:as/stes of the residual heat removal system. -' uses ti true rwsiduai heat renuval Lys t x motor-driven ccatri fugal pumps e unvey water from Ln pression poul to (nu ec3ctor vessel through three separ nozzles. The residual heat re . 11 system pur.ps recti n power from alternatin urrent power buses having standby powor sa e bacKap sLpp*y.

Two residual he emoval systes pump motors and the associated automat. -o tor-ope ra b a valves rec ' e alternating current power from one bus, while tne low pressure ce sprav stem pump and the other residual neat removal system pero rotor anc valves ceive power from another bus.

ti.3.2 perfomance Evaluation In $cction 6.3 of the GE55AR-233 :1555 Safety Analysis Report, the General Elver.ric Company provided an analysts of the energency core cooling syst4uis astno tne assw:p-tions and calculational techniques described in 10 CFR Part 50, acction 50.4o lod Appendiz : to 10 CFR 30. The analysis was performed using the General Electric t.cou, <

evaluation models described in Topical Report f!EDE 20566 (Dra f t), " Loss-of w. ir t

( Accident aid Emergency Core Cooling Models for General Electric Soiling ilater J *4cters.

Suositted in August 1974, and the General Electric Company refill /roflood cati.e uw (supplement to the SAFE code description) transmitted to the itRC oy letter. J. L.

Gyorey to V. Stello, Jr., dated Decemoer 20, 1974 The General lectric Ccroany . as suomitted an additional report by letter front G. L. Gyorey to 'J. Stello, Jr., Jatua August 25, 1975. that discussos the way in wnfen the General Electric Comoany refluo.4 model is used in the analysis of t, oiling water reactors with in inroud low ;;ressura coolant injectiut system injoction, the background of our review of the General Electric Company ar'ernency core cooling systeet models is described in the ilRC staff Safety (valuation Report issi.ao 1.i .:o.uec-tion with tne order dated December 27, 1974 for operating Jot vp collino s. ate, reactors. The bases for acceptance of the principal corticns of 6ne General fle:tNc Coagany evaluation model are set forth in the ."RC staff's Status Recurt or Octci.er tid 4 and the $uoplement to the Status Report of flovemcer 1974, wnicn are ref arencas in tM Cacencer 27,1974 !afety Evaluatinn Resort. Tooother, tne Oect.acer 27,197a sar.,tj tvalua6 ton Report on operattng clants and tne Status Reoort :mi * *s lucoli. rent :escott the basis for our acceptance of tne emergency core cooling s> s ' . esolvat'uo ""-

tho General Electric Comoany's evaluation codel, in .:xoinati.,n .i:n tne ni.nt i e:'ft

..irameters, constitutes 'an accec:*tle emergency core coling sys.eci esalvatam ..4. ..

In confomance with Aopendia R . -) CFR Part 50 aad is apoli:acle to GES M 2 3 a o type nuclear power plants.

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Ouring the course of our review, we concluded that adof ttonal break sizes snosto te

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analyzed to suostantiate the break spectrum curve.

We also re'q uired tnat otr.er orsak 4 locations, i.e., steam line, feedwater line, and core spray Itnes be studied to suostan-y tiate that the limiting break location was the recirculation line. is part of tne g loss of-coolant accident analysts. additional SWR /6 single failure sensttivity analyses

. if were performed to evaluate the effects of a single failure that could cause any manually-controlled electrically-operated valve to move to a positica enat adversely affects the

,{ emergency core cooling system. The analysis showed that enese failures are less snere

$ than those considered for the emergency core cooling systa.a 2nalysis.

t

,j We also investigated the effects of flooding caused by a loss-of coolant accident. By letter from W. O. Gilbert to J. F. Stolz, dated August 11. 1975, the General Electric 4  !.

Company submitted the results of a study on emergency core cooling syste:: valves sithin the containment. The results show that all emergency core cooling system valve r.ctv s

,} which must be operable during and af ter a loss of-coolant accident are loc::ad outsloe i

the containment and will not become submerged due to the occurence of a loss of caolant j accident. Therefore, neither the short term requirement nor the long term cooling

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capability is affected by suomergerice effects.

J ,

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The staff was informed by the General Electric Company that certain calculational errors

.'." had been discovered during t9e General Electric Company reverification procram .nfen

[' may affect the performance nf the GE55AR 238 NSSS emergency core cooling system. k' i' have received General Electric Company letters (E. O. Fuller GE to 0. F. hss'. Jr..

NRC. dated January 19.1977t A. J. Levine to 0. 8. Yassallo. NRC. dated January 41. "l .

4, 1977t and A, J. Lavine to 0. r. Ross, Jr., dated February 7,1977) whien descrite ti.e I calculational changes made 'cr GE55AR 238 NSSS emergency core cooling systam perfor*.'nco I ,

evaluation along with revised analyses. The calculational cranges nere foetif teo .s.

1 (1) Correct the aster Jaye setpoint ti. SAFE for hfgn cressure core sorsy systani ini tiation.

(2) Modify the core power in the REFl.000 code.

4

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, (3) Correct the design basis accident break area.

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(4) Correct the guide tube thermal resistance, f i (5) Correct the initial core quality.

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(6) Correct the vaporization calculational constants in the REFLOOD coce.

4 h j,j Oy letters fram A. Levine. General Electric Comoany, to D. 45s, fr. , dacea 0,:::cer :

Jhj 1179 amf January 20. 1977, the General Electric Comoany tito eq.ested s :. e;s "' : -

REFLOOO code.' The staff has reviewed tnis request and finas it occaotaole. L e s s '= :

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I ev&luation for this REFLOCO ::c- modification is orovideo in a letter fr:.i C. 453, ;r '

MRC co M. Goller, NRC. dateu ~:oruary 18. 1977

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In addition, the staff concluded that the General Electric Company incorrectly modeled j '-- 1 counter-current flow-limiting effects on the emergency core cooling system water .

entering the core. Counter-currenc flow-limiting ef fects within the fuel channels i raises the prwssure differential between the upper and lower plena compared to tne l. am pressure differential that would be prudicted if only single pnase steam flow is assumed Q,;

in the channels. This added pressure differential from the two-onsse flow considerattan i f tends to reduce the flow of spray water between the by,iass reglun 2nd the lower plenum.

• O; The General Electric Company REFl.000 computer programs did not account for the c.o-pnasa pressure drop in the fuel bundle; consequently, they predict nonconservative times d:e 'I. .

core reflood. The staff has reviewed avaliable information on counter-current two-gnase *Y pressure drops in pipes for representative hydrsulic diameters. Based on this informa- Y" tion, we required that the General Electric Company include a one pound per square inen 6 correction in the calculation of driving head to predict bypass flow to the lower slenwn X until more exper.mntal tvidence is available to support an appropriate model, h

.j After incorporating all carrections, the General Electric Company suc:nitted a loss-of- I coolant accident reanalysis and rep wted the results in letters frcm 4. .evine, Gr.naral **

Electric Company to 0. Ross, Jr., N**.. dated February 14, 1977 ano I. Fuller. Genersl P.

G Electric Company to 0. '!assallo, NRC, dated Feoruary 17, 1977. l' The results of the Appendix K calculation for GESSAR-238 N$$$ (basec cn maximun N average planar linear heat generation rate of 12.27 kilowatts per f:ot) show a ; eau .

3-cladding temperature of 2038 degree. ranrenheit; a peak local oxidatton af less ta.:a .

two percent; and a maxtrum core average hydrogen generatien of less enan 0.14 oer:ent for the worst break assuming a failure Jf the low pressure ecolant Injection systd .(

diesel. The previous creak spectrur' suomitted in May 1975 for GE51AR 233 NS$$ clants :a '

was based on unity local peaking factars for all rods in the limiting bundle. For this previous analysts, the maximum "near heat generation rate selected was 13.2 d kilowatts per foot (this can ce interpreted as a maximum average planar heat genwestion  ;. l rate of 13.4 kilowatts per foot). Using a maximum everage planar linear neat generstien rate of 12.27 kilowatts per foot, based on conservative aaposure dependent local . - -

peaking factors, produces a break spectrum of the same general shape ut of lower magnitude than previously submitted, with the largest break si:e ylaliting tr.e higrest k.'

peak cladding temperature. ,88'"

we nave reviewed tne analysis of the emergency core cooling systam performanca suceit:M by the General Electric Comoany for GESSAR-238 NSSS and concluoe enat tne analysis as s.

performed wnolly in conformance witn tnw requirements of 10 CFR Part 33, Section 50.4o(3. [jl The GE55AR-235 N555 erergency core cooling system perfomance issares :enfomanca ita .," l (1) the peak cladJing temcerature Ifmit of 2200 degrees Fanrennet t. (2) tne 244i. .11 th\

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cladding oxidation limit 17 percent of total :ladding thickness t,ef0ra Juication, .1) ,,.- 4 ene maximum nyorogen generstidn core- lJa limit of one percent of tre total meta'. in @

"r the cladding, (2) the core geometry term cooling recuirement of maintair

, suing imenaole to 00oling.' 3no (3) t .P IGn; ,(.

. ac:eotaole core temcerat re: Ind Jacay can  :

removal.

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Analysis was not provided for emergency core cooling system performance *during resctor S operation with one recirculation loop out of service. Therefore, reactor operation '

I under such conditions will not be authorized until the cecessary anal ses

/ have tc!*t f performed, evaluated, and determined to be acceptable.

% Ouring our review of the analyses of BWR/5 and SWR /6 reactors, we expressed 3 cencarn

]g 9 y-r .. relating to recirculation flow control valve closure curing the oesign easts loss-:f-coolant accident. The results of a General Electric Ceccany tensitivity stuc/ ta 4

'd l t evaluate the effects of fast closure of a recirculation flow control valve ::inci; ant h f with the design basis loss of-coolant accident and worst postulated erergencj core

.Lhk,j cooling system failure were submitted in a letter dated April 25. 1975 from A..'+.i o to V. Stcllo of the NRC staff. The results of this sensitivity study snow Cnat t'e

,M c.f calculated peak cladding temperature remains below 2200 degrees Fanrenheit.

With regard to the emergency core cooling system reanalysis with the enanges nota-

p above, the staff did not require the sensitivity study related to recirculation / law 4

q control val e closure to be reanalyzed. The General Electric Comoany nas staceo - st.

j h) (1) the valvs n.ill fall "as is" with loss of of fsite power. This is, in cart. La o s

"""*'i spring-loadec valve design feature wnich locks the valve in the "as is' position in t.*e event of loss of hydraulic pressure for any reason; (2) The valve is s'esigneo to ,sose

• ,l only to a position wnich permits 30 percent flow; and (3) curing a loss of< cot ant

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i accident the high containment; pressure signal shifts t'io recirculatio' fl'w c c:ntrol q valve to the manual mode from the automatic mode. The Genar:4 Electric Co'rpany nss g also stated that all electrical systems for the recirculatton ' low control valve ::eracter.

g fr are outside the containment and thus are not sucject to Ene Ic's-of-coolant ic:f w g{ envi ronmen t. The previous sensitivity study of fast clcsure of the rectreulati:n 'b

, control valve (100 percent closure) coincident with design : asis loss of conlart .:-:caat  ;

• ir$' and a worst-case postulated emergency core cooling system f ailure snowed 4 eak : 344ta; '

temperature increase of less than 100 degrees Fanrenheit. This aesign feature is cstrg reviewed by the staff on the Wm. H. Zimmer Nuclear Power Statten. 'Jnit 1, Copet 'ia.

f 50-358 at the final design stage. We exoect tnat the resolution eacne: is 2 et .' t #

wt3 that review will be acolteable to GE55AR-238 N555.

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Even if a revised analyses with postulated recirculation ' low control valve :1:swrs l 1

.o yields a peak cladding temperature greater than 2200 degrees r a hrennett, :r i f in, o' M ,

p the other criteria of Section 50.a6 of 10 CFR Part 50 are exceeoed. a reauctice .e . r. , l it permissible value of the maximum average planar linear neac generation -sta r:n :-

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included in the technical specifications to mitigato costalited recir...liti:n - i

. j control valve ef fects.

J N. We conclude that the probacility of valve motion as a c:nsecuence of s i:ssef e a; tnt accident is small; however, any change reouired in the salie cesign or emercancj :-?

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cooling system analysis because of recirculation valve motion can e leciemeace2 U part of the final lesign stne of review.  ;

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m In summary, we concluoe that the emergency core cooling system for GESSAR 233 1555 meet

' all the criteria of Section 50.46 of 10 CFR Part 50 end the requirements of Appenot.t 4 to 10 CFR 50 and is acceptante. .

l 6.4 Vservice insoection of Class 2 and 3 Ccmoonents

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The ESSAA-238 NSSS design scopo includes cortain Class 2 Jnd 3 connponents. Ihe

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Gener Electric Company has stated in 'the application that tne GESSAR 238 NSSS uestd

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incorpo tes provisions for access for the inservice Inspection of all Class 2 anip3 ,

component XI ;f l.

the /cerica within Societythe design Engineers of Mechanical scope of GESSAR-238 Soiler and pressure Vessel NSSS Code.,/in acc:rcance with ,

addition, the neral Electric Company has specified in Section 1.10 of tnyGESSAR.236 NS$$ Safety Ana sis Report as an interface requirement that Joaquate a ess must be provided in the b lance-of-plant design to pemit inservice inspectic of OES!AR-233 NSSS Class 2 and 3 omponents n accordance'wt*h Section XI of cne Merican iociety sr

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Mechanical Engineers oiler an, pressure Vessel Code. We conc9 1 de enat this is 2ccro acte. We will review he dataals of the inservice inspection program for riass 2 ana ,

components in indivioua appli;4tions that reference GESSA (238 35*.S.

6.5 Main Steam Isolation valve eau ce Control System

( \

The main steam isolation valve leakage control stem is designed for the p..rpose 1f controlling an( reducing the r> . ase of fissi products to tne ens t ronment via ene main steam line following a postu ted los of coolant accident. The main staam isolation valve leakage controi sys c sists of two incepancent systems. Tne inboard system is ocwered frere sie el trical division and tr.a c.,ttoaro s/stes freat tee otner division of the balance-<.,: pl ne rqency power supplf . The Outboars systerti is connected to all steam lines be n the utboard main steam isolation valve ano CN downstream balance-of-plant mam steam blo valve. All steam lines are ::nnectac 9 parallel to the outboard bl d neader. Suc on is drawn on ene space cetween ene o tcr main steam isolation valv and the Mock valv by blowers ano .tnausted te 5 out latn],

volume served by the D ance.cf. plant standby 3 s treatment sf stas. The incoari $/s s ,

is connected between he main steam isolation vai St. Indivijual bleed tir.es and associated valvin re provided for each steam lin the a.tnaast is 31rectea ta a comon header a then discharged by a blower to a b loin; tol.ue served oy tne standby gas t atment systeve. .

The oper ton of the system is limited by a series of pres ure senso.s and tiirers nt:'

s serve interlocks designed to preclude system actuJtion ;r{3r to the pressare 'n na main team lines decaying to the pressure for wnich the main 1,tsam isolatten .41,4 1 age control systas is oesigned to coerate. The interloca1 41 5o ;, recluse ::ntin24*

oeration of any portior> of the main steam isolation valve taaxa :ontrol syscam ..r :.

falls to achieve a suoatmosoneet; ::nditinn in its respectrie steag line $f ter i :rsiet time, in addition, an interlocs i t provided to prevent operit t y'. o. 2n Inotis;6..

Inocard main steam isolation v,. laamage c:ntrol system. .r. lass t eb:rret;,c';;tn; -

main steam isolation valve ins- ine centatnn:ent is falt . :so. Thhmainste 13

3-1

, CONTENTION 3 The NRC Staff does not move for Summary Disposition on Contention 3, but rather moves that the Board either dismiss Contention 3 or combine and/or incorporate it into Contantion 16, item 3, since the pool swell l l

in Contention 16(3) is basically duplicative of the pool swell men-tioned in Contention 3.

This conclusion is strengthened by the fact that Intervenors' witness, Mr. Dale Bridenbaugh, stated in the enclosed Intervenors' Exhibit A-6 at C16-4, of their November 30, 1976 Answer to Staff interreaatories that although pool swell is listed as an item of concern in Contention 3, it was repeated in Contention 16 only for " completeness." l Therefore, since Intervenors' witness has stated that the pool swell )

1 issue as discussed in Contention 3 is the same as the pool swell l restated in Contention 16, the Staff requests that for purpose of simplifying the procedure of the hearing, Contention 3 be eliminated as repetitious of the pool swell issue in Contention 16.

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3-2 STATEMENT OF CONTENTIONS

-3. Intervenors contend that the Applicant has not adequately demonstrated that the structures and canponents within the suppression pool have been designed to withstand the hydrodynamic forces of a high vertical water swell which results from the postulated Design Basis Accident for Black Fox 1 and 2.

16. Intervenors contend that the Applicant has not established the integrity of the Mark III containment in that the following items have not yet been resolved:

1. Vent Clearing

2. Vent / Coolant Interaction k 3. Pool Swell

4. Pool Stratification

5. Pressure Loads and Flow Bypass l

INTERV BORS' MOVEMBER 30, 1976 ANSWERS TO STAFF INTERROGATORIES FordR SERV. ICE CO. OF OKI).HOMA

' . . (2501 ,

BLACK FOX' NUCLEAR' STATION Concencien 16

1. Intervenors ~ concend "that the Applicant has not established.

che integricy of the Mark III containment in that the followdng l items have not yet been resolved'i .

(1) Vent Clea=ing .

(2) Venc/ Coolant Interaccion .

(3) Pool Swell (4) Fool Stracificacion 1

(5). Pressu=a Loads and Flow Bypass " l

2. The NRC Interrogatory requested that no additional informa-( cien would be required other chan ,co 1(a) scace '

x cha na=es, addresses and professional qualificacions of the i

persons.upon whose views Ince:venors rely to substantiate each contencion; '(b) author's cicle, date of books , ca:c:s , papers ,

etc. relied upon to substanciac'a views; (c) describe all indspe donc calculacions, physical and mathematical nodels ; (d) sec j l

forth specific faces, each person na=ed in 1(a) relies upon ,

l co subscanciace views.'

3. I, Dale 3ridenbaugh, am a graduate engineer choroughly f 2~d liar with the design, construction and operacion of nuclear generaci plancs. I received ny 3.S. in nachanical engineering f c::i che South Dakota School of Mines and Technology in 1953. For che past cwency-c ee years until February 19 75 I worked as an engine

_st A Z - cte-L -

with the Ganoral Electric Company on a wide variety of most '  ;

i aspects of power generation-equipment design,"manufactura and operations . During the last ten of those twenty-two years 11 was is management positions in the area of the monitoring, of I

performanca of nuclear power plants all over the world and is.

the development and implementation of solutions to operational ,

l problems of those plants. During the ten. months prior to my

• l 1

resig=ation in Februa=y I also was serving as c:anager of Ganaral \

Electric's Maric IQntainment Assessment program, an assessmant of the safety and design of twenty-five nuclear plants designed ,

and constructed by tha General Electric Company is the Unicad States. This position required the management,of Utility Liason and of Uuclear Energy Division Resources to assass the adequacy of understanding of containment accidant.dynadic

\- phenomena and of the containment design adequacy for continued  ;

1 operation "of the planc.

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For the pa'st several months I ha e been a partner and consultant on energy and environmental matters with ,MH3 Technical Associates. In this capacity I have been retained by

'the Incarvenors to assist in the Black Fo:c hearing process.

Additional details concerning my education and qualifications ,

are included in Attachment 1.

4. I have,reviewad the information submitted with the Applicant's PSAR concerning the phenomena listed in the five subparagraphs of the Contention. Ihere appears to be no substantive coc=ent

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- C16-2

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on 'any of chsso its.ms. in the Applicanc's submitted docu-s

- mancacion and.the rafarancas contained in Volume IV, pages  :

'6.2-138 through '-140 do not appear cc -report. any~or the results obtained from the General Electric Mark III casc facility in San Jose, Califcznia, nor describe the extensiva analyses of such data particularly wich.respace to info::=ation obtained during. the lactar part of 1974 and in 1975.

It is assantial thac che phenomena liscad a's well as ,

others identified during the casting program be choroughly casted and underscoed ask taken 'into consideration in the design of cha Mark III containment as . subs.cancial forces and affects can be experienced during the Design 3 asis Accident. This could lead to failure of the contahment syscan due to design defects that might exisc. Briefly su-mm-ized, I

(

che following is of concarn: . #

(1) Vent cleaitta:: Vane clearing rafars to the phanc=ana of lowering the water laval in the annulus bat eaan the weir wall and the suppression pool wall to uncover the .

vents bacween the dry well and the suppression pool.

Ihis can impace on the function of the pressura supp-rassion containmane and can i=part loads cc the contain-sent structura dua cc (1) che afface of the cina it takes for the vant to clear, (2) the associacad pressura buildup in the dry wall, and (3) the pressura pulsa that is inparted ince the suppression pool water, delivering structural leading forces en the suppression pool material.

- C16 - . .,

(2) Vent /Coolanc Incorrecion. As the ecoam bchts s released from the primary syscam into che dry well is forced by chit dry wall 'pressura cli:fough -

t.ca vant and into . the suppression pool vacar, a pulsating or vibracory steam condansing incarface is set up. ;This. davalops a pressura .

pulse in the suppression pool water. This same pulse can cause a significant loading -

imparted to th suppression pool structural componanes. ,

(3) Pool Swell. ' The pool swell phenomena is discussed in Contantion 3. For purposes of completanass, it will be restated here as the leading phenomena resuldng from the rapid. for~cing of the dry wall non-condansibla

( at=csphere through the suppression pool vents,  :

1 davaloping a large bubble undar the surface of the supprassion pool, and accelerating significant slugs

- of vacar rapidly upkard in the supprassion pool spaca. l These rapid moving slugs of watar can i= pace on i 1

1

, structuras and ce=ponents in and above che suppression  !

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pool area. This phenomena occurs in appro:cimacaly 1.5 1

,saconds and, therafora, has the potential for davaloping {

- destructive forcas in loads on such cocponents.

(4) Pool Stratification. This phenocena has to do with the development of local thermal s::acificacion throughout che suppression pool as a resui: of (1) the discharge of staam in:o the suppression pool _.

- C16 _ _

, dischargo of steam for extsnsivs period of tima dua

/ co excanded blowdowns of stasm through reliaf valva discharga lines. I'c'is pess'ible that 'such loi:al

-l heating : could causa boiling and unacceptable loading  ;

l en the suppression pool and contain= ant structura.  :

.(5) Pressura loads and Flow Eveass. Two separata pheno-mana a a groupad under this category. They have to do with concerns regarding (1) pulsacing or pressu're loads that can be davaloped in. the suppression p6cl and againse che cone =4nmant sc=ucturas as a rasule .

of a continued blowdown of relief valvas through their connaccad discharga lines , and (2). of the possibility of bypassing flow around the suppression pool as a result of disturbanca of the suppression pool surface, sicher through local bo114"g, -

asymmatrical wava ganaration, sais=ic slosh, or other phenomena' that would affect the coverage of the vents with water during the DBA.

(6) Davalcemanc of Leading Critaria. Cuca the aforeman-cioned phenomena have been casted extensively, a significant ' engineering affort is required to c:ans-

. slaca such phenomena data into usabia loading valuas ,

1 for use in the detail design of the struccuras and .

components. Description of this stap in the design evolucion and verification does not appear to be described in che Applicancs ?SAA or' raferences. Such documentacien should be prepared and choroughly reviewed by the NRC S taff and t{.e ACRS , and the

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l results ofesuch reviews nada availabla :o in: ares:ed In tervanors . - d 6 l_ , _ _

All of the foregoing, and a' number of other hydrodynamic phenomena have baan and are being inves-tigated by the Applicant's vendor. It appears that an incomplaca presentation of the caseing program, tase results, and necessary design basis loading .

conditions has baan submittad ce show chac the Maric III ccacainment is adequate to daca. Complaca informacion should be submitted so that adequata raview can be mada.

5. Attached as partisane information to this subject is a copy of the Tastimony of Bridanbaugh, Ecbbard and .

Minor before the Joint Committaa on Atomic Energy ,

- in Washington', D.C. on Feb=uary 18th, f976. This

.i Tancimony contains information parrinanc' ce che contanciens made on concainmane caseing and adequacy.

Also rafaranced as partinent infor=acion is the lattar from N. Karr, ACRS Chairman to Honorable Di; cia.

Las Ray, AEC Chai: man, Raport on the Rive: Band S tation, Unies L and 2, Jan. 14,1975. This letter summarizes plans and status of casting of the Mark III

- Containment design and associacad phenocana.

- CT6 1 o

. 5-1, Contention 5 I. The Contention Intervenors contend that the applicant has not adequately demonstrated that the reactor pressure vessel supports and pedestal for Black Fox 1 and 2 can withstand the loads resulting from the design basis require-ment of 10 C.F.R.. , Part 50, Appendix A, Criterion 2 relating to earthquakes.

II. Argument Criterion 2 of 10 C.F.R. , Part 50, Appendix A provides:

Criterion 2 Design bases for pmtection against naturr.1 phenomena. Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perfonn their safety functions. The design bases for these structures, systems, and conponents shall reflect: (1) appropriate consideration of ,,

the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of nonnal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety function's to be perfonned.

The acceleration value for the safe shutdown earthquake for Black Fox is.32g. This matter was the subject of extensive testimony at the

- environmental and site suitability hearings. (See discussion starting-at page 18 of the "NRC Staff Proposed Findings of Fact and Conclusions of Law on Environmental and Site Suitability Matters" filed on January

.5-2 .

9, 1978.) The enclosed Staff affidavit of Mr. H. E. Polk describes the procedures and criteria set forth in the PSAR which have been reviewed by the Staff and have been found to be in compliance with Criterion 2 of Appendix A to 10 C.F.R. , Part 50 in that stresses resulting from loads considered will not exceed values allowed under applicable codes. The Staff review found that the Applicant would follow the criteria listed below in designing the reactor prinssure vessel supports:

A. A free field ground acceleration value of .129. Polk Aff. at' . 3.

B. The allowable stresses with respect to the pedestal are

' specified in the American Institute of Steel Construction Specifications for the steel portions and in.the American Concrete Instutute Standard ,

x 318-71 for the concrete portions. Polk Affidavit at .u. 4. . i C. The reactor pressure vessel (RPV) skirt is classified, j designed, and constructed according to the rules of Section III of the j American Society of Mechanical Engineers Boiler and Pressure Vessel Code. Balance of Plant component supports which attach to the RPV support skirt will be designed in accordance with Section III, Subsection NF of the ASME Code. Polk Affidavit at ". 4 Thus, the procedures and criteria to which the Applicant has committed assure that the RPV supports and ~ pedestal will withstand the loads resulting from the design basis requirement of 10 C.F.R. , Part 50, Appendix A, Criterion 2, relating to earthquakes. Polk Affidavit at .4

.,-e. . . , , . , - - , _. - - -

5-3 For this ritason, Intervenors' Contention Number 5 should be dismissed I as lacking any matarial facts to litigate in the safety hearing.

III. Statement of Material Facts As To Which There Are No Genuine Issues A. The acceleration value for the safe shutdown earthquake for Black Fox, (Jnit 1 and 2 is .12g.

B. The fact that the BFS pressure vessel support will be designed in accordance with the allowable stresses set forth in American Concrete Institute Standard 318-71, the American Institute of Steel Construction Specificatic~ and ASME Code Section III, Subsection NF, guarantees that the pubin alth and safety, as well as Criterion 2 of l Appendix A to Part 50, will be protected.

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. l UNITED STATES OF AMERICA l NUCLEAR REGULATORY COMMISSION l BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

)

PUBLIC SERVICE COMPANY OF OKLAHOMA )

ASSOCIATED ELECTRIC COOPERATIVE,

• Docket Nos. STN 50-556 INC. AND WESTERN FARMERS ELECTRIC STN 50-557 )

COOPERATIVE

- )

(Black Fox, Units 1 and 2) )

AFFIDAVITS OF H. E. POLK i AND 1 J. M. KOVACS CONCERNING CONTENTION 5 i

liy name is H. E. Polk. I am employed by the Nuclear Regulatory l Comission as a Structural Engineer in the Structural Engineering Branch.

)

I have been employed in this position since November, 1974. My qualifi-cations are listed on the attached qualification sheet.

