Text

Trnscript of 820615 Hearing in Hauppauge,Ny.Pp 4,322-4,519. Supporting Documentation Encl
	ML20054F525
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Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	06/15/1982
From:	; Atomic Safety and Licensing Board Panel
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	ISSUANCES-OL, NUDOCS 8206170067
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{{#Wiki_filter:- NCCI.ZAR REGU*ATORY CC.E SS!CN BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the .%N d: : LONG ISLAND LIGHTING COMPANY : DOCKET NO.' 50-322-OL l (Shoreham Nuclear Power Station) : O DATI: June 15, 1982 PAGES: 4322 - 4519 - AT: Hauppauge, New York f f - l CY

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4322 lll 1 UNITED STATES OF AMERICA 2 NUCLEAR REGULATORY COMMISSION 3 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4 -- ---- -----------x 5 In the Matter of: : 6 LONG ISLAND LIGHTING CCMPANY a Docket No. 50-322-OL 7 (Shoreham Nuclear Power Station 8 - ---------------- : 9 Thi.rd Floor, B Building 10 Court of Claims a 11 State of New York 12 State Of fice Building 13 Hauppauge, New York h 14 Tuesday, June 15, 1982 1b The hearing in the above-entitled matter 16 convened, pursuant to notice, at 10:30 a.m. 17 BEFORE: 18 LAWRENCE BRENNER, Chairman

   '                                      Administrative Judge 19 20                     JAMES H. CARPENTER, Member 21                     Administrative Judge 22                     PETER A. MORRIS, Mem ber 23                     Administrative Judge

_ 24 WALTER H. JORDAN, Assistant to the Board 25 Administrative Judge , kh 1

 -t                                                              ALDERSON REPORTING COMPANY,INC, i                                        400 VIRGINIA A\E., S.W., WASHINGTON. 0.C. 20024 (202) 554-2345

4323 Y h 1 APPEARANCES: 2 On behalf of Applicants 3 T.S. ELLIS, III, Esq. 4 ANTHONY F. EARLY, Esq. 5 Hunton C Williams 6 707 East Main Street 7 Richmond, Virginia 23212 8 On behalf of the NRC Regulatory Staff: 9 DAVID A. REPKA, Esq. 10 RICHARD J. RAWSON, ESO. 11 EDWIN REIS, Esq. 12 Wa sh in g ton , D.C. 20555 13 On behalf of Suffolk Countys 0 14 LAWRENCE COE LANPHER, Esq. 15 Kirkpatrick,, Lockhart, Hill, 16 Christopher C Philiips

      -2       17              1900 M Street,         N.W.

Washington, D.C. 20036 { 18

 ,             19 2

20 21 2 22 23 24 25 L e ' ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345 ___ d

4323-A h 1 C_ g,E T_ E E T_ S_ 2 WITNESSES: DIRECT CROSS REDIRECT RECROSS g 3 Edward T. Burns, 4 George F. Dawe, George Garabedian, 5 Pio W. Ianni, Vojin Joksimovich, 6 Robert M. Kascsak, Paul J. McFuire, 7 Paul W. Rigelhaupt and a . ae 8 By Mr. Ellis 4341 g By Mr. Lanpher 4357 10 (AFTERNOON SESSION... page 4392) 11 Edward T. Burns, George F. Dawe, 12 George Garabedian, 13 Pio W. Ianni, Vojin Joksimovich, S 14 Robert M. Kascsak, Paul J. McGuire, 15 Paul W. Rigelhaupt and David J. Robare (Resumed) 16 By Mr. Lanpher 4392 17 18 Resolution of SC Contention 28(a) (iii)/ SOC Contention 19 7A(3) and SC Contention 27 (.c) / SOC Contention 3 (c) . . . . . . .page 43 29 20 Portions of LILCO's Testimony on SC/SCO 7B that LILCO Does Not Object to Striking; Errata Sheet LILCO 21 Testimony on SC/ SOC Contention 7B and SOC Contention 22 19B; Testimony of Messrs. Burns, Dawe, Garabedian, - Ianni, Joksimovich, Kascsak, McGuire, Rigelhaupt and 23 Robare for LILCO regarding Suffolk County /Shoreham Opponents Coalition Contention 7B and Shoreham Opponents 24 Coalition Contention 19(b) with attachments 1 thru 9....page 4346 25 O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., W ASHINGTON. D.C. 20024 (202) 554-2345

4323-B e

 $                         1                                             C_ O E T.,E E T,S.,

2 Additional FSAR Cites for LILCO's Testimony on SC/SCO

                          ,              7 B and SOC 19 (b ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . p age 4 3 5 5 4

5 RECESSES : 6 Noon - 4391 7 g Afternoon - 4454 9 10 11 12 13 O 14 15 16 17 18 19 20 21 22 23 24 25 O ALDERSON REPORTING COMPANY,INC. 400 VIRGINIA AVE S.W., WASHINGTON. D.C. 20024 (202) 554 2345

I 4324 i i O 1 enactsnissa 2 (10:30 a.m.) . g 3 JUDGE BRENNER: Good morning. We have a few

.)

4 preliminary matters and then we will go into our ruling-5 on the motion to strike portions of 7B, and then we will 6 proceed with the testimony after that. 7 1 would like to introduce with great pleasure 8 Judge Walter H. Jordan who is with us on the right. 9 When we get to Riverhead, we will have more room for all l 10 of us, and I do not think I need to detail his 11 background again. It is set forth in our order of May 12 28, 1982. And Judge Jordan will be assisting the Board 13 as both a technical interrogator and informal assistant O 14 on various issues through the case, including the issues 15 involved in Contention 7B. 16 The Board, -- on another matter, the Boa rd 17 yesterday received staff motion to strike a portion of 18 the testimony of intervenor's witnesses on SOC 1 19 Contention 9, Notification of Disabled Safety Systems. l 20 I would like to set a schedule for response by the other 21 parties. It can be an oral response, either this week 1 1 22 if possible, although if you prefer, we can do it at the 23 very beginning of next week. 24 Mr. Lanpher, would Suffolk County be handling 25 the lead on the response? O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4325 () 1 MR. LANPHER: No, sir. 2 JUDGE BRENNER: Could you cor.municate with 3 counsel for SOC that I would like to schedule -- I would 4 like a written respon3e this week if possible, next 5 Tuesday up here at the latest, or an oral response if 6 they would prefer. It is not a very involved motion and 7 we could handle it orally. And I am willing to schedule 8 it at their convenience, either at the beginning of the 9 day or the end of the day, this week or Tuesday morning, 10 so that should give them a lot of leeway. 11 If LILCO wishes to take a position on the 12 motion, perhaps they could coordinate through Mr. 13 Lanpher and SOC and we could do it all at the same time. O 14 MR. EARLEY: Yes, sir. 15 JUDGE BRENNER: In chambers on Friday we 16 discussed with counsel for all the cognizant parties the 17 possible extension of the schedule for the filing of 18 testimony now due on June 22. And I would like to get 19 that on the record if a date has been agreed upon. I 20 think the date we all focused upon was June 29th. Is 21 that correct, Mr. Lanpher? 22 MR. LANPHER: That is correct. That is 23 agreeable with Suffolk County, sir. 24 JUDGE BRENNER: All right. We, of course, 25 will give the extenstion to all parties, and this O . ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4326 () 1 involves the testimony on Suffolk County Contention 16, 2 Anticipated Transients without Scram, and the four 3 contentions, Suffolk County 12 through 15 involving QA 4 and QC matters. That testimony is now due here on 5 Tuesday, June 29th -- I am sorry -- in our offices in 6 Bethesda on Tuesday, June 29th. 7 We will not be in session that week, and we 8 vill not look favorably on a f urther extension. Ac the 9 parties know, that date was important because the Board 10 will have time to devote to the testimony that week. 11 We had testimony due yesterday, and at the 12 time I left the office the only testimony I had received 13 was the staff's. I did receive c message that LILCO's O 14 would be delayed because they had some word processing 15 problems. I know nothing of any testimony from Suffelk 16 County or 50C. 17 MR. LANPHER: Suffolk County filed testimony 18 yesterday on both issues. I was not down there. I 19 asked that it be messengered, if it was completed, by 20 the end of -- by 3:30 -- actually to get it to you, and 21 apparently tha crunch of the office did not allow it to 22 be hand-delivered late yesterday. But we did file it. 23 JUDGE BRENNER: It may have come af ter I lef t 24 yesterday. All richt, we will look for it on our return. 25 MR. LANPHER: On both issues, we did file i l l ALDERSON REPORTING COMPANY. INC. ) 400 VIRGINIA AVE S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4327 rm (-) 1 testimony. 2 JUDGE BRENNER: There were cross examination 3 plans due today, and I imagine we will receive them (~ } 4 today. The staff had filed theirs early at some point 5 last week, as I recall. 6 Tomorrow, probably tomorrow morning, the Board 7 would like te discuss schedules for the filing of 8 testimony on security issues and also, schedules for 9 discovery and the filing of testimony on on-site 10 emergency planning, or more accurately, LILCO's 11 emergency planning. And, in fact, if the parties can 12 agree on a discovery schedule for emergency planning, we

7. g 13 would not mind hearing that, but we realize the notice

\J 14 is short, sithough you should have been thinking about 15 as you approached the 22nd, and we will give you time 16 beyond tomorrow, if you have not fully agreed on such a 17 schedule. But we at least want to discuss the possible 18 parametes of the schedule. 19 HR. LANPHER: Judge Brenner, don't we have a 20 schedule for testimony on security issues? My memory is 21 that it is July 22 or somwhere in that range. 22 MR. ELLIS: I think that is correct. 23 JUDGE BRENNER: You are all one up on me this 24 morning. Yes. I did not transfer it over to my 25 calendar but I have it on the mast list as July 20. All ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE S.W., WASHINGTON. D.C. 20024 (202) 554-2345

4328 I () 1 right. I guers we will limit the discussion to 2 emergency planning tomorrow morning. 3 MR. LANPHER: That will be taken up first 4 thing, then? 5 JUDGE BRENNER: I would prefer that, unless 6 the parties need additional time for some details. o 7 MR. LANPHER: If I could get back to the Board 8 after the break or after lunch, I think someone other 9 than myself would like to be involved in that. 10 JUDGE BRENNERs Mr. Brown? 11 MR. LANPHER: Mr. Brown or Mr. McMurray. 12 JUDGE BRENNER: Are they here in town? 13 MR. LANPHER: No, but I believe Mr. Brown is 0 14 coming up. I just want to check his exact schedule. 15 JUDGE BRENNER: All right. Don't bring them 16 up just for that. As long as they are going to be here 17 this week, in any event, let us know and we will 18 schedule it accordingly. 19 MR. LANPHER: Okay, thank you. 20 JUDGE BRENNER: But I would like not to do it 21 the last day of the week. 22 MR. LANPHER: Fine. I will phone at lunch and 23 let you know. 24 JUDGE BRENNER4 Are there any minor 25 miscellaneous matters that have to be discussed this O l ALDERSON REPORTING COMPANY,INC, 400 VIRGINI A AVE., S.W., WASHINGTON, D.C. 20024(202) 554 2345

4329 l O i m-ning, in eddition to the ones 1 ouet covered 2 2 MR. EARLEY: Judge Brenner, we handed you this g 3 morning a copy of the signed Resolution of Concerns on 4 the iodine monitoring issue. And if it is appropriate, l 5 if we could have that bound into the record this morning. 6 JUDGE BRENNER: Fine, we can do that. I think ) 7 on Friday we considered binding it in at the time we 8 receive the report, but since it is ready -- and as far 9 as transcript time goes, it is in proximity to our 10 discussion of Friday morning -- let's bind it in at this 11 point. 12 (The Resolution of Concerns on the 13 iodine-monitoring issue follows:) 14 15 16 17 18 19 20 21 22 23 24 v 25 O ALDERSON REPORTING COMPANY. INC. 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

a 6 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION (J~ Before the Atomic Safety and Licensing Board In the Matter of )

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

(Shoreham Nuclear Power Station, ) Unit 1) ) RESOLUTION OF SC CONTENTION 28 (a) (iii)/ SOC CONTENTIOi4 7A(3) _AND SC CONTENTION 27(c)/ DOC CONTENTION 3(c) This agreement among LONG ISLAND LIGHTING COMPANY () (LILCO), SUFFOLK COUNTY (SC) , the SHOREHAM OPPONENTS CO-ALITION (SOC) and the NUCLEAR REGULATORY COMMISSION STAFF (NRC S taf f) resolves SC's and SOC's concerns regarding SC Contention 28 (a) (iii)/ SOC Contention 7A(3) and SC Con-tention 27(c)/ SOC Contet. tion 3(c). Both sets of contentions involve post-accident iodine monitoring equipment. SC and SOC are willing to withdraw SC Contention 28 (a) (iii)/ SOC Contention 7A(3) and SC Contention 27(c)/ SOC Contention 3(c) in return for LILCO's agreement to do the following:

1. LILCO will provide SC and SOC with a complete description of the post-accident sampling O system for the station vent (lDll-PNL-126).

This shall include information from the vendor's

manual, diagrams concerning the iodine monitor, the vendor's description of shielding at the () sampling station, and information concerning the counting equipment to be used to count the iodine samples.

2. LILCO will provide information or analysis to SC and SOC which document the accessibility of the station vent post-accident iodine monitor skid (1Dll-PNL-126), covering: (a) radiation levels at the location of this skid for two controlling accident conditions (design basis LOCA and Rod Drop Accident) and (b) personnel M

expogure estimates during filter change-outs.

3. LILCO will provide to SC and SOC an itemized quantitative assessment of factors contributing -

to the uncertainty of the release quantities of gaseous iodine-131, in particular: (a) flow rates (1) station vent (2) sampling auxiliary pump skid for Dll-PNL-126 (3) pump skid for Dll-PNL-126 (b) collection efficiency (c) plate out (time-history plot of transmission () factor for step increase in iodine concentration), and (d) uncertainties associated with the measurement of sample. l

LILCO shall provide the information listed above not later than July 15, 1982, except as may be extended by consent

 .. of SC and SOC, which shall not be unreasonably withheld.

It is understood that this resolution is without prejudice to the right of SC or SOC to submit a contention in the emergency planning proceeding which would contest the adequacy of the accura ^y of iodine monitoring at Shoreham. Assuming LILCO performs steps 1-3 above in accordance with this resolution, SC and SOC agree that the scope of an iodine monitoring contention in the emergency planning proceeding would not conte.et the details of the iodine monitoring system or LILCO's compliance with NUREG-0737 or Regulatory Guide 1.97 with respect to iodine monitoring but rather may allege that the accuracy achieved in iodine monitoring is not satisfactory to meet the requirements of 10 CFR Sec. 50.47, Part 50, Appendix E or NUREG-0654. It is understood that any such ccntention will be submitted by June 22, 1992, or within 14 iays after the information to be delivered pursuant to steps 1-3 has been received by the County, whichever is later. Respectfully submitted,

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                                         $}iOREHAM OPPONENTS COALITION DATED:     %       gg

4330 () 1 JUDGE BRENNER: All right. We can turn to 7B 2 testimony and reisted matters. We are denying Suffolk

-  3 County's motion to strike some of the portions of 4 LILCO's testimony on 7B related to the probabilistic 5 risk assessment, except for those portions which LILCO 6 ggreed to strike with a t least one adjustment, as we 7 will note at the end.

8 The motion was filed in writing by the county 9 on June 7. LILCO responded orally on the record the 10 morning of June 10; the staff filed a written response, 11 received the morning of June 14, as invited and 12 permitted by the Board, and we have taken all of these 13 things into account in our ruling. O 14 We agree, in essence, with LILCO's points that 15 the bulk of the portions of its testimony objected to by 16 Suffolk County are relevant to rebut allega tions raised 17 by the county's witnesses, in direct written testimony 18 and in response to questions. 19 Furthermore, the testimony is well within the 20 bounds of reasonable; in f act, perhaps expected response 21 to a contention which alleges that certain new 22 methdologies should have been, but were not, employed by 23 LILCO in assessing and classifying equipment, taking 24 into account systems interactions. 25 In short, it is open to LILCO to present O ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WASHINGTON. O C. 20024 (202) 554 2345

4331 () 1 testimony that it has included current state-of-the-art 2 methodology in its PRA, in addition to its testimony 3 that its more traditional analysis is also adequate. 4 Showing, in LIICO's view, a systematic way of l 5 classifying safety componen ts and analyzing systems l l 6 including systems interactions. 7 Some of the confusion, and the main basis for 8 the County's motion to strike, stems f rom the Board 's t 9 observa tion in our confirmatory discovery order of March 10 30, 1982, which, among other things, granted l l 11 intervenor's motion to compel a deposition on and 12 production of the draft PRA. We granted the discovery, 13 as noted at page 3 of that order, within the scope of O 14 the request for deposition by the county, which we 15 quoted on page 1 of th e ord e r, a s a request to inquire 16 into "the scope objectives, methodology, description of l 17 research and analytical tasks involved in the 18 probabilistic risk assessment (PRA) currently being l 19 performed for LILCO by Science Applications, Inc., as l 20 well as LILCO 's intended utilization of the PR A. " l l 21 The testimony by LILCO not being struck comes 22 well within that description. We did attempt to clarify 23 on page 2 of that March 30 order that we did not intend, 24 by the grant of~the discovery to the county, to get into 25 the details of the correctness of the analysis in the O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

4332 () 1 PRA. This was primarily because of our ruling on 2 admi tted reworded Contention 7B, which emphasized that 3 it was a methodology contention with insufficient 4 specificity to inquire into particular problems, other 5 than the leeway we indicated with respect to examples. 6 This is distinguishable from the case where 7 PRA is being relied on by the utility and/or the staff 8 to support its safety conclusions, and also, where there 9 are specific admissible contentions detailing alleged to defects in the PRA. 11 Page 2 of our March 30 crder stated that the 12 draft PRA would not be used to litigate the correctness 13 of either the application of the methodology in the O 14 draf t PR A or the correctness of the draft PBA results of 15 probabilities and consequences of accidents. The term j 16 " correctness of the application of the methodology" may l 17 be ambiguous by itself, and I will attempt to clarify it 18 in a moment. However, we believe our intent was fairly 19 clear from our overall order admitting Contention 7B, 20 and also, the entire order of March 30 granting the 21 discovery to the county of the PRA. 22 The term " application of the methodology" can , 23 invoke two distinct concepts. One is the correctness of 24 the qualitative analysis performed pursuant to a defined 1 25 methodology; the other is a more overall determination ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S,W., WASHINGTON, D.C. 20024 (202) 554 2345

4333 () 1 of scope and content to ascertain what methodologies are 2 used, and in fact, to ascertain whether two different (*g 3 parties are talking about the same thing by their use of V 4 the same label. 5 The former, that is, the correctness of the 6 qualitative analysis performed, is what we did not want 7 to get into. The latter, concept of the determination 8 of scope and content to ascertain just what 9 methodologies are being applied, is within the scope of 10 the contention and clearly within the scope of rebuttal 11 to the intervenor's allegations and testimony. 12 As stated by Dr. Burns, one of LILCO's 13 witnesses, on page 86 of the testimony, we are O 14 interested in the methods used in the PBA and how those 15 compare with other available technologies. 16 As we have attempted to indicate, we do not 17 think that this was a close call, but if it had been, we 18 might have been inclined to grant some leeway to LILCO 19 in the choice of its presentation of the case, 20 commensurate with what we consider to be the very g rea t 21 lee wa y we permitted the county in the presentation of 22 its direct case. 23 Speaking of close calls, we get to the matter 24 of the two attachments. The county has moved to strike 25 Attachment 2 and Attachment 4 of the testimony. We are ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4334 () 1 not going to strike them , b ut they are close calls. 2 They are marginally the type of material that experts in 3 the field may rely upon. They are not published 4 analyses; they are letters. However, given the purpose 5 for which th e testimony purports to use them, we think 6 it is fair to leave them in, and we are capable of l 7 giving them the weight they deserve, depending on where , 8 the examination on them, if any, goes. 9 Attachment 2 is referenced on page 31 of the 10 testimony as background to the promulgation of ANS-22, 11 the precursor of ANSI ANS-52.1. Putting aside the 12 PWR-BWR distinction, which we think is explained in the l 13 testimony and in Mr. Ellis's response to the motion to O 14 strike; that is, the BWR approach was based on the PWB 15 approach, but even putting that aside, we somewhat share 16 with the county uncertainty as to what purpose there is 17 in including that portion of the attachment. We are not 18 discussing the last two pages, which is the forward. It 19 is just not clear what point it is being used to 20 support, bat we will leave it in. If it embellished, we 21 can deal with it. If it is not, it will not mean much 22 as it is. 23 Attachment 4 is referenced on page 55 of the ( 24 testimony. It is a letter to the NRC from the Nuclear 25 Power Committee of the IEEE. We understand the O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

l 4335 O ' reteveace or war tais ettecameat 1e deiae tacteded- ra l 2 essence, LILCO wants to show that they are not the only 3 ones complaining, that at least somebody else in the 4 industry hits problems with the Denton memorandum. 5 To that limited extent, it is relev ant . If 6 there is any attempt to carry it beyond the point; that 7 is, the letter by itself demonstrates the validity of 8 these complaints, we will not ascribe weight for that 9 proposition. LILCO's witnesses presented here will have I 10 to support whatever disagreements they have with the ! l 11 concept outlined in the Denton memo. But it is relevant l 12 to show tha t LILCO is not the only member of the 13 industry who has the problem. 14 15 16 17 18 19 20 21 22 23 b) 24 25 O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D C. 20024 (202) 554 2345

4336 () 1 In addition -- and somewhat in passing -- the 2 attachment 4 may be pertinent te inferences, depending 3 on where the examination goes, on the status of the IEEE g 4 Standard P-827, which the County's witnesses, I believe 5 primarily Mr. Hubbard if I recall correctly, referred tc 6 a number of tim es . And this standard is discussed, at 7 least the status of the standard is alluded to, in this 8 letter from the Committee that should be certainly 9 cognizant as to that status. 10 And it is a relatively recent letter, I 11 believe May 1982, although I do not have it in front of 12 me, as contrasted with attachment 2, which dates back to 13 the early 1970's, the first portion of attachment 2. 14 With respect to the list of matters which 15 LILCO agreed to strike, as I stated at the outset, those 16 portions will be stricken with a t least one adjustment l 17 and then a comment or two. The typed list that LILCO 18 provided me indicates that, on page 80, they agree to i 19 strike the last sentence of the first paragraph. 20 As I look at that paragraph which begins on 21 page 80, I have a recollection that -- well, I guess it 22 is fair to say LILCO indicated a "maybe" as to the first 23 sentence. We think the first sentence is very much of 24 the same stripe as the last sentence, and consistent 25 with what we infer to be LILCO's concession in the ALDERSON REPORTING COMPANY,INC, 400 VIRGINTA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4337 () 1 portions it agreed to strike, I guess I would like to 2 inquire of LILCO, shouldn ' t tha t sentence no out also? g 3 The sentence is "My preliminary evaluation 4 has not iden tified any non-safety-grade system the 5 failure of which by itself and without multiplc failures 6 in the safety-grade systems would cause core 7 d eg ra da tion . " 8 MR. ELLIS: Judge, I think you are correct, in 9 fact, that it's in the' same sense. On the last 10 sentence, I see our note indicated that it was the 11 entire last sentence, and if that is what we said that 12 is fine. 13 I thought what I had said in argunent is that 14 the competent state of the art ef fort remark should 15 remain, but that the portion which says " confirms the 16 safety adequacy of the Shoreham design" was a conclusion 17 which should go. That was my recollection. But I think 18 the list that we handed to the Board did not make that 19 discriminating distinction. 20 JUDGE BRENNER: All right. Let's stay with 21 the first part first. In addition to the list we 22 provided, we would add the first sentence of page 80. 23 But I think the list should be bound in with the 24 testimony, and perhaps you can add first and last 25 sentence, that portion of the last, and in addition the ) /~T l \/ ALDERSON REPORTING COMPANY,INC, 400 VIRGINI A AVE S.W., WASHINGTON, D.C. 20024 (2C2) 554-2345

4338 () I testimony should physically have the portions crossed 2 out. 3 Your next point anticipated my next comment. 4 If you want to leave that portion in, you will not solve 5 your problem just with that, becau.se 'here are three. 6 others that are very similar that you take out. And I 7 did want to make the observation about them , and you may 8 have indicated tha t you wished to leave that portion in 9 on the record, I do not remember. 10 Standing by itself, the statement which you 11 just alluded to, "It is my professional opinion that the 12 Shoreham PBA is a competent state of the art effort," 13 that could mean either the details of the quantitative O 14 analyses, which we do not want to get into, or it could 15 stand for the proposition that the current methodology 16 was employed, which would be okay. 17 As long as it is ambiguous, I think it should 18 come out. Had it been clearly for the latter point, it 19 could have stayed in. But it is tied up in the context l 20 of other portions before and af ter it and in fact in the 21 same sentence that you are striking, and I just do not 22 want to get into that wordsmithing on a motion to i 23 strike. 24 I will point out the two other places where 25 these same comments apply, and that is on page 85, I O ALDERSoN REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTCN, D.C. 20024 (202) 554-2345

4339 () 1 believe it is. Yes, the last sentence there which you 2 are striking. In fact, you are striking the entire 3 paragraph. The same comment would apply. ggg 4 In addition, although somewhat differently 5 worded, th e second f ull para'raph on page 109, which you 6 are agreeing to strike, which contains a listing of 7 three conclusions. The first conclusion is similar to 8 what we just discussed. That is, the conclusion is, 9 "The Shoreham PRA is an adequate evaluation of the 10 systems interaction issue within the current state of 11 the technology." 12 It appears closer to the part that we are 13 permitting inquiry into. Hovaver, it is not perfectly 14 clear and I would note the conclusion by itself would 15 not mean much in any event. Whether or not that 16 conclusion gets upheld in the proper context is going to 17 depend on a lot more than just that one sentence either 18 way. 19 So it will be out. But I wanted to make that 20 comment in an attempt, perhaps one doomed to failure, to 21 shed further light on the distinction that we have drawn 22 in ruling on the motion to strike. 23 (Pause.) 24 I think with that we can launch into the 25 presentation of LILCO's testimony on this contention. O ALDER 50N REPORTING COMPANY, INC, 400 VIRGlNI A AE, S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4140 1 Let's go off the resord. 2 (Discussion off the record.) g 3 JUDGE BRENNER: All right, let's go back on 4 the record. 5 There was one matter I forgot to mention. I 6 d o ha ve a list of Staff documents and LILCO documents 7 which we could stipulate in. I do not want to do it 8 right now, but I want to make sure we can do it first 9 thing tomorrow morning. That is, if SOC's lawyer feels 10 he has to be here for that, he should be here because, 11 unless there is some problem with the stipulation that I 12 do not anticipa te, we will just do it first thing in the 13 mornino. And please remind me if I forget. 14 All righ t. Gentlemen, before you get too 15 comfortable, could you all please stand. 16 Whereupon, 17 EDWARD T. BURNS 13 GEORGE F. DAWE 19 GEORGE GARABEDIAN 20 PIO W. IANNI 21 VOJIN J0KSIMOVICH 22 ROBERT M. KASCSAK 23 PAUL J. McGUIRE /~\ V 24 PAUL W. RIGELHAUPT 25 DAVID J. ROBARE, O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE.. S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4341 ( \_)/ 1 called as a witness by counsel for Applicant, having 2 first been duly sworn by the Chairman, was examined and ('T 3 testified as follows: 4 (Discussion off the record.) 5 7UDGE BRENNER: Let's go back on the record. 6 MR. ELLIS May I proceed, Judge? 7 JUDGE BRENNER: Yes. 8 DIRECT EXAMINATION 9 BY MR. ELLISs 10 C With the panel's permission, I would like now 11 to introduce the panel individually and then proceed to 12 introduce the testimony. First, Mr. Rigelhaupt, would 13 you state four full name and address for the record, \ / 14 please, sir. 15 A ' WITNESS RIGELHAUPT) My full name is Paul 16 William Bigelhaupt. My residence address is 37 17 Royalston Road, Wellesley, Mass. 18 Q And with whom are you affiliated, Mr. 19 Rigelhaupt? 20 A ( WITNESS RIGELH AUPT) I am affiliated with the 21 Stone E Webster Engineering Corporation in Boston. 22 0 Til right. Mr. Garabedian, would you nov 23 state your f ull name and residence address for the 24 record, please. 25 A (WITNESS GARABEDIAN) My full name is George D U ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE., S.W WASHINGTON, D.C. 20024 (202) 554-2345

4342 l l l ( 1 Garabedian. My residence is 130 Fulton Street, Boston, 2 Massachusetts. () 3 0 And what is your affiliation, sir? 4 A (WITNESC GARABEDIAN) I work for Stone E 5 Febster Engineering Corporation. 6 0 And while we are at it, let me go back and ask 7 both Mr. Rigelhaupt and Mr. Garabedian, are your 8 professional qualifications found in the attachment to 9 the testimony, the 7.D testimony that was prefiled on to behalf of LILCO? 11 A (WITNESS RIGELHAUPT) That is correct. 12 A (WITNESS GARABEDIAN) That is correct. 13 0 All right. Mr. Dave, would you give your full O 14 name and residence address for the record, please. 15 A (WITNESS DAWE) Yes. My name is Geoge F. 16 Dave, Jr. My residence address is 21 Strawberry Hill 17 Lane, Danvars, Massachusetts. I am employed by Stone C 18 Webster Engineering Corporation in Boston. 10 0 And are your professional qualifications also 20 found in the attachment? 21 A (WITNESS DAWE) Yes, sir, they are. 22 0 All right. Mr. Bobare, would you now sta te 23 your f ull na me and residence address, please. 24 A (WITNESS ROBARE) My name is David J. Robare. 25 My residence address is 1448 Carnot Drive, San Jose, O ALDERSON REPORTING COMPANY,INC. 400 VIRGINI A AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4343 1 California. I am with the General Electric Company. 2 0 Mr. Ianni, would you do the same, please, 3 sir. 4 A (WITNESS IANNI) My name -- 5 0 Oh, excuse me. Let me ask Mr. Robare, are 6 your professional qualifications also included in the 7 ittschment? 8 A (WITNESS ROBARE) Yes, they are. 9 0 Thank you. Mr. Ianni, would you state your 10 full name and residence address for the record. 11 A (WITNESS IANNI) My name is Pio W. Ianni. My 12 residence address is 1759 Newmark Court, San Jose, 13 California. And I am with the General Electric O 14 Company. 15 0 All right. Mr. Kascsak, would you state your 16 -- are your professional qualifica tions attached to the 17 testimony? 18 A (WITNESS IANNI) Yes, sir. 19 0 All right. Mr. Kascsak, would you now state 20 your name and address, please. 21 A (WITNESS KASCSAK) Yes. My name is Robert M. 22 Kascsak. My residence address is 3 Hobart Court, 23 Huntington, Long Island, and I an employed by Long 24 Island Lighting Company. 25 0 All right. M r. McGuire -- are your O ALDERSON REPORTING COMPANY, INC, 400 VIRGINTA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345 l j

          #-_m.a.          a A -      A u.  *_     -._-.+u   A 4300

() 1 professional qualifications attached ? 2 A (WITNESS KASCSAK) Yes, they are. 3 0 Thank you. ( ]) 4 A (WITNESS McGUIRE) My name is Paul J. 5 McGuire. My residence is 1815 North Bengal Road in 6 Metairie, Lousiana, and I an employed by the United 7 Energy Services Coapany. 8 Q Are your professional qualifications attached 9 to the testimony? 10 A (WITNESS McGUIRE) Yes, they are. 11 0 All righ t. Dr. Joksimovich, would you state 12 your full name and residence address for the record, 13 please. l () 14 A (WITNESS J0KSIMOVICH) My name is Bojin 15 Joksimovich. The address is 406 Hidden Hills Lane, 16 Escondido, California. 17 0 And what is your professional affiliation? 18 A (WITNESS J0KSIMOVICH) I am manager of the San 19 Diego office for NUS Corporation. 20 Q Are your professional qualifications a'tiached 21 to the testimony? 22 A (WITNESS JOKSIMOVICH) Yes, they are.' 23 0 Thank you. Dr. Burns, would you give us your 24 full name and address, pleace. 25 A (WITNESS BURNS) Edward T. Burns, 275 Blossom O ALDERSON REPORTING COMPANY. INC. 400 VIRGINIA AVE., S.W., WA?lliNGTON, D.O. 20024(202) 554 2345

4345 () 1 Valley Drive, Los Gatos, California. And I am 2 associattd with Science Applications, Incorporated. And 3 sy qualifica tions are in the testimony. (} 4 Q All right. Now, I would like to ask Mr. Dave 5 as spokesman for the panel: Mr. Dave, do you have a 6 list of corrections to the testimony? 7 A (WITNESS DAWE) Yes, sir, I do. 8 MR. ELLIS: These are the corrections, Judge, 9 which have been distributed to the panel and the 10 Intervenors, and slso to the reporter. 11 BY HR. ELLIS: (Resuming) 12 0 I would like each of you to answer now 13 indidivually , if you would, please. Do you have -- have O 14 you had an opportunity to review the testimony, and do 15 you have with you today the testimony on behalf of LILCO 16 on contention 7(b) and 19(b)? Mr. Dave, you may -- 17 A (WITNESS DAWE) Yes, we do. 18 0 Again, answer each of you individually, 19 beginning with Mr. Rigelhaupt, if you would. Is the 20 testimony as corrected, together with the attachments, 21 true and correct to the best of your knowledge and 22 belief, given your involvement in the testimony? 23 A (WITNESS RIGELHAUPT) That is correct. 24 A (WITNESS GARABEDIAN) That is cor rect. 25 A (WITNESS DAWE) Yes, sir, it is. O ALDERSON REPORTING COMPANY,INC. 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4346 ( 1 A (WITNESS ROBARE) Yes. 2 p, (WIINESS IANNI) Yes. (~} %J 3 A (WITNESS KASCSAX) Yes, it is. 4 A (WITNESS McGUIRE) Yes, it is. 5 A (WITNESS J0KSIMOVICH) Yes, it is. 6 A (WITNESS BURNS) Yes, sir. 7 MR. ELLIS: With that, Judge Brenner, we would 8 move the admission of the testimony and the attachments, 9 the two volumes, into evidence, with the stricken 10 portions as ruled on by the Court and the corrections as 11 supplied to the court. I believe the court reporter has 12 been given a corrected copy. , 7, 13 JUDGE BRENNER: All rich t. Let's sequence b 14 them with a list of the portions that LILCO has agreed 15 to strike, hopefully annotated, with that addition of 16 the first sentence that we discussed, then the errata 17 sheet, then the testimony, and then the attachments. 18 MR. ELLIS: Yes, sir. 19 JUDGE BRENNERa In the absence of any 20 objections beyond those which we have ruled upon, we 21 will admit these into evidence as requested. 22 (The documents referred to, LILCO witnesses' 23 testimony regarding contentions 7(b) and 19(b), with 24 attachments, follow ) 25 O (_) ALDERSON REPORTING COMPANY,INC, 400 VIRGINI A AVE.. S W., WASHINGTON, D.C. 20024 (202) 554 2345

j_ ! 4 [If Portions of LILCO's Testimony on SC/ SOC 713 That LILCO Does Not j ; Object to Striking (_) Page 79, last paragraph f,est a m) Page 80, last sentence of the first paragraph Page 84-85, laut paragraph on page 84, including carry over to page 85 Page 108, last sentence Page 109, first centence Page 109, lact sentence of the fi rs t full paragraph Page 109, second full paragraph Page 117, laut sentence in second paragraph ("Iteaponse " ) (' ') Page 118, last sentence i I s

 +/cr ERRATA SiiEET LILCO TESTIMONY ON SC/ SOC CONTENTION 7B AND SOC CONTENTION 19 (b) v PAGE                     NATURE OF CIIANGE 18  Third p gagraph, second line:

Change the word " combination" to " containment" 31 First full paragra g 14th line: Change the word " sustaining" to " retaining" 42 Fourth line from top: Insert "I" after "(CAR" First_ paragraph, last sentence should read as follows:

                 "In this regard, QAR II is selectively supplemented with requirements such as for surveillance, QC records or radiographs based on engineering judgments."