The purpose of my affidavit is to address intervenor contention 5 regarding the adequacy of the reactor pressure vessel pedestal to withstand the loads resulting from the design basis requirements of 10 CFR Part 50, Appendix A, Criterion 2 relating to earthquake. .

l

My name is J. M. Kovacs. I am employed by the Nuclear Regulatory Commission as a Mechanical Engineer in the Mechanical Engineering Branch. I have been employed in this position since July, 1974.

____g. qtialifications are listed on the attached qualification sheet.

The purpose of my affidavit is to address intervenor contention 5 regarding the adequacy of the reactor pressure vessel supports skirt to withstand the loads resulting from the design basis requirements of 10 CFR Part 50, Appendix A, Criterion 2 relating to earthquake.

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Intervenors contend that the applicant has not adequately demonstrated that the reactor pressure vessel supports and pedestal for Black Fox 1

& 2 can withstand the loads resulting from the design basis requirement of 10 CFR part 50, Appendix A, CMteHon 2 relating to earthquakes.

CM teMon 2 requires: I (1) AppropMate consideration of the most severe of i the natural phenomena that have been histoM eally ,

reported for the site ano surrounding area, with l sufficient margin for the limited accuracy quantity -

and peMod of time in which the histoHcal data -

have been accumulated, 4 (2) appropMate canibinations of the effects of nomal and accident conditions with the effects of the natural phenomena.

Crite on 2 s part (1) has been meet with regard to earthquakes. The i applicant for the Black Fox Nuclear plant has referenced the General i ElectMc GESSAR 238 Standard Steam Supply System (Docket STN 50-550) which shows that the reactor pressure vessel support skirt, a standard l plant component, is designed to resist the effects of an earthquake

' up to 0.30g ground acceleration. The pedestal, which supports the reactor pressure vessel skirt is designed to 0.12g. The design

( basis for the Black Fnx station is 0.12. In regard to cHterion 2 ,

subpart (2) this itystem (skirt.and pedestalTwill Be desicned to withstand the effects of earthquake loads in combination with otner loads as specified in our Standard Review Plan 3.8.3. These combinations include j dead loads, live loads, thermal effects, and pipe reactions duMng nomal  !

operating or shutdown conditions; earthquake loads; pressure, thermal i and pipe reactions due to postulated pipe breaks (including asymmetric  ;

loads); and jet impingement and missile impact loads. l t

The earthquakt loads are computed for the interface between the pedestal 3 and the reactor vessel support skirt. These loads are usually presented in the fem of a response spectra. The procedure necessary to produce this response spectra is outlined in the next paragraph.

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' The reactor _ building is modeled as discrete masses connected by spMngs  !

which are computed from the physical pr.operties of the structure. In- 1 cluded in this model is the NSSS (which includes the reactor pressure i vessel) which is also modeled as masses and springs. The entire model  !

is supported by a set of spMngs and dampers which represent the foundation l mateMal. The spMng and damper values are computed using the physical properties of the foundation media. The model is subjected to an earth-quake time history whose response spectra envelopes the required ground response spectra as specified in Regulatory Guide 1.60, and is referenced to the site ground acceleration value of 0.12g. This required ground .

response spectra was produced using a number of recorded earthquakes and

~

represents a conservative ground motion. The time history of each mass is computed using the model and earthquake time history. The time history H is used to generate the response spectra for each mass coint which is in turn used to design the vaMous-components which are attached to that mass l Saint. PMor to _use, the response spectra is broadened 10% at the ceaks j

1 The applicant has committed to the use of the above procedure to design the reactor' pressure vesse'l supports and the pedestal. This procedure ,

has been used in many nuclear power plants in the past as well as in l many commerical structures. Use of this procedure satisfys requirements

.of driterion 2 with regard to earthquakes.

The pedestal is designed as a seismic Category I structa asing the i acceleration values from the response spectra in combination with other loads described above to produce the total load environment. The allow- l able stresses are specified in the American Institute of Steel Construction Specifications for the steel protions of the pedestal and in American Concrete Institute Standard 318-71 for the concrete portions. These stresses are conservatively limited to ensure that the pedestal will withstand the imposed 1c.sts.

The reactor pressure vessel (RPV) skirt is classified, designed, and constructed according to the rules of Section II of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. The BFS NSSS vendor has committed in the GESSAR 238 SAR to consider and evaluate all loads due to plant operating conditions, anticipated transients and design basis events including SSE in the appropriate combinations In addition,for the l all NSSS systems, components, and component supports. i structural interface requirements for the RPV support skirt are clearly i delineated in Section 1.10 of *.he GESSAR-238 SAR. The applicant has stated his comitment in paragraph 3.8.3.4 of the BFS PSAR to design balance of plant component supports which attach to the RPV support skirt according to the rules of Section III, Subsection NF of the ASME Code as stipulated in GESSAR-238, figure 1.10.1-1.

The comitments contained in the BFS PSAR and.the GESSAR-238 NSSS SAR jointly constitute an acceptable basis'for satisfying criterion 2 of the general design criteria.

I have read the foregoing affidavit and swear that it is true and' ,

j accurate,to the best of my knowledge,

/(MN ,

'H .~ E . Po l k l

l Subscribed and sworn to before me i l this f p. day of su/9 1978. [AM#0//<)

sa 9h. Ne b N6 qry Public // "

grarar of /7'M.w iv? * \

Quw:w of O!*v?ChhA'y My Commission Expires: 7-/-f' l

I have read the foregoing affidavit and swear that it is true and accu: ate to the best of my knowledge.

d/ W

. ft. Kdvacs Subscribed this f1 and sworn to befo(re meday of /a /y ,1978. v/h.ko um kN Notary Public ,

Sr,4ITE OF /77,GRy4nrv0

/7?ca 7 Gosna***f C~rtf My Comission Expires:,7-/-/V  !

- e p P' l

_--_______m_ _ _ _ _ _ _ _ _ _ _ _

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HAROLD E. POLX PROFESSIONAL QUALIFICATIONS STRUCTURAL E1GINEERINGSRANQ DIVISION OF SYSTEMS SAFETY

1. FFICE OF MUrtEAR REACTOR REGULATION I as a Stmetural Engineer in the Structural EngineeMng Branch, l l

Division of Systems Safety, Office of Nuclear Reactor Regulation, U.S. Nuclear Regualtory Commission, Washington, D.C... I am )

responsible for reviewing safety analysis reports with rejard to structures and seismic analysis for nuclear power plants assigned to ..

I joined the Division of Systems Safety in November 1974. I have i served as Structural Reviewer for the safety reviews of Hartsville Nuclear Power Station Black Fox Station, Arkansas Nuclear One Unit 2. Yellow Creek Nuclear Plant and New England Power 1 & 2

^

profer,t.'

- I have a Rachelor of Civil Engineering (1958) and graduate study in Structural Engineering. My 18 years expeMence includes 2 years of aircraft stress analysis with the Martin Co., 8 years of struc-tural analysis and flight performance with the Boeing Co. on the i

Minutaman Missile program and the Apollo project which landed the fint men on the moon. I joined the NRC staff after completing over 4 years of seismic dynamic analysis of nuclear power plants with

. the Bechtel Power Corp. of which the last 2 years was a Supervisor of a seismic analysis group.

I am currently a mader of the American Nuclear Society, ANS2.2/2.10 Working Group on Seismic Instrumentatic1.

l 1

y PROFESSIONAL QUALIFICATIONS JOHN M. KOVACS

. - -- -- U. S. NUCLZtJL PILMmKY CONMISSION MECHANICAL ENGINEERING BRANCE DIVISION OF IZCHNICAL REVIZW m

. . I as a Mechanical Engineer, my ped m responsibilities being to review and evaluate the design criteria for mechma 4 =1 components, the dynamic analyses and testing of saf ety related systems and components and the criteria for protection against the dynamic effects assocuted with postulated failures of fluid systems for nuclear facilities.

e

- I was graduated from the University of Southern California in 1953 with a e

B.S. degree in Civil Engineering and in 1967 was awarded a M.S. degree in Applied Mechanics from the California State University at Sacramento.

From 1953 to 1955 I served as an enlisted man in the Army of the United States.

I was employed by the Northrop Corp., Hawthorne, California ,from 1955 to 1956 as a Structures Engineer in the Basic Loads ar.a Design Criteria Group, my primary responsibilities being to datermine airloads and inertia forces associated with airplane and rocket vehicle dynamics. 1 From 1956 to 1958 I was employed by the Zenith PJ stic Co. , Cardena, California as a Structures Engineer engaged in el analysis, design, development and testing of glass reinforced plastic constructions,

)

principally, radomes for military and civilian aircraf C. 1 l

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. .: - :: :. :.:. . . - .: = : : -

2_

~ .. . .

I was emplo red by the Aerojet General Corp.', Sacramento, California i

from 1958 f.9 1968 as a Group Supervisor in Polaris TEM Propulsion Systen l 1

I Development Program. During this period I was responsible fvr structural analysis, design and testing of 113 *.-weight 5 metal and glass filament wound pressure vessels, superheated structural and mechanical components such as rocket nozzles and thrust vector sontrol devices. Additional responsibilities I included development of analytical methods such as application of finite

- - element techniques to the solution of practical engineering analysis and design problems.. l 1

From 1968 to 1969 I was employed as a Senior Research Speelalist by the l

Boeing Company, Seattle, Washington. While at Boeing, I was responsible for development of new graphite fiber technology and its' application to practical aircraft structural systems and components.

From 1969 to 1974 I was employed by the Aerojet General Corp., Sacramento, ]

California as an Engineering Specialist. During this period my responsibilities included structural analysis, design and testing of nuclear vessels, piping, pumps and valves for the NERVA Nuclear Rocket. Additional responsibilities included analysis and design of hull structures, propulsion machinery, flexible seals and steering devices for naval surface effects ships and air cushion vehicles.

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. 6-1 .

CONTENTION 6 I. In Contention 6, Intervenors maintained that the Applicant did not show compliance with Criterion 2 of Appendix A to 10 C.F.R. Part 50 with respect to tornadoes. Criterion 2 provides that important safety structures, systems and components "shall be designed to withstand the effects of natural phenomena such as

. . . earthquakes, tornadoes . . .." It further provides that the design bases for the plant must reflect the most severe natural phenomena reported in the area (including tornadoes) and combine these loads with normal and accide . condition loads.

Ik. ARGUMENT i

The deposition of Intervenors' witness, Mr. Greg Minor, and the contention itself show that Intervenors' Contention 6 on tornado

~

loads can be. reduced to the following essential elements:

1. ability of the containment to withstand postulated loads and missiles including tornado loads,(Contention 6, Parts a and b);

2. the ability of the new and spent fuel pools to withstand l l

postulated loads, Contention 6 (c) and (d).

As will be discussed below, the enclosed NRC Staff affidavit by Mr. Polk shows that there are no disputed facts left to litigate in the area of tornado loads.

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6-2~-

Containment The Polk Affidavit indicates that the reinforced concrete shield bufiding, which surrounds and protects the steel containment, is designed to with-stand the effects of the design tornado, cyclonic winds, missiles and associated pressure drop. Polk Affidavit at l'. The Polk affidavit indicates that' the BFS steel containment structure is designed to withstand l an internal pressure total load of 15 psig, more than five times the 3 psig pressure drop postulated for the design tornado pressure transient. l 4

For this reason, the Staff believes that pressure drops from tornadoes do not present a litigable safety issue for Black Fox.

i

\ New and Spent Fuel The Polk affidavit also indicates that new fuel will be stored in the Seismic Category I new fuel pool which is located in the Seismic Category I fuel building which is designed for a design basis tornado of 360 MPH and associated missiles which include a 4,000 pound automobile traveling at 100 MPH. In addition, since the pool structure itself and the fuel building are seismic Category I qualified, Mr. Polk concludes that no safety problems due to' tornadoes or missiles are possible in the fuel building or new fuel pool. Likewise, the spent fuel pool, which is also located in the Category I fuel building, is also protected aaainst maximum tornado pressure drops and possible missile penetrations.

3

6-3 The deposition of Mr. Greg Minor, I'1tervenors' proposed witness on spent fuel pools, indicated that while he was concermed about the possibility of high density fuel pool configurations (no high density fuel rack configuration has been requested by Applicant),

he (1) had not reviewed the Black Fox pool design, (2) was not sure of. the tornado load in the Tulsa area, and (3) had no facts within his possession that would show that fuel rack spacing or design would present a hazard to the Black Fox Station. Tr. 122.

While Intervenors' witness did indicate that he had questions related to Itans 1 and 4 of Criterion 61'of Appendix A to Part 50

(

• (Tr. 121-124) , that assertion is outside the' scope of the conten-tion as it was admitted. in this proceeding.

For the reasons stated above, the NRC Staff requests that the Contention 6 issue concerning the effect of tornadoes and missiles on containnent and on fuel pools be dismissed.

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~6-4' .

III. Statement of Material Facts About Which There Are No Genuine Issues

1. The BFS steel containment structure is protected by the concrete shield building which is designed to withstand the design basis tornado, cyclonic winds, missiles and associated pressure drop. The steel containment is designed to withstand internal pressure which is five times the maximum tarando pressure transient.(15 psig). l

2. Seismic Category I new and spent fuel pools, housed wi hin the Seismic Category I fuel building, are designed to withstand the combined effects of natural phener,?na such as earthquakes and tornadoes.

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UNITED' STATES OF AMERICA .

NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

PUBLIC' SERVICE COMDANY OF OKLAHOMA ASSOCIATED ELECTRIC COOPERATIVE, Docket Nos. STN 50-556 INC. AND WESTERN FARMERS ELECTRIC STN 50-557 COOPERATIVE.

(Black Fox, Units 1 and 2) )

AFFIDAVITS OF H.E. POLK ,

CONCERNING CONTENTION 6' 4

My name is H. E. Polk. I am employed by the N'.! clear Regulatory Commission as a Structural Engineer in the Structural Engineering Branch.

I have been employed in this position since November,1974. My qualifi-cations are listed on the attached qualification sheet.

I The purpose of my affidavit is to address intervenor contention 6 l regarding the adequacy of the Containment and the Fuel Storage facility to withstand the loads resulting from the design basis requirements of 10 CFR Part 50, Appendix3 A Criterion 2 relating j to tornadoe s.

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CONTENTION 6 .,

The intervenors contend that the applicant has not adequately demonstrated comoliance with 10 CFR Part 50, Appendix A. Criterion 2 for Plack Fox 1 and 2 with sc4pect to tornadic phenomena related to :

a. Missile penetration of the containment

b. Rapid exterior atmospheric pressure. transients or excursions on the containment.

c. Protection of new fuel

d. Protection of spent-fuel storage facilities.

The NRC staff will show that Criterion 2 has been met with regard to the tornado phenomena.

A. Missile Penetration of the Containment The pressure boundary of the G.E. Mark III containment system is the steel shell that is located inside the seismic Category I shield building. The con-crete shield building protects this steel shell from the external environmental loads such as the postulated tornado missile loads. The applicant has committed to comply with the wall thicknesses required by the staff for the shield building to stop the tornado missiles. Compliance with this requirement ensures that the containment structure will not be breached by the postulated

(' tornado missiles. The staff's requirements are basedron two major'f011 schle tests of reinforced concrete slabs. One test was conducted by the Calspan Corporation in 1974, sponsored by the Bechtel Corporation. The other test was conducted at Sandia Laboratories in 1976, sponsored by the Electric Power Research Institute. The results of these tests are available in the literature. The NRC staff used the results of these tests to establ'3h minimum thicknesses necessary to prevent missile penetration or other ,

daraaging effects.

B. Atmospheric Pressure Transients The reinforced concrete shield building, which surrounds and protects the steel containment is designed to withstand the effects of the design tornados (incompliancewithR.G.1.76)rapidatmosphericpressuredropof3 psi l in 2 seconds and subsequent repressurization of 3. psi in 2 seconds. This '

l load is combined with the tornado cyclic wind pressure and the tornado missile impact to detennine the full load environment. The shield building is then designed in accordance with the ACI Standard 316-/1. The steel containment is designed to withstand an internal pressure of 15 psig which is more severe than the tornado atmospheric pressure transient of 3 psig.

C. Protection of New Fuel The fuel is received by the utility and stored in the new fuel pool which is located in the fuel building. The building is a seismic Category I structure and is designed for the design tornado and associated missiles

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which includes a 4000 lb passenger car' traveling at 100 feet per second and 743 lb 12" diameter schedule 40 pipe traveling at 210 feet per second.

.The building is constructed of reinforced concrete walls and roof which conform to the staff thickness requirements as delineated in Section A of this resportse. Therefore, the new fuel will be protected from the tornadic phenomena and will satisfy the requirements of 10 CFR Part 50, Appendix A, Criterion 2.

D. Protection of'5 pent-Fub1'Starage" Facilities The spent fuel storage facility is housed in the fuel building. The facility consists of the reinforced concrete spent fuel pool and its associated heat removal equipment. The pool and equipment are protected from the tornadic phenomena by virture of being inside of the fuel building. The building is described in Section C of this response.

Therefore, the spent-fuel storage facility requirements of 10 CFR 50 Appendix A Criterion 2.

I have read the forego.ing affidavit and swear that it is true and accurate to the best of my knowledge.

, c>b H. E. Polk Subscribed and sworn to before me this / / day of , 1978.

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4 HAROLD E. POLK PROFESSIONAL QUALIFICATIONS STRUCTURAL ENGINEERING BRANCH OIV15 ION OF SYSTEMS SAFETY OFFICE OF NUCLEAR REACTOR REGULATION I am.a Structural Engineer in the Structural Engineering Branch, Division of Systems Safety, Office of Nuclear Reactor Regulation, U.S. Nuclear Regua1 tory Commission, Washington, D.C.. I am responsible for reviewing safety analysis reports with regard to structures and seismic analysis for nuclear power plants assigned to me.

I joined the Division of Systems Safety in November 1974. I have served as Structural Reviewer for the safety reviews of Hartsville Nuclear Power Station, Black Fox Station, Arkansas Nuclear One Unit 2, Yellow Creek Nuclear Plant and New England Power 1 & 2 project.

I have a Bachelor of Civil Engineering (1958) and graduate study in Structural Engineering. My 18 years experience includes 2 years of aircraft stress analysis with ti!e. Martin Co., 8 years of struc-tural analysis and flight performance with the Boeing Co. on the ,

Minuteman Missile program and the Apollo project which landed the first men on the moon. I joined the NRC staff after completing over l 4 years of seismic dynamic analysis of nuclear power plants with the Bechtel Power Corp. of which the last 2 years was a Supervisor  ;

of a seismic analysis group. f i

I am currently a member of the American Nuclear Society, ANS2.2/2.10 Working Group on Seismic Instrumentation.

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Contention 7 I. Contention Intervenors contand that in order for the Applicant to meet 10 C.F.R. Part 50, Appendix A, Criteria 3, Black Fox 1 and 2 must utilize cables with fire retardant insulation.

II. Arcument Criterion 3 of Appendix A to 10 C.F.R. Part 50 provides:

Criterion 3 -- Fire protection. Structures, systems and .

components important to safety shall be designed and located to minimize consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and heat resistant materials shall be used wherever practical

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throughout the unit, particularly in locations such as the

. containment and control' room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and componsnts important to safety. Firefighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.

As shown by the following enclosed Affidavit of Robert J. Giardina and James D. Bahn, the Applicant has committed to use fire retardant cable insulation which exeeds the minimum requirements of Regulatory Guide 1.120 and Branch Technical Position ASB 9.5-1, Appendix A. Thus, all cable used in Black Fox will meet or exceed IEEE Standard 383 for flame type tests. Giardina and Behn at 2.

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l Since the Black Fox Plant will utilize electrical cable insulation which passes the IEEE flama resistance criteria, the NRC Staff believes

'liat this contention should be dismissed.

i III. Statement of Material Fact About Which No Issues Remain To Be Litioated.

1. Applicant has comitted to the use of electrical cables at BFS with fire retardant insulation which complies with. current IEEE Standard 1 383 (IEEE Standard for Type Test of Class IE Electric Cables, Field and Connections for Nuclear Generating Stations).

( Z. The use of materials complying with IEEE standards will ensure that criterion 3 of Appendix A to 10 C.F.R. , Part 50 (use of non-combustible and heat resistant matarials wherever practical throughout the reactor)

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will operate to minimize the probability and effect of fires at BFS.

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i UNITED ~ STATES OF AMERICA . l J

NUCLEAR REGULATORY COM4ISSION -

BEFORE THE ATOMIC SAFETY AND LICENSING BOARQ_ ,

In the Matter of p ,

i Pubite Servica Company of I Docket Nos.: 50-556 Oklahoma j 50-557 l (Black Fox Station. )

Unit Mos.1 & 2 ) .

. ... 1 AFFIDAVIT OF ROSERT J. GIARDINA AND JAMES D. SEHN ON CONTENTION 7

1. My nama is Robert J. Giardina. I am employed by the Nuclear l Regulatory Comission as a Reacto- Engineer in the Auxiliary Systems Branch. 'I have been employed in this position sinca l

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September,1974. M'y qua:Lifications ara 3istad orr the attached Qualiffcations Sheet.

2. My name is James D. Bahn. I am employed b'y Gage-Babcock &

Associates, Inc. , as a fire protection and safety consultant.

I have been employed in this field for the past eight years.

I am presently providing fire protaction consulting to the ,

Auxiliary Systems Branch of the NRC. My qualifications are ifstad on the attached Qua.itfications Sheat.

3. The purpose of our. affidavit is to address the intarvenor Contantion Number 7 regarding the fire protection l

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program for cables, cable tray separation, and for the plant, respective-These contentions relate. to the compliance of the Black Fox 19 Fire Protection Program with the requirements in Appendix A CMterion 3 to 10 CFR Part 50. Our affidavit will show that the Black Fox l Fire Protection Program is acceptable and will meet the requirements l

of 10, CFR Part 50, Criterion 3.

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4. Contention Number 7 ,

The Intervenors contend that in- order for the applicant to meet 10 CFR Part 50, Appendix A, Criterion 3, Black Fox, Unit Nos.1 and 2 l l

must utilize cables with fire retardant insulation.

Response: )

( The position,. stated in Regulatory Guide 1.120, " Fire Protection Guidelines for Nuclear Power Plants," an'd Branch Technical Position ASB 9.5-1, Appendix A. " Guidelines for Fire Protection for Nuclear Pcwer Plants Docketed Prior to July 1,1976," oIt fire retardant I

l insulation for cables used in nuclear plants is that " electric cable

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• constructions should, as a minimunr, pass the current IEE Standard 383, (IEEE Standard for Type Test of class IE Electric Cables, Field and Connections for Nuclear Generating Stations) flame test" . These tasts are used to screen cable insulation for fire The applicant has committed in his " Fire l

propagation properties. 1 Hazard Analysis Construction Permit Stage Reference Report 16," dated 1

November 17,1977, to use cable insulation that passes as a minimum the flame test in IEEE Standard 383 as modified by Nuclear ~nergy Liability The -;

Property Insuranca Association's (,NELFIA) 210,000 ETU fire tast.

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. . , 3 original IEEE Standard 383 specifies 70,000 BTU fire test. .The-Black Fox couritment test exceeds NRC Requirements.

Therefore, the staff requests that Contention Number 7 be sumarily disposed as there is no material issue of fact.

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4 ile have read the foregotng affidavtt and swear that it is true and accurate to the best of our knowledge. l b  ; . mb R6bartJQir.. tina

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Sehe dhed and sworn to before me this.f Da f N ye/f7P

_O - - l Notary Fuolfc Q' .

My ConnT3ston Expires b /, / 9 7 P .

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• Robert 3. Giardina Pmfasstonal Qualifications ,

Auxtl1ary Systems Branch .

OtvTsfon of Systems Safety Office of Nuclear Reactor Ragulation i

I am a Reactor Enginear in the Auxtltary Systams Branch in the Offica of Nuclear Reactor Regulation, U S. suelear Regulatory Cennission. In this position I perform tachnical reviews, analysas, and avaluations of reactor plant features pursuant to the construction and operation of reactors.

Education I recafved a Bachelor of Science Degrea in Mechanical Engineering from Orexal University in T971. In 1974 I recafvad from Drexel University a Mastar of Science Degree,in Mechanical Engineering with specialization in the area of Thamal and Fluid Sciencas and a Mastar of Scf anca Degree in Engineering Management with specialization in the area of desaarch and .

Development and Engineering Management (. Corporate i.avel). The research paper for the Engineering Management Design was entitled, " Technology Assessment." Since 1974 I have taken a number of coursas on PWR and BWR

.. System Opacation, Reactor Safety, Systam Ral f ability, Fault Tree Analysts, and Ftra Protaction.

DMMeu My experienca includes eight years of Enginaaring in the design, manufacture, and tasting of Shipboard Mechanical Systams and Components at the Philiadelphia Naval Shipyard. These systams and components included cooling water systams, propulsion systams, fire pmtaction systems, hydraulic systam, and ventila-tion . systems.

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I joined the Auxiliary Systems Branch of the Comission in 1974.

Since joining the Comission, I have perfomed safety evaluations err spent fuel pool expansions for five plants as well as provided input

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to or revised the Division of Operating Reactors position on spent fuel pool expansion, The Environmental Impact Statement o.n Spent Fuel Storege, the Regulations on Independent Spent Fuel Storage Facili. ties (10 CFR Part 72) and Regulatory Guides and Standard Review ,

Plans on Spent Fuel Storage'(on and off-site). I have been actively involved in the investigation of the steam generator feed water hammer

- problem and was a member of the Commission's Water Hammer Task Force.

- I am presently reviewing or have reviewed and evaluated the auxiliary systems of five nuclear power plants as wn11 as coordinated f. ire pro-taction reviews for three nuclear power plants.

I have successfully passed the test for and was awarded the certi-ficate of Engineering-In-Training.

Creanizational Membershics:

I am a member of Pt Tau Sigma - National Honorary Mechanical Engineer-ing Fraternity and of the American Management Associations. I am an Associate Member of the American Society of Mechanical Engineers.

I am an active member in the Boy Scouts of America and a Brotherhood Member in the Order of the Arrow, of that organization.

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.. .. i power plants,:and telephone comunication cantars, and tasting their installed fire protectfon systems.

4 From 1971 to 1972, I workad for GTE Automatic Electric, a connunication and talegraphic manufacturer, as a Fire Protaction and Safety Enginear.

My duties included all aspects of fire protaction and safety as they relata to an industrial occupancy. ~ *

• i During a short perfed of time in 1972, I was also employed at Martam .

Electric Company, mariufacturai of Ball equipment doing bastcally as my e

previous employment.

In 1972, I startad working for Gaga-Babcock and Associates, as a fire protaction and safety consultant. My major activities have baan in the area of (1) surveying and evaluating existing fire protaction systams for various industrial facilftfes to datarmine their adequacy and recomand necessary improvements and (2) design of fire dataction and suppression systems associated with (1) above. ,

l I am presently providing fire protaction consuiting to the Auxiliary Systams Branch of the NRC.

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8-1 Contention 8 I. Contention Intervenors contend that in order to meet 10 C.F.R. Part 50, Appendix A, Criterion 3, the Applicant must separate the cable trays, including those in the cable spreading room,so as to prevent a recurrence at Black Fox 1 and 2 of the type of fire which took place in the cable spreading room at Browns Ferry.

II. Argument Criterion 3 of Appendix A to 10 C.F.R. Part 50 requires that structures, systems, and components important to safety should be designed and located so as to minimize, consistent with safety requirements, the probability and effect of fires or explosions. As is shown by the following enclosed affidavit of Messrs. Giardina and Behn, the Applicant has made a commitment to comply with the requirements of Regulatory Guides o? 1.120 and 1.75; IEEE Standard 384; Branch Technical Position ASB 9.5-1 Appendix A, including Sections D.3(c) and F.3 so that cable trays are separated in each unit of Black Fox Station in the cable spreadinp rooms,as well as the plant,as required to:.2, meet 10i.CFR Part 50, Appendix A, Criterion n c- u3.-,nc i'a,.. ,-

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- In each. separate cable spreading room the cable tray separation distance $

both horizontal and vertical, .will meet the standards set forth above and will separate safety divisions one'and four in one cable spreading room from the redundant safety systems two and three in the other spreading -

room . In addition , a fixed automatic deluge

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system will' be . installed with a manual hand hose backup further mini-mizing the effects'of any fires within the cable spreading rooms.

Giardina and Behn-at 2.

l For the reasons stated above, the NRC Staff believes that the Appli-cant has adequately reduced the probability and effects of a fire at the proposed BFS Station. For that reason, no facts remain to be litigated and Intervenors Contention 8 should be dismissed.