O(~N Second paragraph, first line: Insert the word " essentially" before " identical" - 55 Fourth and fifth lines from tog: l Insert a period after " mind" and delete "when vjewed in this light, but this fact has not been ignored a t Shoreham. " 59 First line at top of page: Change "233.90" to "223.90" 65 Second paragraph, second line: Change the word " offers" to " offer" Second paragraph, fifth line: ('h sg Change the word " examines" to " examine" i l l l

i 1 l l i i PAGE NATURE OF CilANGE r i I i 66 Second full paragraph, third line: Change the word " trees" to " tree" ! T'T f, U 68 First full paragraph, seventh line: j Change the word "interrelationshiop" to " interrelationship" \ l 77 Third paragraph, ninth line: ! l Insert the word " trees" after " fault /cVent" l 128 Item (E) Change " Table 3.2-1" to " Table 3.2.1-1" l c l 144 Second paragraph, first line: i i l Strike the word " unqualified" ' Insert the phrase "that are not safety related" after the word "RCIC" 145 Last line on page: O Strike the word " low" , 146 First f ull paragraph, first line: , Change the word " diversity" to " redundant" j t 149 Penultimate line: i, Change the word " valve" to " system" 152 Sixth line from top: (in parenthetical) ' Strike the word "at" Insert the phrase "resulting in drywell" before I the word " temperatures" f t 172 Last paragraph, first line: I Insert the word "not" after the word "is" O . i l

4 ATTACl! MENT 1 -- PROFESSIONAL QUALIFICATIONS WITNESS PAGE NATURE OF CilANGC George Garabedian 1 second paragraph, seventh line: Change "1978" to "1968" Pio W. Ianni 3 l'ou r th line: Change "ECCW" to "ECCS" 3 First full paragraph, second line: Change "with" to "within" Paul J. McGuire 1 Third full paragraph: Add the following at the end of the paragraph:

                                                                                             "In 1970, I was slated to be one of the chief engineers at O                                                                                           Shoreham and worked on Shoreham for eight months."

Paul W. Riegelhaupt 1 Third paragraph, f o u r t.h line: Change "1940" to "1948" 4 First full paragraph, second line: Change "nd" to "and" and "19 years" to "11 years" ATTACIIMENT 5 Delete page 5. t 0 I _ _ _ _ _ _ _ _

Y f[ff LILCO, May 25, 1982 UNITED STATES OF AMERICA I NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

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

(Shoreham Nuclear Power Station, ) Unit 1) ) TESTIMONY OF EDWARD T. BURNS, GEORGE F. DAWE, GEORGE GARABEDIAN, PIO W. IANNI, VOJIN JOKSIMOVICH, ROBERT M. KASCSAK, PAUL J. MCGUIRE, PAUL W. RIGELHAUPT and DAVID J. ROBARE FOR THE LONG ISLAND LIGHTING COMPANY REGARDING

)                SUFFOLK COUNTY /SHOREHAM OPPONENTS COALITION CONTENTION 7B AND SHOREHAM OPPONENTS COALITION CONTENTION 19(b).

I

LILCO, May 25, 1982 hff UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

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

(Shoreham Nuclear Power Station, ) Unit 1) ) TESTIMONY OF EDWARD T. BURNS, GEORGE F. DAWE, GEORGE GARABEDIAN, PIO W. IANNI, VOJIN JOKSIMOVICH, ROBERT M. KASCSAK, PAUL J. MCGUIRE, PAUL W. RIGELHAUPT and DAVID J. ROBARE O FOR THE LONG ISLAND LIGHTING COMPANY REGARDING SUFFOLK COUNTY /SHOREHAM OPPONENTS COALITION CONTENTION 7B AND SHOREHAM OPPONENTS COALITION CONTENTION 19(b)_ 4 I 1 O {

I PURPOSE () This testimony responds to SC/ SOC Contention 7B and SOC Contention 19(b), and responds to the concerns raised by the Board at the close of hearings on May 7. The purpose of the testimony is best summarized by its conclusions: This testimony has demonstrated that LILCO and its contractors have applied a proven, well-established and accepted methodology to the design and clas-sification of structures, systems and components at Shoreham to ensure that the design basis of the plant properly satis-fies the regulatory requirements of the General Design Criteria 1, 2, 3, 4, 10, 13, 21, 22, 23, 24, 29, 35 and 37. The design has been developed and implemented in a disciplined and con-O' trolled manner by GE and SWEC. The design and the design process were developed from extensive experience. Shoreham and its design basis and clas-sification of structures, systems and components is comparable to all contem-porary BWR's, many of which are and have been licensed and operating. The determination of the clas-sification of structures, systems and components was based on the experience of designers in implementing a large body of knowledge reflected and documented in NRC regulations, regulatory guides and industry standards. These gu.idance docu-ments were themselves developed from a systematic approach to nuclear plant design and classification of structures, systems and components. Systems interac-tions were clearly considered in the () design process, with the classification system itself aiding the designers in assuring that the potential for adverse interactions was minimized in the design

process. Further, the design has been checked in many instances with studies L

which investigated specific potential systems interactions. The Shoreham design does include () special attention to non-safety related structures, systems and components. As reflected in this testimony, non-safety related structures, systems and compo-nents are accorded quality assurance and control commensurate with the functions they perform. The inclusion of non-safety related systems in EOP's is based on the sensible principle that operators should be direc-ted to use the full capabilities of the plant to deal with transients and other events because these normal, non-safety related systems are required to be highly reliable and their use in these circum-stances will often make it unnecessary to call upon or challenge the safety related systems. In any event, if the normal, non-safety related systems fail to per-form when called upon to mitigate an {x-}/ event, the full safety related systems are still available to mitigate the event and prevent core damage. There is, therefore, no reason to upgrade a non-safety related system to safety rela-ted simply because it is mentioned in an EOP. Analysis confirms that the set of safety related systems is an adequate and sufficient set to protect the plant from design basis accidents. The use of non-safety related systems in EOP's is an appropriate use of the plant's full capa-bilities in these circumstances and demonstrates again that the plant's capa-bilities take it beyond the design basis analysis. Thus the design basis for Shoreham has been correctly and conservatively established. In implementing the design 73 basis, structures, systems and components

 ~ ,)    have been properly classified considering their importance to safety and reliabi-lity. The items to which GDC 1,  2, 3, 4,

10, 13, 21, 22, 23, 24, 29, 35 and 37 apply are properly identified at Shoreham and the conformance by Shoreham to these CDC's is demonstrated in the FSAR and has

 /~')     been approved by the NRC Staff in the b        SER.

In addition to all of the foregoing, LILCO, on its own initiative and without regulatory or NRC Staff requirement, has commissioned a level 3 PRA for Shoreham. This PRA has thus far confirmed the safety adequacy of the plant design at Shoreham. More particularly, the Shoreham PRA indicates that no unique, adverse interactions between non-safety related and safety related systems exist at Shoreham and that the probability of core melt from the operation of Shoreham is as extremely unlikely as it is for Peach Bottom Nuclear Plant as confirmed by the WASH 1400 study. In summary, therefore, Intervenors' allegations that Shoreham employed no

^)

sj adequate methodology for classifying sys-tems and that systems interactions were ignored at Shoreham are both entirely without basis. As this testimony shows, an adequate methodology has been used to classify structures, systems and compo-nents at Shoreham and systems interac-tions have been appropriately addressed at Shoreham. O (/ _3 L

1 l ATTACHMENTS

1. Professional Qualifications l

2. Information Relating to ANS-22 1

3. FSAR Figure 7A-26

4. IEEE Correspondence

5. Attachments to the Testimony of Dr. Vojin Joksimovich

6. Table 3.1 of the Testimony of Dr. Edward T. Burns

7. Selected Portions of the Standard Review Plan, NUREG-0800 (Revision 1, July 1981)

Selected Portions of the Proposed O 8. Shoreham Technical Specifications l

9. Figures Regarding Reactor Water Level Instrumentation 1 i

l O

I AUTHORSHIP ATTRIBUTION O

In general, responsibility for the testimony may be summarized as follows

! Eaction II.A Pio Ianni David Robare Section II.B Paul Riegelhaupt George Dawe George Garabedian i Section III George Dawe

David Robare Pio Ianni George Garabedian Section IV.B David Robare

() Section IV.C George Dawe George Garabedian Paul Riegelhaupt Section IV.D Robert Kascsak Section IV.E George Dawe David Robare Section V George Dawe David Robare Robert Kascsak Section VI.A Dr. Vojin Joksimovich Section VI.B Dr. Edward T. Burns Section VI.C Robert Kascsak O

-i-

l Section VII.A Paul McGuire Section VII.B David Robare George Dawe Section VII.C David Robare George Dawe Section VII.D David Robare Section VII.E George Dawe David Robare Section VII.F George Dawe David Robare The introduction and conclusion to the testimony is sponsored by all witnesses; the introductions to each i section are sponsored by the respective authors of the

section.

[} Professional qualifications for the witnesses are included in Attachment 1. . l 4 i i 4 l 4 f O 1

l TABLE OF CONTENTS O I. Introduction..................................... 5 II. General Electric and Stone & Webster Design and operating Experience......................... 7 . A. General Electric............................. 8 B. Stone & Webster.............................. 20 III. Methodology for Classification of Systems at Shoreham...................................... 27 A. Use of Design Basis Analysis................. 27 B. American Nuclear Society Standard 22 and Nuclear Safety Operational Analysis...... 30 C. Regulatory Guides 1.26 and 1.29.............. 35 fs D. Applicable Regulations....................... 38 b E. Operating Experience......................... 40 IV. Treatment of Non-safety Related Structures Systems and Components................................... 41 A. Introduction................................. 41 B. General Electric.............................. 42 C. Stone & Webster.............................. 44 l ! D. LILCO........................................ 48 E. Response to Intevenors' Use of Denton Memorandum............................ 50 V. Consideration of Systems Interaction at Shoreham..... 56 (a) Pipe Failure and Internal Flooding........... 56 () (b) Missiles..................................... 57 (c) Fire Hazard Analysis......................... 57 (d) Cable Separa. tion............................. 57

(e) Failure Mode and Effects Analysis (FMEA)....................................... 58 O) \- (f) Electrical Bus Failures...................... 58 (g") Control System Failures...................... 59 (h) High Energy Line Break....................... 59 (i) Probcbilistic Risk Assessment................ 60 (j) Heavy Loads.................................. 60 (k) Analysis of Industry Experience.............. 61 (1) Pre-operational and Startup Testing.......... 61 (m) Protection Systems........................... 63 (n) Scram Reliability............................ 63 (o) Common Mode Failures in . Protection and Control....................... 64 O (p) Water Level Instrumentation.................. 64 (q) TMI-2 Implications........................... 64 (r) Shoreham PRA................................. 64 VI. Shoreham PRA A. Testimony of Dr. Joksimovich................. G5 B. Testimony of Dr. Burns....................... 87 C. Testimony of Mr. Kascsak.................... 121 4 VII. Further Response to Specific Points Raised in Intervenors' Testimony...................... 129 A. Use of Emergency Operating Procedures for Classification of Systems............... 130 () B. Alleged Improper Classification of Four Systems............................. 141

1. Rod Block Function...................... 142

2. RCIC System............................. 144

i

3. High Water Level Trip................... 146

(} 4. Turbine Bypass.......................... 147 C. Consideration of Systems Interactions in Reactor Water i l Level Instrumentation....................... 151 D. Classification of the Shoreham Standby Liquid Control System....................... 160 E. Alleged Inconsistencies in FSAR Table 3.2-1................................. 162 F. Level of Detail Presented in FSAR Table 3.2.1-1.......................... 170 VIII. Conclusion...................................... 174 O 1 O

I. INTRODUCTION {} This testimony is filed in response to testimony filed by Suffolk County and the shoreham Opponents Coalition on their Contention 7B, dealing with classification of structures, sys-tems and components and the possibilities of adverse systems interactions. It also responds to SOC Contention 19(b) dealing with systems classification. And it addresses the areas of interest expressed by the Board at the close of hearings on May 7, 1982. - The essence of Suffolk County's and SOC's contentions may be summarized as follows: 7- 1. that the designers and builders of Shoreham failed V) to use an adequate methodology for classifying systems;

2. that the designers and builders of Shoreham have failed to comply with Regulatory Guide 1.26, Revision 3 and portions of Regulatory Guide 1.29, Revision 3, for classifying systems;

3. that the designers and builders of Shoreham failed to apply appropriate quality assurance to non-safety related systems because they do not use the classification of "impor-tant to safety";

4. that the designers and builders of Shoreham failed

(]) to use an adequate methodology for identifying systems interac-tions; and i l

5. that, because of inadequate classification, the de-

{} signers and builders of Shoreham failed to comply with GDC'1, 2, 3, 4, 10, 13, 21, 22, 23, 24, 29, 35 and 37.* None of these allegations is accurate. As this testi-many shows:

1. The designers and builders of Shoreham did use an adequate methodology for classifying systems. The classifica-tion scheme used at Shoreham is consistent with well-established industry standards and with that used at other licensed BWRs.

2. As part of the classification methodology for Shoreham, Regulatory Guide 1.26, Revision 1, and Regulatory Guide 1.29, Revision 1, were complied with in the classifica-tion of systems. There are no significant differences between revision 1 and revision 3 of these regulatory guides.

3. It is not necessary to give independent significance to the term "important to safety" in classifying structures, systems and components. At Shoreham, non-safety related systems have been designed and constructed with quality standards com-mensurate with the systems' importance to plant safety and re-liability. This type of classification scheme is consistent with that used at c her nuclear power plants and was approved by the ASLB in the TMI-1 restart proceeding.

({}

4. The design of Shoreham provides several levels of control and protection features. The equipment needed to

assure protection from severe accidents and to assure safe {} shutdown following severe natural events is classified safety related. Other equipment used to control transients and nor-

mal plant operations is designed to lesser but still rigorous i

l standards.

5. Systems interactions were considered during the design process for Shoreham.

6. A substantial number of studies have been conducted that consider specific types'of systems interactions. These studies confirm that significant adverse systems interactions are not likely to occur at Shoreham.

7. A detailed Shoreham-specific probabilistic risk as-sessment has been conducted which further demonstrates the unlikelihood of significant adverse systems interactions at

! Shoreham.

8. Shoreham structures, systems and components are properly classified and therefore Shoreham does comply with GDC 1, 2, 3, 4, 10, 13, 21, 22, 23, 24, 29, 35 and 37.

i II. GENERAL ELECTRIC AND STONE & WEBSTER DESIGN AND OPERATING EXPERIENCE The following two sections consider the General Electric and Stone & Webster design experience and process, as well as () operating reactor experience, as they relate to sytems classi-fication and systems interaction.

A. General Electric {} Concern for appropriate classification of structures, systems and components and for potential systems interactions is not new to the General Electric Company's process for de-signing Boiling Water Reactors. Since the inception of GE's nuclear energy activities some three decades ago, concern for appropriate systems classification and for foreseeing and mini-mizing potential adverse interaction effects of all kinds has been a central aspect of GE reactor design. This concern mani-fests itself both in GE's organizational structura and disci-pline for reactor design and in GE's use of reactor operating experience in design improvement. The results, discussed in more detail below, are twofold: first, an. organization struc-tured to encourage communication among and peer review by all concerned parties so as to stimulate understanding of the rela-tionships among reactor systems; and second, the capability to assimilate information gleaned from nearly four hundred reactor years of BWR operating experience. Structures, systems and components are appropriately classifiedand[possibilitiesofadversesystemsinteractionare minimized in Bofling Water Reactors through the very nature of the GE engineering organization and its long history of design () and operating experience. These are discussed in the following paragraphs.

9_

1. Disciplined Design The design of structures, systems and components is con-

trolled within the various GE design organizations to assure that safe and reliable performance of products and services are supplied. The design control processes are documented in prac-tices and procedures which establish the responsibilities and interfaces of each organizational unit, inside and outside of GE, that has an assigned design responsibility. The practices and procedures include measures to assure that design require-

ments are defined and design activities are carried out in a planned, controlled and orderly manner.

The philosophy of design discipline also extends to in-() structions and procedures, document control, purchasing, mate-rial control, process control, and inspection. The more sig-i i nificant activities related to appropriate classification of structures, systems and components and to anticipation and prevention of adverse systems interactions are discussed below. Licensing input from NRC regulations is provided by the GE licensing organization which assigns a plant dedicated licens-ing engineer who has access to all levels of the GE organiza-tion. The independent quality assurance organization within 3E, as well as its customers, audit the GE engineering organi-(]) zation to ensure that the design process and interface proce-dures are followed.

2. Design Interface Control To provide assurance of structure, system, and component interface compatibility, thus minimizing opportunity for ad-verse interaction, design documents are distributed for infor-mation, review and coordination by affected design organiza-tions. The responsible engineer is-required to have his design documents reviewed for interface compatibility by individuals having responsibility for the interfacing documentation in order to assure that there is no conflict in the design objec-tives and that the product resulting from the interfacing designs will function as planned. Design documents are also furnished to the plant owner and/or his agents to provide for O interface compatibility review and coordination by Owner / agent design organizations such as the architect-engineer. These documents d scribe quantitatively what the interfacing organi-zation must meet to ensure functioning of the systems. By ne-cessity this includes system interactions.

3. System Design Design specifications and data sheets containing design basis and other data for a specific plant are developed by the design engineer and issued to the responsible design organiza-tions in the early months following the receipt of an order.

The design controlling documents provide the basis for detailed ({} systems, structure, and component design and classification and I typically include the system and structure design

specifications, piping and instrumentation diagrams, process . diagrams, functional control diagrams, and instrument engineer-ing diagrams. The design specifications, data sheets, and design con-trolling documents incorporate the design and safety require-ments for each plant. This designs process includes the proper classification of structures, systems, and components. These designs are subject to independent design verification within engineering. The various en'gineering organizations within GE are responsible for the design and design control activities for the GE BWR. Engineering personnel are authorized to define and prepare performance parameters and to document the design of systems and equipment. They obtain necessary internal engi-neering interface consultation and services as required. They provide final design approval in accordance with documented engiteering practices and procedures. Responsibility for interrace control with the plant owner is assigned to the re-sponsible project or program manager through whose office all engineering interface documents pass. This communication is further enhanced by the matrix technical management approach in which lead systems engineers are assigned to each system. They ensure that all design aspects affecting the system, including O other 1= terr ce . ere co= 1aerea-

In addition to the design specifications, data sheets (} and design control documents, engineering organizations issue general standard design specifications which establish standard requirements for designing components which satisfy the system and structures requirements. These standard design specifica-tions identify applicable codes, standards, regulations, and other requirements to be utilized to assure compliance with safety criteria, quality levels, and other specific require-ments which have been imposed to obtain acceptable quality, safety, and reliability.

4. Design Verification Design verification is a process of independent peer review of designs against design requirements to confirm that the designer's methods and conclusions, including classifica-tions, are consistent with requirements,'and that the resulting design is adequate for its intended purpose. All BWR product designs and each application thereof are verified. Design ver-ification is performed and documented by persons other than those responsible for the design using the method specified by the design organization. The methods of design verification include one or more of the following: design review, qualifi-cation testing, alternate or simplified calculations, and

() checking. When qualification testing is used as the sole means of verifying design adequacy, a prototype or initial production g unit is tested under the most adverse operational conditions expected to be experienced by the product.

4 5. Team Design Review A team design review is a broad, formal, independent evaluation of designs by persons other than those directly re-sponsible and accountable for the design. These activities are ongoing reviews of designs, selected by engineering management, to evaluate the adequacy of product designs including concepts, the design process, methods, analytical models, criteria, sys-tems, materials, applications, or development programs. When appropriate, design reviews ~are used to verify that product designs meet functional, contractual, safety, regulatory, in-dustry codes and standards, and corporate requirements. The team's technical competence will fall into three broad cate-O gories: (1) broad design experience on similar products, (2) specialized technical expertise such as in heat transfer, materials, structural analysis, etc., and (3) a functional expertise such as manufacturing engineering, product service legal, etc. This is another process in which systems classifi-j cation and potential systems interaction issues are frequently

      )otlighted oftentimes because of the broad perspective and experience of the reviewers.

6. Design Change Control Following issuance of engineering documents, a disci-O plined design change procedure io implemented with controls d

\   commensurate with those applied to the original design.

Measures for documenting and rectifying errors and deficiencies

in the design and for determining and implementing corrective actions are prescribed. The end result, after all changes have been incorporated, is reflected in accurate drawings, specifi-cations and other documents which fully describe the final design for equipment supplied. The control procedure requires that every change must be documented, design verified, approved by the responsible engineer, and reviewed by the appropriate interfacing organizations. The responsible engineer is charged with the responsibility for defining all other design documents affected by the change, and for resolving and coordinating changes with other engineers whose documents are affected. Engineering Change Notices (ECN's) are identified, O issued, and controlled in accordance with documented proce-dures. Copies of ECN's are distributed to design interface personnel, including the responsible engineer who approved the change, to the project manager, and to engineering, manufac-turing, purchasing, and QA personnel in other organizations, as appropriate. Design changes affecting interface conditions between Owner-supplied and GE-supplied equipment are identified and reviewed by the project manager with the Owner or his designated reprasentative. Such documented changes are trans-mitted by the project manager to the Owner or his designated representative such as AE for implementation of design and (]) field changes, as appropriate, in his interfacing scope of sup- ) ply. Similarly, design changes initiated by the AE are

s . w g transmitted to the GE project to assure proper consideration ;;

                                                                                    ~
                                                                            ~

fs and interface control. < - (_) Complex design changes affecting multiple' design groups are first approved by engineering and then ' are rev'iewed and approved by a standing Change Control Board. This board' con- '- sists of senior personnel with the technical perspective'to assure that interfaces are properly addressed. - - The emphasis on interfaces is intentional, and formai steps are taken to ensure all inter-system and inter-dis-ciplinary effects are accounted for. This function is further - emphasized by the Lead Systems Engineer, whose job is.to ensure;

that his specific system properly interfaces with all others-O and with the plant as a whole. Although not conducted mathema-

tically, extensive systems interaction assessments.are.made in this process by virtue of knowledge and experience of the - s , engineers involved. ,

7. GE Design and Operating Experience When General Electric decided to enter the field of com-mercial nuclear power in the mid-fifties, the Company alreadyl had extensive experience as a contractor to the Atomic Energy Commission. GE had been running the vast Hanford operations ,

since 1946. This included eight reactors, fuel. fabrication {) plants, reprocessing plants and waste disposal facilities. GE also was one of the major contractors for the Aircraft Nuclear i Propulsion program, which included the operation of reactor f s y-,- e- w - - + - m m a a T'

t j experiments at the National Reactor Testing Station in Idaho. The Krio11s Atomic Power Laboratory in Schenectady has been and

O continues to be one of two major laboratories developing reac-tors for the U.S. Navy. At the experimental station in West Milton, New York, General Electric pioneered the first reactor containment vessel. The importance of independent reviews to assure safety

has been recognized from the outset in GE's commercial nuclear business. GE organized a Technological Hazards Council in 1956. The Council included senior technical experts from

throughout the Company. Its function was to review the safety of GE's commercial reactor designs, and to advise management on O the overall risks involved in participating in the commercial nuclear power business. The Council defined such parameters as I

the probability of an unconstrained fission product release and i the maximum release which needed to be considered for risk ap-praisals of BWR plants. Their estimates compare well with the more comprehensive results of recent GE Probabilistic Assessment reports. Many of GE's early safety studies are reflected in the fundamental design principles of the BWR, and illustrate the fact that safety has always played an integral role. For ex-i s l ample, characteristics of General Electric BWRs to this day are the small size of GE fuel bundles, relatively low power density and the large number of control rod drives, all of which impose f>

economic penalties but enhance the inherent safety of the BWR. Since GE's earliest plant, Dresden 1, forced-circulation BWRs O have been equipped with emergency cooling systems which have gone well beyond regulatory requirements. GE research, development and testing programs have covered all features of the BWR. In component test facilities GE personnel have tested full scale components such as control rod drives, safety relief valves, j et pumps, flow control valves, steam separators and~ dryers even though some of the components are purchased from other vendors. In addition, full scale main steam line isolation valves were tested under acci-dent conditions at a utility power plant. In the heat transfer and fluid flow facilities many boiling heat transfer tests have been conducted and are continuing on full scale fuel bundles for both steady state and transient conditions. For the emer-gency core cooling systems GE pioneered not only the design, but also the test programs, involving full scale fuel bundles in which spray and flooding heat transfer were studied as well as full scale spray distribution both in air and in steam. A scaled mockup of the BWR which includes the core, vessel, jet pumps, recirculation loops and ECCS systems was also used to validate the ECCS analytical models and also to study ECCS sys-tems interaction effects. The BWR fluid flow facility is a full scale segment of the BWR used to study the effects of fluid flow on the reactor internals, i.e., the system effects on the hardware inside the pressure vessel. ( I

The pressure suppression system including safety _ relief valve discharge has also been thoroughly tested, including full O scale segment tests of Mark I,. Mark II and Mark III type of containments in which LOCA design loads were determined for use by the Architect Engineers and General Electric. GE licensees overseas have, over the last ten years, independently conducted tests of their own in all the areas discussed above. The resulting data base has been available to GE and is used in GE's design work as appropriate. For ex-ample, the basic SRV quencher design, some of the containment ' loads, as well as some heat transfer and fluid flow data have come from overseas. Because of this type of independent, thorough and inten-Con'tenmnt sive testing, the BWR pressure suppression system cc binet :n-

is the most thoroughly designed and fully tested reactor system in the world today. GE's commitment to testing continues.

This same commitment to safety is engendered within the engi-neering organization at all levels and ensures.that the BWR is a safe and reliable product. In 1968, at the time the Shoreham Nuclear Power Station design work was started, General Electric had designed and placed into operation seven nuclear power plants. Also at that time, General Electric was designing 23 additional nuclear power plants, domestic and foreign. All of those 30 plants representing almost 16,000 megawatts of electric generating

capacity, have since been placed into successful, safe service.

Today, worldwide, 41 BWR reactors of GE design, representing -

       . almost 26,000 megawatts of electric generating capacity, are in operation, and 30 additional units are in the design and con-struction phase.

All of this design and operating experience has been brought to bear on Shoreham. Engineering design is an itera-tive process in which experience feedback is a key ingredient. There are thirteen domestic BWRs of the same design as Shoreham that are now operating and five overseas. This is the 4th gen-eration of BWR and the experience gained from these as well as from the previous designs (i.e., BWR 1, 2 and 3), has provided O confidence that undetected adverse systems interactions of safety significance are not likely to exist at Shoreham. Such adverse interactions, if they exist, are most likely to have been discovered through the practical experience of plant oper-ation over nearly 400 reactor years of BWR experience at 41 l , operating BWRs worldwide and indeed some have surfaced over the years.

8. Conclusions l The foregoing describes the processes by which struc-l tures, systems and components are classified and systems inter-l

(} actions are addressed within the General Electric Company. The elements of adverse systems interactions pertaining to safety have received consideration and have been largely

                                       !     precluded in the design of nuclear power plants including Shoreham. The effectiveness of the systematic design and re-view process employed has been continuously validated by nearly 400 reactor years of experience, failure modes and effects ana-lyses, and PRA's on plants similar to Shoreham. These results confirm that the design process has included proper systems classification and considered systems interactions and there-fore has produced a design that is unlikely to entail to poten-tial adverse system interactions.

B. STONE & WEBSTER , This testimony addresses the Stone & Webster Engineering Corporation (SWEC) organization, procedures and experience that enable it to classify systems, structures and components cor-rectly for a nuclear power project with respect to safety, and to anticipate and avoid through appropriate plant design sys-tems interactions that could interfere with the safe operations of the plant. To ensure proper consideration of safety classification and system interactions, the SWEC project team is structured to (1) communicate effectively all of the technical information being developed for the project, (2) be highly organized to use and implement the flow of technical information, (3) be pro-(J

~%

vided with feedback of information from other plants which sheds light on classification and systems interaction, and l

finally (4) have written procedures which are audited to ensure that the above three elem-nce t.re utilized in a systematic man-ner. SWEC utilizes an interactive matrix organization for its departments, divisions and projects to: (1) assure that the experience gained from past projects is effectively transferred to the next project, and (2) provide the multiple disciplines required for complex nuclear projects. The project is formed by assigning team leaders and support personnel from the per-manently organized departments and divisions for the time period required by the project. It is significant that key members of the Shoreham project team have remained in place O over relatively long periods of time, thus providing continuity of technical concepts and information. The project maintains communication with, and receives feedback from, all levels of the SWEC organization. To achieve l l this a management sponsor reporting to the office of the chair-man and president, and an engineering department sponsor re-porting to the director of engineering, are assigned to the project. The Engineering Department is divided into fourteen technical divisions. Each project lead engineer is responsible to his Division Chief as well as the Project Engineer for the {) technical correctness of his work, including considerations of safety classification and system interactions. The Engineering \ _ .- . _ . _ _ _ _ _ _ _. ___. __

Depar ment and its divisions provide the engineers on the project with administrative and technical procedures and guide-lines that cover virtually every phase of project organization and operation. These include procedures which distribute tech-nical information and control the preparation and checking of drawings, specifications, logic diagrams, wiring diagrams and all other products of engineering design. Specific guidelines are issued on design methods, regulatory guide positions and many other licensing issues. This systematic approach to the design process provides great assurance that safety classifica-tion and consideration of system interactions are correctly accomplished. The Engineering Assurance Division performs fre-quent, periodic audits to ensure that all calculations, speci-fications, drawings and other products of the design process conform to the procedures discussed above. The Licensing Division provides regulatory information and feedback from other dockets to the corporate departments and nuclear projects, including Shoreham. It assigns licensing engineers to these projects to assure correct interpretation and application of the regulations, as well as to assist in preparation of licensing documents. The Engineering Assurance Division is responsib.le for providing feedback to the Shoreham project concerning identification and resolution of technical problems emanating from nuclear projects all over the world. Similar information is also received from and sent to the NSSS t

vendor. In this way additional information on systems interactions are promptly taken into consideration. To assist the engineers on the project in carrying out their responsibilities, including those in safety classifica-tion and system interaction, a large group of division techni-cal specialists, technical consulting engineers and specialty organizations are utilized. These technical areas include: heat e2 change, steam plant heat balance, systems standards, shielding, nuclear safety, vibration analysis, piping, valves, pumps, HVAC, radwaste chemistry and processing, water treatment and waste disposal, meteorology, demographics, terrestrial biology, seismology, geology, and others. The organization discussed above has evolved over the last 20 years to meet the needs of the nuclear power industry. The basic concepts have been utilized to design nuclear power plants of all types, starting with Shippingport and Yankee Rowe and continuing to the most recent ongoing projects such as River Bend and Millstone 3. SWEC has participated in the design of PWR, BWR, heavy water-pressure tube, HTGR and LMFBR i nuclear power plants. Nuclear power stations designed by SWEC prior to, or in parallel with, Shoreham are well known for their reliability and safety record. These include Connecticut Yankee, which has generated more power than any other single [ nuclear unit in the world, Maine Yankee, which has had the longest full power run (392 days) of any unit in the world, and I l l 1

the Fitzpatrick power station. This excellent record of safety and reliability reinforces the fact that the SWEC methods of organization, communication, procedures and information feed-back do result in correct safety classification and considera-tion of systems interactions. A number of inherent organizational capabilities con-tribute to.SWEC's ability to anticipate and account for poten-tial systems interactions in the design process. These in-clude: (1) the organization of the corporation and the engi-neering department, (2) the use of written company standards and procedures, (3) the implementation of highly refined com-munication techniques, (4) the provisions for information feed-O back from plants being designed, constructed, licensed and op-erating around the world, (5) participation on virtually all applicable codes, standards and technical committees, (6) intensive, periodic technical reviews at all levels of the com-pany, (7) audits by engineering assurance to assure all company procedures and standards are being followed, and (8) checks by the Licensing Division to assure all regulatory requirements are being met. All of these provide a high level of confidence that systems interactions are properly considered. Obviously, completed nuclear plant designs could not have extensively used the relatively new and developing techni-(]) ques such as those mentioned by the SC/ SOC witnesses. It should be recognized that the techniques discussed by those r

witnesses are limited checks of the adequacy of design and system interactions, but are not a substitute for the experi-ence, organization, quality assurance and engineering process used to assess system interactions. The high level of relia-bility and safety on nuclear power plants designed by SWEC give confidence regarding the adequacy of SWEC design techniques and procedures in considering system interactions. The safety classification of systems, structures and components begins at the earliest stage of the design process. At that time the checks suggested by the SC/ SOC witnesses could not be readily made for lack of detailed technical information. But this early safety classification enables the designers to consider system interactions from the beginning of design until its completion. Of course, input technical information from the NSSS vendor is also part of this process. Composite draw-ings of the containment and other safety related systems, with safety related systems, structures, and components identified, materially aid in evaluation of system interactions long before construction of the plant is started. Significant quality requirements also apply to nonsafety related systems, structures and components. All of these sys-tems, structures and components are designed, fabricated and {} constructed to high industrial quality standards as well as applicable codes and requirements. The same highly qualified personnel in the SWEC Quality Assurance Department who work on \

s safety related portions of the plant also provided many of the inspection and assurance services for all other parts of the plant. Many of the nonsafety related parts of the plant are important for reliable generation of power and these items are designed and built to very high quality standards. There is therefore a high probability that these items would be operable in many instances following an accident or plant transient. For example, SWEC-designed fossil power plants have survived strong earthquakes and were able to continue operation during and after the seismic event. l I (:) l l l l t

III. METHODOLOGY FOR CLASSIFICATION OF SYSTEMS AT SHOREHAM O The methodology,used for classification of systems, structures and components at Shoreham involved the application of regulations, regulatory guides, industry standards, design basis evaluations, and design and operating experience. Each of these is discussed below. These elements were applied in the systematic, controlled design process described in Section II above. A. Use of Design Basis Analysis The nuclear power industry has been guided by a philoso-() phy known as " defense-in-depth" since its beginning. This phi-losophy applies as both an overall guide and a specific guide. In nuclear power plant design, defense-in-depth is applied through three levels of safety. These can be stated as: (1) provide a well engineered plant that operates reliably, (2) provide protection against operational transients due to equip-ment failures or malfunctions, and (3) notwithstanding the pro-tection provided by levels 1 and 2, provide multiple back-ups such that no undue risk is presented to the public as a result of postulated, unlikely accidents. The application of this third level of safety is what is referred to as the design basis approach. Unlikely accidents are postulated and used to determine those features of the

plant that will be necessary to provide mitigation. The (~' baseline NRC regulation defining necessary mitigation is 10 CFR d Part 100, which establishes the off-site accident dose guide-lines which must be met on a site-specific basis. The features of the plant that are designed to mitigate design basis acci-dents, the third level of safety, are called " safety related." Other features of the plant are "non-safety related." This is how nuclear plant design, including that of the BWR, has evol-ved. The approach has been used by both industry and the NRC. This approach was applied to Shoreham in its initial design prior to receiving its Construction Permit. The safety requirements defined in 10 CFR 100 Appendix A O serve to define more clearly the safety related features of a nuclear plant, including Shoreham. Safety related systems, structures and components are those necessary to assure: (1) integrity of the reactor coolany pressure boundary, (2) capa-bility to achieve and maintain safe shutdown of the reactor, and (3) capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guidelines in 10 CFR 100. In assessing design basis accidents, the safety func-tions of the systems, structures and components described above () are considered. These safety functions include: (1) emergency reactivity control, (2) emergency core and containment heat removal, and (3) containment isolation, integrity, and cleanup.

The systems, structures and components relied upon to provide these functions, and those necessary to support them, are the safety related set. One way in which the multiple backups of the defense-in-depth philosophy are provided is the single fai-lure criterion of 10 CFR SO Appendix A. To ensure that the plant could withstand the required applications of single fai-lure, the systems are designed with single failure in mind. Much of the process of designing and classifying systems had been completed for numerous light water reactor plants prior to Shoreham. It was this experience which led to the compiled guidance contained in such documents as ANS-22 (now issued as ANSI /ANS-52.1), Regulatory Guide 1.26, and Regulatory O Guide 1.29, which were applied to Shoreham for specific classi-fication notation. Other aspects of defense-in-depth are applied in the design basis approach. The Loss of Coolant Accident (LOCA) which is used to determine site suitability in accordance with 10 CFR 100, exemplifies this fact. The emergency core cooling systems (ECCS) are designed to satisfy the requirements of 10 CFR S 50.46, which limits fuel damage, and thus radioactive releases from the reactor core, in the event of a LOCA. Notwithstanding this ECCS design, the LOCA effects are assessed assuming the much larger radioactive source term as specified /} in 10 CFR 100. These source terms could only result from an accident beyond that prevented by the ECCS design. This

results in additional conservatisms in the design of the containment system and its supporting systems. The design basis accident analyses demonstrate that mitigation, as required by 10 CFR 100, is provided by the safety-related structures systems and components at Shoreham. These analyses of design basis accidents include the single failure as defined in 10 CFR 50 Appendix A and applied by the General Design Criteria. In conclusion, the design basis accident analysis per-formed for Shoreham only relies upon safety related systems and i components for acceptable accident mitigation. In this manner the proper system classification was confirmed for Shoreham. l B. American Nuclear Society Standard 22 and Nuclear Safety Operational Analyses ANS-22, " Nuclear Safety Criteria For the Design of i Stationary Boiling Water Reactor Plants," establishes a disci-plined and systematic method for defining nuclear safety re-quirements for a BWR. It sets out functional safety require-ments for design, is responsive to NRC regulatory requirements and industry technical requirements, and provides a uniform basis for design safety requirements to be reflected in licens-ing documents. ANS-22 was used in establishing the classifica-(} tion of structures, systems and components for Shoreham. The equipment classification table in the Shoreham ESAR (Table 3.2.1-1) was structured to provide a description of these

classifications with content and format similar to the presentation in ANS-22.

   .         ANS-22 evolved from a similar ANS standard developed for PWRs. In 1965, the AEC approached industry with the request that industry develop criteria to implement the AEC draft General Design Criteria soon to be issued for the first time.

There were a number of objectives. Key among these was obtain-ing industry-wide agreement and providing a method for uniform judgment of protection afforded public safety. An important guide for development of the Standard was to consider the com-plexities and interactions in relating various systems and com-ponents to one another. In 1970, the American Nuclear Society (3 began to develop the standard for BWRs. The PWR standard formed the basic structure for the BWR standard (ANS-22), but the BWR standard was expanded to be not limited to pressure t'ebuning saatain ng components as was the PWR standard. By 1972, the draft standard ANS-22 was available and in use. ANS-22 has

 ,   since been approved by the American National Standards Institute, and is identified as ANSI /ANS-52.1. As stated in each ANSI standard, "an American National Standard implies a consensus of those substantially concerned with its scope and content."   The NRC (formerly AEC), GE , and SWEC have been

(} active in the development of this BWR standard from the begin-ning. A more detailed history of this standard's history is provided in Attachment 2. ,

During the period of the development of these standards, GE performed a comprehensive examination of the safety aspects of a BWR. This effort, called the Nuclear Safety Operational Analysis (NSOA), was a systematic method that identified the sequences of events that must be considered for a BWR and those systems and components that must operate. By application of the NSOA and of previous design and operating experience, the ANS-22 Subcommittee was able to identify and classify systems that are safety related. To fully identify and establish the requirements, re-strictions, and limitations that must be observed during plant operation, plant systems and components must be related to the O' needs for their actions in satisfying the nuclear safety opera-tional criteria. These relationships are displayed in a series of block diagrams that were developed for this generic NSOA as shown in Appendix 7A of the Shoreham FSAR. Attachment 3 shows an example of such a diagram. The operating states in which each event is applicable are first indicated. Then, for each event, a block diagram is prese..ted showing the conditions and systems required to achieve each essential safety action. The block diagrams show only those systems necessary to provide the safety actions such that the nuclear safety operational criteria are satisfied, [} The total plant capability to provide a safety action is gener-ally not shown, only the minimum capability-essential to C

satisfying the operational criteria. Thus, the diagrams depict all essential protection sequences for each event. Once all of these protection sequences are identified in block diagram form, system requirements are derived by considering all events in which the particular system is employed. The analysis con-siders the following conceptual aspects:

1. The BWR operating state.

2. Types of operations or events that are possible within the operating state.

3. Relationships of certain safety actions to the unac-ceptable results and to specific types of operations and events.

4. Relationships of certain systems to safety actions and to specific types of operations and events.

5. Supporting or auxiliary systems essential to the operation of the front-line safety systems.

6. Functional redundancy (the single-failure criterion applied at the safety action level).

1 Each block in the sequence diagrams represents a finding of essentiality for the safety action, system, or limit under con-sideration. Essentiality in this context means that the safety action, system, or limit is essential to satisfying the nuclear safety operational criteria. Essentiality is determined through an analysis in which the safety action, system, or limit being considered is completely disregarded in the ana- {) lyses of the applicable operations or events. If the nuclear safety operational criteria are satisfied without the safety action, system, or limit, then the safety action, system, or i i \ .

limit is not essential, and no operational nuclear safety requirement would be indicated. When disregarding a safety action, system, or limit results in violating one or more nuclear safety operational criteria, the safety action, system, or limit is considered essential, and the resulting operational nuclear safety requirements can be related to specific criteria and unacceptable results. Thus, the system or component would be classified safety related. Stated differently, a system or component needed to fulfill a safety action would be classified safety-related. With the information presented in protection sequence block diagrams, auxiliary diagrams, and commonality of auxil-O iary diagrams, it is possible to determine the functional and hardware requirements for each system. This is done by consi-dering each event in which the system is employed and deriving a limiting set of operational requirements. This limiting set of operational requirements establishes the lowest acceptable level of performance for a system or component, or the minimum number of components or portions of a system that must be oper-able in order that plant operation may continue. The comprehensive, systematic analysis used in the NSOA's formed one of the bases for ANS-22 and both, in turn, formed part of the methodology for classification of systems at [} Shoreham. \ _ . _ . . _ _ _ _ _ _ - . _ _ . . _

C. Regulatory Guides 1.26 and 1.29 The methodology and criteria specified in Regulatory

 . ~s Guide 1.26 Rev. 1 was used for determining quality group clas-sification for fluid system components. Similarly, Regulatory Guide 1.29 Rev. 1 was used for designating which structures, systems and components were required to be seismic Category 1.

As stated in FSAR Sections 3.2.1, 3.2.2, and Appendix 3B, the structures, systems and components of Shoreham were classified in accordance with Regulatory Guides 1.26 and 1.29. Although the design of Shoreham had commenced before these reg-ulatory guides were available, they reflected a consideration of the same elements that went into the classification of sys-O tems at Shoreham. Thus, in large measure, when the guides were issued, Shoreham was in compliance with them. Efforts were then made to fully conform the Shoreham classification scheme with the regulatory guides. As noted in FSAR Appendix 3B, Shoreham meets Revision 1 of Regulatory Guide 1.26 and 1.29. Shoreham's compliance was reviewed by the NRC Staff and is documented in the Shoreham Safety Evaluation Report, NUREG-0420, in sections 3.2.1 and 3.2.2. In SOC Contention 19(b), it is alleged that Shoreham does not comply with Regulatory Guide 1.26, Revision 3 and with Regulatory Guide 1.29, Revision 3 with respect to control room [} habitability and radioactive waste systems. As a result of these alleged failures, SOC argues that Shoreham does not \

comply with 10 CFR 5 50.55a, 10 CFR Part 50, App. A, GDC 1 and r3 2, and 10 CFR Part 100, App. A.

%)

A review of the differences between Revision 1 and Revision 3 of each of these regulatory guides was conducted with the following results.1/ Regulatory Guide 1.26 Paragraph Change Significance i C.1 Footnote identification None-identical footnote is now "3" vs "2" C.1.a Footnote identification None-identical footnote is now "4" vs "3" C.1.c 4th line, added "and "And bypass" clarifies bypass" and footnote 5 boundary of Group B O portion of steam system, but the bypass valves remain excluded from Group B - no significant change Per footnote 5, the Shoreham turbine stop and bypass valves are subject to appropriate QA (see FSAR Table 3.2.1-1, pg. 13 of 24, item XXXI.5). i 13th line, added "the Clarifies the credit shut-off valve or" available for a shutoff valve in main steam or feedwater systems. No impact on Shoreham as feedwater shutoff valve and all piping / valves within are Quality Group A, exceeding Reg. guide 1/ These results were provided to the parties in response to SOC interrogatories.

position. C.1.e footnote identification None-identical footnote O is now "4" vs "3" C.2.a Footnote identification is None-identical footnote now "4" vs "3" 9th line, deleted "either" None-changed from "either/ substituting "(1)"; 10th or" criterion to "and" line, deleted "or" substi- criterion which is less tuting "and (2)" inclusive C.2.b Footnote identification is None-identical footnote now "4" vs "3" C.2.c. Footnote identification is None-identical footnote now "4" vs "3" and "6" vs ff 4 ff C.2.d General rewrite combining None-application of Reg. and modifying paragraphs Guide 1.3 meterology does C.2.d and C.2.3 from Rev. not alter Shoreham class-() 1. Change acts to (1) delete radioactive waste ifications (even for radwaste systems) management systems from the scope of the guide and (2) impose Reg. Guide 1.3 meteorology on the offsite dose calculations. Regulatory Guide 1.29 C.1.n Modified to address only None - no change with the control room and its respect to control room associated equipment habitability systems including equipment used to maintain personnel habitability and equipment environ-ment. Portions addres-sing failure of other structures, systems or components which could

    -        cause incapacitating injury to control room occupants have been moved to Paragraph C.2.

\

C.1.p General rewrite combining None - radioactive and modifying paragraphs waste systems no pg C.1.p and C.1.q from Rev. longer within scope

 \m)             1. Change acts to (1)          of Reg. guide 1.29 delete radioactive waste management systems from the scope of the Reg. Guide and (2) impose Reg. Guide 1.3 meteorology on the off-site dose calculations.

As can be seen from the above, there were no instances in which changes from revision 1 to revision 3 of the guides would dictate a change in Quality Group classification or seis-mic category. D. Applicable Regulations () The following regulations were complied with as part of the methodology for classifying structures, systems and compo-nents used at Shoreham:

a. 10 CFR Part 100 The guideline radiological dose levels provided in 10 CFR Part 100 were utilized to establish the adequacy of engin-eered safety features provided at Shoreham to mitigate the radiological effects of accidents that were postulated for site suitability considerations.

b. 10 CFR Part 100 Appendix A The criteria provided in this regulation for specifying those structures, systems and components that needed to be de-signed to remain functional under safe shutdown earthquake

\ _-

ground motion were applied in designating which structures, systems and components needed to be classified as Seismic Category I. ,

c. 10 CFR Part 50 Appendix A The General Design Criteria were utilized to develop the overall safety criteria and standards applied to the design of Shoreham. As discussed above these criteria are satisfied, in part, through the joint industry and AEC/NRC effort in estab-lishing overall safety criteria for boiling water reactors.

d. 10 CFR Part 50 Appendix B The criteria of Appendix B were utilized to develop the quality assurance programs applied to all activities affecting O the structures, systems and components of Shoreham that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. The appendix is applicable to all activities which affect the safety related functions of such structures, systems and compo-nents.

e. 10 CFR 50.55a This regulation specifies the codes and standards to be used for components of the reactor coolant pressure bound-ary, inservice inspection requirements and the plant protection

{} system. In addition, guidance is provided for applicable code editions and addenda. This regulation was utilized for deter-3 mining applicable codes and standards for the design of Shoreham. (

h E. Operating Experience r^g Shoreham is a BWR 4, a member of the fourth generation V of GE BWRs. There are 18 BWRs of earlier design (BWRs 1, 2 and

3) that are and have been safely operating for some years.

Moreover, there are 18 additional BWRs of the same design as Shoreham that are and have been operating safely for some time. In total, there are over 400 reactor years of safe operating experience at 41 GE BWRs. All of this operating experience has been brought to bear on Shoreham and serves to validate and confirm the design and classification of systems at Shoreham. Also worth noting is that Shoreham's systems classification is essentially similar to tgat of the LaSalle Nuclear Plant, one O of the most recent plants to obtain an operating license. The classification of Shoreham's structures, systems and components is comparable to that of the 18 BWR 4s now licensed and oper-ating and of all previous designs as well. O

IV. TREATMENT OF NON-SAFETY RELATED STRUCTURES, SYSTEMS AND COMPONENTS O O A. Introduction All of the plant systems, including non-safety related systems, have been examined and evaluated for their signifi-cance to total plant function. The relationship of the non-safety related systems to overall plant safety has not been ignored. Non-safety related systems typically have a very important role in reliable power generation. They receive quality assurance and technical treatment commensurate with the goal of a reliable, and therefore safe, power plant. Further, the major non-safety related systems are addressed and de-O scribed in the Final Safety Analysis Report which documents the results of these efforts. For example, Chapter 3 of the FSAR discusses major plant structure s , Chapter 7 discusses instru-mentation and controls, Chapter 8 discusses electrical power systems, Chapter 9 discusses auxiliary systems, Chapter 10 dis-cusses steam and power conversion systems, and Chapter 11 dis- , cusses radioactive waste management systems. In each case, non-safety related portions of the plant are described, along with the appropriate design bases. The information provides the considerations which were employed during the design pro- {) cess for the proper integration of the non-safety systems into the plant design and serves to support the classification table provided in FSAR, Section 3.2.

B. General Electric r' Within the General Electric scope of supply, each struc-(s/ ture, system and component is evaluated and receives quality assurance commensurate with its intended safety or power gener-Z ation reliability function. QAR I (QAR,is an internal GE QA designation that is equivalent to LILCO QA Category I) applies the requirements of 10 CFR 50 Appendix B to safety related i structures, systems and components. 'QAR II (equivalent to 1 LILCO QA Category II) applies quality assurance requirements to j non-safety related structures, systems and components based on function, complexity and importance to reliable power genera-tion. In this regard, QAR II is selectively supplemented with 1 Such e.s for survestlence, QC records or radooyap)p3 the requirements 1cf QAR !!! thrcugh 2AR XIV based on enDineer-ingjudggmentj - General Electric requires an, identical degree of engi-neering design and engineering quality assurance for all struc-tures, systems and components independent of safety classifica-tions. Therefore all design activities including analyses, documentation, review, verification, change control and records fully comply with the highest standards of design control. As described above, the quality assurance requirements for procurement or manufacture of structures, systems and com-ponents in QA Category II are specified by the design and qual-(~} ity control engineers based on an evaluation of the function, complexity and importance to reliable power generation. For

                                                                         ,.            s I

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 )                                                                                                                                s e

most structures, systems and components in QA Category I'I, the ' program addresses most of the criteria in 10 CFR 50 Appeddix B .

     )
       . to an appropriate level of detail commensurite with their im-portance to safety and reliability. General Electric regnires               '

that seller and seller's suppliers have a program in place, including:2/

Quality Assurance Program QA Manual ' Program Audits Access for Inspection and Audit Notification Points Seller's Document Submittals to Buyer Disposition of Seller Deviations Final Inspection and Check of Records Buyers QC Release Notification for Release. () Within General Electric the quality contro' program requires: s Pre-production Review - Notification Points - Product QC Checklists QA Records List. The program also requires that Quality Assurance Records be - kept including: QA Records List ' Preparation of Records Revision Control Approval of Records Binders of Records Final Radiographs. l These requirements assure a high quality of design and i manufacture for non-safety grade systems and components. O 2/ These requirements would also be applied to components manufactured by General Electric. l t

C. Stone & Webster

As stated in FSAR Table 3.2.1-1, there are two LILCO Quality Assurance categories. QA Category I applies to safety related structures, systems and components which must meet the quality assurance requirements of 10 CFR 50, Appendix B. QA l Category II applies to r.on-safety related structures, systems and components which meet the quality assurance requirements defined in specifications.

The fact that non-safety related structures, systems and components are assigned LILCO QA Category II does not mean or imply that they do not receive quality assurance treatment.

g Within the Stone & Webster Engineering Corporation scope of supply, each structure, system and component is evaluated and receives quality assurance commensurate with its intended func-tion. SWEC utilizes three quality assurance categories, with SWEC QA Category I being equivalent to LILCO QA Category I, and l

SWEC QA Category II and III together being LILCO QA Category II used in the FSAR. That is, SWEC QA Category I consists of structures, systems and components to which 10 CFR 50, Appendix B ', applies. The requirements for structures, systems and com-l penents classified as SWEC QA Category II or III are determined l on the basis of function as described below. l SWEC QA Category I consists of structures, systems and components, or portions thereof, whose failure or malfunction could cause a release of radioactivity that coula endanger l

public safety. This category also includes structures, systems

             -~                                                                      and components vital to safe shutdown, the removal of decay heat, or necessary to mitigate the consequences to the public of a postulated accident.                        Thus, it includes those structures, systems and components necessary to assure the integrity of the reactor coolant pressure boundary, the capability to achieve and maintain a safe shutdown condition, and the capability to prevent or mitigate the consequences of accidents which could result in potential off-site exposures comparable to the guide-lines of 10 CFR 100.

SWEC QA Category II applies to structures, systems and components, or portions thereof, which are essential for the reliable generation of electric power but which are not essen-tial for safe shutdown. Equipment or systems which contain radioactive material, but whose failure could not release quan-tities sufficient to endanger public safety, are included in this category. l SWEC QA Category III applies to structures, systems and components thereof, or portions thereof, which are not essen-tial for the reliable generation of electric power, and which do not contain radioactive material or whose failure or malfun-1 ction could not result in the release of radioactive material.

          /                                                                                                      Activities affecting virtually all non-safety related
                }

structures, systems and components at Shoreham, within the SWEC _ scope, are controlled by specifications. Each specification

clearly identifies the assigned QA Category (II or III) defined above. In preparing a specification, the responsible engineer selects tha appropriate QA Category based on the function in-volved. He further specifies, in detail, the quality assurance requirements for the activities controlled by the specifica-tion. Pertinent industry codes and standards, or other docu-ments are also fully identified. Prior to issuance, the speci-fication requires a number of signed approvals including, but not limited to, the Preparer, the Lead Engineer, the Project Engineer and the appropriate Specialist. Quality Assurance Departmen approval is also required for QA Category II or III specifications that require either inspection actions by the QA O Department or submittal of quality assurance documentation, such as inspection, test or examination records. The responsible engineer preparing a specification for Shoreham starts from a master specification. Master specifica-tions are prepared and maintained by SWEC specialists, and rep-resent SWEC's best judgment based on experience, regulation, industry standards and the function of the equipment. If a master specification does not exist, the appropriate specialist provides a similar previously used and approved specification along with any additional guidance required. The preparer uti-li=es the master specification to develop a Shoreham specific (]) specification subject to the reviews and approvals previously described. The Shoreham specification can include special h

requirements beyond those of the master specification, such as those which may be imposed by project procedures or by LILCo. The entire process, including project and master speci-fications, is controlled by the SWEC Specification Preparation Manual, Engineering Assurance Procedures and supplementary Shoreham project procedures to ensure each specification is complete and correct. Quality is further assured by the use of approved bidders lists. In this way, only vendors approved by the SWEC Purchasing Department, Procurement Quality Assurance Division and appropriate specialists, are used. A vendor must demonstrate a history of proven performance in order to be in-cluded on this listing. Vendors, once placed on the approved O list, continue to be periodically surveyed to ensure continued acceptability. SWEC requires that QA Category II and III systems, structures and components be designed, procured, constructed and tested in accordance with applicable codes and standards and good design and construction practice. Although compliance with 10 CFR 50, Appendix B, is not required for QA Category II and III, the 18 criteria of Appendix B represent the principles or elements of a comprehen-sive QA program. SWEC does apply those principles to its QA {} requirements for Category II and III work commensurate with the specific activities performed.

The quality assurance principle of auditing is one ex-ample which SWEC applies to QA Category II and III activities. As part of the overall QA Program, QA Category II and III acti-vities are audited by various departments and divisions of SWEC to assure that quality requirements are in compliance with pro-cedures, sper.ifications, drawings and applicable code and stan-dards. Procurement and construction activities are audited by the Quality Assurance, Cost and Auditing Division; engineering and design activities are audited oy the Engineering Assurance Division and audits of sellers are performed by the Procurement Quality Assurance Division. These are the same SWEC organiza-tions performing the audit functions for QA Category I activi-O ties. D. LILCO l In order to insure that all plant systems and components l operate reliably, and to provide added assurance that~the non-safety related components and systems perform their intended functions, the Shoreham construction effort has in place estab-lished quality programs and requirements for construction acti-vities relcting to the fabrication and installation of these l non-safety related systems and components. l l \

In the piping area for example, the small bore piping in field designed in accordance with design specifications for seisn. : and thermal expansion criteria and the proper installa-tion of this piping is verified by non-manual construction per-sonnel. In addition, non-safety related piping systems are designed in accordance with appropriate industry codes and, again, the proper installation of this piping is verified by construction inspections. Further, welding procedures have been developed for all of the plant's piping systems and welder qualification, training and testing are required before the craftsmen are permitted to install these piping systems. All weld filler material used is QA Category I. In some cases, ( radiography is required for non-safety related welds and visual inspections are also employed using construction non-manual personnel qualified to assess the adequacy of the welds. For non-safety related equipment and components in the electrical area, construction inspections are required in many instances on significant components (electrical panels for ex-ample) in order to insure proper installation and proper func-tioning. The cable pulling criteria utilized for safety and non-safety related cable are the same. All cable, both safety and non-safety related, is bought and inspected to Category 1 requirements. Further, the termination procedures which are (]) utilized for safety related work are also utilized for the non-safety related components and the qualification requirements

for the electricians physically doing the work (primarily terminators) are the same as for the safety related work.

In the structural area, controls are also exercised on the placement of non-safety related concrete, including taking and evaluating samples of the concrete for each placement (pour). Procedures are also in place for construction inspec-tions of structural steel, insulation, and painting activities in non-safety related areas. These activities which assure the quality of the entire plant, both safety related and non-safety related, gain their impetus from LILCO's desire to assure that the plant will oper-ate reliably. This program also assures the quality of those O systems which can be utilized to minimize challenges to the plant's safety related systems. E. Response To Intervenors' Use of Denton Memorandum Citing the Harold Denton memorandum dated November 20, 1981, Intervenors contend that Shoreham's classification of structures, systems and components is faulty because it does not explicitly include a category of structures, systems and components that are important to safety but not safety related. This assertion is incorrect. While it is true that Shoreham, () and other licensed operating BWRs, do not use the term "impor-tant to safety" in the manner suggested by the Denton memoran-dum, the end result is the same and consistent with GDC-1

because, as this section of the testimony shows, non-safety related equipment at Shoreham is designed, fabricated, erected and teated to standards commensurate with the nature of the function to be performed. This point is further made in the following paragraphs while the nature of standards applied to non-safety related equipment at Shoreham is set forth in the other parts of this section of the testimony. At Shoreham, as at licensed nuclear power plants gener-ally, certain structures, systems and components are classified

   " safety related" (or its equivalent " safety-grade"). The term
   " safety related" is derived from Appendix B to 10 CFR Part 50 and Appendix A to 10 CFR Part 100. Appendix B to 10 CFR Part O   50 states that it applies to all activities affecting the
   " safety related functions" of structures, systems and compo-nents. Appendix B requires all license applicants to implement comprehensive quality assurance programs to apply to structures, systems and components that prevent or mitigate the consequences of postulated accidents that could cause undue risks to the health and safety of the public.