III. - Statement of Material Facts About Which No' Issues of Fact Remain To Be Litigated

1. BFS will use separate cable spreading rooms for each unit.

2. Cable trays within each unit will be separated both vertically and horizontallyin accord with the Standards. In addi-tion, the cable spreading rooms will be separated by a 3-hour fire barrier (separating redundant safety systems 1 and 4 from systems 2and3). A fixed automatic deluge system with a manual backup will be provided in order to minimize the effects of a fire.

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UNITED STATES OF AMERICA '

. . . NUCLEAR REGULATORT COMMISSION )

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BEFORE THE ATOMIC SAFETY AND LICENSING BOARQ_- ,

l In the Mattar of , [ ,

f Public Service Company of $ Docket Nos.: 50-556 Oklabona h 50-657 (Black Fox Station, Unit Nos.1 & 2 J .

..l AFFIUAVIT OF ROBERT J. GIARDINA AND JAMES 0. BEHN l ON CONTENTION 8

1. My nama is Robart J. Gf ardina. I am employed by the Nuc1aar Regulatory Comission as a Reactor Engineer in the Auxiliarf j

Syste.:.x Branch. I have been employed in this position since Septembar,1974. My qualffications are listad on the attached l Qualifications Shast.

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2. My name is James 0. Bahn. I am employed by Gage-Babcock &

Associates. Inc. , as a fire protection and safety consultant.

I have bean employed in this field for the past eight years.

I am presently providing fire protection consulting to the Auxfliary Systems Branch of the NRC. My quaiffications are ifsted on the attached Qualtffcations Sheat.

3. The purpose of our affidavit is to address the intarvenor Contention Number 8 regarding the fire protection -

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program for cables, cable tray separation, and for the plant, respective-These contantions relata to the compliance of the Black Fox 19 Fire Protection Program with tne requirements in Appendix A Criterion 3 to 10 CFR Part 50. Our affidavit will show that the Black Fox Fire Protection Program is acceptable and will meet the requirements l

of 10, CFR Part 50, Criterion 3. .,__

Contention Number 8 4.

The Inteneners contend ^. hat in order to meet 10 CFR Part 50, Appendix A, Criterion 3, the applicant must separate the cable trays including those in the cable spreading room so as to prevent a recurrence at the Black Fox, Unit Nos.1 and 2 of the type of I

fire which took place in the cable. spreading room at Browns Ferry.

Responsa , , , , , , ,

The NRC 'dasign critaria for cable tray separation for both the cable spreading rooms and the nuclear plant is stated. fn Regulatory Guida 1.75, "Physfcal Endependence of Electric Systens," which endorses IEEE Standard 384, "Crftarfa for Siparatton of Class IE. Equipment and Cfreutts",

which is for electrically induced cable tray fires and Regulatory Guida l

1.120, and Branch Technical Position -ASE 9.5-1, Appendfx A for both I

exposure and electrically induced ffres. The appiteant has committed l fn his " Fire Ha::ards Analysis ConstructTon Permtt Stage Rafarenca Report I 15," to the following for each of the alack. Fox units, which are not intarconnected. .

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1. In the cable spreading rooms the cable tray separation dis- .

tances both vertical and horizontal will be in accordance Safety with Regulatory Guide 1.75 and IEE Standard 384.

divisions 1 and 4 will be separated from redundant safety divisions 2 and 3 by a 3-hour fire barrier in acco$fance with

- Regulatory Guide 1.120 and Branch Technical Position ASB, 9.5-1 Appendix A. In addition, a fixed automatic deluge sys-tem will be installed with manual hand hose backup in accordance with the guidelines of Regulatory Guide 1.120 and BTP 9.5-T Appendix A, Sectiors 0.3(c) and F.3.

2. In the balance of the plant tha cable tray separation dis-tances will be in accordanca with Regulatory Guide T.75 and In addition hose stations and. portable . f IEE Standard 384. i extinguishers will be provided in areas of low cable concen- l tration and automatically activated water deluge system (. fixed) will be provided for areas of high cable concentration in 1 l

accordance with Regulatory Guide T.120 and BTP 9.5-1, Appendix A,Section 0.3(c).

3. Fire barriers in the cable trays will be provided in accordance with BTP 9.5'-1, Appendix A Section 0.3(a) after the detailed design is. detemined.

These cosinttments. sho7dEriaIiy reduca the probahility and consequences Therefors, the. staff of the type of fire which.took e.aca at Browns Ferry.

Tequests that ContantTon Numhar 8 he. sumarily disposad as there is n _

matarial issua of fact.

4-Ida hava read tha foregoing affidavit and swear that it is true and accurate to the best of our knowledge.

b j RobertJGiardina

=-r h. l as 0. Bahn Subscribed and sworn to before me this # Da f b ye/f7P Q'

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Notary PunIfc My Commission . Expires , b /, / 9 7 P - .

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Robert J. Giardina Professional Qualificattons ,-

Auxiliary Systems Branch ,

Divfsfon of Systens Safety Office of Nuclear Reactor Regulation l

I am a Reactor Enginear in the Auxt1Tary Systems Branch in the Offica of Nuclear Reactor Regulation, U S. Auclear Regulatory Consission. In this position I perfom technical reviews,. analysas, and avaluations of reactor plant features pursuant to the construction and operation of reactors.

CducatTon I recatved a Bachelor of Science Degree in Mechanical En;inaaring from Drexal University in 1971. In 1974 I recafved from W University a

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Mastar of Scienca Degree in Mechanical Engineering ,,ecialization in the area.of Thannal and Fluid Scfancas and a. Master o,f Scienca Degraa in Engineering Mariagement with specialization in the area of Rasaarch and Development and Engineering Management (Corporats i.evel). The research paper for the Engineering Management Design was entitled, " Technology Assessment." Since 1974 I have taken a numbar of courses on PWR and BWR System Operatton, Reactor Safety, System Ralf abfif ty, Fault Tree Analysis, l

and Fire Protaction. ,

Expertenes itr experience includes eight years of Engineering in the design, manufacture, and tasting of Shipboard Mechanical Systams and Comeonents at the Philiadelphia Naval Shipyard. These systams and components included cooling water systamr, propulsion systams, fire protection systems, hydraulic systam, and ventila-tion. systems.

I joined the Auxiliary Systems Branch of the Comission in 1974. i Since joining the Comission, I have perfomed safety evaluations on spent fuel pool expansions for five plants as well as provided input l to or revised the Division of Operating Reactors position on spent fuel poc'T expansion, The Environmental Impact Statement on Spent  !

. Fuel Stoi age, the Regulations on Independent Spent Fual Storage l Facilities (10 CFR Part 72) and Regulatory Guides and Standard Review , l Plans on Spent Fuel Storage (on and off-site). I have been actively involved in the investigation of the steam generator feed water hamer problem and was a member of the Ccmmission's Water Hammer Task Force.

I am presently reviewing or have reviewed and evaluated the auxiliary l systems of five nuclear power plants as well as coordinated fire pro-taction reviews for three nuC dar power plants.

- I have successfully passed the test for and was awarded the certi-ficate of Engineering-In-Training.

Oreanizational Membershies: I I an a member of Pt Tau Sigma - National Honorary Mechanical Engineer-ing Fraternity and of the American Management Associations I am an Associate Member of the American Society of Mechanical Engineers.

I am an active member in the Boy Scouts of America and a Brotherhood Member in the Order of the Arrow, of that organization.

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. i Professianal -Qual tffcations - .. ,

James D. Sehn Gage-Babcock & Associatas, Inc.

My namarts James 0. Bahn. I as amployed as a staff enginear at Gage-Babcock and Associatas, Inc.,135 Addison Avenue, Elmhurst, ITifnofs.

Gage itabcock and Associatas is undar contract with Sandfa Laboratorias to pro'vida fire protaction consultation to the Auxfitary Systems Branch of the NRC. I as assisting the Auxfitary Systams Branch in reviewing -

the ffra p.4 Lion programs of nuctaar power plants of which Black Fox.

Untt Nos.1 and Z ts one. Tha Aux 11ky Systams Branch is respenstble-for ravfaring reactor 1icansa epplteattons and evaluating the design of auxfitary sii6, including. the ff re protaction sys*ms, of the nuclear l

power plant with respect to nuciaar safaef.

I attanded Illinois Institute of Technology in Chicago, Illinois, and I have recafved a B.S. Degree in Ftra Protaction Enginaartng in 1970.

also attanded varfous industrial seminars dealing mostly with fire safety.

I have a total eight years, of work axperf anca with major predacts as stated below.

I was first angloyed at Underwritars Laboratory in Chicago during the sumusers of my junfor and sanfor years at IIT, tasting fira resistanca ratings of various walls, doon, columns, etc.

From 1970 to 1971, I was amoloyed at Marsh and McLancan, tri Chicago defcg:

bastcally fire protactfen inspection of various indus: Mal sitas in the Chicago area, including factcHas, of*icas, hospitals, fossf1 fueled i

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power plants:and telephone connunicatTon cantars, and tasting tnefe installed fire protectfort systems.

Fres 1971 to 1977, I wortad for GTE Automatte Electric, a connunication and telegraphic manufacturer, as a Firs Protaction and Safety EngTnear.

My duties included all aspects of ftea protaction and safety as they relata to an industrial occupancy. --

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cing a short period of time fn 1972, I was also employed at Wartam

! diactric Company, manufacturer of Ball equipment doing hastcally as my prevfous empicyment.

In 1972, I startad working for Gage-Bahccek and Associatas, as a fire protaction and safety consultant. My major activities have been in the area of (1) surveying and evaluating existing fire pretz Men systams for varfous industrial facilities to datamina their adequacy and raccanand necessary fmprovements and,(2) dasign of fire dataction and suppresston systans associated with (1) above.

I as presantly providing fire pretac Ton c:nsul:Tng t:: the Auxilf ar.r Systems Srsnch of the NRC.

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Contention 9 d

I. Contention Intervenors contend that the Apolicant has not desi.,ned an in-depth fire protection system for Biack Fox 1 and 2 which complies with 10 CFR Part 50, Appendix A, Criterion 3. The affidavit of Messrs Giardina and Behn on Contention 9 is the Staff view of the indepth fire protection system described for Black Fox.

II. Argument -

Criterion 3 of Appendix A to 10 C.F.R. Fart 50 provides that structures,

, systems and c'omponents important to safety be designed and located so as ,

to minimize the probability and effect of fires and explosions. Based on a review of the infonnation and commitments contained in the Black Fox ,

PSAR and amendments, Messrs Giardina and Behn found that the Applicant had described the principal architectural and engineering design criteria necessary to reduce the probability and effect of any fires or explosions at the plant. The facility,as presently designed, has sufficient design flexibility so that engineering improvements will not be precluded by allowing construction to begin. In addition, the Staff was not able to identify any items which would require research and development. Giardina and Behn Affidavit at 2. Consequently, the Staff was able to conclude that any additional protection requirements would be successfully resolved before the completion of construction of the two Black Fox Units and that

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the plant can be-constructed and operated at that locality without undue risk to the population. Giardina and Behrr Affidavit at 3.

Messrs G1ardina and Bahn then go on to describe the fire protection revinw that will be done before the issuance of any operating license and which will ensure the actual implementation of the fire protection plan which will be in effect prior to operation. They state that the plant is expected to further improve the BFS fire system and that any new fire related design change or improvements will be required to be incorporated if they occur. Giardina and Behn affidavit at 4.

III. Statement of Material Facts About Which flo Litiaible Issues Remain

1. As outlined in 'the PSAR, the BFS fire protection plan adequately i

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describes the principal architectural design criteria which will form the basis of the fire system.

2. The fire protection system, as described in the PSAR and Fire Protection Analsyis, is sufficiently flexible so that later engineering improvements can be made.

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3. The Fire Hazards Analysis, as presented, will not require any l

research and development to be completed before the operating license is issued.

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4. That prior to the operation of the plant, another thorough review of the Black Fox fire protection system will be made to ensure l that all improvements and regulatory criteria which have occurred since _

the issuance of the CP will be incorporated, thus furnishing the last leg of the assurance that there will be an in depth fire protection l

system for BFS. I l

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1 UNITED STdTES '0F AMERICA NUCLEAR REGULATORT COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARQ ,

In the Mattar of Pubite Service Company of Docket Nos.: 50-556 Oklahoma 50-557 l l

(Black Fox Station, i Unit Nos.1 & 2 ) .

1 AFFIDAVIT OF ROBERT J. GIARDINA AND JAMES D. BE*4N ON CONTENTION 9

1. My name is Robert J. Giardina. I am employed by tha Nuclear Regulatory Commission'as a Reactor Engineer in the Auxiliary

, Systems. Branch. I have been employed'tn this position sinca j September,1974. My qualifications are Ifstad ort the attached Qualifications Sheet.

2. My name is James D. Behn. I am employed by Gage-Babcock &

Associates. Inc. , as a fire protection and safety consultant.

I have been employed in this field for the past afght years.

I am presently providing fire protection consulting to the l Auxiliary Systems Branch of the NRC. My quaiffications are l

. ifstad on the attached Qualifications Sheat.

3. The purpose of our affidavit is to address the intarvenor Contantion Number 9 regarding the fire protection l e

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program for cables, cable tray separation, and for the plant, respective-ly.

These contentions relate. to the compliance of the Black Fox Fire Protection Program with the requirements in Appendix A CMtarion 3 to 10 CFR Part 50. Our affidavit will show that the Black Fox Fim Prd i.fon Program is acceptable and will meet the requirements of 10 CFR Part 50, Criterion 3.

4. gentention Number 9 The Intervenors contend that the applicant has not designed an in-depth fire protection system for Black Fox, Unit Nos.1 and 2,  !

which complies with la CFR Part 50, Appendix A, Criterion 3. l

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Response

The staff presented its evaluation and conclusions with respect to the fire protection system for Black Fax Station Units 1 and 2 in Section $.6.1 of the Safety Evaluation Report (SER) dated June,1977. The Safety Evaluation Report statas that there is. ,

- sufficient design flexibility in the plant to allow future design changes if required and that "the design (of the fire protection systam) as presently proposed, meets General Design Critarion 3,

' Fire Protection,' and applicable guidelines in effect prior to issuance of Branch Technical Position ASB 9.5-1, and Appendix A, thereto, and for the construction permit stage, we find it i

acceptable." Therefore,. we could defer the detailed fire protec-tion system review until after issuanca of the construction pemits.

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Our rev[e[was5a' sed on the infomation and czmitnents submitted in the PSAR, including all amendments and letters, which were sufficient to permit us to make a finding pursuant to 10 CFR 50.35(a), that the fire protection system design is adequate for the leveT of.revfew necessary at the construc<

tion pemit stage. The basis for this conclusion is as follows:

(1) The applicant has described the principal architectural and engineering design criteria with respect to plant fire protection, and has identified the major features or com-ponents' incorporated therein with respect to the plant fire protection systems for the protection of the health and '

safety of the public.

l (2) The presently designed fa,cility has sufficient design flexibility to allow implementation of any design changes

- 1. hat may be necessary to assure compliance of Black Fox Station Units 1 and 2 with Appendix A to the Branch Technical .

Position ASE 9.5-1,. and Regulatory Guide 1.120 and which may Teasonably be left for later consideration.,

(3) There are no safety questions associated with the features l

of the fire ptJtection system that requires any research and development.  ;

On the basis of the above, there is reasenable assuranca (4) that:

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(t)'anyadditfenla protection req'ufruments regarding the fire protection 5,ysting wf1T be satisfactorily resolved at or before the latest dated stated in the appiteation for completion of construct j4 n of Black Fox, Unit Nos.1 and 2.

(if th respect to the fire protection system, and taking into consideration the site criteria contained in 10 CFR Fart 100, the proposed facilities can be constructed and operated at ,

the proposed location without undue risk to the health and .y safety of the public.

The review of the Black Fox station's Fire Protection Program wili be perfonned in accordance with the guideifnes of Regu'latory Guide 1.120 The applicantiss and Branch Technical Position ASB 9.5-1, Appendfx A.

coamitted by letter dated May 25, 1977, to provide a fire hazards analysis and a comparison of his fire protection program to'the guidelines of The applicant submitted I Branch Technical Position ASB 9.5-1, Appendix A.

on November 30,1977, a report entitled, " Black Fox Station Fire Hazards This report Analysis, Construction Pemit Stage, Reference Report 16."

includes a Fire Hazards Analysis and the comparison to the guidelines of A review of the report Branch Technical Position ASE S.5-1, Appendix A.

wvealed that all of the requested positions had been addressed and that the report was acceptable for staff review. When the detailed fire protection' system review is completed, further requirements may be imp l

' to further improve the capability of the fire protection system to prevent

_-___-____a--____-._m r -r a- -

unacceptable damage that ma result from a fire. This review sill be done on a timely basis and prior to the issuance of the operating license so that the applicant can effectively incorporata any design changes .in the final. design.

Therefore,thestaffrequeststhatContentionNumber9bedisposed as there is no material issue of fact.

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f We hava read the foregoing affidavit and swear that it f* true and accurate to the best of our knowledge.

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R6bartJGiardina A m0. Bahn

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.Subscrthed and sworn to before me this. f a f N ye/f7P

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p' Notary Publfc My Comission Expiras b /, / 9 7 P . ,

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Robert J. Giardina Professtonal Qualificatians ,-

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Auxil'tary Systams'Branc'h ,

Ofvtston of Syst es safety Offica of Nuclear Reactor Regulation I as a Reactor Enginear in the Auxiliary Systams Branch in the Office of -

Nuclear Reactor Regulation, U S. Auclear Regulatory Connission. In this position I perfom technical reviews,. analysas, and evaluations of reactor plant featurer pursuant to the construction and operation of reactors.

Education I recafved a Bachelor of Scienca Degree in Mechanical Engineering from Drexal University in T971. In 1974 I recatved from Draxal University a Mastar of Scienca Degree in Mechanical Engineering with specialization in the area.of Thamal and Fluid Sciences and a Mastar of Scienca Degraa in i

Engineering Management with specialization in the area of Resaarch and Development and Engineering Management (, Corporate i.evel). The research paper for the Engineering Management Design was entitled, " Technology l

)

Since 1974 I have taken a number of coursas on PWR and SWR l Assassment."

l System Operation, Reactor Safety, Systen Ralf abtitty, Fault Tree Analysts, and Fire Protection.

Expertence My experienca includes eight years of Engineering in the design, manufacture, and tasting of Shipboard Mechanical Systems and Components at the Phfif adalphia:

Naval Shipyard. These systems and components included cooling water systems, propulsion systas, fire protection systems, hydraulic system, and ventila-tion. systems.

. _g_

I joined the Auxiliary Systems ' Branch of the Comission in 1974..

Since joining the Comission,'I have performed safety evaluations on spent fuel pool expansions for five plants as well as provided input n ta_ar_ revised the Division of Operating Reactors position on spent fuel pool expansion, The Environmental Impact Statament on Spent Fuel Storage, the Regulations on Independent Spent Fuel Storage Facilities (10 CFR Part 72) and Regulatory Guides and Standard Review ,

hin~s~5 Spent Fuel Storage (on and off-sita). I have been actively involved in the . investigation of the steam generator feed water hamer problem and was a member of the Comission's Water Hammer Task Force.

I am presently reviewing or have reviewed and evaluated the auxiliary systems of five nuclear power plants as well as coordinated f. ire pro-taction reviews for three nuclear power plants.

I have suecassfully passed the test for and was awarded the carti-ficate of Engineering-In-Training.

Organizational Memberships:

I am a member of Pt Tau Sigma - National Honorary Mechanical Engineer-ing Fratamity and of the American Management Associations. I am i an Associate Member of the American Society of Mechanical Engineers.

I am an active member in the Boy Scouts of America and a. Brotherhood Member in the Order of the Arrow, of that organization.

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l. . .*

I Professtonal Qualifications" .

James D. Behn Gage 2= M ck &. Associates, Inc.

l My namacts James D. Bahn. I, as employed as a staff enginear at Gage-i Babcock, and AssocTatas, Inc.,135 Addison Avenus, Elmhurst,111tnots.

Gage itabcock. and Associatas is under contract with Sandia Laboratsrfes i to pre'vTda fire protecttert consultatton to the Auxf1tarf Systems Branch  :

of the NRC. I an asststing the Aux 11tarf Systems Branch in reviewing -

tha fire protaction programs of nuclear power plants of which Black Fox,  ;

Unit Nos. I and Z ts one. Tha Aux 11f ary Systems Eranch is respensthle

' for reviewing reactor Itcansa appiteations and evaluating the design of aux 11tary 5346, including. the fira protaction systams, of the nuclear power plant with respect to nuclear safety.

I attandad Illinois Instituta of Technology in Chicago, Illincts, and recatved a B.S. Degrea in Fire Protaction Engineering-in 1970. I have also attanded various industMal seminars dealing mostly with fire safety.

I'have a tntal eight years of work axpertenca with major projects as stated below.

I was first empleM at Underwritars Laboratory in Chtcago duMng the summers of my junior and sanfor years at IIT, tasting fire resistanca ratings of varfous walls, doors, columns, ate.

Frem 1970 to 1971, I was employed at Marsh and McLannan in Chicago defog:

bastcally fire protaction inspection of various indus: Mal sitas in the Chicago area, including factoMas, of"icas, hospitals, fossf1 fuelad

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power piants:and telephone comunication cantars, and tasting their installed fire protectiert systems.

Fma 1971 to 1g72, I worked for GTE Automatte Electric, a comunication and telegraphic manufacturer, as a Fire protection and Safety Enginear.

My duties included all aspects of fies protection and safety as they relata to art industria1 occupancy.

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During a short period of time in 1972, I was also employed at Wartam Electric Cogany, madufacturer of Ball equipment doing hastcally,as my prvifour empicyment In 1972, I startad working for Gaga-Babecek and Associatas, as a fire protaction and safety censultant. My major activities have been in tha area of (1) surveying and avaluat$ng axisting fire protaction systams for various industrial facilfttes to datamina thair adaquacy and reconnand

~

necassary tgrovements and (2) design of fire detaction and suppressten 516 associated with (1) above.

I as presantly providing fire protaction consuittng *a the Auxiliarf Systans Branch of the NRC.

1

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l0-1 Contention 10 I. Contention The Intervenors contend that the Applicant and the Regulatory Staff have not demonstrated that the Black Fox quality assurance program will comply with the pertinent portions of Appendix B to 10 CFR Part 50 in the design and installation of the followir g equipment:

a. Pressure vessel

b. Control rods

c. Reactor protection system

d. Emergency core cooling system

e. Ra'dwaste equipment (both liquid and gas)

[

f. Dry'well wall

g. Weir wall

h. Containment shell

)

II.. Argument The Applicant has described the quality assurance program for the design The and construction of the Black Fox Station in Chaoter 17 of the PSAR.

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latest submittal is in Amendment 10. The~Stiaff reiiewed ihii ofigiliaI suWnittal to ascertain whether each requirement of Appendix B to 10 CFE Part 50 was adequately addressed. Questions resulting from this review were responded l l to by the Applicant, and the description of the quality assurance program i

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10-2 was changed correspondingly. The initial review compared the Applicants' original comitments to a checklist which was the foreninner of the Casarission's Standard Review Plan for quality assurance during the design and construction phase of nuclear power plants (Section 17.1 of M1 REG-75/087).

Amendment 10 has been campared to the Acceptance Criteria found in the Standard Review Plan. The Acceptance Criteria cover each of the eighteen criteria in Appendix B to 10 C.F.R. Part 50, including such things as the quality assurance organization of the Appifcant and his principal l

. contractors as well as the program for controlling design, procurement,  ;

l l

documents, equipment, special processes, inspections, and tests. As stated in the Standard Review Plan, the program is considered in comp 11ance with Appendix 8 to 10 CFR Part 50 and is deemed acceptable when it meets the Acceptance Criteria. The Applicants program was reviewed and found to comply with the acceptance criteria. Spraul at 3.

III. Statement of Material Facts As To Which There Are flo Genuine Issues

1. As demonstrated in the enclosed affidavit of Jack Spraul, the l quality assurance program for the Applicant, the reactor vender, and the architect engineer meet the requirements stated in 10 C.F.R. Part 50, Appendix B.

l

2. For 'each item listed in Contentien 10, the PSAR describes, in a manner commensurate with proteccion of the public health and safety, the quality assurance program to be apolied.

l UNITED STATES OF AMERICA t

NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

[

In the Matter of the' Application of l Public Service Company of Ok1ahoma, l Associated Electric Cooperative, -

Inc. Docket Nos. l STN 50-556 l and Western Farmers Electric Cooperative STN 50-557 l l

(B1ack Fox Units 1 and 2)

AFFIDAVIT OF J. G. SPRAUL ON CONTENTION 10 1

d 1 e

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- 1 CONTENTION 10 l AFFIDAVIT OF J. G. SPRAUL My name is J. G. Spraul. I an employed by the Nuclear Regulatory Commission as a Nuclear Engineer'in the Quality Assurance Branch. I have been employed in this position since January 1974. My qualifications are listed on attached qualification sheet.

The purpose of my affidavit is to address intervenor Contention 10 regarding the acceptability of the Black Fox quality assurance program for a nuclear plant. Contention 10 relates to the compliance of the Black Fox quality assurance program with the requirements in Appendix B to 10 CFR Part 50. My affidavit will show that the Black Fox quality assurance program is not unacceptable due to nonconformance with the requirements of Appendix B to 10 CFR Part 50.

The intervenors contend that the Applicant and the Regulatory Staff have not demonstrated that the Black Fox ' quality assurance program will comply with the pertinent portions of Appendix B to'10 CFR Part 50 in the design and installation of the following equipment:

a. Pressure vessel

b. Control rods

c. Reactor protection system

d. Emergency core cooling system

e. Radwasta equipment (both liquid and gas)

f. Dry well wall

g. Weir wall

h. Containment shall' During the discovery process, it became apparent that the intervenors' concern was not only that the listed equipment would not be under the quality assurance program but also that the quality assurance program  ;

does not meet the requirements of Appendix B to 10 CFR Part 50.

As will be shown below, the listed equipment is under the quality assurance program, and-the quality assurance program meets the applicable requirement.< in Appendix B to 10 CFR Part 50.

The Commission's criteria for evaluating the suitability of a quality '

assurance program for a nuclear plant are contained in 10 CFR Part 50.

These criteria, as defined in Appendix B, include the concept of applying pertinent requirements to all activities affecting the safety-related functions of structures, systems, and components that prevent or mitigate

. . . , . - _-- .r

the consequences of postulated accidents that could cause undue risk to the health'and safety of the public. As used in Appendix B, " quality assurance" comprises all those planned and systematic actions necessary to provide. adequate confidence that a structure, system, or component will perform satisfactorily in service.

Part 17A.1.2.2 of the Black Fox PSAR states that the Black Fox quality assurance program applies to all items identified in Table 3.2-1 (GER) of the PSAR as having 10 CFR 50, Appendix B applied. Table 3.2-1 (GER), in Chapter 3 of the PSAR, refers to GESSAR for the reactor system.

The list below shows the specified quality assurance program requirements for the equipment listed in the contention:

(a) Pressure Vessel: Table 3.2.1 (p3.2-9, 11/26/75) of GESSAR 238 shows that the reactor vessel, item I.1, will be constructed under the quality assurance program meeting the applicable requirements of 10 CFR 50, Appendix B.

(b) Control Rods: Item I.7 of the same reference shows the control rods will be constructed under a quality assurance program meeting the applicable requirements of 10 CFR 50, Appendix 8.

(c) Reactor Protection System: Section 3.2.).2.1 (item a, p3.2-4, 11/26/75) of GESSAR 238 lists the reactor protection system as Safety Class 2. Table 3.2.2 (p3.2-25, 11/26/75) of GESSAR 238 shows that Safety' Class 2 items are constructed in accordance I with the quality assurance requirements of 10 CFR 50, Appendix B. j (d) Emergency Core Coolina System: Section 3.2.3.2.1 (item f, p3.2.-4, 11/26/75) of GESSAR 238 lists the emergency core cooling systems as Safety Class 2. Table 3.2.2 (p3.2-25, 11/26/75) of GESSAR 238 shows that Safety Class 2 items are constructed in accordance with the quality assurance require-ments of 10 CFR 50, Appendix 8.

l (e) Radwasta Equipment (both liauid & cas): The Staff requires only specific Appendix B requirements be applied to radwasta equipment. Special quality assurance provisions (less than total Appendix B) are required. These provisions are listed in Part VI of the Effluent Treatment Systems Branch Technical l Position 11-1 which forms part of section 11.2 of the Commis-sion's Standard Review Plan (NUREG-75/087). Chapter 11 of the .