Thus the safety related group of structures, systems and compo-nents are those required for the performance of specific criti-cal safety functions such as design basis accident prevention and mitigation and safe shutdown. These safety functions'are (]) more specifically set forth in Appendix A to 10 CFR Part 100 which defines the Safe Shutdown Earthquake and requires that

f

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    /                                                                       ,
 !                                                                          1 1

j these safety related structures, systems and components be designed and built to survive this event and ensure the follow-(]

     \_/
                                                                           )

ing functions: l l l I (1) the integrity of the reactor I coolant pressure boundary; 1 l (2) the capability to shut down the l reactor and to maintain it in a safe shut- j l down condition, (3) the capability to prevent or mit-igate the consequences of accidents which I could result in potential off-site expo-sures comparable to the guideline expo-sures of this part. 10 CFR Part 100, Appendix A. Thus, at Shoreham and other nuclear power plants, the ()

/'\

safety related class of structures, systems and components are those required to perform the specific critical safety func-tions listed above. And only this safety related set of struc-tures, systems and components must be designed, fabricated, erected and inspected in accordance with the quality assurance 1 i requirements of Appendix B.3/ This has been done at Shoreham. l As a result, functions of Appendix A 'o c Part 100 for design basis accident prevention and mitigation and protection of the public are capable of being performed by structures, systems and components which have been subjected to the stringent t j , quality assurance requirements of Appendix B. \ t

     '% )

3/ Appendix B does not use the term "important to safety," but rather explicitly applies only to " safety related" func-tions.

I t At Shoreham, as elsewhere, the remaining structures, systems and components, i.e., those not required to be f-)x t u

            " safety-related," are, as other portions of this testimony show, accorded quality standards and assurance treatment com-mensurate with their function. No explicit separate category of structures, systems and components that are not
            " safety-related," but "important to safety" is used at Shoreham. But the same result is still achieved by applying to those structures, systems and components involved in normal operations and transient control, requirements which, although perhaps less stringent than Appendix B, are still quite rigorous. The absence of a distinct, separate category of O    structures, systems and components that are "important to safety," but not safety related, leads to essentially the same result as that suggested in the Denton memorandum. The ration-ale for accepting less stringent standards for the non-safety related set of structures, systems and components is the re-duced consequences which result from failure of these systems.

In the event these non-safety related structures, systems and l components fail, they are backed up by fully safety grade l structures, systems and components capable of mitigating or preventing the resulting event. Only safety related equipment is relied upon for design basis accident analysis as set forth {} in Chapter 15 of the FSAR.

All this does not mean, however, that non-safety related equipment may never be used, in addition to safety related equipment, to deal with or mitigate accidents or transients. On the contrary, such equipment may be, and should be, used.4/ Operators should use the plant's full capabilities to deal with problems because the normal systems are designed and built to be reliable and can be effective in mitigating events, thus reducing the need to call upon the safety related systems. If these normal systems fail the safety related systems are there to handle the situation. The point is, this normal, non-safety related equipment, while not subjected to the full requirements of Appendix B, is nonetheless designed and constructed to a O range of high standards commensurate with its function. Much of this non-safety related equipment plays an important role in the reliable operation of the plant. Reliability is essential to the economic generation of electricity and this ie an impor-tant reason to encure the quality of these non-safety related structures, systems and components. Reliability is also linked to safety. A reliable plant is a safe one because operational problems are less likely to occur and if they do, the reliabil-ity of the normal, non-safety related structures, systems and 4/ For a further discussion of the rationale for using O' non-safety related systems in accident and transient mitigation, see the testimony of Paul McGuire, Parc VII A below. I 6

components may serve to eliminate the need to call on the safety systems or reduce the load on these systems. The clas-sification and treatment of non-safety related equipment at Shoreham has been done with this in mind,when vieaed in thi: 14ght, but this fact hes net been ignored at Shorehea. In summary, the Denton memorandum uses the term "impor-tant to safety" in a manner inconsistent with historic industry and NRC practice.5/ Shoreham and other currently licensed nuclear power plants have not explicitly used a category of equipment that is "important to safety" but not safety related. This difference in terminology leads to no substantial differ-ence in result in the case of Shoreham because, at Shoreham, O structures, systems and components that are not safety related must still meet high standards of design, construction and quality control commensurate with their function. 5/ The Denton memorandum does not specify what design and quality assurance treatment should be applied to structures, systems and components that are "important to safety," but not safety related. The memorandum does not define the class or concept of "important to safety" except by the definition "any-thing that contributes in an important way to safety." This raises many questions particularly since the Denton memo Os clearly states there is no intent to develop or dictate new requirements. The fact that the Denton memorandum has raised many questions is demonstrated by recent correspondence from the Nuclear Power Committee of IEEE. See Attachment 4.

V. CONSIDERATION OF SYSTEMS INTERACTIONS AT SHOREHAM ( As indicated in Part II of this testimony, concern for potential systems interactions permeates the basic process used by GE and SWEC for design, manufacture and installation of sys-tems, structures and components at Shoreham. This section de-scribes a number of studies and programs which are either spe-cific to Shoreham or pertinent to Shoreham and which concern systems interactions. The resalts of these studies demonstrate that potential interactions were adequately considered in Shorehar.t's design. The changes that did result from these stu-dies further assure that adverse systems interactions will not () occur. A brief description of these studies follows: (a) Pipe Failure and Internal Flooding: The effects of pipe failures in both safety related and non-safety related piping were studied inside and outside of primary containment. The results of these studies are presented in FSAR Section 3.6 and Appendix 3C respectively. The dynamic and environmental effects, including flooding, which result from high energy line breaks and moderate energy cracks were determined and shutdown capability demonstrated. In the case of both dynamic and environmental effects, (]) these are studies of spatial interactions.

(b) Missiles: The potential for, and effects of, both f3 internal and external missiles were studied for V Shoreham. The objective was to demonstrate contain-ment integrity, shutdown capability and prevention of loss of coolant accidents given the dynamic effects associated with missiles. Internal and external missile sources were considered with the results being documented in FSAR Sections 3.5 and 10.2.3. These are spatial interaction studies. (c) Fire Hazard Analysis: An evaluation of the Shoreham fire protection features was conducted and submitted () on the Shoreham docket as the " Fire Hazard Analysis l Report" (LILCO letter SNRC-181, Novarro to Boyd, May 5, 1977). This study considered plant fire areas and the consequences of fires with and without active fire protection. This is a spatial interac-tion study. (d) Cable Separation: An analysis of cable separation in the primary and secondary containments was per-formed to demonstrate that a fire that disabled all

cables and raceways (and therefore associated equip-ment) in an entire designated area would not prevent

[} safe shutdown even assuming a concurrent loss of offsite power. The analysis is a spatial

                           -S8-i interaction study. The results of the analysis were submitted to the NRC (LILCO letter SNRC-S32, Novarro to Denton, February 10, 1981). The analysis will be confirmed using as-built drawings.

(e) Failure Mode and Effects Analysis (FMEA): An FMEA was performed on each balance-of-plant safety re-lated control circuit. The FMEA identifies all con-trol circuit component failures which could defeat the systems function, assures that each failure mode is detectable and assures that the single failure criterion is met. The FMEA will identify unaccepta-() ble interactions between redundant trains of a safety related system. The results of the original FMEA were submitted to the NRC in response to Regulatcry Guide 1.70, Revision 1 (LILCO letter SNRC-82, Burke to Snell, February 10, 1976). (f) Electrical Bus Failures: In response to SER Open Item 46 (and I&E Bulletin 79-27), an analysis was performed to demonstrate that following loss of any safety related (1E) or non-safety related (non-1E) electrical bus supplying power to safety or (} non-safety related instruments and control systems, safe shutdown would be attained. The results of the study are presented in the FSAR (response to Request

A 2)3.90) and show safe shutdown is attained using only 1E systems, given the loss of any single bus. This is a study of functional interactions of power supplies and equipment. (g) Control System Failures: In response to SER Open Item 47, an analysis is ongoing to demonstrate that the failure or malfunction of any power source or sensor providing power or signals to two or more control systems (non-safety related) will not result in consequences outside the bounds of Chapter.15 analyses (LILCO letter SNRC-595, McCaffrey to () Denton, July 16, 1981). This is a functional inter-action study. (h) High Energy Line Break: In response to SER Open i ! Item 48, an analysis is ongoing to supplement the I existing high energy line break studies (LILCO let-ter SNRC-577, Novarro to Denten, May 27, 1981). The original studies considered, among other things, environmental effects on safety related control sys-l l tems. This new study will consider the effects of harsh enviroments created by a break on non-safety related control systems. The purpose is to confirm ("} shutdown capability following the break by demon-strating that the consequences of the break are l l

[ l bounded by Chapter 15 analyses when the effects of resulting harsh environments on any non-safety re-O. lated control system are considered. As in the case of the original pipe break study, this is a spatial interacticn study. (i) Probabilistic Risk Assessment: PRA's have been per-formed to assess the core damage probability and risk associated with Limerick, a BWR/4 Mark II plant, and the BWR/6 standard plant with a reference Mark III containment. The positive results of these studies indicate the effectiveness of the generic ) () BWR design, including the consideration of systems interactions. i (j) Heavy Loads: In-response to SER Open Item 59, information on the handling of heavy loads at Shoreham was submitted to the NRC (LILCO letter SNRC-596, McCaffrey to Denton, July 17, 1981). The I study of heavy load handling is performed to ensure

that there is no unacceptable impact on safety re-i lated equipment if a heavy load is dropped.

Additional analyses are ongoing to study heavy load {} handling in all safety related areas of the plant in response to NUREG-0612. This is a spatial interac-tion study. ]

(k) Analysis of Industry Experience: LILCO has estab-lished an Independent Safety Engineering Group (ISEG), which, in addition to other activities, will perform in-depth reviews of plant operating charac-teristics, programs, and experiences at Shoreham and at other nuclear power plants. These will include review of Licensee Event Reports (LER) generated at Shoreham, and Significant Event Reports (SER) and Significant Operating Experience Reports (SOER) de-veloped as a result of the Institute of Nuclear Power Operations (INPO) SEE-IN Program. This-pro-gram, which was endorsed by the NRC (see Generic O Letter No. 82-04), provides a mechanism for central collection and screening of all events from both U.S. and foreign nuclear plants. ISEG review of the SER and SOER will identify any incidents involving systems interactions. These will be evaluated for applicability to Shoreham, and for assessment of whether any corrective action is appropriate for Shoreham. (1) Pre-operational and Startup Testing: The pre-operational and startup test programs will give O LILCO an opportunity to verify that adverse systems interactions will not occur. Several of the tests

are particularly useful in this regard. Just prior to fuel load an integrated electrical test is per-formed. This test verifies that Emergency Core Cooling Systems and the emergency power sources per-form their design function in various degraded modes of electrical power distribution. Throughout most of this test, power is secured to systems and con-trols not powered from the safety related busses. A LOCA signal is simulated and the ECCS and Diesel Generators are verified to perform their design functions in various electrical lineups. The startup test program also incorporateg a loss of O offsite power test which verifies that, with the loss of non-safety electrical busses and their associated systems and controls, the Reactor Protection System will prevent violation of neutron flux and thermal power limitations. The RCIC and HPCI systems are additionally verified to function under these conditions without manual assistance. Startup testing at Shoreham will cover operation from low power at rated temperature and pressure up through 100 percent power. Major testing, as de-() scribed in the FSAR Section 14.1.4.5, will be per-formed at approximately 25, 50, 75 and 100 percent power. The power ascension test program permits

verification of the behavior of the various 7~ interacting plant systems during integrated opera-kJ tion under various conditions of power and flow. Many of the tests will involve inducing transients, such as major pump trips or power level perturb-ations, to demonstrate the responses of affected systems. Details of the power ascension tests are described in FSAR Section 14.1.4.8. (m) Protection Systems: An evaluation was made of the BWR 4 reactor protection, ECCS and reactor isolation systems relative to the requirements of IEEE-279, (} including aspects of separation and system interac-tion. Selected failure modes and effects analyses were performed as part of this study. (n) Scram Reliability: The effects of both random and common cause failures on scram reliability were studied for the BWR 4 using a failure modes an'd effects analysis of all contributing components. This is a study of functional interactions among the components and systems which affect the ability to scram. f

(o) Common Mode Failures in Protection and Control Instrumentation: A study was made of the BWR 4 re-i sponse to various operational transients and acci-dents in the presence of common mode failures which result in loss of the primary protection system ini-tiation signals. This study utilized the systematic methodology of the Nuclear Safety Operational Analysis to evaluate potential systems interactions. (p) Water Level Instrumentation: A study was made of systems interactions relative to Shoreham reactor I water level measurement. This study systematically () addressed both the causes and effects of potential water level measurement errors due to heatup of the water level sensing lines. l (q) TMI-2 Implications: A study was made of BWR 4 sys-l tem performance following accidents and operational transients in light of the TMI-2 experience. This study considered numerous aspects of functional in-teractions between both safety and non-safety sys-tems. (r) Shoreham PRA: The Shoreham PRA looks at the func,-

         }      tional interaction of safety and non-safety related systems. This PRA is dealt with in Part VI below.

t h

VI. THE SHOREHAM PRA A. Testimony of Vojin Joksimovich, Ph.D.

1. Introduction This testimony, written by a member of the Peer Review Group for the Shoreham PRA, addresses the subject of " systems interactions" in the framework of the ongoing Shoreham Probabilistic Risk Assessment study regarding Contention 7B in the ASLB Hearings. As part of Contention 7B, Suffolk County and Shoreham Opponents Coalition claim that LILCO has failed to utili=e improved techniques that are now available for safety

(') classification such as PRA, Failure Modes and Effects Analysis, systems interaction analysis and dependency analyses. Furthermore, the claim is that systems interactions which may have a significant impact on plant safety have gone undetected. As this testimony shows, this claim is not well-founded. In view of the fact PRA methodology and its use as a safety assessment tool are relatively new, Sections 2-4 offer / a general PRA discussion emphasizing its effectiveness in deal-ing with ambiguously defined " systems interaction" issue. Sections 5-6 examined specifically the Preliminary Draft of the Shoreham PRA Study, again emphasizing the " system interactions" O issue. Section 7 addresses the status and some preliminary results of the Study viewed from the standpoint of a member of i

l 1 the Review Board. Section 8 provides some specific responses to the intervenors' criticisms, while Section 9 draws the con-( }) clusions.

2. General PRA Methodology and Objectives PRA quantifies the probabilities and consequences i

associated with accidents and malfunctions by applying proba-bility and statistical techniques as well as in and ex-plant consequence evaluation methods deemed acceptable by the practi-tioners. It examines all the pertinent information available and widens the historical basis by using data from actuarial

events in combination with logic models to predict the fre-

      ) quency and consequences of events which have not happened, but which could cause accidents. A compendium of acceptable methods is contained in the PRA Procedures Guide.p/     Although co-sponsored by the NRC, the guide is still in draft form and cannot be viewed as final NRC sanctioned guidelines in PRA methodology.

One of the most important features of PRA is the taxon-omy and comprehensiveness with which potential accidents are analyzed. The event / fault treep analysis tools assess the sig-nificant ways in which the plant might fail. A wide spectrum of pertinent accident possibilities and consequences are p/ NUREG/CR-2300, Volumes 1 and 2, PRA Procedures Guide, April 1982.

analyzed for potential hazard to the public. Assessment aims at being comprehensive in covering significant risk contribu-tors. The best safeguard for insuring necessary comprehensi-veness is peer review. By using the PRA approach, a safety assessor is able to put into better perspective the contribu-tors to various accident sequences and risk, and thereby iden-tify the needs for additional safety features, if any, improved equipment reliability and, where necessary, areas of research and testing. The first step in the PRA Methodology is identification of initiating events which, when combined with additional plant failures, have the potential for the release of radionuclides to the environment. Once initiating events are t' fined, logic trees are con-structed to display responses of the plant in terms of the pos-sible sequences of events that could then follow. With the use of the plant design, operating procedures and various types of information from systems analysis, an expected sequence of events is defined given an initiating event. The next step is to identify possible variations from the expected sequence i which consist of various combinations of successes and failures of the plant functions or systems designed to control, prevent {) or mitigate releases of radioactivity. The construction of event trees is the most fundamental aspect of PRA since this is where the quantitative aspect of the assessment is formulated.

This involves the use of engineering experience and judgment. () Each branch of the tree represents one possible outcome of the initiating event which is studied in terms of the likelihood (or frequency of occurrence) and consequences. , Event trees can be viewed as a modified form of the decision trees employed in the decision tree analysis.7/ The fault tree is a logic model of the various parallel and sequential combinations of faults that will result in the occurrence of the predefined undesired event. The faults can be events that are associated with component hardware failures, human errors or any other pertinent events leading to the unde-('T sired event. A fault tree thus depicts the logical inter-V relationshigp of basic events that lead to the undesired top event. It is a complex of entities known as " gates" which serve to permit or inhibit the passage of fault logic up the tree. The gates show the relationship of events needed for the occurrence of a " higher" event. The " higher" event is the out , , put of the gate, the " lower" events are the " inputs" to the gate. Thus, gates can be viewed as somewhat analogous to switches in an electrical circuit. The fault tree enables the system failure probability (or unreliability) to be derived in terms of the component failure probabilities. The fault trees O 7/ H. Raiffa, " Introductory Lectures on Making Choices Under Uncertainty," Addison-Wesley, 1968. \

are employed to assess failure probabilities of the various' ( }) reactor systems, and hence their functions, which appear as branch points in the event trees.

Another important aspect of PRA is the assessment of uncertainties associated with the probabilities and conse-quences of each sequence of events. Since a large majority of the event sequences studied have not occurred, and indeed are not likely to occur, it is often difficult to establish the probabilities and consequences with sufficient accuracy to make point estimates fully mean..agful. For this reason, many param-eters in a risk assessment are treated as statistically dis-tributed parameters, so that computations of probabilities and consequences involve the statistical combinations of distribu-tions.

3. PRA and Systems Interactions The " systems interaction or interactions" has become a bu -word meaning different things to different people and in the process confusing practically everybody.

The NRC-sponsored Sandia Studyg/ recognized correctly that " systems interaction is in a sense, a subset of system problems sometimes called common-mode or common-cause O g/ G. J. Boyd, et al., Final Report-Phase 1 Systems Interaction Methodology Applications Program, NUREG/CR-1321, SAND 80-0384, April 1980.

failures." Further, the study puts forth a definition of a () system interaction as "an event or sequence of events causing two or mdre components to fail to perform their function, in-creasing the likelihood of an undesired event." It also states that a system interaction may be characterized by the cause initiating the interaction and the connection which permits the interaction. Another NRC sponsored study states that the connotation of an adverse intersystem dependence is inherently part of the definition of systems interactions.9/ The failure of at least one of the basic safety functions is the first essential char-acteristic of an adverse systems interactions. The basic

safety functions are:

1. To maintain the primary coolant.

2. To transfer heat from the reactor to the ultimate

, heat sink.

3. To render and keep the entire core subcritical.

4. To maintain the integrity of the containment and control radioactivity releases.

Yet another NRC sponsored study offers a definition of system interaction as "a sequence of events leading to the vio-lation of at least one basic safety function as a result of two () 9/ F. Coffman, et al., "The Development of Interim Guidance on Systems Interactions, Proceedings of the International ANS/ ENS Topical Meeting on Probabilistic Risk Assessment," September, 1981, page 626.

or more component failures involving a dependent failure (common-mode or common-cause)".10f In the study proposed for Indian Point 3 the following definition is offered: " System interactions are those events that affect the safety of the plant by one system acting upon one or more other systems in a manner not intended by design with emphasis on non-safety / safety types of interactions."11/ It is obvious that there is no single definition which can be unambiguously interpreted. The majority of them are - vague; hence, it was felt appropriate to offer a more focused definition fully appropriate at the generic level and not nec-essarily meant to be Shoreham plant specific. Systems interac-l tion is a subset of dependent failures (common-mode or common-1 cause). Dependent failures are adequately defined in the PRA 1 Procedures Guide employing the classification displayed in Attachment 5. By virtue of offering this definition the sys-tems interaction assessment is encompassed within the framework of a PRA study employing the same tools such as event trees. Event trees are in fact extension studies of systems interactions. , 10/ J. J. Lim, H. P. Alesso, "On Issues Important to the Development of a Systems Interaction Evaluation Procedure," () l ' Proceedings of The International ANS/ ENS Topical Meeting on Probabilistic Risk Assessment, September 20-24, 1981, page. 635. 11/ Power Authority of The State of New York Indian Poiht 3- , Nuclear Power Plant Systems Interaction Study, December 1981. l l

1 There are several steps in a PRA study in which depen-dent failures are taken into account: (])

1. Selection of initiating events.

2. Definitions of accident sequences (event-tree con-struction).

3. Systems failure analysis (fault tree construction).

4. ' Accident sequence quantification.

The plant event trees delineate the accident sequences leading to core vulnerability (damage) or core melt. The re-sulting accident sequences define the plant systems important to public safety considerations. The fault trees are employed to assess the failure probability for each function or system () displayed as a branch point in the event trees. Hence, the event trees in principle adequately account for intersystem dependencies given representative spectrum of initiating events while dependencies on common support system in principle are adequately accounted for in the fault trees. Methods for dependent-failure analysis recommended in the Procedures Guide 12/ are state-of-art with due emphasis on shared support systems, environmental conditions and human in-teractions. This is illustrated in Attachment 5. O W See note 6 above. C- _ _ . _ _

The four types of intersystem dependencies (Types 2A, () 2B, 2C and 2D) are analyzed by means of event trees, fault trees, or a combination of the tso. A simple event tree (Attachment 5) will be employed for illustrative purposes to illustrate functional dependencies (Type 2A). Suppose that system 2 is not called upon unless system 1 fails. This would be reflected in the second event

 ,    tree displayed where "NN" denotes "not needed."    Another ex-ample is the case where system 2 is only capable of operating in conjunction with the successful operation of system 1. Such a condition could be the result of some physical interaction (Type 2C) that takes place when system 1 fails. This is re-flected in the third event tree displayed where "IM" denotes
      " impossible."

To illustrate accounting for dependencies of Type 2B (Attachment 5), suppose that the fault trees developed for sys-tems 1 and 2 are found to contain the same component failures, 1 A and F, as basic events. Components A and F are shared-l equipment dependencies and can be treated by being incorporated into the event tree displayed. The PRA methodology disregards the labels or grades (

      " safety-related" and "non-safety-related" systems. It eval-() uates performance of the systems entirely on their engineered or reliability merits. In the simple event tree displayed (Attachment 5), system 1 could be " safety" grade while system 2

could be "non-safety" or vice versa. In general, the PRA approach extends well beyond the " design basis" events and the (])

     " single failure criterion."                                      This is equally true of the Shoreham PRA.

4. PRA Study Levels PRA's can be performed at several levels of scope depen-ding on the objectives and perspective sought. The PRA Procedures Guide has identified three discrete levels of scope.

1. Systems Analysis

2. Systems and Containment Analysis

3. Systems, Containment and Consequence Analysis A level 1 PRA consists of an assessment of plant design and operation focused on the accident sequences that could lead to core-melt, their basic causes and frequencies. The results are a list of the most probable core-melt sequences and insights into their causes.

A level 2 PRA, consists of an assessment of the physical processes and the response of the containment in addition to the scope of a level 1. In addition to estimating the core-melt frequencies it predicts the time and mode of the contain-ment failure as well as inventories of radionuclides released () to the environment. As a result core-melt accidents can be categorized by the severity of the release. \

A level 3 PRA assesses the transport of radionuclides () through the environment and assesses the public health and economic consequences of the accident in addition to performing the tasks of a level 2 PRA. The results are generally pre-sented in the form of a " risk curve" depicting the frequency of various consequences. An assessment of external events may be included in any of the three levels described. The selection of the events depends on the site but includes such events as plant fires, internal and external floods and earthquakes.

5. The Shoreham PRA and Systems Interactions The on-going plant specific Shoreham PRA Study 13/ has acdressed the issue of system interactions by virtue of con-structing and quantifying the plant event / fault trees which include the treatment of common-cause initiating events, inter-system dependencies (functional, shared equipment, physical interactions, human interactions) and intercomponent dependen-cies.

Loss of off-site power and internal flooding in the re-actor building are good examples for common-cause initiators. 13/ SAI-001-83-SJ, Probabilistic Risk Assessment, Shoreham Nuclear Power Station, Long Island Lighting Company, Preliminary Draft, March 1982.

Practically every plant event tree with associated fault (p trees can be viewed as an example for functional dependencies. The study evaluated 19 plant event trees; e.g., event tree for sequences initiated by turbine trip, loss of feedwater, loss of condenser vacuum, inadvertent opening of relief valve, large, medium and small LOCA, etc., as well as 5 containment. event trees. The plant event / fault tree construction process was supported by development of dependency matrices. Shared equipment dependencies were examined by virtue of employing system level fault trees supported by dependency matrices; e.g., HPCI/RCIC/feedwater, low pressure injection systems ADS /LPCI/CS, electric power and instrumentation, ser-vice water system. The study examined 10 system level fault trees. Human interactions were also treated within the frame-work of fault / event tree construction, e.g., common-mode misca-libration, maintenance errors, maintenance unavailability, etc. Intercomponent dependencies were addressed by virtue of assessment of common cause or dependent failure probabilities; e.g., diesel generators, LPCI and CS pumps, HPCI/RCIC, etc. Examples for spatial or physical interactions would be: impact on ECCS from internal flooding, impact on ECCS from () interfacing LOCA, impact on ECCS from containment leakage, etc. 1

I Like any other PRA study the Shoreham study considered () multiple failures, went well beyond design basis accidents and did not distinguish between labels of safety and non-safety systems, but assessed systems unavailabilities on the basis of their engineered merit or reliability characteristics. It appears that the contractor has screened the BWR op-erating experience from the standpoint of unusual occurrences that migh t create potential for adverse systems interactions. In additi an, two plant walkdowns with LILCO personnel were con-ducted. To enhance the completeness of its PRA, LILCO should continue its search for hidden dependencies by using operating experience of similar or surrogate plants. LILCO's program (described in Part V(a) above) to review Nuclear Power Plant experience under the charter of the Independent Safety Engineering Group (ISEG) with emphasis on unit design and con-figuration similar to Shoreham provides assurance that this objective will be fully satisfied. In addition, LILCO's com-trees mitment to maintain fault / event, current provides an invaluable

 ,    mechanism for avoiding undetected                                         dependencies from plant changes or modifications.

O

6. Level of Effort O In its preliminary draft form, the Shoreham PRA study is equivalent to a Level 2 study. Upon completion of ex-plant consequences it will be raised to Level 3. Numerous determin-istic assessments of common-cause initiators or systems inter-actions have been conducted, e.g., pipe break and internal flooding evaluation,14/ missile evaluation,15/ turbine missile evaluation,1g/ fire hazard analysis,17/ hurricane study,18/

etc. Probabilistic evaluations have been so far limited to aircraft crashes 19/ and internal floods.20/ Generally speaking, major elements for the Shoreham ( study such as plant systems analysis, in-plant accident pheno-menology and containment event trees represent state-of-the-art in the PRA methodology. 14/ FSAR App. 3C. 15/ FSAR 3.5 Ig/ FSAR 10.2 1_7f SWEC Report, FHAR. 1@/ FSAR App. 2N () 19/ "A Study of the Probability of An Aircraft Using the Grumman Peconic River Airport Colliding into the Shoreham Nuclear Power Station," PLG, April 1976. 20/ Shoreham PRA, Appendix G.

1

1 The Shoreham PRA effort and methodology exceeds the NRC I i () developed IREP methodology which is equivalent to level 1 defi- l nition from the PRA Procedures Guide. The Shoreham PRA is level 3. )

7. Status and Results of Shoreham PRA The SNPS PRA is in progress. The systeme. and contain- I ment analyses are in the preliminary draft form while the ex- l plant consequence part of the study has been initiated. The )

schedule for completion of the study is addressed in LILCO's portion of the testimony. So far, five Peer Review Group meetings have taken place, including two at the site. Numerous comments have been provided to the contractors and LILCO. The Peer Review Group process has been much more interactive than in other recent cases known to the writer. The comments have not yet been fully formalized since the Peer Review Group is not scheduled to complete its review before the year end.

         ^n my-part I have not cbserved any cpecial risk Outl-     s which would suggest that Shoreham BWR stands out f om other BWR's subjected to PRA scrutiny like Peach Bottom. This also indicates that no unique non-safe   / safety systems interactions O, exist. The mean value of core-melt       fr\

equency~4s comparable to p/ that of Peach-Bottom analyzed exhaustively in the Rea r Safety 4*ndyr-WASH-1400.

My preli "inary-evahathn-has-not-identified-any-non-safety-grade system 7he-f ai-lure-of-whichr-by-itsel-f 7 =A "4^- out - mu Ltiple-fei-lures-in-the-safety-grade-syc tc=c , .cu14-eeuse-cora-degradation. I have reviewed the entire Shoreham PRA draft as it currently exists and I have participated in the five peer review sessions to date concerning the PRA. Alth: ugh my-review is continuing, it is my professiciiel vgiulou thet the Sher-eham-PRA ic a competent-state-of-the-art-effort which ceii-firms-the-safety-adequacy-of- the--Shorehain design =.

8. Response to Intervenors' Allegations (a) SNPS PRA Allegedly Does Not Consider Systems O Interactions With respect to page 66 of Intervenors' prepared testi-mony, the Intervenors are totally incorrect stating that the Shoreham PRA does not take systems interactions into account.

On the contrary, the Shoreham PRA exhaustively scrutinizes such events. With respect to pages 66 and 67, the Intervenors are totally incorrect stating that Indian Point 3 PRA, like the Shoreham PRA, "did not address failures and between-systems dependencies." The Shoreham PRA does address these. On page 66 of the Intervenors' prepared testimony refer-ence is made to qualitative techniques such as commonality dia-grams, dependency matrices and FMEA's. It should be recog-nized that these techniques provide no new information, but only a different format to event / fault trees and as they are only qualitative in nature, they do not provide a method for focusing on important public safety considerations. In my ex-perience, when used in isolation, these formats are of limited usefulness for assessing public safety, and are useful only as n design tools. Appropriately, SAI, the Shoreham PRA study con-(_) tractor, has extensively used the dependency matrices to sup-port event / fault tree construction in the Shoreham PRA. FMEA's have been employed by both the reactor vendor and the A/E. On page 67 of Intervenors' prepared testimony, they say, "the dependency analysis PRA should be complemented by a physi-cal survey or walkdown such as performed at Diablo Canyon."

The usefulness of plant walkdowns cannot be denied as a way to check design adequacy or identify spatial dependencies among (') systems. The SAI and LILCO engineering staff have conducted such walkdowns. It is, however, far from clear how the , Shoreham public safety features would be enhanced by a Diablo - l Canyon type of walkdown. The Diablo Canyon plant is somewhat unique due to proximity of the Hosgri Fault and the resulting upgrading of the plant design. It should also be noted that the NRC staff conducted a l survey of the state-of-the-art of methods that could be em-ployed for systems interaction reviews. Three National Laboratories (Battelle Memorial Institute, Brookhaven and Lawrence Livermore) aided in performing the survey. Their j final reports with recommendations are documented.21/ These, l plus continuing efforts, have not identified any new method-ology. The NRC staff's final conclusions on this might be sev-eral years away, according to the ACRS.22/ In my opinion,.the Shoreham PRA approach provides a meaningful and efficient, if not the only, framework, for examining the safety significance of the " Systems Interaction or Interactions" issue. One of the NRC sponsored studies seemed to have arrived at the same or similar conclusion: " Systems interaction analysis can be encompassed within probabilistic risk assessment analysis using O. the same tools as in PRA, but extending the analysis to place a more detailed emphasis on dependent failures including greater l consideration to support systems, shared environmental condi-l tions, and dynamic human error."23/ One can safely conclude l that SNPS PRA provided the best means for addressing the issue. l l 21/ A. Buslik, I, Papazoglou, R. Bari, Brookhaven National - l Laboratory, " Review and Evaluation of System Interaction Methods," USNRC Report NUREG/CR 1901, January 1981; P. Cybulskis, et al., Battelle Memorial Institute, " Review of Systems Interaction Methodologies," USNRC Report NUREG/CR-1995, January 1981; J. Lim, R. McCord, and T. Rice, Lawrence Livermore National Laboratory, " Systems Interaction: State-of-the-Art Review and Methods Evaluation," USNRC Report NUREG/CR-1859, January 1981. () 22/ USNRC, ACRS, P. Shewmon to N. Palladino letter,

Subject:

Report on Systems Interactions study for Indian Point Nuclear Generating Station Unit 3, March 9, 1982. j 23/ Lim, McCord and Rice, note 21 above. i 1

(b) SNPS PRA Is Allegedly Deficient Because It Does Not Consider Fire, Sabotage or Earthquake An assessment of external events requires the use of specialized methods to address important factors not usually encountered in the assessment of internal events. Such a spe-cialized method has not yet been developed successfully for the probabilistic treatment of sabotage events. It has been out-side the scope of any PRA study the writer is aware of, includ-ing the PRA Procedures Guide. The selection of any external event for a detailed risk assessment depends on its frequency of occurrence, magnitude, proximity and consequences. The PRA studies conducted to date have treated the external events to varying degree of detail. Some studies have excluded ext'ernal events altogether. Some have treated them in a " bounding analysis" manner. Some stu-di<as have been motivated by external events.24/ The degree of uncertainty in estimating risk due to external events tends to be greater than that associated with internal events. These stem from less experience, use of relatively new ar..lytical techniques and greater reliance on engineering judgment and expert opinion. O 24/ Pacific Gas & Electric, 1977.

Screening of external events to select the significant

   'T ones constitutes the existing practice. The first step is to

('_/ list all external events specific to the site and plant.25/ . Screening criteria are subsequently established.26/ As a part of this process predominantly deterministic studies are revie-wed for the applicability. Stone & Webster conducted a fire hazard analysis for Shoreham with an objective of establishing sufficient separation of systems / components in the reactor building such that the postulated fire would not prevent safe shutdown. Another example would be studies related to the plant geology and seismology.27/ () (c) SNPS PRA Allegedly Does Not Consider Multiple Failures Intervenors incorrectly suggest that the Shoreham PRA does not consider multiple failures. The probabilistic ap-proach in general considers a stream of multiple failures. SNPS PRA study is no exception to this and hence the contention has no substance. f- 25/ PRA Procedures Guide, Table 10-1, page 10-4 and 10-5. V 26/ Ibid, page 10-10. 27/ SER, Section 2.5 Geology, Seismology and Geotechnical Engineering, NUREG-0420. -

9. Conclusions As indicated already, the term " systems interaction or interactions" has become a buzz-word meaning different things to different people. For this reason, a definition is offered in this testimony which provides appropriate focus. Systems interaction'is a subset of dependent failures (common-mode or common-cause). Dependent failures are adequately defined in the PRA Procedures Guide. Attachment 5 reproduces the classi-fication employed in the Guide. By virtue of offering this definition, the systems interaction assessment is encompassed 1

within the framework of a PRA employing the same basic tools. , () Hence, conduct of a PRA study is the best method available for i

         .                                                              1 l    resolution of the issue.

! Dt-itc the f&ct that-the-Shoreham Pm is stui lu 1 i gress, no spe - 1 risk outliers have been observed whic ould l suggest that the Sho am BWR stands out from ot - BWR's sub-jected to PRA scrutiny. T 's also indicatp that no unique adverse interactions exist betwe a non-safety / safety related systems. The mean value for c txfrequency is comparable l x to that for Peach Bottom- nalyzed exhaustively in the Reactor l I _/ This is significant b ause it shows t Safety Study, [WASE I400. i that the obability of core melt from the operatio Shorebah is as extremely unlikely as it is for Peach Bot om as qcnfi rmed-by-WASHmOO-The1horeh am-P RA-h a s-not--iden t/ 4 l l

any-non-safet-y ral atari =ystam, tha f ailure-of-which, by itself- ]s and multiple failures of safety related sys

                       ~
                                                             , would cause core degradation. In my pro espitnal opinion the Shoreham PRA st;udyd a competent state-of-the-a         fort which co             a--

f ety-adequacy-of-the--plant-design. Thus, the Suffolk County and Shoreham opponents Coalition testimony has incorrectly arrived at the conclusion that LILCO has failed to utilize improved techniques like PRA, and that systems interactions which may have a significant impact on plant safety have gone undetected. LILCO's initia-tion of the study, despite the absence of a regulatory require-ment, was a far-sighted move contributing substantially towards resolution of the " systems interactions" issue. O

B. Testimony of Edward T. Burns, Ph.D O 1. Introduction Long Island Lighting Company has initiated a probabilis-tic risk aseossment of the Shoreham Nuclear Power Station to provide added assurance that the health and safety of the pub-l~ lic are protected. The PRA can be used in a variety of roles including the evaluation of potential systems interaction. Science Applications, Inc. has performed the systems evaluation along with the in-plant containment analysis and the following discussion regarding systems interactions is based upon these investigations and calculations. The following discussion addresses the applicability of the Shoreham PRA in the evaluation of potential systems inter-action issues. Necessarily some background material is also presented to provide a clear definition of the terminology and the present state of the technology.

2. Discussion of Probabilistic Risk Assessment Methods and Objectives A PRA on the Shorham plant has been performed for LILCO.

Through the use of probabilistic risk assessment (PRA) methods, LILCO has taken the initiative to address several issues in-() cluding a positive step regarding the solution of the systems interaction issues. This section discusses the methods used in the PRA and how these compare with other available techniques

for the evaluation of potentially adverse systems interaction. The PRA affords LILCO the following: (])

               -   A quantitative assessment of the risk associated with the plant.
               -   A logic model of plant systems which can be evaluated both qualitatively and quantitatively.
               -   A framework within which systems interactions can be identified and assessed.

The PRA methodology is a systematic approach to the identification of postulated accident sequences and the fail-ures which can cause these accident sequences. The technique used in the Shoreham PRA to evaluate system response during postulated accidents is event tree / fault tree methodology. The () PRA involves the assessment of the plant far beyond the design basis, and is used to assess the ultimate capability of the plant to protect the plant and the public. 2.1 PRA Methodology The Probabilistic Risk Assessment performed for the Shoreham plant provides basic data for a utility risk manage-ment program. Methods similar to those used in the Reactor Safety Study (WASH-1400) are used in the Shoreham Study. The i l following evaluations are performed as part of the Shoreham PRA. () - Quantitative evaluation of accident initiators and the possible course of accidents following the initiators In-containment analysis of radioactive source terms possible following the accident sequences 1 1 I

Containment analysis to assess mechanisms and potential locations for containment failure O - Separate evaluation of potential consequences to the p<blic. Currently, PRA guidelines are being developed (1).2g/ The Shoreham PRA is consistent with the draft guidelines, how-ever, no sanctioned NRC guidelines presently exist. In a plant specific risk assessment, the probability (frequency) of accidents and then subsequent effects (conse-quences) are assessed. The three major tasks are: Phase I Determination of the probebility of radio-active release Phase II Determination of the magnitude of the radi-(3 oactive releases for each unique accident (_/ sequence including the radioactive species l and release time Phase III Determination of the consequences of a ra-dioactive release to the environment or the public After the probability and consequences of each accident sequence are determined for each accident or initiating event, the individual results are summed to obtain the overall proba-bility versus consequences. The purpose of this discussion is to focus on the eva-l luation of system performance as included in the Shoreham PRA. Therefore, the following text will be directed at only event O tree feu1t tree efferte. 2g/ See the references in section 7 of this testimony. l l

The application of probabilistic safety analysis to the evaluation of reactor systems provides a method of determining (]) the systems interaction as a function of event sequence and of estimating the likelihood of failures which can interfere with the ability of the safety systems to maintain a desired func-tion such ar providing adequate core cooling. The methodology used in the Shoreham study is a fault tree / event tree analysis, similar to the approach used in the Reactor Safety Study (WASH-1400). Fault trees are a set of logic diagrams describ-ing the potential equipment failure modes which could disable a system or group of systems. The analysis begins with the selection of initiators. Because of the " defense-in-depth" concept utilized in nuclear plant design, it is highly probable that a given accident se-quence will be terminated before it can affect the public. However, because of the complexity of systems available to pre-vent radioactive releases in nuclear units, diagrams are used to show the progression of accident sequences and the impact of each sequence on the plant's capacity to prevent releases. These diagrams are called event trees. The quantification of accident sequences determined from these diagrams depends on the probabilities developed at each decision point in the event () tree. To determine the probabilities at the decision points (nodes) of the event trees, fault trees are prepared to model the system interactions and the conditional probability of system and multiple system failure.

Event Tree / Fault Tree Analysis (ET/FTA) is a formalized () deductive analysis technique that provides a systematic ap-proach to investigating the possible modes of occurrence of a defined system state or undesired event. The fault tree model of a plant or system has been used as a logical method of dis-playing and evaluating component and system interrelationships. One of the principal benefits of ET/FTA is that it offers a convenient format for assessing the reliability of redundant systems with multiple failures or system unavailabilities. The failures can be of several types: multiple independent fail-ures, cascade failures, or dependent failures. ET/FTA consists of two major steps: (1) the construc-(w]

 \s tion of the event trees and fault trees, and (2) their evalua-tion. The reliability of combinations of systems used to re-spond to demands on the plant safety equipment are determined through the following steps:

Understanding of the system interactions which will prevent the occurrence of an undesired con-sequence Defining the system functions and success cri-teria Reducing the system description (including con-trols and interfaces) to fault tree format down to the detailed component level Incorporating operating procedures into a fault

.)

tree model Combining test and maintenance procedures and technical specifications (limiting conditions for operation and surveillance requirements) into the analysis \

i 1 Quantification of the fault tree events () - Incorporating the probabilities into the event trees. I The event trees used in this study provide the basic 1 tool for displaying the various accident sequences considered, and for relating the probabilities of radioactive release on a f consistent basis. Probabilities for the events shown on the , event tree are estimated by a system fault tree analysis to identify system functional dependencies, components or human l interactions that may contribute to failures of systems and functions, and to quantify the probability of these system i failures under accident conditions. ]

                              )             The event trees provide a framework for linking together the results of the fault tree analyses.        Functional failure probabilities for a system are derived from evaluation of the I                                    applicable failure modes of the system, given the initiator and 4

accident sequence. The same functional failure can be strongly

dependent upon the type of initiating event or accident se-l quence. Determination of accident sequences requires precise success criteria'and consideration of the conditions under which the system is called upon to perform.

In a manner similar to WASH-1400, the Shoreham PRA is performed on a realistic basis. Equipment capability, success criteria, and event sequences are modeled realistically to de-termine, as acurately as possible, the expected course of 1 events and conditions. 1

2.2 Shoreham PRA Perspective () The implementation of the systems analysis portion of the PRA methodology at Shoreham can be summarized as follows: The analysis is based upon existing technology and evaluation techniques, i.e., fault trees and computer codes. The methodology has been applied successfully in nuclear power plant evaluation to identify po-tential dependent failure problems since WASH-1400 (1973). The Shoreham PRA includes all plant systems regardless of classification (safety and non-safety) which may affect public risk due to accidents. The PRA provides LILCO management and engineer-ing staff important insights into the potential contributors to risk. Certain exclusions have been superimposed on the PRA based upon LILCO's judgement that sufficient i deterministic analyses have been performed to ensure adequate public protection against these events (e.g., seismic and fire). Currently the NRC is sponsoring a program to develop interim guidance for the conduct of pilot reviews to identify systems interactions (2, 3). Reviewing the limited information available on the interim guide indicates that the systems in-teraction studies are to be solely and explicitly aimed at uncovering hidden dependencies that would degrade a function. A Brookhaven National Laboratory report (6) states that the () following definition for system interactions is appropriate: 1 1

a situation where the likelihood of an undesired event is increased due to the (~} relationship between two or more compo-U nents. This definition is quite general and subject to widely varying interpretation: The definition of systems interaction as interpreted by BNL (6) under contract to the NRC includes the following fea-tures:

a. the mechanism of the interaction (the initiating cause and the relation which permits the inter-action),

b. the probability that such an interaction will take place, and f_s

c. the consequence of the interaction in terms of (v) its impact on the overall level of risk.

The types of systems interactions which have been clas-sified thus far include the following (2): Functionally coupled systems interactions result either from the sharing of components between sys-tems or through physical connections between systems including electrical, hydraulic, pneumatic and me-chanical. Spatially coupled system interactions result from the proximity of systems to one another within the l plant. For example, a steam leak could short out an electrical junction box across the room from the steamline. A systems interaction results is based on this spatial coupling. Inherent to a spatial I coupling is the concept of spatial domain. Typical spatial couplings involve water, steam, fire, explo-sion, radiation, or pipe whip. The domain over (,) which a coupling can realistically occur will vary depending upon the barriers in the plant. For ex-ample, water leaking from a line in one room may affect equipment in adjoining rooms. But a high pressure pipe whip will affect only systems in the l room within reach of that pipe.

Human coupled systems interactions are special since the operators could influence all systems 18 the (' plant. To better focus the reviews, the guidance excludes human error and sabotage from systems in-teraction reviews. Systems interaction review will assume the operator follows procedure when interact-ing among systems and the procedure is correct. The focus is a fault within one system that induces the operator to influence another, otherwise independent system in the unsafe direction. To illustrate let us postulate a failure with a power supply that causes instruments to display spurious readings to the operator who is misled into influencing another system. 2.3 Comparison of PRA with other Methodologies PRA is an emerging technique for which guidelines are currently being developed (1). The technique has been success-fully applied to nuclear power plants beginning with WASH-1400. () In addition, there are other methods for treatment of systems interaction which have just recently been identified but are 1 yet to be demonstrated as effective tools. These methods in-clude system walkdowns and dependency matrices.29/ In this section a brief summary comparison is made among the analysis 1 options which may be available. Principally these methods in-clude: s 29/ These methods are also suggested by the Shoreham () Intervenors. This section shows that these techniques have indeed been recognized by LILCO and are used in the Shoreham PRA. However,'a complete definition of these methods and their applicability to systems interaction has not yet been achie-ved.

1. Event tree / fault tree methodology as described in the PRA Guide (1), IREP, or the Shoreham PRA -

() 2. Dependency diagrams The Systems Interaction analysis (3) has as its sole purpose the identification of potential functional and human coupling or interfaces which could defeat multiple systems.- In each of the studies performed to date (5, 6, 2) to establish a preferred methodology, the criterion to be used in measuring the acceptability of a system interaction analysis has not been clearly established or discussed other than in reference to a potential safety goal or risk reduction. Since qualitative tools such as dependency matrices and commonality diagrams can-() not, by themselves, be related directly to these goals, a basic difficulty exists with these methods when used outside the framework of PRA. In other words, there are no criteria on which mechanisms are to be considered, on what probability is unacceptable, or on which consequences may be unacceptable. The PRA on the other hand has a much broader charter since it must identify failure modes of systems and multiple systems, evaluate these (qualitatively and quantitatively), and assess public risk associated with such failures. However, in order for the PRA to effectively accomplish its purposes it must treat the systems interactions issue in a thorough manner.

Support for the use of PRA as a basis for the perform-ance of systems interaction studies can be found from published, studies of those who have investigated the available methodolo-gies (3): Risk analysis can satisfy the requirements of systems interaction if greater emphasis is placed on the sources of dependent (. common cause or common-mode) failure - shared support systems, shared environmen-tal conditions, and human errors. Consider for a moment the relationship of the NRCs IREP program 30/ to both SI and the Shoreham PRA.31/ The Interim Reliability Evaluation Program (IREP) has involved five plants, Crystal River (B&W), Arkansas-1 (B&W), Millstone-1 (BWR), (} Browns Ferry (BWR), and Calvert Cliffs-1 (CE). In IREP, the emphasis has been on gaining insight concerning the plant's safety. This has been accomplished initially by using event tree / fault tree methodology augmented by such things as depend-ency matrices that relate auxiliary and support systems to the front line systems, where front line systems are defined as those systems which appear on the systemic event trees. Once the f ault trees -are completed, the support system fault trees are merged with the front line system fault trees. Thus a I front line system fault tree is complete with no transfers in. 3_0/ Level 1 PRA I 31/ Level 3

The idea is to concentrate on the dependencies among front line r systems and the support systems to identify systems interac-O] tions. IREP has a limited scope and must be viewed as not pro-viding an overview comparable to the Shoreham PRA. The Shoreham PRA is a Level 3 PRA (1) compared with IREP which is a Level 1 PRA. Therefore the Shoreham PRA provides additional informa-tion to LILCO regarding the safety of the plant beyond that which would be obtained from a IREP study. To the extent that systems interactions are addressed in a typical IREP study, so has the Shoreham PRA incorporated that level of detail. The Shoreham PRA has incorporated each of these methods to a level O of detail comparable to the IREP program. Since there is no formal guidance or criteria on the level of detail to be ap-plied to these methods, the completeness of these techniques can only be judged on a relative basis with respect to compara-ble analyses. These techniques were recognized as potentially useful and therefore have been incorporated directly in the PRA as additional input in the logic model structure and evalua-tion. SAI judges that fault tree / event tree methodology is the best available technique for augmenting the existing determin-(} istic evaluations and NRC regulations to ensure that systems interactions are exposed and potential areas of concern are identified. Other independent assessments of possible

l l methodologies have concluded that event tree / fault tree methods f are preferable to others available (1, s, 7, g).

3. The Application of the Shoreham Fault Tree / Event Tree Methodology to Systems Interactions Evaluation 3.1 Introduction to the Shoreham Plant Specific PRA Utilities and the NRC are continuing to develop methods, techniques and criteria to ensure that the redundancy built into nuclear plants is not defeated by subtle systems interac-tions which may be overlooked in the design, testing and opera-tion of a plant. Thus far the methods used include:

single failure criteria, defense-in-depth design concept, diversity of systems to fulfill a safety func-tion, selected deterministic evaluations on special topics such as fires and seismic events, and FMEAs. Over and above these techniques LILCO has made use of event tree / fault tree techniques to gain additional insight and perspective regarding the potential for as yet unidentified systems interactions affecting the public safety. The use of event tree / fault tree logic models is the best available method for identification and evaluation of systems interaction (5). (} This methodology represents a highly structured framework that has the following desirable attributes; it is

e o , systematic flexible (-)g

                   -   reproducible                                            .

Within this framework each system can be describe ~d in the logic models in a step-by-step examination to a fine level of detail. . The event tree / fault tree methodology augments the other more deterministic approaches used by LILC3 to ensure plant safety. A major advantage of the use of the event tree method-ology is that it provides an overview of the postulated.acci-dent sequences and the plant functions required to maintain, ' public safety. This overview is useful in assessing the impact

    ) of sequence dependent functional failuretiof systems.         In addi-
                                                                                        /'

tion to the functionally induced failures there are also sy' stem interfaces, support systems, human interactions, and other sys = tem interactions which are accounted'for in accident sequen'es c via the system level fault trees. Section 3.2 summarizes sev-eral examples of the techniques used to identify these interac-tions. There are a total of 19 event trees developed and used ) in the Shoreham PRA compared with 4 in.NASH-1400; and there are ,

)!

10 detailed system fault trees (i.e., similar to WASH-1400). Each has been developed for, and is tailorc;* to, the plant spe- {} cific condition at Shoreham. 1 e \

                                                                                          -100-LILCO has chosen the PRA methodology to assist in iden-tifying any systems interaction which could compromise the pub-lic safety. While other methodologies have been discussed in the literature, these methodologies have not been completely
                                                       . f o rmalized , accepted by the NRC, or applied to the evaluation of a nuclear plant.

The PRA framework provides a method of systematic review

                                                     ' which facilitates the identification of systems interactions that affect core melt.        The PRA carries the analysis of systems interactions and dependent failures far beyond that required in design. basis analyses.       Multiple component and system failures
                                                      'are included to ascertain the ultimate capability of the plant.

1 () Both safety and non-safety systems and interfaces are included. The dependencies induced or aggravated by environmental degra-dation are also incorporated into the PRA. In summary then, ! the knowledge to be gained from the event tree / fault tree re-view includes, but is not limited to the following: Identification of sources and types of poten-tially adverse systems interactions by exploring dependencies down to the component level includ-ing auxiliary / support system dependencies for both safety and non-safety systems. Identification of system and system combination failures for these accident sequences postulated to lead to core melt. (]) - Determination of commonalities and interconnec-tions in identified front line systems or func-tions caused by systems interactions.

                                      -101-In addition event tree / fault tree techniques have the
   -  advantage of allowing both:

qualitative investigation to identify the groups of systems interactions which may adversely affect the plant's sr.faty; and quantitative investigation to evaluate whether the systems interaction have any important safety significance. 3.2 The Shoreham-Specific Methodology Used in the Identification of Systems Interaction

             -The application of event tres/ fault tree methodology at Shoreham has been performed with the recognition that these logic models must account for potential systems interactions in order to accurately assess the risk to the public. To ensure

(~T that the event tree / fault tree logic models are utilized to V their fullest potential to identify systems interaction, the following aids have also been employed in the analysis of the Shoreham plant: Plant Walkdown Containment interaction asessments Review of Systems interaction evaluations (SI) BWR Licensee Event Reports (LERs) and otter operating experience data A detailed engineering review Accident initiators Existing LILCO deterministic analysis

                  -   Dependenc:y Matrices i

                                                         -102-3.2.1  System Walkdown System walkdowns are a useful tool available to the ana-lyst to identify spatial dependencies among systems.                   The Shoreham system walkdowns were performed as part of the PRA.

The walkdowns were performed by the technical group responsible for the system fault tree analysis and the LILCO engineering , staff. The input from the walkdown was directly implemented in ! 1) the construction of the fault tree models, and 2 ) the iden-tification of potential independent multiple system failure modes (i.e., system interaction). The walkdown involved the following principal steps: Fault tree analysts prepared for the walkdown by

 /^^                                reviewing the FSAR, system descriptions, P& ids,

^ k- and operating procedures. Checklists were then prepared to be filled out during the walkdown to assess potential adverse environment on system represented in event trees. The purpose of the walkdown was specifically designated to identify system dependencies and interfaces which could fail multiple safety or non-safety systems. All systems identified as potentially useful in mitigat-ing postulated accident sequences were investigated. The scope of the walkdown can be gauged by considering that each of the two walkdowns required two days at the Shoreham site and in-volved five fault tree analysts and five LILCO system / operating

engineers. This should be coupled with the preparation time of 15 man days and the report writing time of 10 man days. A

                                   -103-total of 65.1 man days was involved. In my professional opinion the walkdowns performed at Shoreham were adequate and appropriate for the use of walkdown techniques in a PRA ana-lysis.

3.2.2 Containment Interactions The Shoreham PRA takes advantage of another benefit not generally available through other methodologies, specifically, the interaction with containment and the containment failure modes under degraded core conditions is established. In many cases containment integrity and its interaction with shutdown systems can be crucial in the assessment of public risk. Level _s 1 PRAs such as IREP do not adequately address the containment interaction issue. PRA's of Level 2 and above are the ones which explicitly address these dependencies. Shoreham is a Level 3 PRA. 3.2.3 Systems Interaction Analysis Another technique used in the Shoreham PRA to support the event tree / fault tree approach as a system interaction identification technique was to review the following studies which have also examined nuclear power plants and the potential for common cause or systemic interaction:

1. WASH-1400

(} 2. Clinch River Breeder Project (CRERP)

3. Limerick PRA

4. Reactor Safety Study Methodology Application Program (RSSMAP)

                                      -104,

5. IREP

(' P

6. Oconee PRA

7. Big Rock Point PRA

8. Special studies (i.e., DC power reliability, diesel generator reliability, anticipated transients).

Through the review of these previous and currently on-going evaluations, SAI has identified a number of potential system interfaces and dependencies which were then incorporated into the Shoreham PRA framework in a plant specific manner so that they could be reviewed in a consistent framework along with other contributurs to risk. 3.2.4 BWR LERs and Operating Experience () Another method employed in the search for systems inter-action is to make use of the BWR operating experience data (e.g., LERs) to establish if usual occurrences which have actually occurred in the industry may shed some light on the potential for adverse systems interactions at Shoreham. The operating experience must be screened not only for serious event sequences but also precursors which are bound to occur with a higher frequency. With this information the . malt tree analyst may consider adding additional failure modes not previ-ously considered or incorporating intersystem dependencies if the plant specific design is subject to similar dependencies as , described in the LER.

                                -105-Along with the review of other SI studies, the Shoreham PRA has utilized the review of LERs and other operating experi-ence data to formulate both fault trees and event trees espe-                 .

cially potential coupling of systems through operator error or auxiliary system dependency. 3.3 Examples of Systems Interactions Investigated in the Shoreham PRA LILCO through the use of the Shoreham PRA methodology has identified both systems interaction potential and the effects of multiple failut as. The areas where the most success has been achieved through the application of event tree / fault tree methodology include identification and quantification of the following: Functional Coupling 32/ Shared system dependencies Human Coupling among systems Intercomponent dependencies Spatial Dependencies . In an effort to highlight some of the systems interac-tions which are explicitly identified and evaluated in the 32/ Functional Decendencies. These are dependencies among systems that follow from the plant design philosophy, system O capabilities and limitations, and design bases. One example is a system that is not used or needed unless other systems have failed. Another is a system that is designed to function only in conjunction with the successful operation of other systems (1).

                                 -106-Shoreham PRA, Table 3.1 (Attachment 6) has been constructed to summarize some of the examples in each of the above categories.

The conclusion from the analysis can be summarized briefly as follows: Functional Coupling - the event tree provide a useful formalism through which the analyst and reviewer can establish the coupling among sys-tems for each postulated accident sequence Shared system dependencies - both fault trees and dependency matrices (Appendix B of the PRA) are used together to identify system dependen-cies. These identified dependencies are then evaluated for each accident sequence to estab-lish the appropriate interaction level on a se-quence dependent basis Human coupling - The human interface with safety and non-safety systems is a key contributor to O system availability. The human interface is modeled in the Shoreham PRA for: maintenance errors manual initiation instrument miscalibration maintenance unavailabilities.

            . Operator actions are incorporated directly in the fault tree and event tree construction.

Operator actions which may cross system boun-daries affecting multiple systems or redundant divisions are included. Intercomponent dependencies - A step forward in the modeling of component dependencies has been made in the Shoreham PRA. Based upon operating experience data, sinilar components are found to have a higher multiple failure probability than O- would be predicted by virtue of random indepen-dent failures. This is more apparent in examin-ing control rod scram precursor failures and multiple diesel failures; however, the Shoreham PRA has also included other component dependencies in this dependency modeling.

                                   -107-Spatial dependencies - The Shoreham PRA has ex-plicitly addressed some of the important spatial dependencies in the Shoteham plant including:

flooding in the reactor building interfacing LOCA sequences containment leakage during degraded core conditions -

                   -   repair effectiveness.

In addition to these analyses which are included in the PRA methodology, other LILCO deterministic eva-luations have also been performed to ensure that spatial dependencies are not dominant contributors to adverse systems interactions. For a brief de-scription of these see Part V above.

4. Results and Statuo of the Shoreham PRA The Shoreham PRA has been completed in draft form and is presently undergoing peer review. There are currently on-going efforts to provide additional explanation and discussion of the results of the PRA however these are primarily informational in nature and will not change the results or conclusions of the PRA.

The PRA was begun by LILCO as a risk management tool. As such, one of the key features of the analysis methodology is the identification of key contributors of public risk. Previously, dependent failures have been identified as poten-tially important causes of system or multi-system unreliability (9, 10). Therefore, a PRA must address these dependencies or [} the accuracy of the calculated risk would likely be compro-mised. -

                                                                      }
                                    -108-As recommended in several studies (4,     5, 6, 7, 8),

{} Shoreham has performed a systems interaction evaluation using event tree / fault tree methodology within the framework of the PRA. This system interaction analysis provides further confi-dence that there are no subtle dependencies among systems which could lead to multiple failures. The PRA incorporates event tree / fault tree methodology to perform the following: identify potential accident initiators develop event trees to account for multiple sys-tem failures and follow the variety of possible accident scenarios. display the front line systems effect for each () accident scenario incorporate auxiliary system dependencies on front line systems and construct a complete com-puter logic model combining both the front line and auxiliary systems. The result of this process is to obtain a logic model 4 which includes the effects of all systems (safety and non-safety) on the progression of the accident sequences. While systems interactions which could disable multiple systems were identified and evaluated in the Shoreham PRA, these interactions have previously been identified or were pre-sent in accident sequences of very low frequency (and beyond () the design basis) such that the risk to the public is low. Thuc far, the reeults of the draft PRA havc not revealed eye-tame inter-actions givbleins-whinh unnl a "aquire redesign nr

                                       -109-modification.     .^ .ne tlia t way Of stating L121= is Litat-the SMPS-PRA has-notMentitT6d7ny pIant-Uni'que problems-which-are-out--of O  line-with-other-LWR PRA evaluat-iens.

Consicering the state of the art in the evaluation of system interactions, LILCO has made a reasonable attempt to augment current specialized deterministic methods by employing event tree / fault tree methodology to identify subtle systems interactions which may occur during postulated accident se-quences. The PRA commissioned by LILCO ' carries the system in-teraction investigation beyond current NRC regulation require-ments. Thue-f a r in - the-evalu a ti on . no ayntems iuLeca m ens have-been-identified-at-Shoreharc-which-could compromi-se pl=* safety. In-summary. tha following_can_be-stated-concerni Shoreham 7RA: N The Shoreham PRA is an adeqtrate evaluation of the systems interaction' issue within the current state-of-the-technology. No-predominant risk outliers have been identi-died in the Shoreham specific analysis. the plant is as safe as other licensed plant wi thi n -the -to le rance -limi-t s-of-the-ana ly a a a - __

5. Responses to Specific Statements Made in the SC/ SOC-Testimony on Contention 7B This section provides direct responses to specific statements made in the prepared testimony on behalf of Suffolk County and SCC regarding Contention 7B.

                                              -110-SAI believes that the submittal prepared for Suffolk County and Shoreham opponents coalition has some inherent mis-conceptions, inaccuracies, and false conclusions regarding the SNPS PRA and the scope of " systems interactions analysis" now under development by the NRC.           This section addresses Section VIII of the SC/ SOC Contention 7B testimony, specifically the following areas:

(1) There appears to be some misunderstanding on the part of the Intervenor consultants as to the scope and completeness of the PRA. In particular, common-alities and dependencies are directly and implicitly modeled in the PRA. Section 3 of this testimony provides examples of these. (2) The depth of analysis required to uncover every sys-tems interactions problem has not currently been

   ~T                     defined. The subject contention provides no guid-(d ance as to an acceptable level of detail.