Black Fox PSAR (pil.2-1c) indicates the radwaste equipment will meet these special quality assurance provisions.

(f), (g) and (h) Dry Well Wall, Weir Wall, and Containment Shell:

These items are all part of " Containment" wnicn will be construct-ed under a quality assurance program meeting the applicable p -rw -

-,r. r n .,

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i requirements of 10 CFR 50, Appendix B per item XXXVII 1 of the i Black Fox PSAR Table 3.2-1 (p3.2-14a, 3/25/77).

Thus, the PSAR contains commitments that the equipment listed in contention 10 will be under an appropriate quality assurance program.

Regarding the intervenors' concern that the Black Fox quality assur-ance program does not meet the requirements of Appendix' B to 10 CFR Part 50, the information presented below shows that the program does meet the requirements.

The applicant has described the quality assurance program for the design and construction of the Black Fox Station in Chapter 17 of the P5AR. The latest submittal is in Amendment 10. We reviewed the original  !

submittal to ascertain whether each requirement of Appendix B to 10 CFR Part 50 was adequately addressed. Questions resu'1 ting from this review were responded to by the applicant, and the description of the quality assurance program was changed correspondingly. The initial review compared the applicants' original commitments to a checklist which was the fore-runner of the Commission's Standard Review Plan for quality assurance during the design and construction phase of nuclear power plants (Section 17.1 of NUREG-75/087). Amendment 10 has been compared to the Acceptance Criteria found in the Standard Review Plan. The Acceptance Criteria cover each of the eighteen criteria in Appendix B to 10 CFR Part 50 including such things as the quality assurance organization of the appli-cant and his principal contractors as well as the program for controlling design, procurement, documents, equipment, special processes, inspections, and tests. As stated in the Standard Review Plan, the program is considered in compliance with Appendix B to 10 CFR Part 50 and is deemed acceptable when it meets the Acceptance Criteria.

Our comparison of Amendment 10 of the Black Fox PSAR to the Accep-tance Criteria fcund in the Standard Review Plan shows that each item in the Standard Review Plan has been addressed acceptably by the applicant.

Specific concerns presented by the intervenors during the discovery process are answered below.

1. The description of the P50 Quality Assurance Program in the PSAR l fails to adequately define who is responsible for each work element I of the Quality Assurance Program, when the work element is initiated, l and h3 the element is accomplished.

Answer:

Appendix 17A of the PSAR, the quality assurance program for PSO, lists the " work elements," i.e. , each criterion of Appendix 8 to 10 CFR Part 50, and identifies the organization responsible for the 1

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4

. r work elements. Appendix 17A includes the procedural steps involved in performing the work elements. This provides-an adequate descrip-tion of what each work element is, who is responsible, and how it is performed. With reference to when, Criterion 6 of Appendix B requires, and the description in the PSAR provides for, activities to be con-ducted by written procedures which are distributed to and used at the location where the prescribed activity is performed. The PSAR is responsive to this Appendix B requirement by specifying that procedures must be written and, further, that personnel will be trained in these procedures. Thus, it follows that the procedures precede the work elements.

The level of detail provided in Appendix 17A of the PSAR complies '

with the Standard Review Plan and is acceptable to the staff.

2. The Quality Assurance Program of the NSSS supplier, General Electric, as described in Section 17C of the PSAR, is too general to allow detailed comparison with the requirements of 10 CFR 50, Appendix B.

The General Electric Quality Assurance Program appears to fail to demonstrate compliance with Appendix B in the Quality Assurance work elements of design control; control of materials, parts, and components;.

control of s,pecial processes; and cor.rective action.

Answer: )

Appendix 17C of the PSAR, General Electric's quality assurance program, references GESSAR which, in turn, states that the General Electric quality assurance program is described in NEDO-11209-03A.

NEDO-11209-03A, " Nuclear Energy Divisions BWR Quality Assurance Program Description," November 1976, was reviewed and found accept-able by the NRC staff. A letter to this effect, issued by the NRC Chief, Quality Assurance Branch, Division of Project Management, is included in the document. The items mentioned in this concern are described in the following places in NE00-11209-03A:

Work Element Description .

1 Design Control - _

Section 3, pp. 41-46 Control of Materials, Parts, and Components Section 8, pp. 55-56 Control of Special Processes Section 9, pp. 59-60 Corrective Action Section 16, pp. 71-72 ]

'The level of_ detail provided in NED0-11209-03A on how each work element is accomplished complies with the Standard Review Plan and is acceptable'to the staff.

• l

Specific concerns are individually expressed and answered below as items 29 (design control), 30 (control of materials, parts, and components), 31 (special processes), and 32 (corrective action).

3. The PSO and General Electric Quality Assurance Program descriptions in the PSAR do not adequately describe the interface with the NRC's recently initiated Vendor Inspection Program.

Answer:

There is no requirement that the PSAR includes any description of interface with NRC's Licensee Contractor and Vendor Inspection Program, and the Black Fox PSAR describes an acceptable vendor con- I trol program with no reliance on the NRC program.

4. The 10 CFR 50 and NRC regulatory guides do not define the require-ments of the NRC Vendor Inspection Program. Recently quality inadequa-cies in the North Anna Plant, Browns Ferry Plant, and Rancho Seco olant raise serious questions about the adequacy of the whole NRC ality Assurance Program. In particular, questions need to be

.swered about the NRC policy of relying on buf1ders for primary ~

inspections with NRC officials serving as auditors. .

Answer: , ,

The ASLB hearing for Black Fox is not the proper forum for question-ing Commission policy and guidance. 'Such questioning is properly made at rulemaking proceedings or in a petition for rulemaking.

1

5. The statement by P50 in the PSAR that the Manager, Quality Assurance authority is limited in that it only " encompasses all activities up to, but not including, the station's preoperational tests" conflicts with the responsibilities as described in the preceding sen.tences which state that "the Manager, Quality Assurance is responsible for the preparation and Management of the PSO Quality Assurance Program."

The responsibilities of the Supervisor, Quality Operations are not defined in the PSAR. .

Answer:

The statements have been changed in Amendment 10 of the PSAR such that it is now clear that the Manager, Quality Assurance has prime responsibility for quality assurance during the design, construc-tion, and operating phases of the plant.

The responsibilities of the Supervisor, Quality Operations are required in the FSAR, not in the PSAR, since the position has no responsibilities until the operations phase.

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6.- The statement in the PSAR that "the technical requirements to be verified in quality audits are identified by the PSO-BFS Project Director" conflicts with the requirements of 10 CFR 50, Appendix B, Criteria I, that states that personnel performing Quality Assurance functions shall. enjoy sufficient independence from " cost and schedule."

The- Project Director, as defined in Paragrpah 17A,1.1.10, is "respon-sible for... meeting the scheduled completion date set for the facility, and keeping the costs within the budget provided." .

Answer:

Amendmen',10 changed responsibility for identifying technical require-ments to be verified by audits from the PSO-BFS Project Director who is onsite to the P50 Manager, BFS Engineering and Construction, who represents a higher level of management and is offsite. This eliminates any concern of undue cost and schedule pressures but still has this function performed by a knowledgeable individual.

7. The requirements of,10 CFR 50, Appendix B, apply to all safety-related j (Class I) structures, systems, and components. The statement by P50 )

in the PSAR that "the extent to which the individual sections and requirements of the Quality Assurance Program applies will depend

( upon the nature and scope of work to be performed and the importance i to safety of the items and services to be performed" implies that not all Class I items must meet the requirements of 10 CFR 50,

, Appendix B.

Answer:

The quoted statement is in consonance with the words in Appendix B

'of 10 CFR Part 50 which state: "The pertinent requirements of this appendix apply...." Criterion 2 of Appendix B states that "The quality assurance program shall provide control...to an extent consistent with their importance to safety." The quotation in the concern, above, is a clear commitment by PS0 that all Class I items will be subject to the pertinent requirements of the quality assur-

ance program.

8. The PSAR does not describe _the manner by which PSO will assure that

" applicable portions of Black Fox Station be designed, fabricated, erected, inspected, and tested in conformance with the rules of the ASME Boiler and Pressure Vessel Code" since the " jurisdictional authorities of the State of Oklahoma have not adopted the Code."

Answer:

The commitment is made in Amendment 7 of the PSAR that: "The appli-cable portions of Black Fox Station will be designed, fabricated, t -:

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erected, inspected, and tested in accordance with the rules of the ASME Boiler and Pressure Vessel Code,Section III," in spite of the fact that the State of Oklahoma has not adopted the Code.

9. The PSAR is not in accordance with the requirements of 10 CFR 50, Appendix B, Criteria II. Appendix B applies to "all activities affecting the safety-related functions of those structures, systems, and components" while the PSAR states that only " major equipment .

suppliers and prime contractors are required to have written Quality Assurance programs, procedures, methods, and operating instructions to carry out and verify that specified activities and the rtuality of material and workmanship are properly achieved and controlled."

Answer: l l

The PSAR, since Amendment 7, states that: "All safety-related l equipment suppliers and contractors for PSO 'are required...."

10. The requirements for procurement document control, as outlined in ,

Paragraph 17A. 1.4.4 of the PSAR, are not in accordance with 10 CFR l 50, Appendix B, Criteria V. The PSAR states that "the review and  !

approval of such nodificat' ions and changes to procurement documents will be performed by affected technical organizations and the quality

( organizations of PSO and B&V," while Criteria V requires that " changes to documents shall be reviewed and approved by the same organizations that performed the original review and approval unless the applicant designates another responsible organization."

Answer:

The PSAR, since Amendment 7, states that: "The review and approval of such modifications and changes to procurement documents will be performed by the same organizations that performed the original review and approval."

11. The P50 Quality Assurance Policy, Quality Assurance Manual, and )

Quality Procedures Manual review and approval requirements are .

)

inadequate in that there is no statement that the preceding docu-ments will all be reviewed periodically, for instance once a year, i to ensure that all docuw nts are current. The Quality Assurance  !

implementation portions of the Project Procedure Manual should include the Manager, Quality Assurance in the review and approval cycle, with the Quality Assurance Manager designated as responsible j for final interpretation of the Quality Assurance documents, not the i Project Director, as defined in Paragraph 17A.1.5.1.3 of the PSAR.  !

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8-Answer:  !

l There is no. requirement that the documents listed be reviewed periodi-cally to ensure they are current. However, they do come under the i program for document control which requires that participating l organizations have procedures for control of the documents and changes. thereto to preclude the possibility of use of outdated or

- inappropriate documents. Further, acceptability of these documents will be determined by the staff on a sampling basis during routine l inspections of the Black Fox Project. There is also a commitment on l page 17.0-1 that the Commission will be given timely notification of any significant changes in the QA Program.

Table 17A.1-1 of the PSAR shows that the Assistant Vice President has final approval authority of the Project Procedure Manual. Both l the BFS Project Director and the Manager Quality Assurance report to the Assistant Vice President. Thus, any question of interpretation ,

of the quality assurance implementation portions of the Project I Procedure Manual would be resolved at the Assistant Vice President l 1evel. This is acceptable to the staff.

F2. Ths PSAR is contradictory in that Paragraph 2 of Section 17A.1.6.2 7

describes a record system in accordance with that " established in i - auidance doc'iment N45.2.9.", while in Section 17A.1.17 of the PSAR P50 commits '.o a record system "as defined in N45.2.9." Also the PSAR is deficient in that it calls out only the ANSI Standard number

- N45.2.9. The PSAR should define both the standard number and the ~

year; for instance, N45.2.9.-1974.

Answer:

Table 1.9-1 of the PSAR provides applicant's position on Regulatory Guides. With respect to records, the table states that PSO will meet the provisions of Regulatory Guide 1.88, " Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records" (Rev. 1, 12/75). This Regulatory Guide invokes the requirements and guidelines of ANSI N45.2.9-1974.

13. The PSAR requirements are in conflict with Appendix B of 10 CFR 50 in that Appendix B does not yet provide that "those suppliers which have been evaluated under NRC approved certification programs may not require preaward surveys."

Answer:

Criterion 7 of Appendix B to 10 CFR Part 50 requires " source evalua-tion and selection." The Commission issues NUREG 0040 which docu-ments inspections performed by the Commission. The preface to NUREG

++e e- y .- p y n a-

.g.

0040 states that it "contains information which is necessary in establishing a.' qualified suppliers' list; however, the information contained in this document is not adequate nor is it intended to stand by itself as a source of information concerning qualified

_- suppliers."

Thus, there is the commitment in the PSAR that "the Manager, Quality Assurance, shall evalbate each apparent successful bidder's QA program to determine the need for a preaward survey. We expect the Manager, Quality Assurance, to consider the information available in NUREG 0040 as part of this determination.

14. Paragraph 17A.l.7.2.2 of the PSAR is inadequate in that the time ~

period between contractor audits is not described.

Answer:

Page 17A.1-32 of the PSAR states: "T'he Manager, Quality Assurance,

... will have audits performed on such contractors activities during appropriate time periods when significant. quality related activities are being carried out." In addition, PSO has committed in part 17A.1.2.1 of the PSAR to conform to the " Gray Book," WASH-1283, which includes

( ANSI N45.2.12, " Requirements for Auditing of Quality Arsurance Programs for Nuclear Power Plants." Audit frequency is covered in part 3.4 of this ANSI Standard which states: " Applicable elements of the quality assurance program shall be audited at least annually or at least once within the life of the activity, whichever is

- shorter."

15. The PSO plans for surveillance are not clear. Does PSG inte.id that surveillance will be performed on all Class I items "where verifica-tion of procurement requirements cannot be determined by receipt inspection?"

Answer:

Part 17A.1.7.2.2 of the 'PSAR states: " Surveillance 3erformed on items when verification of procurement requirementi be determined by receipt. inspection." In light of the 1; ay limiting modifiers on " items," the staff interprets th) ,atement to mean "all" items that perform a safety-related functt 1.

16. The PSAR states that the " Field Project Director has the responsi-bility for receipt inspection" which is in conflict with 10 CFR 50, Appendix B, Criteria I, which requires that Quality Assurance activi-l ties be independent from the pressures of cost and schedule. The

• I role, if any, of the Manager, Quality Assuranca of PS0 in receipt I

inspection,-is not described in the PSAR. (Note'that the " Field

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i Project Director" is the " Manager, BFS Construction" in Amendment 10 l of the PSAR.)

Answer:

The PSO organization described in Section 17A.1.1 and shown on Figures 17A.1-1 and 17A.1-2 of the PSAR has been. reviewed versus the Standard Review Plan (Section 17.1) and found to meet the require-ments of Appendix B. It is acceptable to the Staff for the design and construction of the Black Fox Station. The authority, responsi-1 bilities, and duties of the Manager, Quality Assurance (Section 17A.1.1.17) and the Manager, BFS Construction (Section 17A.1.1.7) are typical for these positions and are also acceptable to the Statf.

At the plant site, the Manager, BFS Construction has overall responsi-bility. The Superintendent, Quality Control reports directly to the Manager, BFS Construction and has overall responsibility for implement- J l

ing the site quality control program, including receipt inspection (part 170.1.7), acceptance inspection of construction work after ,

acceptance under the contractor's QA program (part 170.1.7), and f

monitering contractors' control of special processes (pert 170.1.9). .

t The Commission's requirement for independence of the PSO site quality control personnel from undue pressure of cost and schedule is satis-fied as follows:

(a) The Site Quality Assurance Superintendent reports to the Manager, Quality Assurrm. He monitors and audits quality assurance and qu 'l activities at the job site (part 170.1.1.7).

(b) The PSD 'nce organization reviews and approves all PS0 qual. oc:dures (and changes) and is responsible for certi app mval of training of P50 quality control p e ..n..) ,,part 17A.1.1.18).

These responsibility assignments are acceptable to the Staf'.

17. The General Electric product quality certification, as described in Section 17 of GESSAR, does not directly reference "the specific codes and standards met by the item," as required by Paragra;;,h 17A.1.7.2.3 of the PSO pcrtion of the PSAR. In addition, the PSAR requires that PS0 be provided.on the certificate of conformance a description of all_ "nonconformances dispositiuned use-as-is or repair," while the GESSAR in Paragraph 17.1.15.2 commits to provide this information to PSO only on "nonconformances to procurement requirements regarding end use." .

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Answer:

Section 17A.1.7.1 states that GE will supply the NSSS equipment and services and that the provisions for GE's control are described in Appendix 17C (NEDO-11209-03A, GE's topical report on quality assur-ance).- Thus, GE will follow the commitments in Appendix 17C which i states that "GE will identify and report ... those nonconformances

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to procurement requirements regarding and use of modules which are dispositioned 'use as is' or ' repair. '" A footnote indicates these are nonconformances that may have a potential adverse effect on the safety, operability, reliability, integrity or maintainability of i the module. I 1

Section 17A.1.7.2 shows that other purchased equipment (exclusive of the NSSS) will be the direct responsibility of ?SO, and nonconform-ance control of this equipment will be in accordance with Section 17A.1.15.

18. The PSO description in the PSAR requires that " contractors will be '

required to maintain materials identification traceable to end use in safety-related components," while the GESSAR only commits to marking material as being acceptable and does not commit to maintain-( ing traceability.

Answer:

In the area of identification and control of materials, Appendix B ,

to 10 CFR Part 50 requires that: " measures shall be established for  ;

the identification and control of materials, parts, and components, '

including partially fabricated assemblies. These measures shall assure that identification of the item is maintained by heat number, part number, serial number, or other appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, erection, installation, and use of the item. These identification and control measures shall be designed to prevent the use of incorrect or defective material, parts, and components." .

l Thus, Appendix B requires that material be identified so that the proper material goes into the proper place. Traceability is not a specific requirement of Appendix B.

The' statement that GESSAR "only commits to marking material as being ,

acceptable and does not commit to maintaining traceability" is in

. error. Both the P50 commitment and the GE commitment meet the Appendix B requirements for control of materials. GE has committed to the following in GESSAR (NED0-11209-03A):

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" Procedures and practicas are establisned and documented which .

provide for the identification and control of materials, parts and l l

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I components. ~ Equipment components, including subassemblies of parts, are identified and controlled by serial, numbers, part numbers,- or heat numbers.

"The identification and control measures provide for relating an

. item of production (batch, lot, part or assembly) at any stage, from ,

material- receipt through fabrication and shipment to an applicable . 1 drawing, specification, or other control document.

" Material is marked to identify inspection status and maintain identification. As applicable, materials traceability is provided in accordance with documented procedures specifically required by the ASME Boiler and Pressure Vessel Code.

" Material identification is verified by the materials organizations prior to . release for assembly or installation. Final verification of correct identification is accomplished by Quality Assurance personnel prior to release for shipping." l l

This is acceptable to the Staff. '

19. The description of control of special processes in Section 17A of the PSAR is not in accordance with the requirements of 10 CFR 50, 1

' Appendix B, Criteria IX. The PSAR does not mention the need for qualified personnel for special processes. The PSAR does not describe who is responsible within P50 for designating or approving special .

processes and for determining the need and approving of the quali- l fied personnel.

Answer:

In the area of special process control, Criteria IX of Appendix B to 10 CFR Part 50 requires that: " measures shall be established to assure that special processes, including welding, heat treating, and nondestructive testing, are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and othen special requirements. "

As stated in 3ection 17A.1.9 of the PSAR, the control of special processes is the responsibility of-the contractors performing the work. For example, GE states that special processes are accom-plished using qualified personnel, procedures, and equipment (NEDO-11209-03A, p. 57). PS0 and its principal contractors will ce respon-sible for approving. vendor QA programs covering special processes and for auditing activities.related to the control of special processes.

This is acceptable to the Staff.

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20. The responsibilities ~of the Manager, Quality Assurance, for preopera-tional testing need to be described in Section 17A of the PSAR, particularly in regard to the responsibility for determining equip-ment acceptance.

Answer:

As noted in the answer to item 5, the Manager, Quality Assurance is responsible for quality assurance during preoperational W ting. He' l has ultimate responsibility for the acceptance of equipet...

21. No mention is made of the 10 CFR 50, Appendix B requirement that-  !

"the test is performed under suitable environmental conditions." l Answer:

The requirement that testing be done under suitable environmental  ;

conditions is covered in part 17A.1.2.7, " Controlled Conditions," of the PSAR. Also, the fourth paragraph of 17.0.1.11 states that-all testing of safety related items will.be performed in accordance with ,

written procedures which establish prerequisites for environmental conditions.

22. There is no PSO requirement for traceability of installation measur- l ing and test equipment to national standards, such as the NBS. The responsibility of the Manager, Quality ~ Assurance, for control of measuring and test equipment needs to be defined in the PSAR. PS0 should specify in the PSAR their method of maintaining test equip-ment traceability in order to allow for reinspection or retesting of equipment inadvertently accepted by PSO with out-of-calibration  !

measuring equipment.

Answer:

The pmgram for the control of installation measuring and test equipment is described in Section 170.1.12 of the PSAR. This section states that ca:ibration standards will be traceable to nationally recognized standards if available, or else the basis of calibration will be preserved. The commitment is also made that inspections and tests performed with measuring and test equipment later found to be.

malfunctioning will be investigated to determine their validity.

The Superintendent, Quality Control, will be responsible for these controls,'and the Manager, Quality Assurance, through the Site QA Superintendent, will verify that the Superintendent, Quality Control is properly carrying out his assigned responsibilities.

23. The procedure described in Paragraph 17A.1.13 of the PSAR is in violation of the requirements of 10' CFR 50, Appendix B, in that the

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- 1 Field Project Director is responsible for the " surveillance and inspection of onsite cleaning, handling, and storage," as well as

" cost and schedule." The Quality Assurance Program responsibilities of the Manager, Quality Assurance, for handling, storage, and ship-ping should.be defined in the PSAR. (Note that the " Field Project Director" is the " Manager, BFS Construction" in Amendment 10 of the PSAR.)

Answer:

At the plant site, the Manager, BFS Construction has overall responsi-bility. The Superintendent, Quality Control, reports directly to the Manager, BFS Construction and is responsible to implement the site quality control program covering activities such as receipt inspection, as well as cleaning, handling, and storage of equipment.

The 10 CFR Part 50 Appendix B requirement for independence from cost l and schedule and the responsibility of the Manager, Quality Assurance l are satisfied as follows: l (a) The Site Quality Assurance Superintendent reports to the Manager, ,

, Quality Assurance. He monitors and audits QA/QC activities at

. thejobsite.

(b) The PSO QA organization reviews and approves all PSO QC procedures (and changes) and is responsible for certifying and approving the training of PSO QC personnel.

24. The nonconforming material control program described by PS0 in the PSAR is not in conformance with the requirements of 10 CFR 50, i Appendix B, Criteria XV. The PSAR does not describe the PSO system "for identification, documentation, segregation, disposition, and notification to affected organizations." Nor does the PSAR describe the P50 system by which " nonconforming items shall be reviewed and accepted, rejected, repaired or reworked in accordance with documented procedures." Finally, P50 management review is required for noncon-formances with "significant cost or schedule impact," while not ,

- requiring routine'P50 management review of major quality problems l

~other than those few cases which are defined as significant deficien- J cies in accordance with 10 CFR 50.55e.

Answer:

The nonconforming material control program at the plant site is ,

acceptably described to meet Appendix B requirements in part l' 170.1.15 of the PSAR. Offsite, most of the nonconformance control

'- program has been delegated to P50's principal contractors and vendors.

PSO retains oversight responsibility, however, in accordance with part .17A.1.15 of the PSAR, and will receive all deviation requests.

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P50 activities.will be performed in accordance with documented procedures contained in the PSO Manuals. Major quality problems are within the scope'of 10 CFR 50.55e.

25. The PSO procedure for reporting significant deficiencies is not in accordance with the requirements of 10 CFR 50.55e. P50 proposes in the PSAR "that each of the prime contractors Quality Assurance pro-grams include provisions for reporting significant deficiencies,"

while 10 CFR 50.55e requires that the " holder of the permit shall notify the Commission of each deficiency found in design and con-struction, which, were it to have remained uncorrected, could have affected adversely the safety of operations." PSO does not require all suppliers of Class I equipment have a field failure feedback and analysis system. Identification of inservice parts failures and an analysis of the cause of the failures is a key ingredient in the corrective action portion of a total Quality. Assurance Program.

Answer:

Amendment 7 of part 17A.1.16 of the PSAR states that procurement documents will require that each contractor's QA program includes provisions for reporting significant deficiencies to PSO and that .

the PSO Manager, Quality Assurance must identify significant deficien-1' cies (per 10 CFR E0.55e) and have them reported to the NRC by the .

Assistant Vice President Nuclear. This includes field failures.

The answer to Item 32 addresses the question of field failure . feedback and analysis in greater detail.

26. PS0 fails to designate who is responsible for the Quality Arsurance records system.

Answer:

Part 170.1.6 of the PSAR states that the Field Document Supervisor under the Superintendent, Administrative Services will be responsible for the site document control center during the construction phase, and Part 17A.1.17 states that copies of all required quality assurance records will ultimately be on file at the plant site.

27. The General Electric Quality Assurance Program, as described in the GESSAR, is deficient in.the implementation portion of four of the eighteen criteria of 10 CFR 50, Appendix B. The specific deficien-cies to be discussed herein include design control, Criteria.III; identification and control of materials, parts, and components, Criteria VIII; control of special processes, Criteria IX; and correc-tive action, Criteria XVI. In addition, the role of the NRC Vendor Inspection Program should be clearly delineated in both the GESSAR, the PSAR, and the applicable Federal regulations.

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Answer:

The general concern' relating to the four criteria are answered in item 2. Specific concerns are expressed and answered individually as items 29 (design control), 30 (identification and control of 1 saterials, parts, components), 31 (special processes), and 32 (cor-rective action).

As noted in the answer to item 3, there is no requirement that the PSAR or GESSAR delineate the role of the NRC Vendor Inspection Program. PSO and GE both describe an acceptable vendor control program with no reliance on the NRC program.

Concerns relating to Federal regulations are properly noted at rule- I making proceedings or in a petition for rulemaking, not at the ASLB hearing for Black Fox.

28. Section 17.0 of GESSAR states that the NRC "will be given timely notification of any significant changes in the Quality Assurance Progrsa." Why doesn't the NRC have to review and approve all changes to the General Electric Quality Assurance Program prior.to their

, incorporation into the General Electric system as ' required in a similar authorization program as described in Section III of the ASME Code? (The remainder of this item is identicaf to items 3 and

4. )

Answer:

The quality assurance program for General Electric is described in NEDO-11209-03A. -The introduction to this document states: "The NRC will be given notification of programmatic changes in the QA Program as described herein, prior to implementation, except changes which do not decreasa'the effectiveness of the' program, or those that reflect organization changes. Organizational changes which affect the QA Program will be reported to the NRC within 30 days after implementation. This QA Program description will be reviewed at least annually and revised, if necessary, to reflect programmatic and organizational changes."

This is acceptable to the Staff.

29. The design control description in GESSAR is inadequate in that insufficient quantitative data is included to verify the adequacy of the General Electric. design verification program. The BWR-6 design family of plants includes many unproven designs which require develop-ment and prototype design. verification testing. Examples include the Mark III containment, flow-induced vibration of the reactor

' internals, stress corrosion of 304 SS piping and control rod collets, j - -e-

emergency core cooling system hydrodynamic characteristics, solid state protection systems, and multiplexing of control rod position from electronics located within the containment's hostile environ-ment. The GESSAR should be modified to include a list of all design verification programs, with a description of the extent of prototype testing, and with the applicable test accept / reject criteria. All too often in'the past the NRC has allowed reactor vendors to do too *

- little prototype testing. Then the NRC is faced with the dilemma of design inadequacies discovered in plant operation after the plant has been made radioactive.

Answer:

The General Electric program for design verification is described in Part 3.9 of NEDO-11209-03A. In the area of design verification, Appendix B to 10 CFR Part 50 requires that~ design control measures provide for such verification. The controls listed in NED0-11209-03A were compared to the Commission's Standard Review Plan and found to meet the Commission's requirements. For example, the commitment is made that when qualification testing is used as the sole means of verifying design adequacy that a prototype or initial production unit is tested under the most' adverse design conditions for the unit 4 being tested. .

There is no requirement that the quality assurance program in the PSAR list the design verification programs, describe the extent of prototype testing, or list accept / reject criteria. These are technical requirements which are not specified by quality assurance personnel.