The remainder of this section is provided in the format of responding to specific interviewer statements which appear misleading or in error. Statement by Intervenors "The results of this analysis are valid to the extent that the systems have been successfully designed as independent and redundant." (page 66). Resoonse This sentence, taken from the preliminary draft of the {} Shoreham PRA and quoted in the SC/ SOC testimony, may have been misinterpreted.

                                       -111-The Shoreham plant has been constructed according to a g3   comprehensive set of regulations and regulatory guidelines.

d These regulations and guidelines are formulated in the context of single failure criterion. The PRA uses these regulations and guidelines and builds upon them to assess the impact of multiple failures due to common cause or dependent-failures. By doing this LILCO has gone beyond existing regulations to identify if any significant multiple failures may occur through systems interaction or common-mode failure. LILCO has used the PRA to examine event sequences involving multiple failures which may affect the safety of the Shoreham Plant. The PRA

    , provides the framework within which to investigate systems in-teraction and potential common-mode failures which may occur despite the design of the Shoreham Plant in a manner consistent with existing regulations.

Statement by Intervenors

               " Interactions were excluded because they have not been systematically identified at Shoreham."      (page 66).

Response

l As a review of the PRA would reveal to Intervenors, this response is false because systems interactions have been syste-matically identified in the Shoreham PRA. Also the effects of {} multiple failures have been assessed. l

                                    -112-Statement by Intervenors
            " Alternative methods exist which would supplement and improve the existing design basis /SRP approach and thus reduce the likelihood of adverse systems interactions." (page 63).

Response

The intent of LILCO has been to implement alternative methods to investigate potential systems interactions. The use of alternative approaches, both deterministic and probabilis-tic, is to supplement the existing NRC design basis /SRP ap-proach. The probabilistic approach initiated by LILCO more

 /3 than one year ago was the use of fault tree / event tree method-U ology.

Statement by Intervenors "There are . . . methods to approach the systems inter-action question which provide for greater assurance that most adverse interactions will in part be identified. Three such currently available approaches tosystems interaction are: de-pendency analysis, fault tree / event tree diagrams, and walk-downs" (page 64).

                                       -113-

Response

() LILCO has used each of these techniques in the Intervenor testimony within the PRA analysis. The following 1 references in the PRA discuss the applicability of each: Dependency analysis: Appendix B to the PRA. Fault tree / event tree diagrams: Section 3 and Appendix B to the PRA Walkdown: Section 2.4 to the PRA. Statement by Intervenors

               " Dependency analysis looks at the various ways that com-
   \I conents and systems depend upon one another.       Some examples of this approach are binary matrices, failure modes and effect analyses (FMEAs), and auxiliary safety systems commonality dia-gram" (page 64).

4

Response

LILCO and its contractors have made use of each of these techniques to establish additional information on potential systems interaction. The results of each of these investiga-i tions have been applied within the integrated framework of the {} PRA to establish an overview for analysts, management, and reg-ulators on the interaction of system / component dependencies on postuated accident sequences. Both licensing-type accidents I

                                                        -114-l                (i.e., with single failures) and potential Class 9 scenarios with multiple failures are included in these evaluation. The PRA includes the equivalent application of these techniques as indicated by the following examples:                                       '

Binary Matrices: Appendix B of the PRA describes the use of dependency matrices to identify shared systems and potential system interactions. FMEAs: The construction of system level fault trees requires the analyst to perform an FMEA. The generic component fault trees provide examples of the applica-tion these generic fault trees were used for Shoreham and are discussed in the open literature. O Commonality Diagrams: A graphical method was not used. Rather Appendix B of the PRA describes the evaluational approach used for each system. The fault tree logic models incorporate these commonalities directly and event tree sequences define the multiple system combina-tions required to operate. These logic models were evaluated as follows:

1. quantitatively for risk estimation and comparison with possible safety goals; and

2. qualitatively to establish minimum component failure

(} combinations which would lead to public risk.

                                            -115-Statement by Intervenors "Using commonality diagrams, dependency matrices and FMEAs, a reviewer can obtain a thorough understanding of the interactions possible in a power plant.            That knowledge is best then applied through the use of fault tree / event tree diagram-ming, such as used in PRA" (page 65).

Response

LILCO has used fault tree / event tree logic models to assist in identifying potential systems interactions which could lead to multiple failures during postulated accident se-quences. The thorough understanding of plant systems interac-() tions required to effectively use these techniques was obtained by the application of the suggested methods, however there are other additional factors which are significantly more important than these formalized techniques. These include:

1. Review of operating experience data (e.g., LER)

2. Use of experienced BWR fault tree analysts

3. Review of the system results by LILCO system and operating engineers 1

4. Review of conclusions by expert peer review group Statement by Intervenors The dependency analysis /PRA should be complimented by a physical survey or "walkdown" (page 67).

                                   -116-

Response

Expert group of PRA analysts accompanied by systems and operating engineers from LILCO performed two "walkdowns" of the Shoreham plant. These walkdowns were both for familiarization and to identify potential systems interaction problems which ' could result in multiple failures as initiators or during degraded conditions present in postulated accident sequences. See Section 3 for further discussion. Statement by Intervenors The best method of resolving the systems interaction issue at any given plant is a combined approach, using depe-O dency analysis /PRA, balanced by a walkdown study (page 68).

Response

LILCO has taken the initiative outside of the licensing regulations to assess potentially adverse system interactions through both of the recommended techniques, i.e.,

1. A level 3 PRA which includes dependency modeling of functional, human, and selected spatial interactions has been performed.

2. Two walkdowns of the plant have been conducted by

(} the fault tree analysts and the PRA reflects the results of these walkdowns. L I

l

                                           -117-Statement by Intervenors
   ,r3           "The EOPs direct operator to utilize equipment which has V

not been classified or qualified commensurate with the function . performed" (page 75).

Response

EOPs appropriately direct operators to use non-safety related systems and components to deal with transient events because use of these systems can be effective in reducing the need to call upon safety-related systems. These non-safety related systems are reliable systems used in the generation of power. The PRA dad nct revcal any non oafety &olated cyteme, (o whether--used--ncrmelly or in-the-EOP 's, that have an unreliabil-i-ty-tha t-would--j eep&i di a e plant s&fety.

6. Conclusion LILCO has engaged in a PRA of the Shoreham plant utiliz-ing fault tree / event tree methodology to identify and evaluate the impact of systems interactions on public safety. This intiative by LILCO surpasses the licensing requirement, but more importantly is viewed by LILCO as a check of the adequacy of the design safety of the Shoreham plant. Each of the sys-tems which might assist in the safe shutdown of the plant are s

(~'l

   '-   addressed in the FRA (both safety and non-safety).          Multiple component failures, system dependencies, and potential systems k __

                                     -118-interaction which could affect multiple systems are identified and their impact evaluated through the use of fault tree and event tree sequence quantification.              System interdependencies have been identified in the Appendices of the PRA and incorpo-rated into the SNPS PRA logic model.              Other potential formats exist for the display of these dependencies, but the Shoreham PRA logic model contains all the pertinent information to iden-tify systems interactions.        The Choreham PRA results te date                     .

h aveneve aled-no-s y s tems-i n te r ar t i o n a -wh i ch-wottid-requ i-re-rede-sLgn_or-modification--or-which-would-compromise ptant safetp O O

                                -119-7    REFERENCES

() 1. PRA Procedures Guides, A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, NUREG/CR-2300, dated September 1981.

2. Coffman, F., et al., The Development of Interim Guidance of Systems Interactions, Proceedings of the International ANS/ ENS Topical Meeting on Probabilistic.

Risk Assessment, dated September 1981.

3. Alesso, H.P., et al., on Issues Important to the Development of a Systems Interaction Evaluation Procedure, Proceedings of the International ANS/ ENS Topical Meeting on Probabilistic Risk Assessment, dated September 1981.

4. Systems Interaction and Single Failure Criterion, MHB Technical Associates, November 1980.

5. Cybulskis, P., et al., Review of Systems Interaction Methodologies, Battelle Columbus Laboratories, NUREG/CR-1896, dated January 1981.

(]}

6. Buslik, A.J., Papasoglou, I.A., Baria, R.A., Review and Evaluation of Systems Interactions Methods, Brookhaven National Laboratory, NUREG/CR-1901, dated January 1981.

7. Final Report -- Phase I Systems Interaction Methodology Applications Program, U.S. Nuclear Regulatory Commission /Sandia National Laboratory, NUREG/CR-1321, dated April 1980.

8. Kelly, J.E., et al., Systems Interaction: State of the Art Review and Methods Evaluation, Lawrence Livermore Laboratory, NUREG/CR-1859, dated January 1981.

9. Reactor Safety Study, WASH-1400, Nuclear Regulatory Commission, dated October 1975. -

10. Reactor Safety Study Methodology Application Program, S.W. Hatch, P. Cybulskis, R.O.Wooton, Nuclear Regulatory Commission NUREG/CR-1659 4 of 4, dated October 1981.

11. Lewis, H.W., et al., Risk Assessment Review Group Report to the U.S. Nuclear Regulatory Commission, NUREG/CR-0400, dated September 1978.

12. Zion Probabilistic Safety Study, 1981. *

                                -120-

3. Testimony of Robert M. Kasesak on Status and Use of Shoreham PRA

( (a) Motivation and Goals The Shoreham PRA study was a self-motivated LILCO under-taking which by its very nature attempts to evaluate plant re-sponse beyond the normal design basis. The study came about through LILCO's monitoring of the development of PRA method-ology and its desire to make use of this emerging technology to assess better the Shoreham design relative to hypothetical accident sequences and their resulting public risk. The initial goals of the Shoreham program were primarily to: (1) assess the Shoreham Emergency Plan by evaluating the Shoreham specific O response to these hypothetical accidents and'their conse-quences, (2) perform an independent design verification to ensure that the Shoreham plant design has no atypical or dis-proportionate elements within its design that dominate risk, and (3) develop reliability / risk analysis capability within LILCO. The program would afford LILCO engineers first hand training and exposure of this technique through experts trained in this field. O

                                     -121-(b) Scope and Schedule

() Simply stated, the scope of the,Shoceham risk assessment is to establish, utilizing established probabilistic risk tech-niques, the probability of occurrence at Shoreham of hypotheti-cal accident sequences and their consequences that dominate risk to the general public. As stated previously, the PRA is viewed by LILCO as a useful independent design verification tool that supplements traditional deterministic safety studies. The risk assessment is being performed in phases and is similar in scope to WASH-1400. Phase I, performed by Science Applications, Inc., involves a fault tree / event tree analysis () and concludes with the establishment of a core vulnerable fre-quency per reactor year of operation for Shoreham-specific accident sequences. In-plant consequence analyses that estab-lish the characteristics of a radioactive release specific to the Shoreham plant are performed in Phase II, also by Science Applications, Inc., utilizing the MARCH / CORRAL methodology. Ex-plant radiological consequence calculations of radioactive releases are to be performed in Phase III.A by Pickard, Lowe and Garrick, Inc. and includes the calculation of public risk with the CRACIT code. Phase III.B represents an emergency plan effectiveness study utilizing the ex-plant consequence model k- established in the risk assessment. Internal event initiators such as loss-of-coolant accidents, transients, and anticipated

                                           -122-transients without scram were considered in detail as part of the probabilistic work.         The Shoreham risk assessment includes the identification of systems and accident sequences that con-tribute to risk and a comparison of risk to WASH-1400.

All aspects of the analysis were fully documented and described to the extent necessary for independent verification (methods and results) by a Peer Review Committee established by LILCO consisting of the following experts: Drs. Norman Rasmussen (MIT), Walton Rodger (Nuclear Safety Associates), and Vojin Joksimovich (NUS). The current schedule for completion of the Shoreham pro-Q'l babilistic risk assessment contractor work is approximately the # end of September 1982. It is envisioned by LILCO that its own review of the work as well as the peer review process will con-tinue through the end of 1982. (c) Evaluation of Results

1. Management Review Once the Shoreham PRA has oeen finalized the report will l be disseminated within the LILCO Management organization. The l report will be assessed by the Nuclear Engineering Department which has acted as the Project Manager of the analysis and
I'T k/ therefore is charged with assessing its findings. After review and further analysis, Nuclear Engineering will forward its

                                                                                                                 -123-conclusions and recommendations to the Vice President-Nuclear after consultation with the Shorebam Nuclear Review Board.33/

This Report will also be assessed by the Peer Review Committee and its chairman the Vice President of Engineering, who will also present the findings of that independent body to LILCO Management. Thus two distinct pathways to senior LILCO management will be utilized in assessing the results of the Shoreham PRA and Emergency Plan Effectiveness study. One is via the Vice President-Nuclear utilizing the Nuclear Engineering Department in conjunction with the Shoreham Nuclear Review Board, and the () second is via the Vice President-Engineering using the re-sources of the independent Peer Review Committee. 33/ The Nuclear Review Board is established pursuant to , Chapter 6 of the Technical Specifications. It is composed of a group of persons with broad expertise in operations, engineer-ing reactor physics, and radiochemistry. The Board's broad O function is to ensure plant safety and to ensure that plant modifications and operational changes do not compromise safe operation of the plant. The Board will review the PRA and con-sult with Vice President-Nuclear concerning any recommendations of the Nuclear Engineering Department. I l

                                     -124-

2. Nuclear Engineering Department (NED) Review

() The role of NED has been one of Project Manager and providing technical support and overview for the PRA. NED focused the scope of the program, prepared the procurement specification and has coordinated the generation of essential technical information for analysis. This organization and its staff engineers and scientists have, by exposure to this effort and earlier training, become closely familiar with this disci-pline, and will perform a final LILCO review to ensure that the initial corporate goals have been satisfied. These goals are first that the Shoreham Plant design can be technically com-() pared (within the uncertainty in the analysis) with WASH-1400, I and second that no atypical or disproportionate " risk outliers" are identified in the Shoreham Plant Design. Once assured that the initial goals of the report are satisfied, NED will focus attention on the findings of the analysis relative to identi-fying dominant risk-contributing events for SNPS. This infor-mation will afford LILCO a special perspective on what events or systems contribute to public risk. Armed with this informa- , tion, LILCO will be better able to judge where improvements in l l safety, if any, would provide the most benefits in reducing public risk. (

                                                   -125-

3. LILCO Maintenance of the PRA h) It is LILCO's intent to maintain the event tree / fault tree models produced for Shoreham in a state that accurately represents the plant design. Plant modifications and operating experience (Reference Section V) will be fed back into the models to ensure that the fault trees and event trees effecti-vely represent current plant design.

Future significant plant modifications will be evaluated for their impact on the system level fault trees and accident sequence event trees. This process will enable NED to better understand and evaluate the impact of plant modifications on the core vulnerable frequency. It will allow for a screening of proposed changes in system reliability from a risk stand-point. In this manner the PRA tools will be used by LILCO engineers as input to the decision making process for future plant modifications. The program identified above has been defined in general terms within the charter of the Nuclear Engineering Department. NED engineers, through exposure to this program, have been trained in this methodology. Some additional specialists will be necessary to support the program identified above. The sta-ffing needs to perform this program are currently under review. J

                                   -126-               '

4. Review of Operating Experience

a. LILCO Program As identified previously in this testimony Section V(k)

LILCO has established a program for the review of Nuclear Plant Operating Experiences. This Program is under the charter of the Independent Safety Engineering Group (ISEG), and is charged with: " assessment of the operating experience of the other nuclear power plants, with particular emphasis on units of design and configuration similar to Shoreham." Procedures are under development to address LILCO review of: (1) Licensee Event Reports generated at Shoreham, (2) Significant Event Reports distributed by NOTEPAD, and (3) Significant Operating Experience Reports (SOER's) distributed through the SEE-IN Program in INPO.

b. Operating Experience Input to the Risk Management Program As already noted, LILCO intends to maintain the fault / event trees models as design tools to evaluate challenges to plant design. These challenges can take the form of poten-tial events (e.g., LER events), plant procedure changes or mod-() ifications. The approach will be to use the LER Review Program to: (1) Assess the impact or similarity of events at other plants to the plant response of Shoreham, and (2) Use the LER

\

j i

                                                                   ,         j l

t l

                                                 '1                    '
                                  -127-f k data base, in particular the Shoreham LER's, to evaluate th'e         -

input failure rate data used within the Shoreham PRA analysis. I In this manner the PRA fault / event tree models can be used to e monitor system reliability and ensure that the model is main-tained to reflect the operating characteristics of the plant. O O \ _ . - _ - . . - -

                                      -128-VII. FURTHER RESPONSE TO SPECIFIC O                 POINTS RAISED IN INTERVENORS' TESTIMONY In their testimony on SC/ SOC 7B, the Intervenors raised a number of specific points concerning systems classification and systems interaction.     .These criticisms can be divided into six general categories:

(A) the alleged failure to use emergency operating pro-cedures for classification of systems; (B) the alleged failure to properly classif'j four sys-tems; (C) che alleged failure to adequately consider systems interactions in reactor water level measurement; (D) the alleged inadequacies in the classification of (]) the standby liquid control system; 3.21-1 (E) alleged inconsistencies between FSAR Table 3,G-4 and i Regulatory Guides 1.26 and 1.29; and (F) the alleged inadequacy of the detail presented in FSAR Table 3.2.1-1. None of these criticisms has any validity as will be demonstrated below. A. Use of Emergency Operating Procedures 34/ 34/ This testimony (Part VII A) was prepared by Paul J. McGuire.

                                     -129-

1. Introduction (g My.name is Paul J. McGuire and I have been a certified G

Senior Reactor Operator (SRO) on the Dresden and Cooper Nuclear . Power Stations and I was the Plant Manager for four years at the Pilgrim Nuclear Power Station. I have studied both the Shoreham Emergency Operating Procedures (EOPs) and the testi-mony relating to EOPs submitted on Contention 7B, dated April 13, 1982, by Suffolk County and the Shoreham Opponents coalition. I am of the opinion as a reactor operator and nuclear plant manager that the classification of the struc-tures, systems and components used in the Shoreham EOPs is cor-rect and consistent with other BWRs, specifically and gener-j O ally, with which I am familiar. I also base this determination on my familiarity with the systems and components which I have personally used to control transients during the startup test program at the Cooper Nuclear Power Station. While at the Cooper Nuclear Power Station, I was a Shift Engineer through the preoperational, initial core load, and power ascension test . programs. The power ascension test program produces transients including loss of offsite power test at 100% power, to prove the plant systems are adequate to safely control the tran-sients. The Cooper Nuclear Power Station is a G.E. BWR 4 Mark (} I and the Pilgrim Nuclear Power Station is a G.E. BWR 3 Mark I and both are, in all important operaticnal aspects, essentially similar to Shoreham which is a BWR 4 Mark II.

                                   -130-In addition, subsequent to the TMI accident, as chairman of the BWR Owners' Group Systems Subgroup for the Short Term Lessons Learned phase, I was involved with reviewing the capa-bility of BWR systems, structures and components to handle abnormal events, including multiple failures. One focus of the Subgroup was to assess whether the set.of safety related sys-tems was a sufficient set of safety systems to prevent core damage, assuming certain failures in these systems. To accom-plish this objective, the Subgroup focused on each safety-related system and considered the impact of its failure on adequate core cooling. As a result of this review, the Subgroup concluded that the safety related systems were the O   only ones required for safe shutdown of the reactor, even as-suming multiple failures, but we also realized that when offsite power is not lost, the reliable equipment used for power production could just as easily meet the same objectives in certe.in instances without challenging the safety related systems. Foy example, in circumstances where vessel water level is decreasing, the level could be restored by the opera-tor taking manual control of the feedwater system to increase feed flow to the vessel. This could be accomplished by in-creasing the speed of running pumps or by starting an addi-l     tional condensate booster pump if one is not in operation. Of

(~} l course, if this is not done or if operator actions do not re-store level and the level continues to fall, the safety systems will automatically initiate at a predetermined low level. l l

                                  -131-On the basis of its review, the Subgroup recommended that the operator be provided with a more systematic approach to controlling transients. In particular, the Subgroup recom-mended development of simple, complete procedures so that the operators coulf use the full capabilities of the plant in deal-ing with problems. Prior to TMI, plant EOPs typically were event oriented rather than symptom oriented, and the majority of operator actions required were only to verify operation of safety related systems. Non-safety related systems were not typically or fully addressed in EOP's prior to the review by the Subgroup.

The essential conclusions of the subgroup were that the O emergency procedures should be symptom oriented and that they should address both safety related and non-safety related sys-tems so that operators would understand and be directed to use the full capabilities of the plant. By using non-safety re-lated systems in the proper sequence, the subgroup concluded that the objectives of the procedures would be met while mini-mizing challenges to safety related systems. Finally, the Subgroup concluded that the existing safety related systems were adequate for safe operation and safe shutdown in the event the non-safety related systems do not function and required no {} additional safety related systems. The same is true for Shoreham. Safe shutdown can also be achieved using only safety related systems in the event of failure of the normal - non-safety related systems referred to in the EOP's. L

                                                     -132-As a result of the Subgroup's recommendations, a ft11ow-on group within the BWR Owners' Group, the Subcommittee on Emergency Procedure Guidelines (Subcommittee), was formed to develop and validate emeregency procedure guidelines (EPGs) for use as guidelines by BWR plants generally. EPGs consider the full capabilities of the plant including both safety related and non-safety related equipment and are symptom or'ented i

rather than event oriented. As the EPGs are an operator's logical approach to dealing with symptoms, they typically start with normally used non-safety related systems. If failures progress in nonsafety-related equipment, the safety related equipment comes into play automatically in the EPGs. If fail-O ures of automatic safety functions occur, manual use is men-tioned. The procedures carry the operator well past the design basis into extreme multiple failure events. The EPGs generally follow the analysis carried out by the earlier Subgroup. The EPGs indicate that a plant's full capability is well beyond the design basis. LILCO has been an active participant in the BWR Owners' Group's EOP Subcommittee and the current Shoreham EOPs are con-sistent with the recommendations of the Subgroup. This back-ground shows why there is no reason to conclude, as Intervenors do, that the mere mention of a non-safety related system in an (G~3 EOP means its classification should be upgraded. Also, the i non-safety related systems may be in the EOPs for reasons J f

                                  -133-unrelated to mitigating the event, such as protecting r-} equipment. If the non-safety related system is involved in mitigating a symptom and, if it fails, a safety-related system takes over. If the non-safety related system performs prop-erly, the safety-related system is not challenged unnecessar-ily.

2. Consideration of EOPs In the prefiled testimony, Intervenors list a number of Shoreham EOPs.35/ I will address separately each EOP listed by intervenors in their prepared testimony with the intention of explaining the purpose of the non-safety related systems as

() they are used in the procedures. As will be evident, in many instances where a non-safety related system or component was identified by the Intervenors as being in an EOP, its function is to protect certain equipment or to accomplish some other purpose unrelated to mitigation of the transient or event underway. Even in those instances where the identified non-safety systems do play a role in mitigating the event or transient, the EOP reccgnizes the possibility that the non-safety system mentioned will not work by relying ultimately on a safety related system. All EOP accident or transient scenarios are bounded ultimately by a safety related system. 35/ While these EOPs have recently been revised in accordance with the EPGs, the revisions also direct the operator to use non-safety related systems and components.

                                      -134-

a. Feedwater/ Level Control System Failure SP 29.006.01 Including SP 29.010.01 " Emergency Shutdown Emergency eS Procedure)

V (1) Description of EOP and Related Event This procedure addresses the operator's action when the Feedwater/ Level Controller malfunctions, causing erratic reactor water level condition to occur in the reactor vessel. This controller instability may only cause a slight swing or it may cause a large swing. It was with these possible scenarios in mind that this EOP was written. In essence, the procedure calls initially for the operator to take manual control of the feedwater system and attempt to remove the cause of the tran-sient. O

 \~#             (2) Systems Listed By Intervenors In Exhibit 4 of Prefiled Testimony In their prefiled testimony, Intervenors included a table, Exhibit 4, that listed for several emergency operating procedures, safety related and non-safety related systems men-tiened in the EOP. For this EOP, the non-safety related sys-tems mentioned in Intervenors Exhibit 4 are the feedwater con-troller, reactor feedwater pump speed controller, low flow con-trol valves, main turbine associated valves and oil pumps and the condensate and condensate booster pumps. Condenser vacuum, as mentioned in the Emergency Shutdown Procedure in Exhibit 4, is a plant condition, not a system or component. The turbine stop valve position switch is part of the reactor protection

                               -135-system and is safety related. The function of each of these non-safety related components in the context of this EOP is discussed below and as noted previously, their function is either unrelated to mitigation of the event or even if involved in mitigation of the event, the non-safety system is not essen-tial to mitigating the event.

In the procedure, the Feedwater and Reactor Feedwater Pump Speed Controllers and Low Flow Control Valves (both non-safety related) are placed in manual control. This is done to isolate the failed controller component from the pump controls. This in itself would end the transient if the swing in water level was not too great. ( \ The Main Turbine and its associated valves (non-safety related) as mentioned in SP.29.010.01 direct the operator to trip, and verify tripped, the turbine components. This protects the turbine by isolating it from steam paths. The motor suction pump is started to provide lubrication during turbine coastdown. These actions are only to protect an expen-sive component and are not required for any mitgation. Also in SP.29.010.01, th'e condensate and condensate booster pumps are being sequentially removed from service as less flow to the reactor is required following the scram. This is done to pro-tect the pumps from a low flow condition which could cause pump damage.

                                -136-

b. Loss of Offsite~ Power (SP.29.015.01 Including SP.29.010.01 " Emergency Shutdown Emergency O Procedure," and SP.29.023.01 " Level Control")

(1) Description of EOP and Related Event In this procedure, it is assumed that all offsite power (138 KV and 69 KV power) is lost, and the diesel genera-tors start and supply power to the essential buses (101, 102 and 103). This procedure addresses the operator's action under these conditions. (2) Systems Listed by Intervenors in Exhibit 4 of Prefiled Testimony __ Intervenors' Exhibit 4 states that the following systems are mentioned in the EOP: reactor recirculation MG set lube oil pumps, TBCLCWS, main turbine associated valves, oil pumps and turning gear system, and condenser vacuum.36/ None of these structures, systems or components plays any role in the mitigation of the event. Operator actions described are to protect the main turbine and the reactor recirculation M-G sets (Subsequent Actions), but do not impact the transient. The starting of the TBCLCWS occurs after power has been restored and therefore is not required for the transient. As stated previously, conden-ser vacuum is not a component. O 36/ SC/ SOC Exhibit 4 indicates on page 4 for the Loss of Offsite Power EOP, a component identified as "cond. (?) water & booter (sic) pump trip." This does not make sense.

                                      -137-In effect, reference to the non-safety related com-ponents in this procedure is a reminder to the operator that
 \

after the transient is stabilized, actions to protect non-safety related equipment are appropriate.

c. Loss of Condenser Vacuum (SP 29.012.01) Including Emergency Shutdown Procedure (SP 29.010.01)

(1) Description of EOP and Related Event In this procedure, it is assumed that the condenser vac-uum is lost either due to loss of the circulating water system or loss of sealing steam to the main turbine. Loss of vacuum will cause the reactor to isolate from the condenser (due to loss of the condenser as a heat sink). The main turbine and reactor feedpump turbines will trip. (2) Systems Listed by Intervenors in Exhibit 4 of Prefiled Testimony The non-safety related equipment mentioned in Intevenors' Exhibit 4 is main turbine associated valves and oil pumps, condensate pumps, and condensate booscer pumps. As in the previous discussions, this equipment is only used after the transient is stabilized and not to stabilize the transient. O l

                                        -138-i                d. Transient With Failure to Scram (SP.29.024.01) and Emergency Use of S.L.C. (SP.29.004.01) Emergency Procedures f-)s (1) Description of EOP and Related Event In this procedure, it is assumed that the Reactor Protection System trips and all cr some of the control rods fail to scram fully. The procedures outline the various steps the operator should attempt including injection of Standby Liquid Control. The Reactor Water Cleanup System (RWCU) will isolate from the reactor via its safety related isolation valves to prevent any boron removal by the RWCU system.

(2) Systems Listed by Intervenors in Exhibit 4 of Prefiled Testimony () In Intervenors' Exhibit 4, the procedure mentions the use of the RWCU system, FW flow (pumps), and turbine bypass valves. RWCU is mentioned in these procedures only to verify it has isolated, for the reason noted above, or to isolate it , manually if the automatic isolation did not occur. All compo-nents associated with that isolation are safety related. The feedwater pumps are not specifically mentioned in the proce-dure. The turbine bypass valves are used after shutdown is achieved to control pressure. This mode of operation is one of three listed alternatives. O

i

                                           -139-                            I

e. Loss of Coolant (SP 29.014.02, and SP 29.014.01)

Emergency Procedures Including Emergency Shutdown Emergency Proceudre (SP.29.010.01) (1) Description of EOP and Related Event In these procedures, it is assumed that the plant suffers a loss of coolant accident (either a large or interme-diate pipe break), and emergency core cooling systems operate to insure adequate core cooling. (2) Systems Listed by Intervenors in Exhibit 4 of Prefiled Testimony Intervenors's Exhibit 4 identifies main turbine associated valves and oil pumps, feedwater pumps, and the ADS timer. The ADS timer is part of the safety related ADS and is () itself safety related. The feed pumps and turbine components appear in the shutdown procedure, but for the reasons previ-ously addressed, they have no purpose in mitigating the LOCA. I 3. CONCLUSION As shown, many of the non-safety related systems men-tioned in the Shoreham EOPs play no role whatever in mitigating the event, but are included in the procedures for such purposes as equipment protection. And in those instances where a non-l safety related system in the EOP can play a role in mitigating i I , the event, there is a safety-related system capable of prevent-i (} ing core damage in the event the non-safety related systems fail. The inclusion of these non-safety related systems in EOPs is based on the principle that operators should be

                                         -140-directed to use all available systems and the use of the normal, non-safety related systems in these EOPs minimizes challenges to the safety related systems.         In summary, there is no reason to contend, as Intervenors do, that a non-safety re-lated system should be upgraded because it is mentioned in an EOP.

I would also like to make the point that many non-safety related systems referred to in the EOPs are impor-tant to the reliable operation of the plant. These non-safety related systems used for the generation of power are normally in continuous operation, demonstrating their reliability.

   , Further, these systems are part of the preventive maintenance
  '> program, their history is trended for performance, and those components that do not perform will be modified or replaced.

In addition, there are codes and standards to which these sys-tems must be built. The testimony supplied by Suffolk County and the Shoreham opponents Coalition implies that there are no standards except for safety related equipment. Based on my experience, this is not true at all. B. Alleged Improper Classification of Four Systems - The SC/ SOC testimony discussed a number of systems which play roles in responding to transient and accident conditions that are allegedly either not shown on LILCO's classification table or not properly classified. Principal among these are i

                                                        -141-four systems which are relied upon to mitigate events in the FSAR Chapter 15 analysis but are not fully safety grade.                         These systems are the Rod Block Monitor, the Reactor Core Isolation Cooling (RCIC) System, the Level 8 Trip and the Turbine Bypass System.                      None of these systems is relied upon for accident mitigation and, as will be shown, each has safety features com-mensurate with the safety function it provides.

1. Rod Block Monitor The rod block function 37/ is designed to prohibit erron-eous withdrawal of a control rod so that local fuel damage does not occur. The rod block monitor (RBM) will initiate a rod block signal to the rod drive control system to stop drive motion and prevent local fuel damage during the worst single rod withdrawal error 3g/ starting from any permitted power and flow condition. Even if local fuel damage did occur, however, it would pone no significant threat of radioactive release.

The RBM's principal objective is to further fuel life by res-tricting rod movement to within defined limits so as to mini- . mise local flux peaking. 37/ The rod block function involves three systems, LPRM, Rod i Block Monitor and Reactor Manual Control System. C.J 3R/ Contrary to the testimony filed by Suffolk County, the Rod Block Monitor does not mitigate the control rod drop or any other accident. 4 w ,,. ,

                               -142--

Although the rod block function does not meet all the requirements of a safety system, it is designed to meet most design principles of safety systems. The RBM input signal from the LPRM is designed to full safety system criteria. The sys-tem is redundant in that two channels of information must agree before rod motion is permitted (only one of the RBM channels is required to trip to prevent rod motion). The system has self-monitoring features with provisions to check the self-monitoring. Loss of power to the RBM will cause a rod block. The following features are included in the RBM design:

a. Redundant, separate, and isolated RBM channels.

b. Redundant, separate, isolated rod selection informa-tion, including isolated contacts for each rod selection push button are provided directly to each RBM channel.

c. Separate, isolated LPRM amplifier signal information is provided to each RBM channel.

d. Separate and electrically isolated recirculation flow inputs are provided to the RBM's for trip reference sig-nals.

e. Independent, separate, isolated Average Power Range Monitor reference signals are provided each RBM channel.

f. Independent, isolated RBM level readouts and status displays are provided from the RBM channels.

g. There is a mechanical barrier between channel A and channel B of the manual bypass switch.

h. Independent, separate, isolated rod block signals Os are provided from the RBM channels to the manual control system circuitry.

                                     -143-During the rod block function, the Reactor Manual g-  Control System receives a signal from the Rod Block Monitor and V

initiates the signal to stop further control rod drive motion, thus terminating the event. Although the Reactor Manual Control System is not designed to full safety system standards, it has been proven to be a high quality system as evidenced by . successful operating experience. The system also has a self-test feature. In addition to the high quality of the RBM design, tech-nical specification surveillance will be incorporated at Shoreham to further assure rod block function operability. A copy of this technical specification is part of Attachment 8. The RBM has a limited safety function in that it prevents local fael damage in the event of a worst-case rod withdrawal error. It is not needed to mitigate any accident. Commensurate with this safety function, the system has many design features cou-pled with technical specificatian surveillance to ensure that the system is reliable and performs properly.

2. RCIC System The RCIC system is a high pressure system which provides core cooling during reactor shutdown by pumping makeup water into the reactor vessel in case of a loss of flow from the main feedwater system. It can also supplement the HPCI system by

(^)5 u providing coolant makeup at high pressure conditions. t i _

                                   -144-As shown in FSAR Table 3.2.1-1, almost all of the RCIC g system is classified as safety related. The only significant NJ area in which the system is not safety related is in its con-trol and instrumentation. Even there, many aspects are safety i

related. The system components which provide the safety func-tionc of detecting low level and injecting water into the ves-sel are qualified for safety related operations. The safety functions of the control and instrumentation are also designed in accordance with safety system criteria. Moreover, the RCIC system is separated in a completely different electrical divi-sion from the HPCI system. gg 7,/,g,,[ The unqu21ificd components of the RCIC, include the baro-O metric condensor whose failure would not preclude systems oper-ation and four control room indicators whose failure would not impact the automatic operation of RCIC. The only other aspect of the RCIC design which does not meet full safety criteria is the single channel high level trip which prevents overfill of the reactor vessel. This does not affect the operation of the safety function of the system. Although the RCIC system is not relied upon to mitigate accidents, it can perf:rm safety functions. The RCIC design and classification is very nearly safety grade so as to assure the capability of the system to perform reliably these func- [} tions. Thus, the system is classified and designed commensur-ate with the safety functions it performs.

                                -145-

3. High Water Level (Level 8) Trip of Main Turbine and Feedwater Pumps

( The level 8 trip signal is assumed to operate in the Chapter 15 transient analysis in the feedwater controller failure event. During this transient, it is assumed that the feedwater controller loses its function and erroneously ini-tiates a maximum feedwater flow. The higher feedwater flow increases the reactor water level. Neutron power will also increase and settle at a higher level but will remain below scram initiation point. Water level would then continue to rise until it reached the high water level (level 8) trip set-point. Normally, the level 8 signal would trip the turbine and () shut down the feedwater pumps to terminate the disturbance. Should the level 8 trip fail, there would be a delay in the trip of the turbine until either manual operator action is taken or until wet steam begins to enter the turbine producing a trip on increased vibration. Analyses show that the effect of level 8 failure does not have a significant impact on the transient event severity. The level 8 trip on the feedwater system is a high qual-ity designed and manufactured system having significant toler-ance to single failures. There are 3 trip channels with inde-pendent power supplies, two on battery busses and one on a 120 VAC instrument bus, so that any single electrical failure is tolerated without any effect on system functions. The ber

                                       -146-vessel water level differential pressure transmitters and other instrumentation and control components associated with the level 8 feedwater pump trip, though not classified safety re-lated, are identical in design and manufacture to the fully safety related components associated with the ECCS and RPS low vessel water level trips.

redund%nt The high quality and divercity features of the level 8 trip combined with the minimal consequences of its postulated failure on transient severity demonstrate the adequacy of its design.

4. Turbine Bypass System The turbine bypass system is used during normal startup and shutdown to pass partial steam flow to the conden-ser. The turbine bypass valves also operate automatically fol-lowing a turbine trip or load rejection.

Following a turbine trip or a generator load rejec-tion, the turbine stop valves or the turbine control valves will close immediately to stop the steam flow to the turbine. The accumulation of steam in the vessel pressurizes the reac-tor. The turbine bypass valves are designed to open automati-cally under such conditions in order to reduce the pressuriza-tion rate by directing some steam (25% of full power) to the condenser. Should the bypass valves fail to open, reactor ves-t (O'T

,       sel pressure would be somewhat higher and the transient impact on the fuel would be increased. Analysis at full power

                                      ~147-conditions shows, however, that bypass failure would increase the change in Critical Power Ratio (CPR), an index relating to the reactor fuel heat transfer capability, by less than 0.08.

The overall effect is a slight reduction of the fuel heat transfer capability. However, the majority of the fuel is still maintained well above the CPR limit criteria. The re-sulting dose effect (if any) does not approach a small fraction of the 10 CFR 100 criteria. The turbine bypass system is described in Section 10.4.4 of the FSAR. As discussed, it consists of two steam lines from the main steam header to the bypass valve chest, four bypass valves, and four steam leads to the condenser, each including a A i kJ pressure reducer at the condenser connection. The bypass valves are controlled by the turbine generator electro-hydraulic control (EHC) system. The power supply to the con-trol system is from 120 VAC uninterruptable instrument and con-trol power for high reliability and plant availability. This power source, although not safety related, is available follow-ing loss of offsite power. In addition, an alternate power source is provided from a shaft driven permanent magnet genera-tor supplied with the main turbine. The steam lines up to, but not including, the turbine {} bypass valves are Quality Group B, QA Category I, Seismic Category I (Table 3.2.1-1, item XXXI.3). The turbine bypass valves are Quality Group D, QA Category II, Seismic Category NA

I 1 i I

                                   -148-                                l (Table 3.2.1-1,  Item XXXI.5). The turbine bypass valves are, however, subject to the extensive quality assurance program of O the supplier, General Electric, Large Steam Turbine Generator.

(GE-LSTG). This program is documented in GE-LSTG publication GEZ-4982A, " General Electric Large Steam Turbine Generator Quality Assurance Program." The EHC system is also subject to GEZ-4982A. The bypass system piping downstream of the bypass valves is not safety related. It is designed, inspected and 1 tested in accordance with ANSI B31.1. This design is in compliance with Regulatory Guide 1.26 Revision 1. It also complies with Regulatory Guide 1.26 Revision 3, including footnote 5. The NRC Staff, in Appendix A to SRP 3.2.2 (Attachment 7), has presented its position with respect to main steam components for BWR plants such as Shoreham. The Shoreham turbine bypass system, as described above, complies with the Branch Technical Position incorporated in the SRP. GEZ-4982A describes the practices and procedures em-ployed in design, manufacture, procurement and testing of the turbine generator system to ensure products of high quality and reliability. As stated by GE-LSTG in GEZ-4982A, the standards and procedures used were based on sound, documented analysis and test results, along with substantial experience, and are more suitable to the specialized components of this system than existing codes and standards for products intended for more

                                    -149-general service. The program outlined in GEZ-4982A includes such measures as detailed design procedures, material certifi-(~)!

s-cation, subvendor inspection, in process quality control, - audits, and record keeping. The program also includes noncon-formance documentation and engineering disposition. The turbine bypass system was field erected under the supervision of GE-LSTG, received quality control under the Shoreham Construction Site Inspection Program, and is subjected to a preoperational test program as oppea.d to acceptance tests. The use of preoperational testing rather than acceptance testing is indicative of the additional treatment given the turbine bypass system in recognition of its function even though it is not safety related. The bypass system is also subjected to the startup test program. The testing philosophy and procedure for Shoreham as well as specific tests involving the turbine bypass system, are summarized in Chapter 14 of the FSAR. In addition to careful design, procurement, installa-tion, and testing of the turbine bypass system, plant operation is subject to operability of the turbine bypass system by Technical Specification 3.7.10. (Attachment 8). This exten- , system sive treatment of the turbine bypass valve is commensurate with (} its intended functions in startups, shutdowns and transients.

                                     -150-C. Consideration of Systems Interactions For Reactor Water Level Instrumentation

1. Introduction The SC/ SOC testimony cited the reactor water level system as an example of a safety. system that could be adversely affected by the failure of a non-safety related system. It also went into other alleged defects (such as lack of diver-sity) in the reactor water level system that were not explic-itly tied to systems interaction. This part of the testimony will address in detail the reactor water level system to show that the design of the system is adequate and that systems in-teractions of .ne sort raised by SC/ SOC will not jeopardize

() plant safety. Figure 1 (Attachment 9) illustrates one of the two sets of cold reference leg reactor water level measurement instru-mentation provided at Shoreham. Reactor vessel water level is measured by differential pressure transmitters which measure the difference in static head between two columns of water. I One column is a " cold" (ambient temperature) reference leg out-side the reactor vessel; the other is the reactor water inside the reactor vessel and the variable leg. The measured differ-ential pressure is a function of reactor water level. The cold reference leg is filled and maintained full of (} condensate by a condensing chamber at its top which continu-ously condenses reactor steam and drains excess condensate back l l l

                                   -151-to the reactor vessel through the upper level tap connection to the condensing chamber. The upper vessel level tap connection is located in the steam =one above the normal water level          -

inside the vessel. Thus, the reference leg presents a constant reference static head of water on the high pressure tap of the d/p transmitter. The low-pressure tap of the transmitter is piped to a lower-level tap on the reactor vessel which is lo-cated in the water zone below the normal water lavel in the vessel. The low-pressure side of the transmitter thus senses the static head of water / steam inside the vessel above the lower vessel level tap. This head variee as a function of re-actor water level above the tap and is the " variable leg" in the differential pressure measured by the transmitter. Lower taps for various instruments are located at various levels in the vessel water zone to accommodate both narrow and wide-range level measurements (see Figure 2, Attachmen* 9). Reactor level indicators and recorders are shown on Figure 3 (Attachment 9). This figure also shows the condensing chamber. Shoreham level instrumentation, including elevations and setpoints, are shown in Figure 4 (Attachment 9).

2. Effect of High Drywell Temperature on Reactor Water Level Measurement Instrumentation O

High drywell temperatures may result from various events such as loss of drywell coolers, a loss of coolant

                                         -152-accidents (LOCA) or an anticipated transient without scram.

The high drywell temperatures resulting from a LOCA are higher than drywell temperatures produced by the other events. The highest drywell temperatures are produced by small (e.g., .01 sq. ft.) and intermediate (e.g., .04 sq. ft.) break accidents resulcun3 rw drywell that discharge steam into the drywell (at temperatures 3 as high as 340 F) for an extended time period resulting in substantial heatup of components / air in the drywell (including reactor water level sensing lines). Thus, the high drywell temperatures resulting from loss of drywell coolers and associated effects are bounded by those caused by a LOCA.

    ,           It should also be noted that although the drywell 4   '    coolers are non-safety grade, their operation is closely moni-tored to ensure proper drywell temperature.      Drywell air tem-perature is maintained during all normal plant operations by two unit coolers, each with four cooling coils and fans.         The reactor building closed loop cooling water (RBCLCW) system is the cooling medium for the cooling coils.      Although the drywell air cooling system is not safety related, the fans, dampers and valves receive power from emergency power supplies to provide continued operation following a loss of offsite power with no            ,

accident signal present. The system is automatically shut down and isolated on an accident signal.

  /
      }

O

                                                                                     -153-Drywell air cooling system performance is monitored in the main control room.                                                    Alarms are provided for a number of parameters, including various area and exhaust high tempera-                                                                          -

tures, RBCLCW return high temperature, and unit cooler high supply air temperature. In addition, primary containment air temperature is monitored by temperature instruments located throughout the drywell. Shoreham proposed Technical Specification 3.6.1.7 (Attachment 8) requires initiation of plant shutdown if the containment average air temperature can-not be reduced to below 135 F within 8 hours. The proposed Technical Specifications have been submitted to the NRC (LILCO letter SNRC-665, Smith to Denton, February 1, 1982).

3. Water Level Measurement Error - Reactor Pressurized When the reactor is at rated pressure / temperature conditions, the error in narrow and wide-range reactor water level indicator recorders and water level safety trip due to high drywell temperature is not significant. Specifically, there would be less than one inch of error due to high drywell ,

temperature in water level instrumentation connected to one of the reference columns and less than 6 inches of error in water level instrumentation connected to the other (second) reference column. The magnitude of the error is dependent on the differ- []) ence in the vertical change in elevation (drops) of the refer-ence leg vs. variable leg water level sensing lines in the

I

                                                  -154-drywell. Equal vertical drops in the drywell of the reference and variable leg sensing lines do not result in water level O             measurement errors because the change in density of water in the level sensing lines caused by high drywell temperature pro-duce equal changes in pressure on both sides of the differen-tial pressure water level sensor (see Figure 1, Attachment 9).

Conversely, if the vertical drops are not equal, a level mea-surement error will result due to increases in drywell tempera-ture because of the unequal affects on the two sides of the water level sensor. The maximum error is equal to 13% of the difference between the vertical drop of the reference and vari-able leg (i.e., on Shoreham 13% of 5.4 inches or .7 inches on one set of level sensing lines and 13% of 44.76 inches or 5.82 inches of error on the second set of level sensing lines).

4. Water Level Measurement Errors - After Reactor Depressurization When the reactor is depressurized, a limited number l of events can result in the drywell temperature higner than the
reactor temperature. If the drywell temperature exceeds the

[ reactor temperature, some of the water in the water level sens-ing lines can be lost due to flashing and boil-aff. Loss of water from the water level sensing lines can result in tempo-rary erratic water level indication snd an indicated water level that is higher than the actual water level.

                                  -155-Specifically, loss of drywell coolers and steam line breaks can result in the drywell temperature exceeding.the re-actor temperature. Steam line breaks can produce drywell tem-               .

peratures as high as 340 F. High drywell temperatures crused by loss of drywell coolers are less than 340*F and hence steam line breaks produce the highest (bounding) potential drywell temperatures. General Electric has conservatively evaluated loss of drywell coolers and many steam break accidents and has deter-mined that immediately following reactor depressurization, for the worst case scenario, flashing may result in a loss of up to' 20% of the water in the portion of the sensing lines located!.in', the drywell. Water in the variable leg sensing line will be replenished by drain back from the reactor. However, there is no similar mechanism to replenish the water lost from the ref-erence leg. Loss of water from the reference leg will result in an indicated water level that is higher than actual. A loss of 20% of...the water in the portion of the reference leg located in the drywell would result in a maximum of 1.9 feet of error in the indicated water level. This error does not pose a safety concern because operating instructions require the oper-ator to maintain the water level in the normal indicated range which is 16 feet above the top of the active fuel (TAF). (} If flashing or conditions for potential flashing occur, the plant operating procedures require the operator to initiate -

                                                                                                                                            -156-the automatic depressurization system (ADS) and flood the reactor with the low pressure ECCS.                                        Flooding the reactor will replenish the water lost from the reference legs and hence re-store proper water level measurement.                                                   The operator would be alerted to the flashing conditions by erratic level indica-tions/ recordings and reactor temperature approaching drywell temperature.

Therefore, if the operators follow plant operating pro-cedures, the error in reactor water level measurement due to high drywell temperatures will not exceed 1.9 feet. As indi-cated above, this magnitude of error does not pose any safety concerns. Plant operating instr' actions also require the operators to initiate drywell sprays if the drywell temperature exceeds prescribed limits. If drywell spray initiation occurs prior to reactor depressurization, the spray will cool the water level sensing lines thus preventing flashing of water in the lines and the associated water level measurement error. In the unlikely event the operators fail to follow plant operating procedures provided to refill the reference legs by flooding the reactor and further assuming the operators do not initiate the drywell spray system, water could continue to be lost from the reference leg by boil-off by persisting high dry-well temperature conditions. Conservative evaluations by General Electric based on worse case assumptions conclude that

                                                -157-all of the water in the portion of the reference leg located in the drywell could be lost due to the combination of flashing and boil-off in more than ten hours. This would result in a     .

maximum water level measurement error of approximately 9 feet in instrumentation connected to one of the reference legs. The other leg would have a smaller error. Thus, the maximum poten-tial water level measurement error is a slowly developing con-dition that provides substantial time (several hours) for the operators to recognize and correct the problem. In addition, even if all of the water in the reference leg were lost due to flashing / boil-off, it would.not be a con-cern if the operator follows the plant operating procedures and maintains reactor level in the normal indicated water level range. This is because water level would then be maintained approximately 7 feet above the TAF. Even in this highly unlikely event in which the water level measurement error in the instrumentation connected to one l of the reference legs is approximately 9 feet, operators would receive a low water level alarm at least 54 minutes prior to ! initial core uncovery from the other reference leg. If the operators disregard this alarm they would receive a second alarm approximately 10 minutes prior to initial core uncovery i {} from the instrumentation with the 9 foot error.39/ 39/ During cross-examination, the Intervenors raised a concern ( about a break in the water level instrumentation reference let. footnote continued l

                                                            -158-

5. CONCLUSION 1

( Considering the limited number of events that can cause the phenomenon, the number of operator errors that must be made and the conservative analysis assumptions described above, the probability of loss of drywell coolers resulting in reference leg flashing / boil-off and core uncovery at Shoreham is ex-tremely unlikely. The Shoreham design, therefore, is adequate to deal with the interactive effect of the non-safety grade drywell coolers on plant safety. O footnote continued First, the issue has nothing to do with classification since the reference leg piping is already classified as safety re-lated. Furthermore, GE has studied the issue and concluded that the accident is bounded by the DBA's already analyzed in Chapter 15 of the FSAR. This conclusion is supported by the NRC. An NRC (NRR) memorandum from Harold R. Denton to Carlyle Michelson, dated October 30, 1981 in response to a draft report on this subject states in part Our initial response to the subject preliminary report was to ascertain if the report raises a question of immediate safety concern. We confirmed that an instrument sensing line malfunction could be the initiating event for adverse control system action and simultaneously affect a limited number of protective system () channels. However, the unaffected protective channels are sufficient to provide all protective functions. On this basis, we determined that the concern raised in the report does not require any immediate licensing action.

                                                -159-D. Classification of the Shoreham Standby Liquid Control System

() The SC/ SOC testimony cited the standby liquid control (SLC) system as an example of a system having classification deficiencies. This part of the testimony shows that the SLC system is classified commensurate with its safety function. The Standby Liquid Control System is a diverse, backup reactivity control system capable of shutting the reactor down l from rated power operation to the cold condition in the ex-4 tremely unlikely event that not enough control rods could be inserted into the reactor core to shut down the reactor. The operation of this system is manually initiated from the control ]() room. The SLCS is a "special safety system" which means it 1 will perform a safety function to the degree specified. When needed, the system will be manually initiated to shut down the reactor and keep the reactor suberitical as it cools. The sys-j tem was not designed to perform the safety function of fast j j scram of the reactor or control of fast reactivity transients. Those safety functions are performed by the reactor protection system working in conjunction with the control rod drive system to provide reactor scram. 4 Because the SLC system only provides a backup reactivity control function, it is not a full safety system. The system nonetheless has been designed to high standards. The SLC i I

     ...,..-r .
                                              ,   - - - _,   - . - . - , , ~               m y --

                                   -160-system provides redundant loops of safety grade active

(~) equipment necessary for boron injection. The redundant loops V are powered by separate power sources capable of being connec-ted to the standby AC power for operation during a station power failure. All of the SLC equipment essential for inject-ing boron solution into the reactor is safety grade equipment. Non-essential equipment, such as test loop, drain and flush lines and SLC tank heater system is not safety grade. The test loop, drain and flush lines are isolated from the main loops by safety grade isolation valves to assure integrity of the main loops. The tank heater system, although not safety grade, is highly reliable. It consists of redundant heaters -- one auto-matically controlled by the tank temperature monitoring system, the other a larger manual heater to back up the automatic hea-ter. Moreover, the heaters are used as a back up to the ambi-ent heat in the reactor building, which will generally be at least 70 degrees Farenheit in the SLCS area during reactor op-eration. Even with failure of all heaters and the ambient heat source, the tank solution will not precipitate immediately. Instruments are provided to monitor the tank solution tempera-ture. Should the temperature drop below a pre-set temperature (11 degrees Farenheit above the maximum saturation tempera-() ture), an alarm would occur alerting the operator to take cor-rective action. The tank solution contents, concentration, and temperature are monitored at least once per 24 hours in

                                               -161-l accordance with Shoreham proposed Technical Specifications 4.1.5.            (See Attachment 8)    The storage tank is also equipped with level sensors and monitors.                                           .

These design features demonstrate that the Standby Liquid Control System is designed and qualified to standards commensurate with its intended function. E. Alleged Inconsistencies Between Table 3.2.1-1 and Regulatory Guide 1.26 and 1.29 (SC/ SOC Table, page 26)

1. Quality Group D On page 24 of their testimony on Contention 7B, SC/ SOC claim that LILCO has been inconsistent with Regulatory Guide 1.26 in the use of Quality Group D. This allegation appears to be based on the language contained in part B of the Guide. LILCO's use of Quality Group D is, in fact, consistent with past and current interpretation and use of Regulatory Guide 1.26. The NRC Staff in the Shoreham Safety Evaluation Report (NUREG 0420) stated:

We have reviewed the classification of the pressure retaining components of those fluid systems identified l in Tables 3.2.1-1 and 3.2.1-3 of the Shoreham final Safety Analysis Report and on the system piping and in-strumentation diagrams in Sections 5, 6, 9, 10 and 11 of the Shoreham Final Safety Analysis Report and we find l them to be acceptable. The basis for acceptance in our review has been the con-formance of the pressure-retaining components of fluid N systems important to safety with general design criter-i {s / ion 1, the requirements of the codes specified in Section 50.55a of the 10 CFR Part 50, and to Regulatory Guide 1.26.

                                    -162-Regulatory Guide 1.26, Revision 1 (in paragraph C.3)

() describes Quality Group D components as those which are not Quality Group A, B or C, but part of systems that contain or may contain radioactive material. Table 1 of the regulatory guide indicates quality standards for Quality Group D to be general industry standardd, as opposed to ASME Boiler and Pressure Vessel Code, Section III, the recognized code for safety related pressure retaining components at the time Revision 1 to the guide was issued. This would not be consis-tent with Quality Group D being a safety related category. Further, a comparison of paragraph C.3 of the Guide to para-(} graphs C.1 and C.2 indicates that the components addressed within Quality Group D are not those necessary for safety re-lated functions. The NRC, in its latest Standard Review Plan (NUREG-0800, Revision 1, July 1981), addresses Quality Group classification in SRP 3.2.2. The SRP states that "(a)n applicant may use the NRC Group Classification system identified in Regulatory Guide 1.26 or, alternately, the corresponding ANS classification sys-tem of safety classes which can be cross-referenced with the classification groups in Regulatory Guide 1.26. " ANS-22,

     " Nuclear Safety Criteria for the Design of Stationary Boiling

, s-p/ Water Reactor Plants," (and ANSI /ANS-52.1 which superseded it) identifies a classification system with three Safety Classes l l

                                 -163-(SC-1, SC-2, SC-3) and Other, Systems, Structures and Components (OSSC). Comparison of ANS-22 and Regulatory Guide 1.26 shows that SC-1, 2 and 3 (within the scope of fluid sys-      -

tems containing water, steam or radioactive materials) cor-respond to Quality Groups A, B and C. Similarly, OSSC cor-responds to Quality Group D. LILCO trectment of Quality Group D, specifically with respect to Quality Assurance and Seismic Categories, is consistent with ANS-22. Cross-referencing of this system to Regulatory Guide 1.26 by the NRC further con-firms that Quality Group D is .4ot a safety-related classifica-tion. {} The SRP further indicates Quality Group D is not applied to safety related components in Appendix B to SRP 3.2.2. Figure B-1 (Attachment 7) of this appendix shows Quality Group D components to be non-seismic Category I. The use of Quality Group D by LILCO is also consistent with the use of this classification in other license applica-tions submitted to and reviewed by the NRC.40/ Neither indus-try nor NRC Staff use Quality Group D to designate safety re-lated Equipment. Thus, LILCO's classification of Quality Group (~T 40/ It is also consistently applied by LILCO to Shoreham. In Table (/ 3.2.1-1 of the FSAR, Quality Group D components are shown as QA Category II in all cases, not "in many instances" as stated in SC/ SOC testimony at page 24.

                                -164-D components as LILCO Quality Assurance Category II, seismic Category NA as described in Table 3.2.1-1 is consistent with Regulatory Guide 1.26.

2. Reactor Water Cleanup System The reactor water cleanup system (RWCU) is an auxiliary system which continuously removes water from the suction side of each recirculation pump, and from the reactor bottom head, to remove solid and dissolved impurities from the recirculated reactor coolant. Within the RWCU, the reactor water is cooled, filtered, and demineralized before being returned to the reac-tor through the feedwater system. A small portion of RWCU is part of the reactor coolant pressure boundary (RCPB), up to and including the outermost containment isolation valve in the suc-tion lines. This portion of the RWCU is safety related pursu-ant to Regulatory Guide 1.26. The remainder of the system is not part of the RCPB and can be isolated from the reactor by two automatically initiated motor-operated gate valves in the suction line and three check valves in the discharge and feed-I water lines with a remote manual-operated gate valve for l long-term leakage control. The motor-cperated isolation valves may also be manually closed utilizing only fully safety grade equipment.

f

                                        -165-RWCU classification is consistent with industry standard
   -   ANS-22 which categorizes RWCU components.outside the RCPB as U,s OSSC (other systems, structures and components) not requiring           .

10 CFR 50 Appendix B QA or seismic design capability. The ANS designation would normally be the equivalent of the Regulatory Guide 1.26 Quality Group D classification and therefore not required to be safety grade. Due to the wording of Regulatory Guide 1.26, the RWCU systems also qualifies as Quality Group C since a portion is connected to the RCPB and is capable of being isolated by two values. However, since no safety function is performed by the RWCU, the components outside the RCPB need not be safety grade, hence the minimal classification required is "C", "II", "NA". In some cases, GE and SWEC have elected to provide equipment in excess of these minimal requirements. Thus, the apparent inconsistancies in Tabli 3.2.1-1 are justified since the mini-mum standards have been met or exceeded for all RWCU compo-nent3. . This same classification approach has been consistantly applied to all BWR's.

3. QA Category vs Seismic Category On page 27 of their testimony, SC/ SOC claim that LILCO Quality Assurance Category classifications and GE seismic i

classifications are inconsistent in ESAR Table 3.2.1-1. SC/ SOC i

1

 ;                                                                      -166-i point to twenty-four instances in which, it says, LILCO classifies a component "QAC I" but GE classified it as " seismic

NA."41/ .

An analysis of the particular items referred to shows that none of the items are improperly classified. In the SC/ SOC table on page 26 of their testimony, two categories of

]                  inconsistencies between the QA and seismic classifications are cited.               Under the "I, NA" heading, there are entries for cable, firestops and waterproof doors.

The cables referenced are those from Table 3.2.1-1 with safety function; thus, they are classified as LILCO QA I. They are consistently classified as seismic NA by choice. This is an appropriate way to categorize cable since seismic qualification j is not applicable to cable as such. The proper performance of cable with safety related function under seismic conditions is not assured by the qualification of the cable, but rather by j the cable raceway (conduit, tray and support systems). Safety i l 41/ This allegation contains two basic misconceptions about the i classification scheme for Shoreham. First, Table 3.2.1-1 represents

LILCO classifications (as explained in footnotes 4 and 5 to the
table). Information was, however, provided by the supplier
(SWEC or GE) in completing the table. This relates to the second misconception that all seismic classifications were provided by GE. In all but one of the twenty-four instances l (Table 3.2.1-1, Item XXII.2), the items referred to are in the SWEC, not GE, scope of supply as shown by the second column on Table 3.2.1-1. Thus, except for the one instance, SC/ SOC are incorrect in stating the seismic classification input is from GE. Similarly, the same misconceptions exist in the SC/ SOC statement on items classified "QAC II" and " Seismic I".

l l l l i

                                     -167-related cable at Shoreham is routed in seismically qualified raceways classified as QA Category I and Seismic Category I.

For clarity, a note to this effect is being added to Table - 3.2.1-1 in the FSAR.42/ Firestops are both QA I and seismic Category I as shown in Exhibit 2 of the SC/ SOC testimony (Page 12 of 24, item XXVII.d). Thus, there is no inconsistency. Waterproof doors are listed in Table 3.2.1-1 (Page 17 of 24, item XLII.13) as Category I, seismic Category NA. These waterproof doors are flood protection doors located in the screenwell pumphouse, control building and diesel fuel pum-phouse as shown in the third column of Table 3.2.1-1. As de-scribed in FSAR Section 3.4, the waterproof doors are provided for those structures containing equipment necessary for safe shutdown of the plant during the Probably Maximum Hurrican (PMH) event. Because their function is to assure safe shutdown 42/ SC/ SOC also raised the question on cross-examination (tr. 1509) about the existence of safety related cables that are not in raceways. The cable referred to is under the reactor vessel in the Source Range, Intermediate Range and Local Power Range Monitoring systems.

This cable is of co-axial design. In order to prevent electrical-noise from interfering with the proper operation of these sytems, preoperational testing at Shoreham will include testing to assess vulnerability to electrical noise. Should these systems be susceptible to noise, modifications will be evaluated and incorporated to eliminate the concern. Possible solutions
{~)N s include the addition of conduit, physical separation and grounding techniques. In any event, the safety related function associated with these systems are fail safe in that a system failure or loss of sensor signals will institute a scram. -

Therefore, seismic qualification of the cables is not required. I

                                      -168-during a PMH event, the waterproof doors are classified QA Category I. They do not need to be classified seismic Category I because their function is not required for the design Basis Earthquake (DBE). The plant is not required to be designed for a concurrent PMH and DBE.