GESSAR includes a list of verification test programs which are con-firmatory in nature. The Commission's technical staff has reviewed these tests and found them to be sufficient.

30. The Quality Assurance Program description in the GESSAR does not specifically describe the program necessary to meet the requirements of 10 CFR 50, Appendix B, Criteria VIII. For example, the General Electric program commitment to maintain traceability of cables in the floor sections.of ,the Power Generation Control Complex (PGCC) is not described. In the past, cables in PGCC floor sections have not been traceable to'the reel nor has test data been maintained for t

each reel. A lack of traceability of wire and cable in an electrical system is equivalent to a loss of control of weld rod in a mechanical component. Section 17C8 of the GESSAR should be revised to specifi-cally describe the Quality Assurance Program-iw lementation for the identification and control including traceability of materials, parts,'and components.

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i Answer: I The answer to concern 10, describing the GE program for the identifica-tion and control of materials, also answers this concern.

31. The General Electric system to define, select and control special processes and special process operators, other than ASME code-related processes, is not adequately described in Section IX of the GESSAR. The description included in the GESSAR does not clearly describe the General Electric Quality Assurance Program to " assure that special processes... are controlled and accomplished by quali-fied personnel using qualified procedures" in accordance with the requirements of 10 CFR 50, Appendix 8, Criteria IX. Examples of potential special processes which should be described in the GESSAR include uranium plating of neutron sensors, hydrogen brazing of neutron sensors, epoxy mixing and cering of seals of containment electrical penetrations fabrication of LPRM' seals, soldering of circuit boards, and crimping and terminating of control cables. The .

GESSAR should be revised to include a description of the control of  !

special processes for non-code processes in accordance with the intent of 10 CFR 50, Appendix B, Criteria IX. A similar commitment and desc'ription should be included for Class IE electrical modules purchased by General Electric such as pressure trecsmitters, process

\ instruments, and relays.

Answer:

The requirements of Criterion IX of Appendix B to 10 CFR Part 50 relating to the control of special processes are given in the answer ,

to concern 19. Related GE commitments, found on page 57 of NEDO-11209-03A, are as follows:

" Procedures and practices are established and documented to provide i assurance that special processes such as welding, heat treating, l cleaning, and nondestructive examination are accomplished under con-trolled conditions. These special processes are accomplished in i accordance with applicable codes, standards, regulations, specifica-tions, design criteria, and other special requirements using qualified personnel, procedures, and equipment.

" Requirements for special processes such as welding, heat treating and nondestructive examinations are specified on design documentation by the engineer responsible for the design. Appropriate inspections and tests to assure control of these special processes are designated in test and inspection procedures.

" Manufacturing and Quality Assurance personnel, procedures and equipment involved or utilized in the . execution or control of special

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processes are qualified to the requirements of applicable codes and standards. Documented evidence of the validity of such qualifications is maintained for all such personnel, procedures and equipment in accordance with applicable codes and standards. Each cognizant QA organization has the responsibility of periodically monitoring these records to assure the continued validity of all such qualifications.

For special processes not covered by existing codes or standards, or where product requir'ements exceed the requirements of established codes or standards, the necessary qualifications of personnel, procedures, or equipment are defined in appropriate documents."

Although examples of special processes are listed in the commit-ments, they are examples only, and the commitments apply to all special processes including those listed in the concern. Also, the commitments apply whether the special process is performed within GE or by a supplier to GE.

The description of GE's program for the control of special processes is acceptable to the staff.

32. Identification of all inservice parts failures and an analysis of the cause of the faTTures is a key ingredient in a corrective action program to upgrade plant designs in accordance with the requirements of 10 CFR 50, Appendix B, Criteria XVI. The General Electric prograr of field feedback, as described in the GESSAR, does not apply to all failures.

In addition, the General Electric corrective action system can be only.as complete as the failure data transmitted by PSO to General Electric. Section 16A of Chapter 17 of the P50 portion of the PSAR needs to be revised to include the PS0 commitment to provide field failure information to all suppliers of Class I equipment. Adequate exchange of field failure data between the applicant and the suppliers is an absolutely essential element in a quantitative measurement system designed to provide corrective action as described in 10 CFR 50, Appendix B, Criteria XVI. ,

Answer:

P50 will not be required to require all suppliers of Class I equip-ment to have a "ield failure feedback and analysis system because this is not a specific criterion of NRC regulations in 10 CFR 50, Appendix 8. However, the need for field failure feedback and analyses has been recognized and is being met in several ways.

First, there is 10 CFR 50.55(e) which requires the NRC receive notification of each deficiency found in design and construction which, if it w'ere to remain uncorrected, could have affected adversely

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i tha safety of operations of the. nuclear power plant at any time throughout the expected lifetime of the plant. During operations, 10 CFR 50.71 and the technical specifications require Licensee Event Reports (LER) which are diseminated quarterly by IE. Finally, 10 CFR Part 21, " Reporting Defects and Noncompliance," is now in effect. This regulation requires directors and responsible officers of firms and organizations building, operating, or owning a nuclear plant or supplying safety-related components (including safety-related design, testing, inspection, and consulting services) to report defects in components which may result in a substantial safety hazard. ,

J Deficiencies which are not generic are handled by the Commission on a case-by-case basis. Potential generic problems are identified by NRC'and can result in circulars or bulletins being sent to licensees  ;

for their information/ action. In addition, abnormal occurrences  !

i (unscheduled incidents or events which the Commission determines are significant from the standpoint of public health and safety) are published in the Federal Register and in an annual report to Congress.

GE's method of field failure analysis is described in Section 16 of

- ' NEDO-1109-03A (p. 71/72) which states: " Continuous surveillance of operating BWR performance is maintained. Periodic contact by NED l

BWR Services personnel with plant operations personnel provides .  :

detailed information ori statistical performance of the plant as well as narrative reports of equipment malfunctions or failures. Automatic data handling systems record and analyze the statistical information.

This information is analyzed by the cognizant design components and carefully aonitored for significant or generic equipment weaknesses.

This provides feedback to the responsible design component to avoid repetition of the same problems. Prompt communication to all opera-tors of similar equipment is provided whenever significant or avoid-able problems are discovered."

Based on our review, we conclude that the quality assurance program l description in the Black Fox PSAR does comply with the Commission's l j

quality assurance requirements specified in Appendix B to 10'CFR Part 50, that each sof the intervenors' concerns has been adequately answered, and that the quality assurance program description is acceptable for the l l

design, procurement, and construction of the Black Fox Station.

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1/78 J. G. Spraul PROFESSIONAL QUALIFICATIONS QUALITY ASSURANCE BRANCH OFFICE OF NUCLEAR REACTOR REGULATION I am a nuclear engineer in the Quality Assurance Branch in the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission. In this position I am responsible for the Commission's review and evaluation of ,

the quality assurance programs proposed for the design, construction, and operation of nuclear power plants.

I received a Bachelor Chemical Engineering degree from the Georgia Institute

'of Technology in 1951. In 1971 I completed the requirements for the Professional Designation-in Quality Control from the University of California, Los Angeles. My nuclear experience includes 2 years of engineering work in gasesus diffusion with the Goodyear Atomic Corporation and 12 years of l nuclear fuel and nuclear power plant component design, manufacture, and test with the Atomics International Division of Rockwell International.

My quality assurance experience includes 2 years as Chief Inspector and 4 years as Director of Quality Assurance at Atomics International, respon-sible for managing the entire quality assurance program.

I joined the Quality Assurance Branch of the Commis'sion in 1974. Since

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joining the NRC, I have reviewed the quality assurance program of 14 l nuclear power plants as well as 12 topical reports on quality assurance from utilities, architect-engineers, NSS suppliers, and constructors. ,

i I am a member of.the American Nuclear Society and a senior member of the American Society for Quality Control. In 1972 I was certified as a Quality . Engineer by the American Sa'ciety for Quality Control. This certification was renewed in 1977. _.

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I have read the foregoing affidavit and swear that it is true and accurate to the best of my knowledge.

. , L-?% . .

J. G. Spraul Sworn before me this k day of M , 1978

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Contention 12 _

I I. Contention Intervenors Contention 12 relates to the issue of compliance of the Black Fox Station spent fuel pool design and construction with the requirements of Criterion 61 of Appendix A to 10 C.F.R. Part 50. The enclosed affidavit of Robert J. Giardina discusses whether material issues of fact remain to be resolved in this area.

II. Argument Criterion 61 requires that fuel storage and waste handlino systems be

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designed to: ,

1. Provide *adeqttate safety under normal and postulated accident ,

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conditions.

2. Have suitable shielding for radiatiori protection.

3. Have appropriate containment, confinement, and filtering systems. l

4. Have a reliable and testable residual heat removal system for ,

decay and other heat removal.

l 5. Prevent a significant reduction in fuel storage coolant inventory under accident conditions.-

As indicated by the Giardina affidavit, the BFS plans to build and operate a standard GESSAR spent fuel pool. That fuel pool will be required to be located in a Seismic Category I tornado and missile-proof controlled leakage building in order to have suitable shielding for the oublic and appropriate containment and filtering systems. The spent fuel pool itself

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+4 e 12-2 will have a Seismic Category I cooling and make-up water system to ensure that a significant cooling inventory remains under all conditions to remove decay and other heat fran the spent fuel which wif t be stored in the pool.

Giardina affidavit at 1-2.

The Staff's review of the GESSAR spent fuel pool desien (and thus the l Proposed Black Fox design) relied upon Reg. Guides 1.13, 1,29, and l

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standard Review Plans 9.1.2, 9.1.3 and 9.4.2 to assure the adequacy of j

fuel pool design, seismicity, spent fuel pool cooling and air ventilation.

In addition, the Plant was review under Reg. Guide 1.117 on Tornado f Design and found to comply with the Commission's regulations. Giardina Affidavit j s

at 2. All of the items listed above then combide to assure that the BFS spent fuel pool when constructed, will be designed to assure adequate safety under normal and postulated accident conditions.

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That a material issue of fact does not remain to be litigated in this area is illustrated by the discovery conducted so far in this case.

In their November 30, 1976 Answers to Staff Interrogatories, Intervenors, when asked for the facts behind their spent fuel pool contention, made no reply to the Staff inquiry. In the May 31, 1978 deposition of Intervenors' proposed witness for contention 12, tir. Dale Bridenbaugh,while expressing a general concern for spent fuel pool design and removal of heat, had not reviewed the spent fuel pool design in the intervening two years and had no i

facts within his profession which showed that fuel spacing or design would be a problem at SFS. Tr. 122. _.

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III. Statement of Material Facts As To Which There Are No Genuine Issues Based on the information listed above, the NRC Staff believes no controversy exists as to:

1. The fact that a Staff review has shown that the BFS spent fuel pool complies with Criterion 61 to Appendix A of 10 C.F.R. Part 50 by reason of the Seismic Category I design of the fuel building and the Seismic Category I design of the spent fuel pool which will be 1ccated therein.

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. UNITED STATES OF AMERICA  !

NUCLEAR REGULATORY CCMMISSION EEFORE TI4E ATCMIC SAFEIT AND LICENSING 30ARD 1

In the Mattar of )

Pubite .Sarvice Conpany of a Oklakma ] Occket Numbers: Sa-536 I 50 6 (Black Fox Station,) l -

Unit Nos. I and Z J 2

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AFFIDAVIT OF ROBERT J. GIARDINA ON CONTENTION 12 l

1. My nama is Robart J. Giardina. I an employed by tha Nuclear Regulator /

Comtsston as a Reactor Engineer fn the AuxfTf arf Systems Eranch. I have been empToyed in this positiarr sinca Septanbar,1974. My qualfff-cations are listad on the attached. Qualification,5heet.

2. The purposa of my affidavit is to address the inter /enor C3ntantion '

i Number 17. regarding adequacy of the design and const$1ction of the alack Fox, Unit Nos. I and Z spent fuel pool. Contention 12 relatas to the compitece of the Black Fox spent fuel pool design and. construction with  !

the requirements in Appendix A, Critarfon 61 te 1G CFR Part 50. My affidavit wfTT show that the design and construction of the Elack Fox spent fuel pool is acceptable and will meet the requirements of la CFR Part 50, Appendix A,. Criterion 61.

3. The intarteners contand that the appitcant has nec adequataly demort-scrated that the desigr and construe: Ton of Black Fox,. Unf t Nos. I and 2 spent fuel pools will comply wf tt 10 CFR ? art 50, AopandD: 1, F.-itarion 61 .

4. The Black Fox spent fuel storage facility, except for the spent fuel area ventilation system, was reviewed and evaluated under the GESSAR-BWR-6 Docket, as part of NRC's standardization -directives. As will be shown below, the spent fuel pools will be designed and constructed so  :

that the requirements of Appendix A, Criterion 61 to 10 CFR Part 50 are met.

5. The Comission's criteria for evaluating the adequacy of the design and ,

construction of spent fuel pools is contained in 10 CFR Part 50, f Appendix A Criterion 61. Additional guidelines for acceptance are given in Regulatory Guides 1.13, " Spent Fuel Storage Facility Design

. Basis;" 1.29, " Seismic Design Clast.ification;" and 1.117, " Tornado Design ,

l Classification." Standard Review Plans (SRP) 9.1.2, " Spent Fuel Storage;"

9.1.3, " Spent Fuel Pool Cooling and Cleanup System;" and 9.4.2, " Spent l Fuel Area Ventilation Systant," are used to evaluate spent fuel pool designs.

6. The spent fuel storage facility for GESSAR was found to meet the guidelines given in the Reguiatory Guides and Standard Review Plans in that the pool would be a seismic Category I (designed for the safe shutdown earthquake) structure, which would be located in a seismic Category I, tornado and missile-proof, controlled leakage building. The spent tuel pool cooling ,

system will provide adequate cooling including a seismic Category I  !

make-up system. Since these met our design guidelines, the spent fuel  !

storage facility was found' adequately designed and acceptable for l l

licensing. j

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7. The spent fuel pool area ventilation sy tem (SFPAVS) was reviewed under the Black Fox Docket in accordance with SRP 9.4.2 and was found to be acceptable. The systen provides a seismic exhuust system and redundant isolation valves at the fuel building containment Loundary and'will main- ,

tain the building at a negative pressure during all operating conditions.

During accident. conditions, the SPRAVS is automatically isolated, and the ventilation and negative pressure are maintained at the stand.-by gas treatment system (SGTS). The filtration of the radioactive materials

- is accomplished at the filter trains of the SGTS. This is a GESSAR reviewed item which was found acceptable. The design met our requirements and, thus, was acciptable.

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8. In addition to the above, the applicant in his Preliminary Safety Analysis Report, has also committed to meet the criteria contained in 10 CFR Part 50, Appendix A, Criterion 61, as well as comply with the guidelines of Regulatory Guides 1.13,1.29, and 1.117, as pertaining to the spent fuel storage facility. Thus, any modifications, including GESSAR modifications, made to the spent fuel storage facility by the applicant, will have to meet these design criteria for approval. .

9. Therefore, the staff sees no material facts in issue to be disposed of in Contention Number 12.

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i I have read the foregoing affidavit and swear that it is- true l and accurata to tne best of my knowledge l J . l.

Rocart d. qCaiardina Subscribed and Sworn to Bef1re me this S'~ Cast af %% /9 7J-f f

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/ l Notary Puolic My Comission Expires ,/e (7P -

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Robert J. Giardina Professional Qualifications

• AuxfTfary Systemn Branch Ofvisfort of Systems Saferf Offica of Nuclear Reactor Regulatfort I an a Reactor Enginear frr the Auxfitary Systams Branch in the Offfca of Nuclear Reactor Regulatton, U. S. Nuclear Regulatory Comissferr. In this posttion I perform technical reviews, analyses, and evaluations of reactor plant features pursuant to the construction and operation of reactors.

Educatfon I received a Bachelor of ScTence Degree in Mechanical Engineering from.

Drexel Urifversitf- in 197T. In 1974 E recafved from Drexel Universtef a Mastar of S'cTence Degree irr Mechanicai Engineering: with specializatforr in l' .

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the area. of ThermeT and. FTufd Sciences and a Master of Science Degree in Engineering Management with specialf:atfort in the area of Research and Development and Engineering Management (Corporate avel). ' The research paper for the Engineering Management Design was entitled, " Technology Assusment.

Since 1974 I have taken- a number of courses. on PWR and SWR System Operation,. Reactor Safety, Syster ReifabfTity,. Fault Tree AnaTysts, and. Ffre Frotactiert.

Experience My expertence includes sign:yearr of Engineering irr the design, manufacture, and tasting of Shipboard Mechanical Systams and Comconents at the Phfifadelphia Naval Shipyard.

T"nese systems and components included cooling star sys:ams, propulsion systams, fire protac:fon systems, hydraulic systems, and verr:tla-tion systems + -

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.s . . < T I joined the Auxfitary Systems Branch of the Comisifon frr 1974. Since joining the Comfrzfon, I have performed safety evaluations on spent fuel

. pool expans i ons f or five plants. as we1T as pmvfdad input to or revised

• • ofvision of Operating Reactors posttfon on spent fuel poor expansferr,

. The Enytmnmental Impac?. Statament on Spent Fuel Storage, the Regulations on Independent Spent FueT Storage Facfif tfes (.10 CFR Part 72) and Regulatory Guides and Standard Review PTans on Spent Fuel Storage (on I

and off-site). I have been actively involved in the investigation of the steam generator feed water hanner problem and was a member of the I am presently reviewing or have Consission's Water Hammer Task Force.

reviewed and evaluated the Auxiliar/ Systems of five nuclear power plants )

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.as well as coordinated fire protaction reviews for thres nuclear power p ants.

I have successfuTTy passed the tast for- and was awarded the certiffcate of Engineering-In-Trafning.

Orcanfrationai Membershfes: l I am a member of Pt Tau Sigma - National" Honorary Mechanical Engineering Fraternity and of the Americarr Management Associattons I an an Associate Member of the Amerfcan Society of Mechanical Engineers. E an an active member in the Boy Scouts of America. and. a Erotherhood. Member in the Order of the Arrow, of that organizatton,

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' Professional-Qua11fications James 0. Behn Gage-Babecek &. Asscciates, Inc.

My namects James 0. Bahn. I am employed as a staff enginear at Gage-Babcmci and Associatas, Inc.,135 Addfron Avenua, Elmhurst, Illinofs.

Gage !!abecck 7.nd Assocf atas is undar centract with Sandia Laboratsrfes j to prc'vfde fire protaction consultation to the Auxf1tarf Systams Branch of tha NRC. I am assist #ng tha Auxiliary Systams Branch in ruvfewing -

tha fire protactfen pregnms of nuclear power plants of which Black Fcx, Unit. Nos. I and Z is one. Tna Auxiliarf Systams Eranch is responsibia for reviewing reactor ifcansa applications and evaluating the desfgn of auxfifarf systams, including. the fire pretacticn systams, of the nu sear-  ;

power plant with rs.spect to nuclear safety. '

I attandad Illinots Instituta of Technology in Chicaso, Illinots, and received a S S. Cagree in Firs Protacticn Enginaaring in 1970. I hava also attanded various industrial seminars dealfng mostly with firs safaty. i I'have a t:tal afght years of work expertenca wi:h major projects as stated below.

I was first amolayed at Underwritars Labontarf in Chicago during the summers of my junior and sanfor years at IIT, :ssting firs resisanca ratings of varfous walls, doors, columns, acc.

Frsm 1970 :s 1.'r/1, I was amoloyed at Marsh anc McLannan. fn Chica;c dcicg' l bastcaliy ffr1 pre:ac:ien inseec icn cf yaricus iacus::1al si:as in :ha l 1

Chicago arsa, including fac: cries, cf'icas, hcspi:als, fcssfi fualac 1

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power piants;:and talaphone comunication cantars, and testing their insta11ad fire protaction systams.

Frsa 1971 to 1972, I worked for me. Automatic Electric, a comunication and talegraphic manufacturer, as a Fire Protaction aad Safety Engineer.

My dutfas Included all aspects of fica protaction and safety as they relate J to an industrial occupancy. .. ..

Dur.ing a short perfod of time tn 1972, I was also empicyed at Wartam Electric Comgany, manufacturer of Ball equipment doing basically as my previous employment.

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In 1972, I startad working for Gaga-Babcock and Associatas, as a fira protaction and safety consultant. My major activities have been in the area of (1) surieying and avaluating axisting fire protac: Ten systems for various industrial facilftf as to datamine their adequacy and etc ...and necassary improvements and (2) dastgn of ffva detaction and suppresston systems associatad with (1) above.

1 I am presantly providing fire protection consulting to the Auxf1tarf Systams 3rsnch of the NRC.

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Contention 13 1

I. Contention ,

Intervenors contend that Applicant's preliminary emergency plan does not adequately comply with 10 C.F.R., Part 50, Appendix E, in that:

a. It does not-aiequately describe what contacts and arrangements have been and will be made with government agencies;

h. It does not adequately describe the matters required by Appendix E,PartII(C),(D),(E),(F),&(G);

c. It does not adequately describe.the matters required by Appendix E, Part II(B) in that there are no state or local plans for coping with emergencies arising out of our connected with a radioactive related emergency;

.- d '. It does not adequately comply with Appendix E, Part II(A) in that only one person (by job classification) will be in charpe of notification.

The failure to have backup or subordinate responsibility would result in the plan being unable to cope with emergencies; and

e. .It does not include adequate plans to evacuate the site as the result of an explosion of a barge carrying explosives on the Verdigris River.

II. Argument 10 C.F.R. Part 50, Appendix.E describes minimum requirements for the l I

Applicants' Preliminary Safety Analysis Report in order to assure compatibility of the proposed emergency plans with facility design features, site layout I

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13-2 and site location with respect to such considerations as access routes, surrounding population distributions and land use. As demonstrated in the NRC Staff affidavit, the Applicant has met each of the established minimum requirements.

Thus the Staff. believes the Applicant's preliminary emergency plan adequately complies with 10 CFR Part 50, Appendix E in that: (a)itadequately describes what contacts and arrangements have been and will' be made with government agencies; (b) it adequately describes the matters required by Appendix E. Part II(C), (D), (E), (F), and (G); (c) there is no requirement that the Applicant describe state or local plans for coping

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with emergencies arising out of or connected with a radioactive related ,

l emergency'at this preliminary licensing stage; (d) it adequately complies with Appendix E, Part II(A) in that it provides a description of the organization for coping with emergencies, and the means for notification in the event of an emergency, or persons assigned to the emergency organiza- l tion; and (e) it includes adequate. plans to evacuate the site as the result of an explosion of a barge carrying explosives on the Verdigris River. l C.R. Van Neil Affidavit, Passim.

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13-3 III. Statement of Material Facts As To Which There Are No Genuine Issues

1. The BFS preliminary evacuation pian adequately address the contacts to be made with local governments and adequately describes the requirements of Part II(C), (D), (E), (F), and (G).

2. There is no requirement in Appendix E at the CP stage that requires a description of state or local plans for coping with emergencies.

3. The BFS preliminary evacuation plan adequately describes the 7

Applicants' organization which is designed to cope with emergencies as required by Part II(A) of Appendix E to 10 CFR, Part 50.

4. That the evacuation plan, which provides for a nuclear accident would be adequate to evacuate the site were a barge to explode on the Verdigris River.

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! ' UNITED STAT 55 0F AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of the Application of Public Service Company of Oklahome locket Nos' Associated Electric Cooperative, In'c.

oTN 50-556 50-5U Western Famers Electric woperative (Black Fox Units 1 and 2)

AFFIDAVIT OF C. R. Van Niel l ON CONTENTION 13 .

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1. My name is C. Richard Van Niel . I am employed by the U. S. Nuclear Regulatory Comission as a Reactor Safety Engineer in the Emergency Planning Branch, Division of Project Management. I have been employed in this position since October 19Ti . .My pimfassional qualifications are described on the attached

< sheets.

2. The purpose of my affidavit is to address intervenor Contention 13 regarding the adequacy of the applicant's preliminary emergency plan. Contention 13 stater that the applicant's preliminary emergency plan does not adequately comply with 10 CFR Part 50, Appendix E in one general area ano in four specific instances. My affidavit will show that the applicant has submitted sufficient infomation to adequately meet the requirements of 10 CFR 50, 1 Appendix E. I will address each specific part of the contention separately. l 1

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CONTENTION 13a l

The intervenors contend that "the applicant's proliminary emergency plan does not adequately comply with 10 CFR Part 50, Appendix E in that:

a. It does not adequately describe what contacts and agreements have been and will be made with government agencies."

The applicant, in Section 13.3 of the PSAR has provided information concerning the preliminary contacts and arrangements made or to be made with local, State, and Federal governmental agencies with responsibility for coping with emergencies.

The applicant states. that initial contacts for assistance in the event of an emergency have been made with the Oklahoma State Department of Health. Arrange-ments will be made with the United States Nuclear Regulatory Coninission, Office l I

of Inspection and Enforcement and the United States Energy Research and Development Office. Administration (now the Department of Energy), Alb j

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in conjunction.with the preparation of the final Emergency Plan, to be submitted in the Final Safety Analysis Report. The Oklahoma State Department of Health has been designated by the Governor of the State of Oklahoma as the principal coordinating agency for radiological emergency response. The applicant has made contact with the Oklahoma State Department of Health, Radiation Protection Division to discuss the development of the Black Fox Station Emergency Plan i and the coordination of the State with other offsite agencies.

The infonnation described above has been judged by the staff to be adequate to meet the' requirements of 10 CFR 50, Appendix E at this stage of the licensing process. We will require more formal contacts, including letters of agreement, i

in connection with the operating license review.

CONTENTION 13b Intervenors contend that " applicant's preliminary emergency plan does not adequately comply with 10 CFR Part 50, Appendix E in that: b. It does not adequately describe the matters required by Appendix E, Part II (C), (D), (E),

(F), and (G)."

Evaluation Report for the Black Fox Station Units 1 and 2, dated The June 1977 (NUREG-0190) .at Sections 12 and 13 detail the staff's evaluation Safety finding and shows- the applicant's conformance with the specific sections of Appendix E Part II. This information is located as follows:

10CFR50 Appendix E Part Safety Evaluation Report II C Section 13.3, paragraphs 2, 3, 5, and 7 II D Section 12.6, paragraph 7, and Section 13.3, paragraph 7 II E ,

Section 13.3, paragraph 7 II F Section 13.2, paragraphs 2 and 4, and Section 13.3, paragraph 8 -

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10CFR50 Appendix E  ;

Part Safety Evaluation Report II G Section 13.3, paragraph 3, 5, and 9 CONTENTION 13c Intervenors contend that " applicant's preliminary emergency plan does not adequately comply with 10 CFR Part 50 Appendix E in that: c. It does not adequately describe the matters required by Appendix E, Part II (B) in that there are no State or local plans for coping with emergencies arising out of or connected with a radioactive related emergency."

The Staff agrees that the PSAR does not describe " state or local plans for coping with emergencies arising out of or connected with a radioactive related emergency." 10 CFR 50, Appendix E Part II B does not require such a description.

CONTENTION 13d j Intervenors contend that " applicant's preliminary emergency plan does not adequately comply with.10 CFR Part 50 Appendix E in that : d. It does not adequately job classificationcomply will w)ith be in Appendix E. Part II (A) charge of notification. The in failure that only one person (

to have backing or subordinate responsibility would result in the . plan being unable to cope with emergencies."

On the contrary, the applicant has made provisions in several avus for backup assistance in emergency situations. This information is presented in Sections 13.1 and 13.3 of the PSAR. The applicant states that, in the event 4 of an emergency, the Shift Supervisor will act as the initial Emergency o Coordinator and will take immediate correctivo action. He will maintain consnunications with the control room and offsite agencies. He will be relieved by the ranking member of station management onsite (the Station Manager, Station Superintendent, or Operations Supervisor) who will be responsible for overall activation of the Emergency Plan. An Emergency Consnittee consisting of the Station Manager, Station Superintendent, Operations Supervisor, Technical Supervisor, Radiation Safety and Chemistry Supervisor, and the Security and Office Supervisor, is activated to evaluate the emergency and to recommend further action. The Station Manager is responsible for the dissemination of public infonnation and contacting local, State, Federal and Public Service Company of Oklahoma officials.

The Shift Supervisor will be licensed as a Senior Reactor Operator and will be supported on shift by at least two additional licensed operators. Written I emergency procedures will prescribe operator actions for emergency situations.

These actions' include placing the reactor in a safe shutdown condition, giving evacuation instructions over the public address system (if required),

notifying offsite personnel, and performing other duties to implement the l emergency plan. .

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This description by the applicant of the chain of connand for taking emergency actions, including notification, provides a sufficient factual basis to meet the requirements of 10 CFR 50, Appendix E, Part II (A).

CONTENTION 13e Intervenors contend that " applicant's preliminary emergency plan does not adequately comply with 10 CFR Part 50, Appendix E in that: e. It does not include adequate plans to ' evacuate the site as the result of an explosion of a barge carrying explosives on the Verdigris River."

The applicant has presented sufficient information to assure the compatibility of the proposed emergency plans with the features of the facility for plant evacuation resulting from any of a number of postulated emergencies, including fire, explosions, radiation leaks or contamination, and natural disasters.

This information is presented in Section 13.3 of the PSAR.

The applicant states that local alanns will provide evacuation signals for buildings and areas around the site. A signal over the public address system (initiated by the control room operator) will serve as an evacuation alarm for the entire site. Evacuation will be directed from the reactor control reom '

which is the primary Emergency Control Center and is designed for_ continuous occupancy under emergency conditions. Assembly points will be established where personnel acc6untability will be performed. The plant access road and State Highways 33 and 88 provide routes for evacuation frzm the site.

This description by the applicant of the proposed actions for plant evacuation provides a sufficient factual basis at this preliminary planning stage to meet the requirements of 10 CFR Part 50, Appendix E.

Earlier testimony by the staff and applicant on Contention 14 showed that the probability of a fertilizer barge explosion is extremely remote and that if one were to occur, it would not adversely effect the operation of 1 the plant at the site (see page 29'of the "NRC Staff Proposed Findings of Fact and Conclusions of Law on Environmental and Site Suitability Matters," l datedJanuary9,1978.) t l

I have read the foregoing affidavit and I swear that it is true and accurate to the best of my knowledge.

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C. Ricnard Van Niel j Sworn before me this m

1. ~' day of a% 1978 ,

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'C'. :R'ICliNRD' VAN NIEL EMERGENCY PLANNING BRANCH DIVISION OF PROJECT MANAGEMENT U.S. NUCLEAR REGULATORY C0lHISSION PROFESSIONAL QUALIFICATIONS i

I presently serve as a professional staff member of the Emergency Planning Branch, Division of Project Management. I review appropriate sections of preliminary and final safety analysis reports for nuclear power plants to evaluate ,the safety-retated aspects of an applicants' emergency preparedness activities. In addition, I assist in the preparation of guidance developed to establish explicit and unifom acceptance criteria for subjects required to be addressed by license applicants relating to emergency preparedness matters. I participate in the development and review of standards, codes and guides.related to operational and industrial safety, and emergency preparedness at licensed facilities.

I attended the University of Rochester and received a Bachelor of Science degree in Mechanical Engineering in 1955. Upon college graduation, I accepted a position as an engineer with the E. I. duPont Company Terre Haute, Indiana.

From November 1955 to March 1959 I was on active duty with the U.S. Navy. <

l In 1959 I returned to E. I. duPont as an Engineer in the Reactor Technology  ;

-Section of the Works Technical Department at the Savannah River Plant, Aiken, South Carolina. In this position, I (1) analyzed special components to be charged to the reactors to insure compatibility and safety, and redesigned reactor components to improve production, safety and economy, (2) acted as liaison and consultant for technical groups conducting experiments in the reactors, (3) followed the operation of the production reactors and made recommendations to increase production and improve reactor safety, (4)

~ followed the installation of new equipment and modifications to the reactor building and to the existing equipment, and (5) prepared procedures for

. equipment checkout and operation, and reviewed and revised existing operating i procedures and safety rules. I was promoted to Shift Supervisor in the Reactor l Department in 1963. In this capacity, I supervised the operation of a nuclear production reactor, assigned and directed reactor and auxiliary operators in their duties, conducted on-the-job training, made recommendations to improve reactor safety, reviewed and revised operating procedures, analyzed malfunctions and authorized repair of equipment and instrumentation. In early 1968.I was transferred to the Raw Msterial Department where I revised and updated existing operating'and emergency procedures, and ' revised building and equipment safety and criticality rules.

In the summer of 1968 I accepted a position as Nuclear Engineer and a year later was promoted to Reactor Safety Specialist with the U.S. Atomic Energy Comission, Washington, D. C. I served from 1968 to 1971 as a professional staff member under the Assistant Director for Nuclear Facilities, Division of 1

l Operational Safety. In this position, I (1) provided technical staff i assistance in the planning, promoting, and coordinating of the health and safety aspects as related to AEC-owned nuclear reactors, (2) conducted appraisals of AEC Field Office reactor safety and emergency preparedness

' activities, to make recommendations. to Field Office management on the conduct

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Professional Qualifications ~

C. Richard Van Niel

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and performance of nuclear safety surveillance activities, (3) evaluated AEC Contractor nuclear safety activities by determining through personal l visits, if pol.icies, guides and. operating practices were being followed. l and (4) recommended new or revised nuclear safety guides, codes, standards, and requirements for governing the conduct of the Division's reactor safety i and emergency preparedness programs.

In October 1971 I transferred to that part of the AEC responsible for the regulation of privately-owned nucleer facilities, and joined the Operational Safety Branch. The Branch was reorganized in April 1974 to the Industrial Security and Emergency Planning Branch, and again in April 1977 to the Emergency Planning Branch, to place increased emphasis in the area of emergency planning. During a brief period (September 1976 to May 1977) I served in the Reactor Safeguards Development Branch, Division of Operating Reactors, where I assisted in the development, evaluation, and coordiantion of reactor safeguards programs and policies.

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15 -

Contention 15 I. Contention Intervenors contend that the Applicant has not adequately demonstrated that Black Fox 1 and 2 will meet the requirements of 10 CFR 50, Appendix  !

I A, Criteria (sic) 31 with respect to utilization of materials and/or procedures which will minimize the probability of intergranular stress corrosion cracking of stainless steel piping at Black Fox 1 and 2.

II. A'rgument_

Criterion 31 of the General Design Criteria is identified in Appendix A to 10 CFR Part 50 and reads as follows: .

Criterion 31- Fracture Prevention of Reactor Coolant and Pressure Boundary.. The. reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, l l

maintenance, testing, and postulated accident conditions (1) the boundary behaves in a non-brittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperature and other conditions of the boundary material under operating, maintenance, testinp, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3). residual, steady-state and transient stresses, ard (4) size of flaws.

Some confusion exists as to Contention 15 since'the contention itself

. states that Intervenors are concerned about the possibility of intergranular stress corrosion cracking of stainless steel piping at Black Fox and cite General Design Criterion 31.' As the enclosed Kane affidavit explains, )

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stainless steel piping is not generally susceptible to brittle fracture l

even in a sensitized condition as set forth in Criterion 31, but instead may be susceptible to intergranular stress corrosion cracking.E The problem of stainless steel cracking (covered in Section 5.2.6.2 of Appendix A to the BFS SER), Mr. Kane relates, is not relevant to brittle fracture (which is a concern for ferritic materials) but is concerned with a depletion of chromium in the grain boundaries of stainless steel which may be caused during the manufacturing of the piping or during subsequent fabrication processes. This precipitation and resulting grain boundary deficienc,y is referred to as " sensitization". Kane Affidavit at S.

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The long range Staff and General. Electric plan of controlling stainless steel piping corrosion cracking wil.1 be through the replacement of stainless steel materials with the materials which have been tested and found to either not be susceptible to intergranular stress corrosion,or through the use.of a qualified fabrication techniques which will prevent sensitized uns,tabilized wrought, austenitic stainless steel from being exposed to reactor coolant. Kane affidavit at 9.

M rittle B fracture of ferritic materials at BFS wil.1 be controlled by pressure-temperature limits to be imposed on BFS via technical specifications at the 01. stage. This-matter is discussed in SER Section 5.3.2.4,~

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Appendix A. Kane affidavit at 4.

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15-3 In de short tem, the General Electric Company proposes, and the Applicant and Staff agrea , to employ service proven materials and metallurgical structures made nonsusceptible to intergranular stress corrosion cracking through special processing steps, thereby avoiding the risk associated with innediate changes to new materials which may not have been test proven. These special processing steps to eliminate intergranular stress corrosion cracking are:

1. Solution heat treatment

2. Corrosion resistant cladding

3. Wald heat input control

4. Reduced inside (diameter) temperature welding. r r .

Kane affidavit at 9-10 In sunnary, the NRC Staff concludes that because authentic stainless steel

- is not susceptible to brittle fracture and adequate short tem and long tem solutions to brittle fracture and stress corrosion crackin~g are available in the BFS time frame, there are no unresolved safety items f which would preclude the issuance of construction permits for Black Fox Station, Units la dn 2. Kane Affidavit at 12. No material facts being 1

in dispute, the NRC Staff believes that Contention 15 should be dismissed.

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15-4 III. Statement of Material Facts About Which No Litioible issues Exist

1. Austenitic stainless steel is not susceptible to brittle fracture even in a sensitized condition. .

2. Applicant has committed to materials for the short term and material testing and replacement for the long term in order to preclude the rsssibility of intergranular stress corrosion cracking in the reactor coolant pressure boundary.

3. The use of. temperature-pressure limitations in the BFS technical

' specifications will preclude brittle fracture of ferrous materials

• in the BFS reactor coolant pressure boundary.

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UNITED STATES OF AMERICA NUCLEAR REGUIAIORY COMMISSION l

BEftXE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

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PUBLIC SERVICE COMPANY OF OKLAHOMA )

ASSOCIATED ELECTRIC COOPERATIVE, INC. ) Docket Nos. STN 50-556 AND WESTERN FARMERS COOPERATIVE, INC. ) STN 50-557

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(Black Fox Station, Units 1 and 2) )

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AFFIDAVIT OF WILLIAM F. KANE INTERVENORS CCNTENTICN 15 My name is William F. Kane. I am employed oy tihe Nuclear Regulatory Connission as a Licensing Project Manager in.the Division of Project Management, Office of Nuclear Reactor Regulation. I have been employed in l

this position since June 1973. My professional qualifications are contained j l

in an attachment to this- affidavit.- This affidavit was prepared by me or l l

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under my supervision:  :

The purpose of this affidavit is to address Contention 15 wnich l reads as follows:

Intervenors contend that the Applicant has not adequately deconstrated i that Black Fox 1 and 2 will meet the requirements of 10 CFR 50, Appendix A, Criteria ,(sic) 31 with respect to utilization of materials and/or procedures which will minimize the probability of intergranular stress corrosion cracking of stainless steel piping at alack Fox 1 and'2.

As presently structured, Contention 15 relates the need to minimize the prooability of intergranular stress corrosion cracking of stainless steel to the requirements of Criterion 31 of the General l Design Criteria which deals with brittle or rapidly propagating i fracture of reactor coolant pressure boundary materials. Although this may have been inadvertent it therefore makes it necessary _

to break this testimony down into two parts:

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1. Fracture Prevention of Austenitic Stainless Steels used in the anactor Coolant Pressure Boundary

2. Minimization of Intergranular Stress Corrosion Cracking of .

.austenitic Stainless Steels Used in the Reactor Coolant Pressure Boundary My affidavit' will show that there are no unresolved matters on Black Fox Station, Units 1 and 2 with respect to either fracture prevention or minimization of intergranular stress corrosion cracking of austenitic stainless steel used in the reactor coolant pressure boundary.

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1. Fracture Prevention *

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Criterion 31 of the General Design Criteria is identified in Appendix A to 10 CFR Part 50 and reads as follows:

Criterion 31' - Fracture Prevention of kactor Coolant Pressure Boundary. The reactor coolant pressure councary snall ce cesigned witn sufficient margin to assure that when stressed'under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a non-brittle manner and (2) the procability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperature and other conditions ,

of the boundary material under operating, maintenance, testing, l and postulated accident conditions and the uncertainties in determining

-(1) material ~ properties, (2). the effects of irradiation on material properties, (3) residual, steady-state and transient stresses, and (4) size of flaws.

As noted-above, Criterion 31 of the General Design Criteria deals with the brittle or rapidly propagating' fracture of reactor coolant pressure boundary materials. For this reason, tne staff's review of Criterion 31 is directed principally toward ferritic materials, eg. the reactor pressure x vessel, which are susceptible to this type of failure. A generally accepted definition of brittle fracture is a fracture with little or no plastic deformation ~. Ferrous materials sucn as carbon and low-alloy steels,

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-3 wuich are characterized by a body centered cubic crystal structure, exhibit an abrupt loss in toughness or brittle fracture when exposed to low temperatures and/or high radiation.

I Our safety evaluation of the reactor coolant pressure boundary materials is provided in Section 5.2.3 of Appendix A to the Safety Evaluation Report for Black Fox Station, Units 1 and 2 and is repeated below:  ;

i te materials of construction of the reactor coolant pressure boundary and connected systems will be in conformance with the requirements of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, including addenda and code cases, appropriate to comply with 10 CFR Part 50. Se residual elements in the ferritic maiorial of the reactor vessel beltline will be controlled in order i to reduce ute sensitivity of the materials to irradiation effects.

The materials that will be exposed to the reactor coolant have been identified  ;

l by specificat.on and all of the materials are compatible with the expected environment, as proven by extensive testing and satisfactory performance in nuclear reactor service'. General corrosion of all materials except carbon and low-alloy steels will be negligible. For those materials, conservative corrosion allowances have been provided in accordance with the requirements of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.

~ me purity of the reactor coolant is controlled to minimize corrosion. The maximum allowable contaminant levels are in accordance with our positions and have been shown by test and service experience to be adequate. The instrumentation provided for the monitoring of reactor coolant enemistry will ce adequate to detect changes on a timely basis to effect corrective actions to restore specified water purity. These controls provide reasonable assurance that the reactor coolant pressure boundary components will be adequately. protected from conditions that could lead to deleterious corrosion and loss of structural integrity of a component.

We have reviewed the materials selection, toughness requirements, and extent of materials testing propcsed by the General Electric Company to provide assurance that the ferritic materials used for pressure-retaining components of the reactor coolant pressure boundary will have adequate toughness under test, normal operation, and transient conditions. S e ferritic materials will meet the toughness requirements of the American Society of Mechanical Engineers Boiler and Pressure vessel Code section III.

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! The fracture tougnness tests and procedures required by Section III of the American Society of Mechanical Engineers Boiler and pressure vessel Code for the ferritic components of the reactor coolant pressure coundary provide l-reasonacle assurance that adequate safety margins against the possioility of nonductile benavior or rapidly propagating fracture can be estaclisned -

for- pressure-retaining components of the reactor coolant pressure boundary.

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l Further discussion of the controls placed on reactor operation to protect against crittle fracture of ferritic materials is provided in Section 5.3.2.4 J Appendix A to the Safety Evaluation amport for Black Fox Station, Units 1 and 2 and is repeated below:

The bases for the following pressure-tengerature limits to be ird on the reactor coolant pressure boundary during operation and tests were reviewed to assure that they will provide adequate safety margins against nonductile behavior or rapidly propagating failure of ferritic components of the reactor coolant pressure boundary as required by Criterion 31 of the General Design Criteria:

1. Pressure-tengerature limits for preservice hydrostatic tests.

2. Pressure-temperature limits for inservice leak and hydrostatic tests.

3. Pressure-temperature limits for, heatup and cooldown operations.

4. Pressure-temperature limits for core operation.

Appendices G and H to 10 CFR Part 50 describe the conditions that require pressure-tengerature limits and provide the general basis for these limits. These appendices specifically require that pressure . 1 paperature limits must provide safety margins at least as great as

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those reconmended in AEpendix G to Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code during heatup, cooldown and test conditions. Appendix G to 10 CFR Part 50 also requires additional safety margins whenever the reactor core is critical (except for low-level physics tests).

Actual operating limit curves cannot be determined at the preliminary design stage because the fracture toughness and other required tests have not been performed on the actual material that will be used.

Required operational pressure-tengerature limit curves, with.tengeratures snown relative to the reference transition temperature, and the casis for determining the curves were reviewed and compared with the acceptance criteria described below.

We evaluated the pressure-temperature operational and test limitations for acceptability using the methods referenced in the American Society of Mechanical Engineers Boiler and Pressure vessel Code and the Welding Research Council Bulletin 175, "PVRC Recommendations of Fracture Toughness."

We have concluded that the reactor is capable of being operated in a manner that will minimi::e the possiblity of rapidly propagating

~ failure, in accordance with Appendix G to Section-III of the American Society of Mechanical Engineers Boiler and Pressure to Section III of the American' Society of Mechanical Engineers Boiler and Pressure vessel Code and Appendix G to 10 CFR Part 50. We will require that the technical specifications for each plant referencing GESSAR-251 (sic) specify reactor operation in accordance witn Appendix G to 10 CFR Part 50.

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Tne use of Appendix G to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section III as a guide in estaclishing safe operating limitations, using results of the fracture toughness tests performed in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code and Commission regulations, )

will assure adequate safety margins during operating, testing, maintenance, and postulated accident conditions. Compliance with these American Society of Mechanical Engineers Soiler and Pressure Vessel Code provisions and C**= ion regulations constitutes an acceptable basis for satisfying the requirements of Criterion 31 of the General Design Criteria.

Contention 15 refers to a problem related to some austenitic stainless steels, intergranular stress corrosion, which is not related to Criterion 31 of the General Design Criteria..

Austenitic stainless steels, such as are used in much of the reactor l coolant pressurg boundary piping f..g. recirculation system piping are characterized by a face centered cubic structure. The austenitic stainless steels do not exhibit the abrupt loss in toughness or brittle fracture characteristic of ferrous materials when exposed to low temperature I

and/or high radiation. l Typically, austenitic stainless steels are furnished in the solution annealed condition in which the carbon in the steel is dissolved in l the austenitic matrix and corrosion resistance is at a maxinum. During l any heating in the temperature range fran 800 to 1600 degrees Fahrenheit, whether this occurs during fabrication, welding, or service, enromium carbides will precipitate most commonly at the grain boundaries. This depletes in chromium tne metal adjacent to the precipitated particles ,

and sensitizes the steel, making it susceptible to localized grain boundary attack. . Witn respect to Contention 15, it snould ce. noted that while

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severely sensitized austenitic stainless steel will be highly susceptible to intergranular stress corrosion, it will not, because of sensitization 1

be susceptiole to brittle or rapidly propagating fracture as addressed by Criterion 31 of the General Design Criteria. The effect of the carbide precipitation which occurs as a result of sensitization is well known to have

' an almost negligible effect on mechanical properties except for some loss in ductility at sub-zero temperatures. See for exangle, " Metal Handbook,"

8th Edition, Volume 1, 1961, American Society for Metals, pages 422 and  !

592, National Bureau of Standards Circula: 520, " Properties of Austenitic Stainless Steel at Low Tenperatures," by V. N. Krivebok, " Low Tenperature Impact of Annealed and Sensitized 18-8," by E. H. Schmidt, Metal Progress,  !

1950, and "Austenitic Cast Stainless Good for f.w-Temperature Applications -

But," by J. W. Juppenlatz, Iron Age, September 4, 1957. Therefore the l

austenitic stainless steel reactor coolant pressure boundary materials, even i in the sensitized condition, are not susceptible to brittle or rapidly propagating fracture for reactor conditions. Further, the controls placed on reactor operation for the protection of ferritic materials will assure that this condition is maintained for the life.of the plant.

In sununary, the staff has concluded there is a reasonable assurance that the Black Fox Station, Units 1 and 2, including those portions facticated from austentic stainless steel will be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a non-brittle manner and (2) the procamility of rapidly propagating fracture is minimized.

The staff 'has also concluded that the design will reflect consideration -

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of severe temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and~the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady-state, ,

and transient stresses and (4) size of flaw.

2. Intergranular Stress Corrosion Cracking l

l Although the staff has concluded there is reasonable assurance that because of the inherent high material toughness of austenitic stainless  !

steel, stress corrosion cracking is, unlikely to cause a rapidly propa-gating failure, the presence of any cracks .is undesirable. To this end the staff has, in its review of the Black Fox Station, Units 1 and 2 application, requested that the applicants comit to measures which will minimize stress corrosion cracking. The. staff has evaluated the information and comitments provided by the applicants in the Black Fox Station, Units 1 and 2 Preliminary Safety Analysis Report and, with respect to intergranular stress corrosion cracking of stainless steel piping, has concluded that it is acceptable for the construction permit stage. Our evaluation of t'his matter for the balance-of-plant, i.e., outside of the nuclear steam supply system, is provided in Section 6.1.1 of the Black Fox Station, Units 1 and 2 Safety Evaluation Report and is provided below.

The a5plicants incorporated the safety analysis from GESSAR-238 Nuclear Island (Docket No. STN 50-447) and GESSAR-238 NSSS (Docket No._ SIN 50-550) to cover engineered safety features materials. ,

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Engineered safety features, pressure boundary materials will

cogly with Section III of the American Society of Mechanical Engineers Code. Nonpressure boundary materials will comply with American Society for Testing and Materials and/or American Society of Mechanical Engineers Code specifications.

The applicants commit'to controls on the application of austenitic stainless steel- for the balance-of-plant which will satisfy the requirements of Regulatory Guide 1.31, " Control of Stainless Steel Welding," (Interim Position) and Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel."

Controls on the thermal insulation used on austenitic stainless steel cogonents of the engineered safety features will be in accordance with Regulatory Guide 1.36, " Nonmetallic Thermal Insulation for Austenitic Stainless Steel." Cold worked austenitic stainless steel will not be applied by the applicants for pressure boundaries of engineered safety features. The applicants commit to iglement the GESSAR-238 Nuclear Island resolution related to controls on the acidity of the reactor containment sprays and emergency, core cooling water following

< a postulated loss <f-coolant accident. The resolution is intended to provide reasonable assurance of freedom from stress corrosion cracking of the austenitic stainless steel components and welds of the containment spray and emergency core cooling system throughout the duration of the postulated accident to co gletion of cleanup.

Should the results of our GESSAR-238 Nuclear Island review indicate that the post-accident acidity of the containment

' spray and emergency core cooling system water is not compatible with i the materials used in these systems, we have concluded that i appropriate design measures such as changes in types of insula-tion and containment coating or waterproofing of piping can be taken to make the design acceptable. We therefore conclude that the material description provided by the applicants is acceptable for this stage of review.

Our evaluation of this matter for the nuclear steam supply system is provided in Section 5.2.6.2 of Appendix A to the Black Fox Station, Units 1 and 2 Safety Evaluation Report and is provided below.

In Septencer 1974, cracking was experienced in the stainless steel piping at Dresden Nuclear Power Station Unit 2, Docket  ;

No. 50-237. Itis was the. first of a series of incidents of intergranular stress corrosion cracking that occurred in eignt General Electric Company boiling water reactors. The _

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cracking occurred in weld heat-affected zones in Type 304 stainless steel recirculation system bypass piping systems and core spray lines. As a result of these incidents, a special task group within the NRC was formed to investigate the causes of the cracking. S e results and conclusions of the task group are given in the staff technical report,

" Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of' Boiling Water Reactor Plants,"

NUREG 75/067, October.1975.

S e NRC task group found that austenitic stainless steel piping in the reactor coolant pressure boundary of boiling water reactors is susceptible to stress corrosion cracking due to tne presence of oxygen in the coolant, high residual stresses, and some sensitization of metal adjacent to welds.

m ey found that such cracks w re expected to be in the heat-affected zones adjacent to welds and not to occur outside these zones where sensitization has not taken place, provided the pipe material is properly annealed. They concluded that the most promising solution to intergranular stress corrosion cracking of austenitic stainless steel is to replace susceptible piping with material that will be less adversely i

affected by oxygenated water.  !

2e General Electric Company formed its own task group to investigate the causes of stainless steel pipe cracking. ,

They have concluded that the NRC task group findings are in l general agreement with the results of their own findints. l Further, the General Electric Company has made the comit- l ment that all areas where intergranular stress corrosion cracking has been found to occur in the past 10 to 15 years of boiling water reactor operation will be replaced with materials which are not susceptible to intergranular strer corrosion or qualified fabrication processing techniques will be used to prevent sensitized unstabilized wrought austenitic stainless steel from being uposed to reactor coolant.

In the short term, the General' Electric Company proposes to enploy service proven materials and metallurgical structures made nonsusceptible to Intergranular stress corrosion cracking through special processing steps, thereoy avoiding any risks associated with imediate changes to new materials.

They propose ene following processes which have been proven by full size pipe tests:

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1. Solution heat treatment

2. Corrosion resistant cladding

3. Welding heat input control

4. Reduce inside-(diameter) temperature welding.

These processes will be applied to newly fabricated Type 304 stainless steel weld joints going into service on service sensitive lines,'e.g., those pipe runs with a previous history of. stress corrosion cracking (core spray, recircula-tion system bypass, and control rod drive return lines).

They also propose to apply the same process inprovements to other lines that have not experienced problems in the past, e.0 , recirculation system lines.

Solution heat treatment of stainless steel welds places them in a metallurgical' condition that is nonsusceptible to stress corrosion cracking. Solution heat treatment is applied in

- .the pipe fabricator's shop in.accordance with procedures

,f that include rapid cooling through the 1800 degrees Fahrenheit to 800 degrees Fahrenheit sensitization range. This prevents carbide precipitation.

Where solution. heat treatment is not practical, weld joint inside. surfaces will be protected with corrosion resistant cladding prior to making the final weld. Tests now in progress indicate that a minimum of eight percent ferrite is effective in preventing corrosion.

Where neither of the above processes is possible, it is propcsed to apply welding heat input control for both shop and field welds. A special process is being developed that

- reduces the inside surface tenperature of the pipe subsequent to the root pass. m is minimizes sensitization and reduces residun1 stresses hi the inside diameter of. piping. Accelerated stress corrosion tests on actual pipes are being performed to qualify these processes. The General Electric Company has taken steps to solution heat treat all piping-in the recirculation lines for BWR/6 plants where the spools have not yet been facticated. This program is limited by the  ;

current pipe facticator's furnace and quench tank capacilities.

l ne General Electric Company intends to.inglement other process

! steps'as they are proven and~become practical. Systematic ,

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qualification will precede the introduction of any changes.

The General Electric Company has proposed changes in materials as. its long. term solution to the austenitic stainless steel -

pipe cracking proolems. The'long term replacement material

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qualification program includes laboratory testing of welded ,

sections *, vendor qualification,' American Society of Mechanical l Engineers Boiler and Pressure vessel Code qualification i and trial introduction into boiling water reactor service. I As successful performance of a new material is demonstrated, use of it for pipe application will be expanded with the ultimate goal of complete -itment for use in boiling water reactors.

We concur with the objectives of the General Electric Company programs and we find the progress and comitments of the l General Electric Company acceptable for the Preliminary Design Approval.- We will continue to monitor the progress of the short and long term programs on a generic basis and will report the results of our evaluation at the final design stage of review.

Subsequent to issuance of the Black Fox Station, Units 1 and 2 Safety Evaluation Report, there have been two principal activities which deal with intergranular stress corrosion cracking of stainless steel. The staff has concluded that these activities should not preclude the issuance of construction permits for Black Fox Station, Units 1 and 2. A discussion of these activities and the bases for this cen-clusion are discussed below.

The first activity involves the publication by the Comission of NUREG-0313, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," dated July 1977.

We have also incorporated the position contained in NU[EG-0313 as Branch Technical Position MTEB 5-7, " Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping" and issued it as a revision to the Standard Review Plan.

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Black Fox Station, Units 1 and 2 was reviewed by the staff concurrently with the develossent of MrEB 5-7. For that reason, the comitments made by the applicants for Black Fox Station Units 1 and 2 remain acceptacle when r= pared to the positions in MIES 5-7.

S e second activity relates to an ongoing investigation by the staff i

related to the potential for sensitization of austenitic stainless steel cast structures, e.g., cast components of the pressure boundaries of valves and pungs, and weld metal. The potential for . sensitization of valve bodies exists during welding, in-process heat treatment, or during I

hardfacing or hardsurfacing of valve seats. The staff is reviewing this 1 I

matter to determine if any changes to current positions are required. )

Current staff positions exempt castings and weld metal from intergranular corrosion testing when they are exposed to %mperatures in the sensitization )

range after solution heat treatment. However, if the sensitization of cast material cannot be avoided and the resulting consequences of that sensitization are unacceptable, alternate materials or fabrication methods for cast austenitic stainless steel components are technically feasible and within the state-of-the art. me staff's schedule for examination of this concern and definiti;n of a proposed course of action will assure that any required changes arising from this concern can be incorporated into Black Fox Station, Units 1 and 2 prior to operation. l I

In sunmary, the staff concludes that, with respect to Contention 15, there are no unresolved matters and that there are no matters with respect to intergranular stress corrosion cracking of stainless steel that should preclude the issuance of construction permits for Black Fox ,

Station, Units 1 and 2.

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WTIl.I.41 F. KAN.E DIVISICN OF PROJECT MMIAGBetr .

U. 5. NUCIJ.AF REGUI.ATCRY CONISSICN PROFESSIONAL CUALIFICATIONS I a a Licensing Project Manager on the technical staff of Light Water Reactors Branch 4, Division of Project Management, U. 5. hbclear Reguhtory Ccmission. As a member of this staff, I have the functicn  !

of analyzing and evaluating the radiological safety aspects of nuclear facilities for which the Camission is responsible, inemd4ng licenses mi authoritaticns for the design approvals, const=ction and cperation of nuclear reac ors and standa:ni nuclear power plant designs. I have the t=-hn4 ~1 responsibility for the management and technical coordina-tion of the safety reviews for the applicatiens assigned to me.

I accepted an appointment with the technical staff of the U. S. Atomic Energy regulatory organ 4-2*n in 1973. I have the primary responsibility for the unagement of the review of r:diological safety matters on five utility applications for nuclear power plants which' include Black Fox

• aticn, Units 1 and 2, C=mnche Peak Steam Electric Station, Units 1 and 2, HarJville hbclear Plant, Units A1, A2, B1 and B2, Nine Mile ~

Point h. lear Station, Unit 2 and River Bend Staticn, Units 1 and 2.

In addition I have similar respcnsibility for three General Electric C:mpany standard design applications, GESSAR-233 huclear Island and the Ei.5SAR-ZS1 and GESSAR-238 huclear Steam Supply Systems. I am also the staff's technical coordinator for the review or the General' Electric r Cz:pany's Mark III containment desip.

! graduated from Widener College in 1961 with a Bachelor of Science Degree in Mechan;. cal Engineering. I performed my post graduate work in hbclear Engineering at Catholic University in the period fica 1971 -

1973. I as a registered professional engineer in the state of Calliersia.

I have a total of 1t yesrs of professional experience, all in the nuclear field. Six of the years I was emoloyed by the At:mic Energy Divisicn cf i the Allis Chalmers Marmfac= ring Company where I' was affiliated with the 1 han4-1 desip and field installation of the nuclear steam supply systen for the Lacrosse Boiling Water Reactor. Fo11cwing this I was i employed by hbclear Associates International Cot;oratien where I worked

- as a consultsnt to the utility isdustry speciali:ing in the areas of cee w m1 desip and analysis of nuclear steam supply system c::mpenents i and quality assursnce. Subsecuent to this I was employed by the l Reactor Fuels Divisien of Nuclear Fuel Services, Inc. , were I specialized l l

in the area of the :nechanical desip and analytical medel dwelcpent for l

' ruclear fuel relcads for light water reactors. -

C am a memcer of the American huelear Society, American 5ccier/ of L

Mechn4r1 Engmeers, :nd the Henorary Society of Tau 3 eta Pi. I am i the author or co-author of several technical publications. l 1

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I have read the foregoing affidavit and swear that it is true and accurate to the best of my knowledge.

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William F. Kane Licensing Project Manager Division of Project Management  :

Subscribed and sworn to before me this // day of July 1978.

'Y ' . L. /,'.u

'V Notary Puolic j g'

'N My Commission Expires 'd./ / / d'l l

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Contention 18 I. Contention Intervenors' Contention'18 maintains that the Applicant has not demonstrated that he is financially qualified to build Black Fox Units 1 and 2. For the re*asons listed below, the NRC Staff believes that ,

3 Intervenors'. contention is without a factual basis.

1 II. Argument 1 10 CFR 50.33(f) requires that there "be a reasonable assurance" that  !

the Applicants will-be able to obtain the necessary funds to construct .

a the facility and a showing that the Applicant has "a reasonable financing i plan in light of relevant circumstances." Kansas City ' Gas and Electric 1

Company (Wolf Creek,. Unit 1), ALAB-462, 7 NRC _(March 9,1978); Publie l

Service Company of New Hampshire (Seabrook Station, Units 1 and 2), CLI-78-1 I

' 7 NRC 1 (January 6,1978).

Information required to be submitted to show i

compliance with the Commission'.s regulations is contained in Append 1x C I

to 10 CFR, Part 50. i l'

The Karlowicz affidavit indicates that all of PSO's common stock is held l; by Central and South West Corporation, and that Central and South West  !

Corporation earned a 13.4% rate 'of return on comon equity for the calendar l

year 1977. The Karlowicz affidavit also shows that PS0 enjoyed a 13.3%

rate of return on common equity for the same period and all of its out- l standing first mortgage bonds issued are rated double A by Moody's Investor Services .Inc..ahd Standard and Poors,' the major securities rating firms. .

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. I 18-2 Karlowicz affidavit at 3 and 6. The ability to finance mortgage bonds, 1 I

to earn a fair rate of return on total capital and connon equity, to maintain satisfactory interest coverage levels, together with a balanced  ;

capital structure, were seen as a reasonable assurance that PS0 would be able to raise its 61% share of the latest $2,042,300 construction v j cost. Karlowicz affidavit at 4-6.

Associated Electric Cooperative, Inc., a six-member cooperative, and Western Farmers Electric Cooperative plan to finance their nearly 22%

and 17% interest respectively in Black Fox with long term loans quaranteed by the Rural Electrification Administration (REA). With essentially no k state regulation of rates for each utility, power contracts between the cooperatives and their customers will ensure rate returns sufficient to maintain reserves, cover all costs of operation and to make payments on account of principal and interest on indebtedness. Karlowicz affidavit at 7.

Since preliminary approval of a guaranteed loan was given to Associated and its expected shortly for Western, erd no substantial derrogatory in-formation as to-the cooperatives' financial situation is known, the NRC Staff believes that with loan guarantee approval Dy REA there will be a l reasonable assurance of the financing of those utilities' part of the Black Fox Stations' Units 1 and 2.

Based on the information listed above, the NRC Staff believes there are no controverted facts which remain to'be' litigated as to the financial i capability of the three Co-Appit' cants to finance their share of BFS. -

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18-3 That Intervenors have no facts within their possession to indicate that the Co-Applicants are not financially qualified is evidenced by their November 30, 1976 Answer to Staff Interrogatories which gave no statement

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as to the factual basis for their contention. In addition, none of the three Intervenors' consultants deposed on May 31 and June 1 and 2,1978 indicated any facts within their possession with which they would oppose the facts set forth above by the Staff.

III. Statement of Fact As To Which There Are flo Genuine Issues  ;

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1. Applicant PS0 has a balanced capital structure, a bond

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rating of double A, satisfactory interest coverage ratios, and has achieved upwards of a 13% return on comon equity.

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2. A letter of intent to commit a loan guaranteed by the Rural l Electrification Administration to Associated is in publication in the Federal Resiger and Western is expected to achieve this status shortly.

3. Both Associated and Western Farmers Cooperative have substantial I control in maintaining an adequate rate of return sufficient to maintain reserves, cover cost of operations and make payments on account of principal and interest on indebtedness by reason of their power contracts

, with their customers.

e e

1 UNITED STATES OF AMERICAN  !

NUCLEAR REGULT0RY COMMISSION l BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

)

Public Service Company of Ok,lahoma, )

Associated Electric Cooperative, Inc.) Docket Nos. STN 50-556 and . ) and STN 50-557 Western Farmers Electric Cooperative )

)

(Black Fox, Units 1 and 2) )

AFFIDAVIT OF MICHAEL L. KARLOWICZ, JR.

CONTENTION 18 1 am employed as a Financial Analyst in the Office of the Assistant Director for Quality Assurance and Operations, Division of Project Management, Office of Nuclear Reactor Regulation, United States Nuclear Regulatory Commission.

( Attached to this affidavit is a copy of my professional qualifications.

This testimony is in response to intervenor's contention number 18, as follows:

"Intervenors contend that Applicant has not demonstrated that he is financially qualified to build Black Fox,1 and 2."

As part of the responsibilities of my position at the NRC, I have been assigned the task of reviewing the financial qualifications of the above captioned applicants. I will discuss the most pertinent factors which I had considered during the course of my review of the reasonableness of the applicants' financing plans, first for the investor-owned applicant -

Public Service Company of Oklahoma, a wholly owned subsidary of Central and South West Company, and then for the cooperative applicants - Associated Electric Cooperative, Inc. and Western Farmers Electric Cooperative.

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This review also addresses the revised financial projection of Public Service Company of Oklahoma, as stated in its June 27, 1978 amendment to the application.

Investor Owned Applicant Rate of Return on Common Equity Of all of the factors which are . considered during the course of review of.

an applicant's financial projections in the determination of its financial qualifications, the a' '"mptions regarding its rate of return earned upon common equity is the . significant. This rate is arrived at by first deducting from grost ,erating revenues all costs incurred in the operation of the business such as operating and maintenance expense, depreciation, l( taxes, interest charges, and preferred dividends. This computation re-sults in net income available to the common stockholder, the " bottom line" of a company's operations. Dividing this figure by a company's' common equity (the investment dollars provided by a company's common stockholders plus accumulated retained earnings) results in the rate of return on common equity. Of all capital provided to a company (i.e., proceeds of long and short term debt, preferred stock, w..] common stock), shareholders of common stock bear the highest risk. While capital costs attributable to a utility by deot_and preferred stock are fixed by' contract, and must be paid at the agreed rate, those dollars earned on common equity represent rer,idual  ;

earnings after payment of the expenses incurred in running the company's operations. Since the applicant is a public utility, its rate of return is set by' the regulatory agencies having jurisdiction over it through the )

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t rates it charges, and unlike its rates, which are fixed, the return on

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common equity is only allowed to be earned and is not guaranteed. Since there exists no absolute' certainty as to a utility's future earnings, one is required to consider its previous level of profitability, the current regulatory environment, management's ability to control costs and attrition, and other relevent circumstances in determining the reasonableness of a projected return on comon equity.

A review of this applicant's financial picture cemonstrates an operating-and regulatory environment which results in an above average level of profit ability. During the last five years, Public Service Company of Oklahoma earned'in excess of 13 percent on its average return of comon equity.

Presently, the Oklahoma Corporation Commission allows this applicant a rate of return on its comon equity in the range of 13.5 to 14.5 percent and approved an increase in its rates to provide for a 14.0 percent return.

Last year it earned a rate of return on comon equity of 13,3 percent.

Concurrently, tr.e parent company of Public Service Company of Oklahoma, Central and South West Corporation, has also realized a satisfactory level of earnings. Last year, the parent company earned a rate of return of 135 percent on its consolidated comon equity. Although the 15.0 percent ,

1 rate of return on common equity utilized by Public Service Company of Oklahoma is somewhat above what the industry has historically earned, j regulatory agencies recently have been increasingly allowing returns around ]

this level in apparent recognition of the higher prevailing cost of capital.

Based on the above, I have determined that the return on common equity

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assumed by Public' Service Company of Oklahoma in its financing plan is within the zone of reasonableness. l Cash Flow

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In the meeting of an applicant's year by year construction expenditures, the first item considered is its cash flow. By reason of certain non-cash expenses - primarily depreciation and deferred federal income taxes','and retained e~arnings, a company may internally generate funds. Construction expenditures and other capital requirements in excess of internal cash generation must be financed externally through new issuances of common stock (this applicant will receive capital contributions from its parent company in lieu of going directly to the outside market to issue common stock),

debt, and preferred stock. Previously, this applicant projected $36 million to be available from its 1977 internal cash generation, which amounted to -

l 21.3 percent of its projected capital requirements for that year. In actual operations for 1977, Public Service Company of Oklahoma generated internally

$56 million, or 29.1 percent of its capital requirements. During the years 1975 and 1976, this applicant generated internally $40.2 million and 548.2 million, or 56.3 percent and 62.9 percent of its respective annual capital requirements. Taking into account these considerations, the 40.5 percent  !

l ratio of internal . cash generation to total capital requirements projected during the construction period of Black Fox, Units 1 and 2, appears reasonable l l

ano attainable.

l Capital Structure In orcer for a utility to ccnduct a viable financing program and maintain the j

- attractiveness of its securities, it must hav.c. a reasonably balanced capital

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-S-structure. The practice of over-borrowing can lead to heightened and un-wanted risk to holders of the utility's long term deot, thereby reducing

'the attractiveness of such securities. An additional effect of over- ,

1 borrowing is the increased leverage which results in extreme sensitivity of the utility's earnings on common equity. Assuming a constant level of l profit, the additional risk occasioned by increased leverage will result 4

in a discounting of the market price of those securities. Under adverse economic conditions, a utility could be placed in a financing squeeze and enjoy little, if any, flexibility in its financing. The interest coverages (see infra) of the company will also decline, and a point can be reached at which it will be barred from the issuance of new debt by indenture

[( restrictions.

On the other hand, by maintaining a reasonable and well balanced capital structure, some latitude will exist in its financing. This will achieve borrow'ng reserve, allowing _ freedom both in the timing and selection of sect r :1es to be issued to meet capital requirements. Moreover, under these j circumstances, its securities will maintain their attractiveness to investors by reason of their lower risk.

Turning our focus to this applicant, et the end of 1977, Public Service )

l Company of Oklahoma had a capital structure consisting of 41.4 percent common l l

equity,12.1 percent preferred stock, and 46.5 percent debt. At the same j time, Lits parent company - Central and South West Company had a capital

' structure of 43.2 percent common equity, 9.2 percent preferred stock, and 47.5 percent debt. In comparing the equity . component of capitalization of

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6-both the applicant and.its parent company to the industry in general (an average of approximately 35 percent common equity), I find their capital structures to be reasonably balanced. In light of the above, the target capital structure.as stated by the applicant in its assumptions underlying the financing program appears reasonable.

Coverage Ratios -

In order to protect the assets mortgaged under a company's debt, a contract (indenture) is made between the company and the bondholders. Although there are many provisions contained in a bond indenture, it predominantly is the coverage ratio test which provides protection to bondholders. This provision precludes a company from issuing more debt should there not be satisfactory earnings coverage of a company's interest obligations. Accordingly, this ratio is a major criterion used by the financial community in making credit decisions with respect to a company's debt. Appropos to this, the first mortgage bonds issued by Public Service Company of Oklahoma are rated double A by both Moody's Investor Services, Inc. and Standard and Poor's.

Looking at the coverage ratio of this applicant for the year ending August 31, 1977, Public Service Company of Oklahoma had interest coverage of 3.98 times, compared with the indenture requirement of two times interest charges.

Interest coverage ratios are projected to be from 3.9 to 5.6 during the period of construction of Black Fox, Units 1 and 2. This range of interest coverage ratios appears consistent with the capital structure, rate of return on common equity, and costs of new capital assumed by Public Service Company of Oklahoma-f in its projections.

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i__.________ _ _ _ _ _ _

l Cooperative Applicants Associated Electric Cooperative, Inc. and Western Farmers Electric Cooperative are membership rural electric cooperatives subject to the jurisdiction of the Rural El~ectrification Administration. For all intents and purposes, these cooperatives are not subject to-the regulatory juris-diction of state utility commissions. Sources of funds necessary to meet the construction expenditures attendant to the participation by Associated i

and Western in Black Fox will be obtained through long term debt guaranteed by the Rural Electrification Administration. As noted in my previous testimony, staff requires copies of the executed loan commitment notices and the executeu joint ownership agreement as condition to the construction permit.

As to the current status of the loan guarantee applications, I have been informed oy Lou'is Pitts, a staff me.ber of the Rural Electrification Admin-istration of the U.S. Department of Agriculture, that the application of Associated Electric Cooperative has been accepted and that a letter of intent to commit the required loan has been submitted for publication l in the Federal Register. He has further advised me that he expects the same for Western Farmers Electric Cooperative by next week, or shortly l thereafter. I have also been informed that letters advising receipt of  ;

the respective application for a loan commitment by the Rural Electri- 1 1

fication Administration are being forwarded to Senator Thomas F. Eagleton - l 1

Chairm6n of the Subcommittee on Agriculture, Rural Development and Related Agencies, Senator Henry Bellmon - Ranking Minority Member, the Senate i '

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Committee or Appropriations, and Congressman Jamie L. Whitten - Chairman of the U. S. House of Representatives Committee on Appropriations.

Conclusion .

Although I have considered other factors and assumptions previously addressed in my prior statement evaluating the respective applicants' financial assurances of obtaining funds to design and construct Black Fox, Units 1 and'2, I believe that the foregoing factors discussed are the most relevant. Each of these factors will interact with each other during the actual financing of the subject facilities. Additionally, my evaluation encompassed the applicants' ability to obt,ain funds necessary to meet all of their capital requirements ,during the period of construction, including construction expenditures attributable to other activities in addition to the costs required to construct the subject facilities. Although I do not consider these financing projec-tions to be a forecast of what will necessarily occur, in my judgment, I believe that they have demonstrated one possible way by which the planned construction program might reasonably be financed. I would realistically expect that the financing plans will change over time to accommodate changing financial and economic conditions. The financing being proposed is in accord with general industry practices and the l assumptions being used, although they are not susceptible to precise measurement against absolute criteria, are in line with what one would expect under the postulated conditions. The revisions to tne financing plan of Public Service Company of Oklahoma, as stated in its June 27, I

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1978 amendment to its application, are reflected in the foregoing discussion and do not affect the cooperative co-applicants. Accordingly, there is no reason to change my previous determination that the applicants

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are financially qualified to design and construct Black Fox, Units 1 and 2, subject to the previous conditions. Since my evaluation of these projections leads me to conclude that they may be characterized as reasonable, I believe that the reasonable assurance standard has been satisfied.

Certification I nereby certify that the foregoing facts are true and accurate to the best of my knowledge.

Michael L. Karlowicz. dr.

Sworn to and subscribed before me this 14th day of July, 1977 j,a y' h1. . Y$ 7m %W 1 Ju, C  % - W f )? . ' 1 ' '

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MlCHAEL L. KARLOWTCZ, JR. PROFESSIONAL QUALIFICATIONS My primary responsibility as Financial Analyst at the U.S. Nuclear Regulatory Commission is the performance of financial qualifications review of applicants during the nuclear licensing process. This review includes an analysis of estimated construction costs or operating expenses, projected financing methods and underlying assumptions, and regulatory climate and trends. In this regard, I have prepared financial qualifications analyses for inclusion in supplements to the Staff's Safety Evaluation Report. My responsibilities also include the monitoring and keeping abreast of the money and capital markets, particularly those affecting the investor-owned electric utility industry. . I received a Bachelor of Science Degree with a major in Mathematics in 1972 from Saint Peter's College. In 1976 I received my Juris Doctorate from the Delaware Law School . Thereafter, I attended the post-

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graduate L. L. M. tax law program at New York University in 1976 throug. I am a menber of the John Marshall Honorary Society of Delaware Law S_ acol, the American Bar Association, and its section of Public Utility Law. I am admitted to the New Jersey Bar, the United States District Court for the New Jersey District, the United States Tax Court, the United States Court of Claims, and the United States Court of Customs and Patent Appeals. My application is currently pending for the District of Columbia Bar admission. Prior to my joining the U.S. Nuclear Regulatory Commission in December 1977, I spent three years with the New Jersey Department of the Public Advocate, Division of Rate Counsel. As both attorney and economist, my responsibilities includedthe representation of the public interest in proceedings involving proposals of increases in rates or discontinuance  ! of service by regulated industries. In this respect, my responsibilities were the presentation of the public position and the cross-examination of expert witnesses representing applicants or intervenors before state  ; regulatory commissions. During the years 1969 through 1974 I was associated with Public Service Electric and Gas Company in the System Planning Department and the Office of the Corporate Economist. There, I was responsible for the conducting of short, medium, and long range studies in both systems and financial planning, preparation of expert testimony, and economic analysis. 4 i k.. 5

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l' \ 19 Contention 19 I. Contention l Intervenors contend that the Applicant has not adequately demonstrated l that Black Fox 1 and 2 will comply with 10 C.F.R. Part 50, Appendix A, Criterion 4, in that the potential . dynamic effects on the containment associated with internally generated turbine missiles have not been adequately considered. i II. Arcument Criterion 4 states: , S$ructures, systems, and components important to safety shall C- be designed to accomodate the effects of and to be compatible with the environmental conditions associated with normal operation, l maintenance, testing, and postulated accidents, including loss-of-Thees structures . systems _ and enmennents shall l Coolant accidents. be anoroorfately orotected aaninst dynamic effects. includia? +ha effects of missiles. eine whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. (emphasisadded). As indicated by the Staff affidavit prepared by Dr. Campe, the primary method of protecting a nuclear reactor's safety systems against low trajectory turbine missiles is the placement of the turbogenerator l i in a peninsular orientation relative to the safety systams -(containment l t auxiliary, control and diesel generator buildings). This turbine orientation places the safety systems outside of the missile strike zone of the unit in question. Thus, because of this placement, the Staff affiant was' able to conclude that missiles generated from one nuclear reactor's' l ( turbogenera' tor..will pose little risk to:its.own safety systems. Camoe affidavit at 2.

[ 19-2 However, since the Black Fox Station is a two unit site, the peninsulary placement or orientation of the turbogenerator for each unit does not

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preclude the possibility of low trajectory turbine missiles from one unit striking safety related structures of the adjacent unit. Therefore, the NRC Staff looked at the probability of missiles from one unit striking the safety systems of the adjacent unit. Campe affidavit at 2, 4. The Staff review of the strike probabilities was performed in tenns of high trajectory; and low trajectory turbine missiles. The independent Staff analysis of the possibility of low trajectory missiles striking another unit's safety system began by conservatively assuming that destructive overspread of a turbogenerator occurs with a frequency of 4 x 10-5 per turbine year. The probability that a missile might hit and damage safety related system in the other BFS unit's control room, containment and auxiliary building was computed to be less than 10 per turbine failure and thus was found to be within the guidelines of Reg. Guid 1.115. This was due in part to the fact that before there can be any damage to a safety system, a missile will have to penetrate two three foot thick concrete walls before leaving the turbine building and an additional two or more foot

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   *<                                                                                               1 19-3                                            j l

thick concrete wall upon entering the containment, control room or auxiliary buildings. Campe affidavit at 4. Thus the Staff comoutation l confirms the accuracy of Applicant's turbine missile analysis which concluded that the possibility of safety related systems being struck , Campe and damaged by turbine generated missiles is extremely small. affidavit at 1-2. 1 i As to high trajectory missiles (those ejected at a nearly vertical ] directions from the turbine generator), the Staff computed that the possibility of damaging the reactor coolant pressure boundary, spend fuel . pool, ; control room and cable spreading rooms of one unit was less than 10-7per turbine year, an ifcceptably low probability. Campe affidavit at 5.

           .       For the reasons listed above, the NRC Staff believes that the Staff's analysis demonstrates.that there will be little possibility for the compromise of safety system from Black Fox missiles and that Contention
       .           19 should be dismissed.

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~ 19-4 III. Statement of Material Facts As To Which There Are No Genuine Issues To Be Litiaated

1. Because of peninsular alignment of each turbogenerator with the safety components of its own unit, there is little possibility that a missile generated at one unit will strike the safety system of the same unit.

2. The possibility of a low trajectory missile from one unit striking another units' safety system is 10-3 per turbine failure, an acceptabily low probability.

3. The possibility of a high trajectory missile from one unit damaging the safety systtm of another unit is 1.5 x 10~8 per turbine year, also an acceptably low probabilit .

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I, 1 UNITED STATES Oc AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD l In the Matter of the Application of ) Docket Nos. Public Service Company of Oklahoma, ) Associated Electric Cooperative, Inc. ) STN 50-556 and -

                                                                   ). STN 50-557 Western Farmers Electric Cooperative              .
                                                                   )

C (Black Fox Units 1 and 2) . AFFIDAVIT OF K. M. CAMPE ON CONTENTION 19

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8-2 In each separate cable spreading room the cable tray separation distances, both vertical and horizontal, will be in accordance with Reg. Guide 1.75 for safety divisions 1 and 4 in one cable spreading room and for the redundant safety divisions 2 and 3 in the other cable spreading room. For the reasons stated above, the NRC Staff believes that the Applicant has adequately reduced the probability and effects of a fire at the proposed BFS Station. For that reason, no facts remain to be litigated and Intervenors Contention 8 should be dismissed. III. Statement of Material Facts About Which :10 Issues Of Fact Remain To Be Liticated

1. BFS will use separate cable spreading roomsfor each unit.

2. Cable trays within each unit will be separated both vertically and horizontally in accordance with the provisions of Reg. Guide 1.75 and IEEE 384. In addition, the cable spreading rooms will be separated by i a 3 hour fire barrier, separating redundant safety systems 1 and 4 from systems 2 and 3, A fixed automatic deluge system with a manual backup will be provided in order to minimize the effects of any fire.

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AFFIDAVIT OF KAZIMIERAS CAMpE S My name is Kazimieras M. Campe. I am employed by the Nuclear Regulatory Commission as a Nuclear Engineer in the Accident Analysis Branch. I have been employed in this position since December 1972. My qualifications are listed on an attached qualification sheet. The purpose of my affidavit is to address intervenor Contention 19 regarding turbine missile risks with respect to the proposed design of the Black Fox Nuclear Station. Contention 19 relates to alleged on compliance of proposed Black Fox Unit Nos.,1 and 2 with 10 CFR part 50, Appendix A, Criterion 4, and contends that the potential dynamic effects

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of turbine missiles on the containment have not been gdequately considered. i 1 My affidavit will show that potential turbine missile effects with respect-to all safety related structures including the containment for the pro-posed Black Fox' Muclear Station have been considered adequately and that l the turbine missile risks in terms of the safety and health of the public are acceptably low. CONTENTI0'N 19 The intervenors contend that the applicant has not demonstrated adequately that Black Fox Unit Nos.1 and 2 will comply with 10 CFR part 50, Apoendix A, Criterion 4, in that the potential dynamic effects on the containment associated with turbine missiles have not been considered adequately. The applicant' has submitted a turbine missile analysis and a conclusion, based on the analysis, that the probability of significant damage to 1

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safety related systems and components by a turbine missile is extremely small so that potential turbine missile consequences are acceptable. These analyses and conclusion's notwithstanding, the staff performed an independent turbine missile analysis. The staff's avaluation which was reported in Section 3.5.3 of the SER (Reference 1) concluded that the probability was acceptably low and stated that:

                                "The Black Fox Unit Nos. 1 and 2 turbogenerators are arranged in a peninsular orientation relative to their respective containment buildings. However, the two-unit plant configuration is interactive in that some portions of safety related structures of each unit              . _ . _ _

(i.e., contairunent, auxiliary, control, and diesel generator buildings) - are. exposed to potential low trajectory turbine missiles from the' adjacent unit's turbogenerator. The staff has estimated the proba-bility that low trajectory turbine missiles will strike safety related j systems and has found it to be acceptably low. Consequently, the 1 proposed plant design is protected. adequately against potential low trajectory turbine missiles." Additional material underlying our evaluation and which supports this conclusion is given below. I

                             ~ The staff uses the criterion given in Standard Review Plan 2.2.3 (Ref-
      -                       erence 2) to determine whether the l' oss of a safety related system from a turbine missile has an acceptably low probability.        This SRP, which is used to cvaluate external man-made events including explosions, ship, and aircraft impacts states that the probability of events occurring                     ,

i which could' lead to potential consequences in excess of the 10 CFR Part 100 I exposure guidelines should be no greater than about 10~7 per year (when ) 1

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realistically evaluated)^. Where tMs diteria is not satisfied, the event of concern snould be considered in the plant design basis. In

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( 3 - addition, Regulatory Guide l.ll5 (Reference 3) states that, assuming a turbine failure has occurred, the probability of missiles striking and

            ' damaging safety related systems should be no greater than about 10-3, This is based on the conservative estimate of turbine failure as 10-4 per turbineyear,(Reference 4).

Turbine missiles can be divided for analysis into low trajectory missiles (those missiles which can strike plant structures directly) and high trajectory missiles (those missiles which are ejected upward in a nearly vertical direction, and thus strike plant structures by falling from above). The staff performed an independent analysis which postulated a turbine failure and which examined the geometric relationships of the plant safety systems to the turbine and considered the ejection paths or trajectories taken by turbine missiles based upon known studies,(Ref-erence4), I Nwi

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Since each turbine for a given unit is arranged in a peninsular orienta-tion relative to the containment and all other safety related structures for that unit, it is nearly impossible for low trajectory turbine missiles from a given unit to strike any safety related systems in the same unit. With respect to low trajectory turbine missiles, therefore, we evaluated the probability of turbine generator missiles from one unit striking the safety related targets of the adjacent unit. The exposed targets for each unit consist of the containment, control room, and auxiliary buildings. The probability of a turbine missile striking these structures given that a turbine failure has occurred are estimated to be 2.9 x 10-3, 9.4 x 10-4, and 1.5 x 10-3 respectively. These estimates are based on a comparison of the solid angle subtended by the target to the total solid angle into which a turbine missile can be ejected. The strike probability for the diesel building was not calculated since its orientation with respect to the adjacent unit's turbine generator is such that a single turbine mis-sile would not be in line with more than one of the three diesels. Thus, on the basis of redundancy and separation, the diesel generator building 4 was not considered to be vulnerable to turbine missiles. Since the low trajectory turbine missiles must penetrate two three-foot thick concrete l barriers prior to leaving the turbine building, and since the containment control

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roorii"and~aUxiliar{b~uirdingValis lpreserit ~ari~ additfonal two' or more feet of l~ concrete, penetration probability associated with these structures is judged to be.less than 0.1. This is based on the cbservation that the i upper limit-for penetration by destruct.tve:overspeed turbine missiles is

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... 1 _ g _ I about 8__ feet of concrete, using the conservative staff formulae referenced in Regulatory Guide 1.115. Consequently, the probability of sustaining damage of safety related systems by low trajectory turbine missiles is less than 10-3 per turbine failure.  ! High trajectory turbine missiles (i.e., those missiles which are ejected in a. nearly vertical direction from a turbine generator) have a low strike probability density. Our estimates of the strike probability density indi-cate that it is on the order of-10-7 per square foot of horizontal target area. The horizontal target area associated with the containment building . l is 14,525 square feet, so that in the event of a turbine failure the proba-  ! bility of striking the containment with a high , trajectory turbine missile C is about 1.5 x 10-3 The probability of damaging the reactor coolant  ; l pressure boundary within the containment, given that a high trajectory. l turbine missile falls 'on top of the containment, is estimated conservatively to be about 0.1 (assuming that every missile that strikes the containment l manages to penetrate the containment roof). This estimate is based on  ! assuming a uniform distribution of missiles falling within the contain-

                 -ment, so that the strike probability within the containment can be evaluated by ratioing the plan area of a target to the total plan area of the con-tainment.      Thus, the total probability for high trajectory turbine missiles      1 damaging the reactor pressure boundary within the containment is about
                             -8                      Similar. results were obtained with respect

1. 5 x 10 er turbine year.

to the other relevant targets potentially vulnerable to.high trajectory

/ turbine _ missiles, namely the. spent fuel pool, the control room, and the

('" cable spreading rooms.

~ 6-1 On the basis of the above evaluation, we conclude that the proposed Black Fox design is adequately protected against turbine missiles. REFERENCES

1. Black Fox Station Safety Evaluation Report, June, 1977.

2. Standard Review Plan, NUREG-75/087.

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3. Regulatory. Guide 1.115, " Protection Against Low Trajectory Turbine Missiles." l l

4. S. H. Bush, " Probability of Damage to Nuclear Components," Nuclear i Safety, Volume 14, No. 3, May-June 1973.

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f XAZIMIERAS M. CAMPE PROFESSIONAL OUALIFICATIONS ACCIDENT ANALYSIS BRANCH , DIVISION OF SITE SAFETY AND ENVIRONMENTAL ANALYSIS I am a member of the Accident Analysis Branch of the Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Comission. My duties include the identification and evaluation of hazards 'a the

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safe operation of nuclear power plants due to accidents external and internal to those plants. Part of these duties involve the evaluation of turbine missile Msks. I have reviewed turbine missile generation,

                   ..rike, and damage probabilities for nuclear plant license applications since 1973. I have personally examined and performed photographic documentation of the Shippingport and Gallatin plant turbine failures.           j I have prepared most of the technical input for the Regulatory Guide 1.115 which addresses protection requirements against low trajectory turbine missiles. In 1975 I co-authored a paper with J. Read of the Accident j         Analysis Branch on the subject of higtf trajectory turbine missile strike      i C     I         probabilities. Currently I am one of the contributing authors to the i         Staff's forthcoming turbine missile evaluation report.                           .

I I. graduated from the University of Connecticut'where I received S.S. and M.S. degrees in Mechanical EngineeMng 1958 and 1960, respectively. Between 1960 and 1962 I completed some advanced mathematics courses at the Rensselaer Polytechnical Institute branch in East Hartford, Connecticut. During this peMod I was employed by Pratt and Whitney at the CANEL Analytical Physics Group as an analytical engineer. From 1962 to 1966 I attended Purdue University, where I received a Ph.D. in Nuclear Engineering. From 1966 to l 1972 I was employed by Hittman Associates, Inc. where I worked in the Radioisotope Department. During this peMod my responsibilities included i I radiation shielding analyses, radioisotopic generator design, and comeutar code development for reactor core physics calculations. Since 1972 I have  ; been emoloyed by the Nuclear Regulatory Comission in the Acciden: Analysis , Branch. [ , . ,w I

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I have read the foregoing affidavit and swear that it is true and accurate to the best of my knowledge,

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a g ... K. . ampe Sworn before me this 10 day of July, 1978 iL4.~fcb4 .b W u~p y4y e ny e _c sm

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4 1 l 66-1 , Contention 66 I. Contention Intervenors Contention 66 maintains that the Applicants' BFS design does not adequately protect the public from the potential consequences of sabotage in that the plant does not require sufficient integrity and safety redundancy. For the reasons listed below, the NRC Staff believes that no litigable sabotage issues remain at this stage of the BFS review. II. Argument While the BFS is still in the design stage (submission of the FSAR and operation of the plant will be several years hence), the NRC Staff believes that Applicant's submissions and connitments to date,will ensure protection from sabotage . The provisions of 10 C.F.R. 550.34(c) require that each applicant for a license to operate a production or utilization facility shall include a physical security plan that addresses vital equipment, vital areas, and isolation zones as well as their plans for compliance with the provisions of 10 C.F.R. part 73. Such a preliminary plan was sutrnitted by the Applicant in 1976, prior to the enactment of the new provisions of Part 73 i - - s

66-2 In addition, by letter of March 3,1978, the Applicant committed to developing a security program for Black Fox that would comply with the These submissions were newly enacted Part 73. Ross Affidavit at 2. thought by the NRC Staff to satisfy the provisions of 10 C.F.R.150.34(c in that the Applicant had provided adequate evidence that it knows and will implement the security concepts embodied in Part 73 before an i l operating license for BFS is issued following a thorough safety review l Ross Affidavit at 3-4. Such a plan, l at the operating license stage. following input and review by the NRC Staff, was thought to meet the requirements of Section 73.55. Ross affidavit at 5. 1 The Staff notes that the deposition of Richard B. habbard on June 1, 1978, revealed no facts within the possession of the Intervenors to contradict the information above. Tr. 114-115. III. Material Facts As To Which No Genuine Issues Remain To Be Resolve

1. Applicants' security plan (submitted in 1976), when coupled with their comitment to comply with the provisions of 10 C.F.R. 573.55, furnish a reasonable assurance that the BFS final security plan and the design of the plant will be adequate to protect the public and plant workers the Plant becomes operational.

against the effects of sabotage wher , l l l - -

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION l l l

                       -BEFORE THE ATOMIC SAFETY AND LICENSING BOARD, In the Matter of                         -)

PUBLIC SERVICE COMPANY OF OKLAHOMA Docket Numbers 50-556 -~ 50-557 (BLACKFOXNUCLEARSTATION, UNIT -NOS.1 AND 2) ) AFFIDAVIT OF WILLIAM J. ROSS

1. I, William J. Ross, am a Reactor Safeguards Analyst attached to the Reactor Safeguards Licensing Branch, Division of Operating Reactors of the Nuclear Regulatory Comission. For the past two years I have managed and performed reviews of applicants' and licensees' site physical security plans provided to protect reactors C at the site against industrial sabotage, theft, and dtversion of

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special nuclear materials.

2. I have prepared the Staff Response to Contention 66.

3. Contention Number 66. The applicant's present design does not adequately protect the public from the potential consequences of sabotage at the Black Fox Plant in that the plant does not require sufficient integrity and safety redundancy to thwart a saboteur.

Response: , ThePublicServiceCompanyofOklahoma(theapplicant)hasapplied I to the Nuclear Regulatory Comission-(the Comission) for a Con-struction Permit to build the Black Fox Nuclear Station, Units Nos.'l'and"2.. .In c'ompliance with.the' provisions'of g50.34_(c) of (

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Title 10 CFR the applicant submitted on May 14, 1976 a physical security. plan. in Amendment #3 to the applicant's Preliminary Safety Analysis Report. This security plan was of preliminary nature and based on non-specific requirements in 10 CFR Part 73 and gui- l dance in Regulatory Guide 1.17. This Regulatory Guide, in turn, referenced the standard " ANSI N18.17 - Industrial Security for i Nuclear Power Plants" developed by the American Nuclear Standards Committee in 1973. On February 24, 1977 the Commission published (42 Fp 10838) new regulations (10 CFR 73.55) that require "each applicant for a license to operate a nuclear power reactor pursuant to Part 50 of this chapter'whose application is submitted after February 24, 1977, shall include in the physical security plan required by 35 0.34(c) the infonnation identified in paragraphs (a) through .

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4 (h) of this section and if such applicant receives an operating license, shall comply with the provisions of this section on receipt of the operating license." By letter of March 3, 1978 the applicant committed to " developing a security program for our Black Fox Station which will comply with the requirements of 10 CFR 73.55."' The applicant further states that he is " emphasizing within PS0 and Black and Veatch Consulting Engineers the responsi-bility to maximize physical security through plant design and

          ~ layout as described in ANSI 18.17 (1973)."

In its implementation of Parts 50 and 73 the staff endeavors to l

ensure that an acceptable level of physical security is developed in parallel with the design and construction of the facility. During the initial phase, application for a Construction Permit, the applicant must provide evidence that all of the concepts of an acceptable security program are understood and will be implemented. The applicant's preliminary security plan that was submiti. un May 14,1976 as part of the Preliminary Safety Evaluation Report for the Black Fox Plant fulfilled this requirement. Subsequent to this submittal the staff and applicant developed lines of com-unication to facilitate transfer of information related to botl) physical and administrative methods of achieving an acceptable level of protection at Black Fox. To a large extent this transfer of information has consisted of the staff advising the applicant of revisions in part 73 and supplying guidance for implementing 373.55. This guidance has included both st#f-generated documents that describe acceptance criteria for meeting the specific require-( ments,of 373.55 as well as information related to physical security equipment and systems.- Included in the staff's guidance has also been a proposed format for the development of a security plan based on the requirements of 173.55. The staff has not required the applicant'to submit a revised preliminary plan since the applicant _ , _ _ has already comitted to complying with the requirements of this new section.

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In actions that parallel those required in g50.35 for the safety considerations,the construction permit will be subject to the limitation that a license authorizing operation of the Black Fox Plant will not be issued by the Comission until the staff has approved the security plan for this site. Consequently, the security plan that will be submitted in conjunction with a Final Safety Analysis Report (FSAR) in the second phase of the licensing process, will be reviewed by the staff against the requirements of 1 appropriate sections of the regulations. Since these regulations are being constantly reviewed and upgraded, when needed, by the Comission the criteria for an acceptable security program at the time of submittal of the Black Fox FSAR may differ from those considered acceptable to meet the requirements of 573.55. The staff believes that by maintaining close contact with applicants during the design and early construction stages the' physical components of a security plan can be implemented in a timely manner and any changes in regulations can be factored into the evolution of the overall plan. In this manner, assurance is gained that the security plan submitted at the FSAR stage will comply with the requirements of pertinent regulations. The staff will then review the security plan and, if necessary, work with the licensee to assure that the final version provides

5 the level of protection that meets the requirements of 173.55 (for superceding requirements) and provides reasonable assurance that the health and safety of the public will not be endangered by operation of the facility. The final step in this licensing process consists of an inspection of the site by the Office of  ; Inspection and Enforcement to ensure that the security plan is , fully operational before an operating license is issued. Until the reactor becomes operational a threat of industrial sabotage is - not valid. It is the staff's position,that an approved security plan developed to meet the requirements of 173.55 currently provid'es an acceptable level of protection, including structural integrity and redundancy of systems and components, against threats of theft of special nuclear material and industrial (radiological) sabotage.

3. I certify that the answers given are true and accurate to the best of my knowledge. .

3 s y' '; . g SUBSCRIBjDANDSWORNBEFOREMEa Ju/q THIS 7 DAY OF 'Jilliam J. Ross h{ , LD N' OY Mildred M. Groff  %  ; y -- On~~~ ~h ' st l ..

JUN 2. B M Vitae of William J. Ross . _._ . My name is William J. Ross. During the past two years I have been a Reactor Safegua.'js Analyst in the Reactor Safeguard Licensing Branch, Office of Nuclear Reactor Regulation, Nuclear Regulatory , Consnission. In this position I manage and perform reviews of site physical security plans provided to protect reactors at the site against industrial sabotage, seizure and theft of special nuclear materials. In addition I coordinate and perform reviews of operating reactor site safeguard contingency plans. These duties cover all aspects, of licensing nuclear power plants involving physical security i- and safeguards and include drafting of regulations, standards, and codes ,and subsequent review guidance as well as implementing these rules. For three years prior to holding this position I was a Project ,, Manager in the Divisions of Operating. Reactors and Environmental Projects. In this position I managed the review and evaluation of safety and environmental features associated with the construction and operation of nuclear power plants. Before joining the Commission in 1973 I participated as a nuclear chemist at tne Oak Ridge National Laboratory for 24 years in research and development programs related to the design of nuclear reactors and the interaction of nuclear related activities and their environments.

A-1-1 Contention A-1 I. Contention Intervenors contend that the Applicant and Staff have not adeouate': analyzed the cause and means of prevention of explosions resulting from hydrogen escaping from the BFS off-gas system which would be similar to the f1111 stone explosion. Enclosed affidavit by Jacques S. Boeoli addresses this subject. i II. Arcument The Black Fox off-gas system'is designed to acconnodate the effects of an off-gas explosion by being designed to a pressure of 350 psio. This is more than 20 times the nonnal operating pressure of the system. Boegli

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affidavit at 3. In addition, the proposac BFS system has a continuous j hydrogen analyzer to monitor and maintain steam dilution such that the mixture will be below the flamable limits for hydrogen and oxygen. A l control room alarm for explosive mixtures provides a reasonable assurance I that the plant operators will know and correct any conditions in which i explosive niixtures could exist in the BFS off-cas systs.m. The Boegli affidavit also states that the Applicant has proposed to use welded pipe connections and bellow steam seals on valves to mitigate any

       -    leakage from the off-gas system.         Liquid seals will automatically reseal the system after any pressure surges,.a: manual valve can be used to refill

A-1-2 the seal and a scienoid valve operates to isolate the seal. These valves fumish adequate protection from continuous leakage and wi11 prevent explosions caused by hydrogen and air accumulations in confined areas.

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The system also has redundant components which permit alternate gas treatnent pathways in case there is damage to the system. Boegli affidavit at 4-5. In response to Intervenors' concern that a " Millstone-like explosion" will occur at BFS, Mr. Boegli notes that the BFS system wil.1 utilize a steam dilution system to neutralize explosive mixtures and will not utilize the air dilution recirculation type system which was installed at Millstone. He further states that no explosions due to leaks from a steam dilution system similar to the proposed Black Fox system have been reported, and that the BF5 off-gas system is sufficient 1f different from the equiement which allowed the Millstone explosion so that a Millstone-like explosion is not expected. Boegli affidavit at 5. However, even if an explosion were to occur at BFS, mitigative measures would prov.ide a reasonable assurance that releases would be a small fraction of those limits con-tained in 10 C.F.R., Part 20.

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N.

A-1-3 III. Statement of Material Facts About Which No Issues Remain To Be Litigated

1. The Black Fox Plant will utilize a steam dilution off-gas system which is. not similar to that of the tiillstone plant.

2. The BFS off-gas system is designed to a 350 gsig pressure limit, more than 20 times the BFS operating pressure.

3. Operating experience at BWR's using steam dilution systems have reported no explosions due to leaks from a system.similar to that of l - Black Fox.

4. The BFS off-gas system is designed with redundant gas treatment pathways and mitigative provisions such that even if an exolosion were to occur, releases would be a small part of Part 20 limits.

N

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD IN THE MATTER OF DOCKET NOS. STN 50-556 PUBLIC SERVICE COMPANY OF STN 50-557 OKLAHOMA, ASSOCIATED ELEC-TRIC COOPERATIVE, INC., and WESTERN FARMERS ELECTRIC C00FERATIVE, INC. (Black Fox Stations, Unit Nos. 1 and 2) AFFIDAVIT OF JACQUES S. BOEGLI ON THE OFFGAS SYSTEM (CONTENTION A-1)

1. My name is Jacques S. Boegli. I am an employee of the U.S. Nuclear Regulatory Commission in the Effluent Treatment Systems Branch, Office of Nuclear Reactor Regulation. I am responsible for the review and evaluation of radioactive waste treatment and effluent control systems for nuclear power reactors. In this capacity, I authored Chapter 11 to the staff's Safety Evaluation Report (SER) for Black Fox Station, Unit Nos. 1 and 2.(NUREG-0190, June 1977),

evaluating the safety related matters in the proposed offgas system. My professional qualifications are attached to this affidavit.

2. The purpose of this affidavit is to address the intervenors' Additional Contention No. 1:

\

                                                                                         "Intevenors contend that Applicant and Staff have not adequately analyzed the cause 'and means of prevention of explosions resulting from hydrogen escaping from the off-gas system. Such explosions are apparently limited to BWR reactors and have associated secondary explosions, e.g., ignition of hydrogen in the base of the effluent release stack."

3. The Board requested the staff to address additional measures, based on operating experience, to prevent offgas explosions, like that which occurred at Millstone, Unit No. 1.

t

4. A sumary of operating experience of BWR off-gas systents is pro-vided in a report, " Technical Report on Operating Experience with Boiling Water Riactor Off-gas Systems," NUREG-0442, NRR-USNRC, l

Washington, D.C., 20555, April 1978. This report describes the generation and processing of "off-gas" in Boiling Water Reactors, the safety considerations regarding systems processing the "off-gas", operating experience involving ignitions or explosions of "off-gas" during more than 100 reactor years of BWR comercial

                        -. operation, and possible measures to reduce the likelihood of future  j ignitions or explosions, and to mitigate the consequences of such incidents should they occur. This report recomends a number of       1 preventive measures and contains IE Bulletin No. 78-03 (February 8, 1978) sent to operating BWR's requiring a review of operating and L                           maintenance procedures to assure proper operation of off-gas systems in accordance with all design parameters. This report concludes that there are no serious design flaws in the engineering of f

4 off-gas systems that meet the acceptance criteria for current design requirements as provided in Section II.3 of Standard Review Plan 11-3,

                        " Gaseous Waste Management Systems," NCREG-75/087, NRR-USNRC, Washington, D.C. 20555, November 24, 1975, which specifies:
                        "The gaseous waste handling and treatnent system should be des'igned to withstand the effects of an explosion, if the potential for an explosive mixture exists. Instrumenta. tion with automatic alarm and control functions should be proviced to monitor the concentra-
                                              ~

tions~of the appropriate gas in portiens of the systems having the potential for containing explosive mixtures. The design , should include precautions to stop continuous leakage paths, i.e., l to provide liquid seals downstream of rupture discs and to prevent l permanent loss of the liquid seals in the event of an explosion." (

5. In the evaluation, Chapter 11 of the SER,-I considered the design

     ~                   and operating pressures proposed by the applicant for the Black

(' - Fox Station, Unit Nos. 1 and 2, offgan system. All components l which will handle potentially explosive gas mixtures of hydrogen and oxygen are designed to a pressure of 350 psig and will operate at 7 psig at startup and less than 2 psig during normal operation. The design pressure of the system is greater than 20 times the normal operating pressure and the staff considers this an adequate design safety margin to withstand the effects of an internal explosion. Based on this staff position, I concluded that the pressure boundary of the system is designed to maintain its integrity in the event of a hydrogen explosion.

6. In the SER for Black Fox Station, Unit Nos. 1 and 2, I also con-sidered the instrument alar and control features. The applicants'

i l design requires'a continuoas hydrogen analyzer to monitor the  ! i l hydrogen concentration and automatically control the steam dilu-tion upstream of the recombiners to less than 4% (by volume) and below the flamable limits for hydrogen and oxygen. Alarm a features are proposed with local and control room indication of off-normal conditions. Therefore, I concluded that the instrumen-tation meets the acceptance criteria above, and will provide , reasonable assurance that no explosive mixtures will exist in the offgas system..

7. In'the SER for Black Fox Station, Unit Nos. 1 and 2, I also con-sidered the design provisions to mitigate leakage from the proposed off-gas system. The applicant has proposed to use welded pipe

( connections and bellow stem seals on valves in the off-gas system.. To prevent gas escape through drains, there will be liquid traps with adequate volume to automatically reseal aftitr a momentary liquid surge, with a drain to the main condenser and a valve to isolate the seal. I concluded that the above provisions provided adequate precautions to stop continuous leakage paths from the off-gas system and consequently to prevent explosions by excluding hydrogen (with air) accumulations from confined areas.

8. In addition, the offgas system proposed for the Bleck Fox Station, Unit Nos. I and'2, is designed with redundant components that will permit alternate paths in the event of ignition or explosion

  'l     q

k. l l

                                    .-. _      _                                 -          . . , , - -   ..a.

within the system. The charcoal beds will be continuously monitored for high temperature and may be isolated in the event of ignition, without loss of treatment.

9. The off-gas system proposed for the Black Fox Station, Unit Nos. 1 and 2, utilizes steam dilution rather than the air dilution-recircu-lation type system, like Millstone, Unit No. 1, that have had operating problems due to recombiner catalyst migration causing internal ignition or explosion problems. The incident reported I

December 13, 1977, at Millstone, Unit No. 1, resulted indirectly  ! l due to an operational problem with the air dilution type system. l The operating experience at BWR's using steam dilution has indicated fewer internal ignitions or explosions. In any event, there have I been no explosions reported due to leaks from a steam dilution type system similar to the off-gas system proposed'for'the Black Fox Station, Unit Nos l'and 2.

10. Based on the above details, it is concluded in the evtluation findings, Section 11.6 of the SER, that provisions incorporated into the applicant's design to prevent uncontrolled releases due to hydrogen explosions are consistent with our acceptance criteria, and therefore present reasonable assurance.that the off-gas i system will function safely. Since the Black Fox off-gas system is sufficiently different from that of Millstone, Unit No. 1, an

, 1

                           ~                     '

6-explosion of the type that occurred at Millstone is not expected at the Black Fox Station, Unit Nos. 1 and 2. Even were such an explosion to occur, the design features of the off-gas system at Black Fox Station, Unit Nos. 1 and 2, would provide reasonable assurance that the releases would not exceed a small fraction of the limits specified in 10 CFR Part 20. I hereby certify that the information above is true and accurate to the best of my knowledge.

                                                            .uwS acqu
                                                                        /Yrfd

5. Boeg11

( Subscribed'and sworn to before me this/d Date of V h /f 7 / -

                                    /                                                                  ,

Notary / Pub lic Sh. - -y

                                                //

My Commission Expires I / /972 . iy a  ; i i l

 ~

JacquesS.Bo$gli Professional Qualifications Effluent Treatment Systems Branch Office of Nuclear Reactor Regulation l My name is Jacques S. Boegli. I an a Senior Nuclear Engineer in the Effluent Treatment Systems Branch in the Office of Nuclear Reactor , Regulation. I attended Case Institute of Technology and received a l Bachelor of Science Degree in Chemical Engineering from Indiana Techni- ( cal College in 1951. In 1952. I received a Master of Science Degree in Chemical Engineering from Kansas State College. From 1955 to 1956 I co.91sted advanced courses in chemical and nuclear engineering at the University of Michigan and applied Health Physics training at the Oak l Ridge National Laboratory. . l From 1953 to 1973 I was employed by the National Aeronautics and Space Administration and held positions as research engineer in heat and mass transfer, design engineer in nuclear reactor coolant, utilities, ventila-tion and radwaste systems, proces systems supervisor, and technical consultant at the NASA Plua frook Reactor in Ohio. In July 1973, I joined the Nuclear Regulatory Countission (formerly AEC) as a Senior Nuclear Engineer in the Effluent Treatment Systems Branch, Division of Technical Review. In this position I am responsible for the review and evaluation of radwaste treatment systems and for the calculation of releases'of radioactivity from nuclear power reactors.

              '!he duties involve developing analytical models and performing calcula-tions on the effectivenes:: of proposed radwaste systems, studying
                          ~

technicalogical improvements and developias criteria governing radwaste processing, monitoring, shielding and handling. e G q S

m

                                                                . UNITED STATES OF AMERICA NUCLEAR REGULATORY C0ft11SSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of.                                           )

PUBLIC SERVICE COMPANY OF OXLAHOMA, ASSOCIATED ELECTRIC COOPERATIVE, INC. Docket Nos. STN 50-556 m 50-59 l' WESTERN FARERS ELECTRIC COOPERATIVE, INC. (Black Fox Station, Units 1 and 2) CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF'S MOTION FOR SUl91ARY DISPOSITION", dated July 14, 1978, in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, this 14th day of July, 1978: e r . j Steldon J. Wolfe, Esq. Mrs. Carrie Dickerson . Atemic Safety and Licensing Board Citizens Action for Safe U. S. Nuclear Regulatory Comission Energy Inc. _ ~ Washington, D. C. - 20555 P. O. Box 224 l Claremort, Oklahoria 74107 i I Mr.' Frederick J. Shon, Member Atomic Safety and Licensing Board Mr. Clyde Wisner U. S. Nuclear Regulatory Comission NRC Region 4 - -- - i Washington, D. C. 20555 Public Affairs Officer l 611 Ryan Plaza Drive l Dr. Paul W. Purdom Suite 1000 l Director, Environmental Studies Group - Arlington, Texas 76011 i Drexel University i 32nd and Chestnut Street Andrew T. Dalton, Jr. . Esq. L Philadelphia, Pennsylvania 19104 Attorney at Law i l 1437 South Main. Street, Room 302 Jos h Gallo Es Tulsa, Oklahoma 74119 Isham,Linco$n&Beale

                             'a i gton                         5b36                   Mrs. Ilene H. Younghein                      l 3900 Cashion Place Michael I. Miller, Eso.                                 Oklahoma City, Oklahoma 73112 Isham, Lincoln & Beale One 1st National Plaza                                  Paul M. Murphy Suite 2400                                              Isham, Lincoln & Beale Chicago, Illinois .60606' g

On'e First National Plaza, Suite 4200 ( L Chicago, Illinois 60603

r , -. _ _ _ _ _ _ _ _ - - 4%F e Atomic Safety and Licensing Atomic Safety and Licensing Appeal Board Board Panel U. S. Nuclear Regulatory Connission U. S. Nuclear Regulatory Commission-Washington, D. C. 20555 Washington, D. C. 20555 s Docketing and Service Section Mr. Maynard Human Office of the Secretary of the General Manager

           ' Commission                                         Western Farmers Coop. , Inc.

U. S. Nuclear Regulatory, Commission P. O. Box 429 Washington, D. C. 20555 Anadarko, Oklahoma 73005 Lawrence Burrell Mr. T. fl. Ewinn Route 1, Box 197 Acting Director 7 Fairview, Oklahoma 73737 Black Fox Station Nuclear Pro. ject Public Service Company of Oklahoma Mr. Gerald F. Diddle P. O. Box 201 General Manager Tulsa, Oklahoma 74102 Associated Electric Cooperative, Inc. P. O. Box 754 Dr. M. J. Robinson Springfield, Missouri 65801 / Black & Veatch P.O. Box 8405 sj 1 \ Mr. Vaughn L. *.onrad Kansas City, Missouri 64114 Public Service Company of Oklahoma P.O. Box 201 . Tulsa, Oklahoma 74102 L. Dow Davis Counsel for NRC Staff

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