Under'the "II, I" subheading of the 'QAC-Seismic" column of the SC/ SOC table, there is reference to pump motors, dryers, TSC building (bidg), displays and subsystems. The pump motor reference is to the recirculation pump motors, and the dryer reference is to the steam dryers in the reactor vessel. These items are not classified "II, I," as shown in Exhibit 2 to the SC/ SOC testimony (page 3 of 24, item III.6 and page 1 of 24, item I.7.b). The pump motors are classified "I, I" and the dryers are classified "II, NA" which are not inconsistencies. The TSC building, displays and subsystems are references to the Technical Support Center (TSC) and the Safety Parameter Display System (SPDS) displays and subsystems (Table 3.2.1-1, page 14 of 24 item XXXIII.1, page 18 of 24 items XLIV.1, XLIV.3, XLVI.3 and XLVI.5). These items are classified QA Category II and seismic Category I in accordance with NUREG-0696, " Functional Criteria for Emergency Response Facilities," February 1981. The NRC Staff established design criteria for the SPDS in {} Section 5.6 of this NUREG. The criteria state that "the total SPDS need not be Class lE or meet the single failure criter-ion," but that its function of aiding the operator shall be [

                                                                        \
                                -169-provided following earthquakes. The five entries called out are classified seismic Category I for this reason and designed to operate following the DBE. The displays identified are       -

those in the control room for operator use. The subsystems identified support the control room displays, and the TSC building houses these subsystems. Since the SPDS is not Class 1E (not safety related), it is classified as LILCO QA Category II. Similarly, the TSC building is correctly classified as QA Category II. F. Level of Detail in FSAR Table 3.2.1-1 The Shoreham FSAR was written to be responsive to (} Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," Revision 1, October 1972. This was the latest revision of the guide available for use during preparation of the Shoreham FSAR. Table 3.2.1-1 is responsive to Sections 3.2.1 and 3.2.2 of this Regulatory Guide. The table summarizes those struc-tures, systems, and components designated seismic Category I in accordance with Regulatory Guide 1.29, as stated in FSAR e Section 3.2.1, and those fluid systems or portions thereof as-signed Quality Group classifications in accordance with Regulatory Guide 1.26, as stated in FSAR Section 3.2.2.43/ l 43/ SC and SOC, in their testimony at page 18, indicate that Quality Group classifications from Regulatory Guide 1.26 were applied by LILCO not only to fluid systems but to others as well. They footnote continued

                                     -170-The table was constructed to be comparable in detail to Table A of ANS-22, " Nuclear Safety Criteria for the Design of Stationary Boiling Water Reactor Plants."     The intent of Table 3.2.1-1 was to indicate the classifications applied to princi-I    pal structures, systems and components at Shoreham.      Thus, it included structures, systems and components other than those specifically addressed by Regulatory Guides 1.26 and 1.29.

However, it was not intended to be a detailed component list for Shoreham, which would be beyond the scope of an FSAR. Table 3.2.1-1 is very similar in scope and content to those tables provided in BWR operating license applications submitted to and reviewed by the NRC for comparable BWR's.44/ footnote continued inferred this from page 3.2-1 of the Shoreham FSAR as quoted on page 18 of their testimony. As can be seen from Table 3.2.1-1 of the FSAR, Quality Groups have been applied consistently and only to components of fluid systems containing water, steam, or radioactive material. The quoted portion of the FSAR only refers to fluid systems. 44/ The Intervenors testified (tr. 1783-4) that they were unable to . determine the classification of the ADS system from the FSAR. The ADS sytem is safety related in its entirety as stated in p v Section 6.3.2.3: "These systems (ECCS] are designed and constructed in accordance with Seismic Category I criteria and Quality Assurance Category I in their entirety." In addition, the ADS is summarized in Table 3.2.1-1 within the nuclear boiler system, Item II (page 1 and 2 of 25). Appropriate subportions are item 2 (vessel, air accumulators), item 10 (safety / relief valves), item 15 (electrical modules with safety function and item 16 (cable, with safety function).

l 1

                                      -171-Table 3.2.1-1 is not the document which establishes or records the detailed classification assignments for all compo-O nents of the plant.        Basic design documents such as parts       -

lists, specifications and drawings contain this level of detail. The preparation and control of these design documents as well as the classifications information they contain, are fully addressed by established, auditable procedures. The design documents are part of the permanent plant record for Shoreham. These are the documents used to control activities such as procurement and QA inspections affecting plant struc-tures, systems and components. Because Table 3.2.1-1 is only a summary of the classifi-cation of Shoreham's structures, systems and components, not every component need be included in the table. In the final i stages of the NRC's review of the FSAR, LILCO was asked to add a number of items to the table. (FSAR request 260.1.) In re- { sponse to SER Open Item 55, LILCO submitted several letters on i this issue. (LILCO letters SNRC-577, Novarro to Denton, May 1 27, 1981; SNRC-595, McCaffrey to Denton, July 16, 1981; and SNRC-596, July l'7, 1981). LILCO responded in the FSAR to Request 260.1 in Revision 23 in October 1981, and included a revised Table 3.2.1-1. The NRC documented the resolution of the issue in SER Supplement No. 1 on page 17-1, stating: { The list of items was reviewed by the techni-cal review branches to assure that safety-related items within their scope of review

                                  -172-fall under the quality assurance program controls. Differences between the staff and the applicant regarding the list have been

(-)s s, resolved to the staff's satisfaction. The list has been expanded to include safety-related items reflected in NUREG-0737,

           " Clarification of TMI Action Plan Requirements," November 1980. Therefore,_the staff has no open items concerning the quality assurance program for operations or to what the program applies.

There is no direct requirement in NUREG-0737 to develop or provide an expanded QA list. Following the TMI-2 accident, the NRC developed an Action Plan to respond to lessons learned from the accident. The NRC collected its total set of TMI-related actions in this action plan, NUREG 0660. Task I.F of NUREG-0660 included, in part, an NRC action to develop guidance r~s k-) for licensees to expand their QA lists. Task I.F, however, was not included in NUREG-0737. not Thus, Table 3.2.1-1 is, required or intended to be a de-tailed compilation of the classification of every structure, system and component at Shoreham. Rather, it is a summary of the classification of principal structures, systems and compo-nents. The table is consistent with the level of detail recom-mended in ANS-22' and is similar to tables provided for compara-ble BWR's.

                                                                     -173-
                                  ~

VIII. CONCLUSION O' This testimony has demonstrated that LILCO and its con-tractors have applied a proven, well-established and accepted methodology to the design and classification of structures, systems and components at Shoreham to ensure that the design basis of the plant properly satisfies the regulatory require-ments of the General Design Criteria 1, 2, 3, 4, 10, 13, 21, 22, 23, 24, 29, 35 and 37. The design has been developed and implemented in a disci-plined and controlled manner by GE and SWEC. The design and the design process were developed from extensive experience. ({} Shoreham and its design basis and classification of structures, systems and components is comparable to all contemporary BWR's, many of which are and have been licensed and operating. The determination of the classification of structures, systems and components was based on the experience of designers in implementing a large body of knowledge reflected and docu-mented in NRC regulations, regulatory guides and industry stan-dards. These guidance documents were themselves developed from a systematic approach to nuclear plant design and classifica-tion of structures, systems and components. Systems interac-tions were clearly considered in the design process, with the O classification system itself aiding the designers in assuring that the potential for adverse interactions was minimized in

                                       -174-the design process. Further, the design has been checked in

{} many instances with studies which investigated specific poten-tial systems interactions. The Shoreham design does include special attention to non-safety related structures, systems and components. As re-flected in this testimony, non-safety related structures, sys-tems and components are accorded quality assurance and control commensurate with the functions they perform. The inclusion of non-safety related systems in EOP's is based on the sensible principle that operators should be di-rected to use the full capabilities of the plant to deal with transients and other events because these normal, non-safety related systems are required to be highly reliable and their use in these circumstances will often make it unnecessary to call upon or challenge the safety related systems. In any event, if the normal, non-safety related systems fail to per-form when called upon to mitigate an event, the full safety related systems are still available to mitigate the event and prevent core damage. There is, therefore, no reason to upgrade a non-safety related system to safety related simply because it is mentioned in an EOP. Analysis confirms that the set of safety related systems is an adequate and sufficient set to () protect the plant from design basis accidents. The use of non-safety related systems in EOP's is an appropriate use of the plant's full capabilities in these circumstances and

i

                                                                      -175-demonstrates again that the plant's capabilities take it beyond

(~3 the design basis analysis. NJ Thus the design basis for Shoreham has been correctly and - conservatively established. In implementing the design basis, , structures, systems and components have been properly clas-sified considering their importance to safety and reliability. The items to which GDC 1, 2, 3, 4, 10, 13, 21, 22, 23, 24, 29, 35 and 37 apply are properly identified at Shoreham and the conformance by Shoreham to these GDC's is demonstrated in the FSAR and has been approved by the NRC Staff in the SER. In addition to all of the foregoing, LILCO,.on its own initiative and without regulatory or NRC Staff requirement, has commissioned a level 3 PRA for Shoreham. This PRA has thus far confirmed the safety adequacy of the plant design at Shoreham. More particularly, the Shoreham PRA indicates that no unique, adverse interactions between non-safety related and safety related systems exist at Shoreham and that the probability of core melt from the operation of Shoreham is as extremely un-likely as it is for Peach Bottom Nuclear Plant as confirmed by the WASH 1400 study. In summary, therefore, Intervenors' contentions that Shoreham employed no adequate methodology for classifying sys-tems and that systems interactions were ignored at Shoreham are {} j both entirely without basis. As this testimony shows, an ade-quate methodology has been used to classify structures, systems

                              -176-and components at Shoreham and systems interactions have been appropriately addressed at Shoreham                           i D

D

LILCO, May 25, 1982 7])5 vl' UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

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

(Shoreham Nuclear Power Station, ) Unit 1) ) TESTIMONY OF EDWARD T. BURNS, GEORGE F. DAWE, GEORGE GARABEDIAN, PIO W. IANNI, VOJIN JOKSIMOVICH, ROBERT M. KASCSAK, PAUL J. MCGUIRE, PAUL W. RIGELHAUPT and DAVID J. ROBARE FOR THE LONG ISLAND LIGHTING COMPANY REGARDING SUFFOLK COUNTY /SHOREHAM OPPONENTS COALITION CONTENTION 7B AND SHOREHAM OPPONENTS COALITION CONTENTION 19(b)

1. Professional Qualifications

2. Information Relating to ANS-22

3. FSAR Figure 7A-26

4. IEEE Correspondence

5. Attachments to the Testimony of Dr. Vojin Joksimovich

6. Table 3.1 of the Testir.ony of Dr. Edward T. Burns

7. Selected Portions of the Standard Review Plan, NUREG-0800 (Revision 1, July 1981)

8. Selected Portions of the Proposed Shoreham Technical Specifications

9. Figures Regarding Reactor Water Level Instrumentation I

9 Attachment 1 Professional Qualifications O O

PROFESSIONAL QUALIFICATIONS {T EDWARD T. BURNS Manager Reliability and Safety Engineering Science Applications, Inc. My name is Edward Burns. My business address is Suite 200, 5 Palo Alto Square, Palo Alto, California. I am employed by Science Applications Incorporated (SAI) as Manager of Reliability and Safety Engineering and have held this position since May 1980. In this capacity, I perform the engineering analysis and logic model construction for SAI's BWR PRA work and supervise the computer evaluation of these models. () I received a Bachelor of Science in Engineering Science fom the Rensselaer Polytechnic Institute, Troy, New York in 1967, and a Master of Science in Nuclear Engineering for Rensselaer in 1968. In addition, I have a Ph.D. from Rensselaer in 1971 specializing in nuclear engineering. Over a five year period (1971-1976) while at the Division of Naval Reactors, DOE, I was responsible for the design review and resolution of assembly and disassembly prob-lems for a prototype reactor core. In addition, I was respon-sible for evaluating recc~ mended mechanical and thermal-hydraulic designs for two advanced reactor cores. Also at NR, ( I was responsible for the design review of laboratory thermal-hydraulic testing to support qualification of computer design l

codes for reactor cores and the minimization of flow-induced vibrations. Subsequently, I joined SAI in 1976 and have since been involved in the following projects:

             --Lead analyst on the recently completed Limerick (BWR/4) PRA;
             --Lead analyst on the Shoreham (BWR/4) PRA;
             --Lead analyst for the probabilistic evaluation of ATWS accident sequences in support of a Utility Group peti-tion to the NRC;
             --Lead analyst on studies for EPRI dealing with nuclear power plant availability and safety;
             --Lead analyst for the data evaluation and reliability model preparation for the San Diego Gas & Electric Company co-generation plant; I have also participated in the PRA analysis of the Clinch River Reactor which included nuclear power plant fire protection system evaluation; cost-benefit assessment of the proposed GE ATWS fault tree development; and quantification for a safety analysis of the Purex fuel reprocessing plant.

I have conducted a bench mark calculation of BWR reac-tors cores during transient events, specifically a qualifica-tion of the MEKIN computer code to calculated BWR core power l ! during a turbine trip transient. fs This calculation required the d development of a consistent set of cross sections and the application of appropriate thermal hydraulic models. l

PROFESSIONAL QUALIFICATIONS O GEORGE F. DAWE Supervisor, Project Licensing Stone & Webster Engineering Corporation My name is George Dawe. My business address is 245 Summer-Street, Boston, Massachusetts 02107. I am employed by Stone & Webster Engineering Corporation (SWEC) as Supervisor of Project Licensing. I have held this position since January 1980. In this capacity, I am responsible for technical and administrative supervision of all licensing personnel assigned g~g to SWEC headquarters projects, including field assignments. My V duties include assuring project awareness of regulatory developments, assuring proper and consistent application of SWEC licensing policy, and consulting with projects and clients on solutions to licensing related issues. In 1966, I received a Bachelor of Science degree with majors in physics and mathematics, from the United States Naval Academy. Following graduation, I completed extensive training from the Naval Nuclear Power Program, including Submarine School, New London, Connecticut; Nuclear Power School, Vallejo, California; and Nuclear Power Training Unit (D1G Prototype), () West Milton, New York. The training involved nuclear power plant theory, design, operation, and maintenance including

specific training and qualification at a land-based naval prototype nuclear propulsion plant. In addition, I have par-O ticipated in Stone & Webster's Continuing Education Department courses in technical and management subjects. Prior to joining Stone & Webster, I served seven years as a commissioned officer in the U.S. Navy, attaining the rank of Lieutenant. I completed two years duty (1968-1969) aboard the USS Will Rogers, SSBN 659, in engineering and operations. My duties included direct supervision of the operation of the nuclear propulsion plant as an Engineering Officer of the Watch, and plant maintenance as Electrical and Reactor Control Division Officer. I was subsequently appointed Director, Core () Characteristics and Reactor Physics Division, U.S. Navy Nuclear Power School, Vallejo, California (1970-1973). My responsi-bilities included scheduling, developiny, upgrading and teaching technical courses as well as supervising, training and assigning division instructors. While on active duty, I suc-cessfully completed the 2-day written and oral examination and qualified for assignment as Chief Engineer on nuclear-powered vessels. I joined Stone & Webster in 1973 as an Engineer in the Licensing Group. In January 1974, I was assigned as Licensing Engineer for the Shoreham Nuclear Power Station Unit 1 (SNPS-1) under construction, and was Lead Licensing Engineer from 1976 to 1980. In this capacity, I was responsible for all licensing

k 4 p:

related activities for SNPS-1 including preparation of the Final Safety Analysis Report and Environmental Report. I have lj had additional assignments at Stone & Webster including 'coordi-I .> nation of the development of company positions for NRC Regulatory Guides and Lead Licensing Engineer for the Special.j , i Projects Group of the Operations Services Divisi'on. In my cur-i t rent position, I am responsible for licensing activities on both nuclear and non-nuclear projects. I am currently the Stone & Webster representative to, and a participating member of, the AIF Subcommittee on Backfitting Requirements. I have over 15 years experience in the nuclear power field. i I hold a certificate as Engineer-in-Training in () Massachusetts by 8-hour examination.

i s 9 -E I I i - l o l I

        >                                  j.

4

r* i PROFESSIONAL QUALIFICATIONS U GEORGE GARABEDIAN Senior Power Engineer, Power Division Stone & Webster Engineering Corporation My name is George Garabedian. My business address is 245 Summer Street, Boston, Massachusetts 02107. I am employed by Stone & Webster Engineering Corporation (SWEC) as a Senior Power E'ngineer in the Power Division and have held this posi-tion since 1974. I am presently the SWEC Project Manager for the Liquid Metal Fast Breeder Reactor (LMFBR) Large Development Plant program as well as Project Manager for SWEC LMFBR Support efforts with'the National Nuclear Corporation, Ltd. I was awarded a Bachelor of Science degree in Chemical Engineering in 1959, and a Master of Science degree in Engineering Management in 1971, both by Northeastern University. In the interim, I completed graduate courses in-nuclear and aerospace engineering at the University of California at Los Angeles and a Nuclear Power Reactor Safety t

                                                -   . 6 Seminar at MIT, in-1918.                          In 1974 and 1975, I participated in seminars devoted to Safety of Liquid Metal Fast Breeder Reactors at MIT and Northwestern University respectively.

[ Prior to joining SWEC in 1967, I was associated with the 9 hydraulics and thermodynamics group of Lockheed-California i 1

1 h $

4 0

Company's Rye Canyon Research Laboratory. I conducted studies on R&D problems associated with hypersonic vehicles, the super- , ()- sonic transport (SST) and advanced military helicopters. Earlier as a Senior Research Engineer with North American Rockwell-Atomics International Division'(AI), I was concerned with the environmental control system for the Apollo Progrc.m; nuclear auxiliary electric power systems for outer space appli-cations; R&D studies for new purification process methods with organic coolants for nuclear power plants; design engineering for the sodium reactor experiment (SRE); and design engineering for the Hallam (sodium cooled) reactor station. I joined SWEC in 1967 as an Engineer with the Nuclear Division and transferred to the Power Division as a Power Engineer in 1972. My previous project assignments include the following: Nuclear Engineer and Assi stant Project Engineer on the Shoreham Nuclear Power Station, Unit 1 (SNPS-1); Lead Nuclear Engineer for the Nine Mile Point 2 (NMP-2) Station; Task Leader for resolving moisture separator / reheater drainage problems for the Nine Mile Point (NMP-1) Station; Licensing Consultant to the Project Management Corporation (PMC) (Chicago, Illinois) and later for the Energy Research and i Development Agency at the Clinch River Project Office (Oak Ridge, Tennessee); and Project Engineer for the Santillan (} Project (northern Spain). l i

Initially, on the SNPS-1 project (March 1967-August 1971), I assisted in gaining the SWEC patent of the prsssure suppres-() sion concept for an over/under-type containment. I was respon-sible for preparation of portions of the PSAR, established basic design criteria, reviewed design layouts for equipment and sys'tems within the reactor building, coordinated potential safety and accident analysis problems with the reactor vendor (GE) and LILCO personnel, and integrated GE-APED supplied sys-t tems within the balance of plant. I attended the public hear-ings (1970-1971), assisted in preparation of direct and rebut-tal testimony, and served as consultant to the LILCO legal staff in analyzing the public hearing record. Later I was Lead Nuclear Engineer for the planned NMP-2 station, and corrected O operational problems with the NMP-1 (September 1971- September 1972). From September 1972 through 1974, as assistant Project Engineer, I was responsible for coordinating engineering efforts of the licensing, power and structural mechanics disci-plines on Shoreham when the project was restarted after a shut-down period. Next as consultant to PMC (June 1974-December 1976) on Clinch River, I assisted the utility industry in licensing efforts of preparation, review, and resolution of {} licensing-related matters for the PSAR. I coordinated efforts of three customer organizations (PMC, TVA and ERDA), three

4-reactor manufacturers (Westinghouse, GE and AI), and a Architect Engineer (Burns & Roe). I functioned as licensing () consultant to the ERDA Office of Public Safety and advised on all major engineering / licensing issues. I served as coordi-nator and spokesman with state and local officials on the socioeconomic effects of the project. I was next assigned as the Project Engineer for the Santillan (BWR) Nuclear Power Station in 1977. Site studies were conducted, a PSAR was pre-pared, and plant arrangement studies were under way when the project was indefinitely postponed. As Senior Power Engineer for the Engineering Manager of SWEC (August 1977-October 1978), I headed a task force for the Engineering Manager of SWEC reviewing then-current applications of Engineering Management Systems to projects. The goal was to develop corporate standards for future SWEC standard plant pro-jects. The task force developed a systematic work breakdown structure and detailed descriptions of SWEC services which cor-relate engineering and design efforts and documents with individual work packages. I headed another task force effort for the Engineering Manager to analyze and evaluate cost and schedule impacts on an ongoing BWR project due to changing nuclear industry regulations, codes and construction practices. i l I also coordinated a similar task force for an ongoing PWR pro-

.'~)     ject.

LJ l l l i

i 4 To date, I have 23 years of experience in nuclear-related. engineering activities. I am currently the Project Manager () responsible for SWEC LMFBR activities for both the U.K. CDFR plant and the U.S. Large LMEBR Development plant. The former requires consulting services to the National Nuclear Corporation Ltd. for the U.K. 1350 MWe Liquid Metal Fast Breeder Reactor, the latter, consulting services for the con-ceptual design phase of the US DOE 1,000-MWe LMFBR Loop-type plant. The latter program requires liaison and coordination among Bechtel, Burns & Roe, Boeing Engineering and Construction Co., five reactor manufacturers, Westinghouse, GE , AI, Combustion Engineering and Babcock & Wilcox, DOE, and EPRI. I am also responsible for preparation and coordination of the O SWEC business plan for Advanced and Noncommercial Nuclear Reactors Market Sector, including the LMEBR and other advanced fission reactor concepts. I have been a member of the ANS B31.3 (ANSI NI84) Code Committee on BWR Emergency Core Cooling and Containment Heat Removal Systems. O

PROFESSIONAL QUALIFICATIONS

!(9V                                    P. W. IANNI Manager, Nuclear Systems Performance Engineering General Electric Company

] My name is Pio W. Ianni. My business address is General Electric Company, 175 Curtner Avenue, San Jose, California and I am employed by General Electric Company. I am l currently Manager of Nuclear Systems Performance Engineering of g3 the Plant Systems section of GE's Nuclear Power Systems Engineering Department. In this position, which I have held since January 1982, I am responsible for directing the overall AJ BWR performance evaluations including the emergency cooling. Since graduating from the University of Nevada with a Bachelor of Science Degree in Mechanical Engineering in 1951, I have been employed by General Electric Company. During my first year with General Electric, I completed graduate level ! work in heat transfer, fluid dynamics and nuclear theory, and in 1957, I completed a year of graduate studies at the Oak ! Ridge School of Reactor Technology. Since late 1952, I have worked as an engineer of nuclear power reactor design in var-ious capacities. l t t

In 1952, I was employed by the Knolls Atomic Power Laboratory for a two-year period as a thermal and mechanical design engineer. During this time, I conducted analyses and bs/ tests relating to the reactor steam generator design. My major work involved evaluating the mechanical design, heat transfer, and fluid flow performance of the emergency cooling and steam generator systems. As part of my service in the Armed Forces, I also spent one year in the Army Reactors Branch where I per-formed a variety of technical evaluation related to the Army compact reactors. I have been with General Electric's Nuclear Power System Engineering Department since early 1956. Throughout my twenty-six years in this Department, I have evaluated the fluid {} flow, heat transfer, and systems pertormance of various reac-tors under design. For five years, I was a technical group leader for the initial and reload design of the Vallecitos Experimental Superheat Reactor, a steam-cooled reactor with boiling moderator. In this capacity, I was responsible for all heat transfer, fluid flow, and stress analyses and related aspects of the safety systems and safeguards evaluations, and assisted on initial startups and core reloads. In April 1965, I became a technical group leader and

manager in the BWR Design Engineering Organization. For seven years in that position, I was responsible for the design and i C: )
l 3

                            . . _ ~ . - _

safeguards analyses conducted to determine heat transfer, thermodynamic, and overall system performance of the Department's Boiling Water Reactors. During this time I was deeply involved in the development of both the ECCd and pres-sure suppression containment of the type employed in Shoreham and subsequent BWRs. In 1973 I managed the overall power plant and piping watkow layout with the containment and assisted the Architect Engineers in the Mark III containment design. From 1975 to 1980, I was Technical Program Manager for the Mark I, Mark II and Mark III containment programs in which the performance and various hydraulic and thermodynamic loads were defined from many test and analysis programs.

  ~%         During 1981 I was a Senior Program Manager in the (J

Projects Department responsible for managing the many changes being imposed on the Control and Instrumentation Department. In January 1982, I assumed my present engineering management responsibilities. I have written papers for the American Nuclear Society i and numerous technical reports in the field of nuclear power. I have been a member of the American Society of Mechanical i Engineers since 1951 and have been active on various committees including Chairman of the Nuclear Engineering Division. I am a r licensed professional mechanical engineer and a nuclear () engineer in the State of California. l , l l

PROFESSIONAL QUALIFICATIONS () VOJIN JOKSIMOVICH Member Peer Review Group - Shoreham PRA Program NUS Corporation My name is Vojin Joksimovich. My business address is Suite 250, 16885 W. Bernardo Drive, San Diego, California. I am the Manager of the San Diego Office for the.NUS Corporation. The office offers consulting services in reliability and risk assessment primarily to nuclear utilities and the Electric Power Research Institute (EPRI). I received a Dipl. Ing. in Nuclear and Electrical () Engineering from Belgrade University in 1961. Subsequently I received a Ph.D in Nuclear Engineering from Imperial College of Science and Technology which belongs to the London University. My thesis on the subject of risk assessment was one of the ear-liest in the field, if not the earliest. Prior to joining NUS in 1981, I had the following pro-fessional experience: General Atomic Company, 1972-1981 Westinghouse Electric Corporation, 1970-1972 British Atomic Power Constructions, 1965-1970 Yugoslav Energoprojekt, 1961-1965 General Atomic - Managed a 40-person Safety, Reliability and Systems Department, responsible for GA's diver-sification (non-HTGR) consulting services. Served as a

consultant on safety goals to the Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI). Among () the clients were the Department of Energy's Office of Nuclear Waste Isolation; SRC International; Royal Dutch Shell; EPRI; NSAC; American Nuclear Insurers; Pickard, Lowe and Garrick, Inc.; Sulzer; the Tennessee Valley Authority and Commonwealth Edison. Previously managed GA's Safety and Reliability Branch. Directed the 45-man-year, DOE-sponsored HTGR risk assessment study known as the Accident Initiation and Progression Analysis (AIPA). Participated-in the preparation of several Congressional testimonies. Delivered many presentations to NRC, ACRS and various state and Federal agencies. Westinghouse - As Lead Engineer for the PWR Division, was responsible for licensing 20 Westinghouse reactors by pre-paring safety analysis reports and providing responses to the AEC and ACRS. British Atomic Power Constructions - In the Control and Safety Analysis Department, was responsible for fault analysis of Dungeness "B" nuclear power station. Developed AGR tran-sient computer code (REAXITRAN), used in many control and safety studies. As group leader, was responsible for overall reactor transient analysis work. O

Energoprojekt - Served with Yugoslav consulting and engineering firm initially responsible for reactor physics.and. O

    \-  kinetics and subsequently reactor safety for various reactor types. Supervised the international team organized to select.

the site for the first Yugoslav nuclear poser plant. Various national and international professional recog-nitions are as follows: Member, Peer Review Committee, PRA Procedures Guide Member, ANS Standards Committee on Containment Isolation Chairman, GA Power Reactor Safety Committee Member, Editorial Board, Reliability Engineering Member, AIF Subcommittee on Probabilistic Risk Assessment Member, API Task Force on Risk Assessment Member, IEEE 5.4 Subcommittee on Risk Assessment Served as Chairman and Organizer for numerous PRA, ANS Nuclear Safety and IEEE Conferences ['~} o a i 1 PROFESSIONAL QUALIFICATIONS

Robert M. Kascsak Nuclear Systems Engineering Division Manager

{} ' Long Island Lignting Company 4 l My name is Robert M. Kascsak. My business adoress is Long

;                  Islano Lighting Company, 175 East Old Country hoad, Hicksv111e, 1
!                  New York.       I am currently the Nuclear Systems Engineering i                   Division Manager.        My responsibilities incluoe overseeing an i

! engineering staff organization capable of analyzing ano coordi-nating activites associated with nuclear plant design, opera-l tion, reliability and safety, including approving Architect Engineer designs and vendor designs and developing an in-house I()

support organization associated with future plant mooifica-l tions.

J I graduated from hanhattan College in 1969 with a Bachelor of Mechanical Engineering. In 1977 I received a Master of Science degree in Nuclear Engineering from Polytechnic Institute of New i York. I have completed training courses in BWR and PWh tech-nology. 1 i' 1 joineo LILCO in 1969 as an Assistant Engineer in the Mechanical and Civil Engineering Department. I workeo on var-l() ious fossil fuel power station projects in the capacity of Associate and Senior Engineer, incluoing tne Northport Power l 1

Station Unit 3 and Unit 4 mechanical engineering designs. From f July 1974 to March 1975, I serveo as LILCO Lead Mechanical Engineer for Shoreham and for the J&mesport Nuclear Power Station. In March 1975 I joined the Shoreham Project Group as cn Assistant Project Engineer, after which I assumeo the responsibilities of Project Engineer. From harch 1975 to l January 1979, I was Project Engineer for bhoreham. In this position I was responsible for the review and approval or design activities preparea by our Arenitect/ Engineer, Nuclear Steam Supply System Vendor and LILCO in-house engineering i departments. I am a registered Professional Engineer in New York State ano a member of the American Society of Mechanical Engineers. O j !O

PROFESSIONAL QUALIFICATIONS i . PAUL J.MCGUIRE Manager of Operating Plant Services United Energy Services Corporation My name is Paul J. McGuire. My business address is United Energy Services Corporation, 8235 Dunwoody Place, Atlanta, Georgia. I am currently the Manager of Operating Plant Services, and I am under contract to the Shoreham Plant Staff as a consultant. I received a Bachelor of Mechanical Engineering degree

        )   from Villanova University in 1967 and a Master of Nuclear Engineering degree from New York University in 1971.       In addi-tion, I have been certified Senior Reactor Operator on both Dresden II and the Cooper Nuclear Power Station.

Subsequent to receipt of my bachelor's degree in 1967, I was employed by the Long Island Lighting Company as an engineer in fossil plants, specifically, the E.F. Barrett Station and the Northport Station. JW 1970, Z was slstec) to be one of t he chef engineers at Shorehrm and Work'ecl on Short ha r>< fo r Bight ns on ths, In 1971, I joined the NUS Corporation as a Senior Technical Associate. In this position, I participated in the startup of Peach Bottom 2 & 3 Nuclear Power Plants.

In 1972, I joined the General Electric Company as a Startup Engineer. I was certified Senior Reactor Operator at both Dresden and Cooper Nuclear Power. Plants and participated as a Shift Engineer, Operations Superintendent and eventually Operations Manager for the startup of the Cooper Nuclear Station. Following my assignment at the Cooper Nuclear Plant, I was an operations consultant for General Electric between 1974 through 1976. I consulted at the-following nuclear plants: Tsuruga, Fuhushima, Monticello, Peach Bottom, Nine Mile Point, Quad Cities, Dresden and the Pilgrim Nuclear Stations. In 1976, I accepted the position of Plant-Manager at the Boston Edison's Pilgrim Station. I was the Plant Manager until () 1980. In 1979, I was elected to the BWR Owners' Group as the Chairman of the Systems Group to address the short term recom-mendations of the Lessons Learned Task Force-TMI. In 1980, I joined the Institute of Nuclear Power Operations on loan from Boston Edison. While at INPO, I was one of the two Department Heads for the Operations Group, as well as a Team Manager for the Evaluation and Assistance Division. I was the Team Manager on three (3) INPO Evaluations. In 1981, I joined the United Energy Service Corporation as the Manager of Operating Plant Services. I have been consult-ing at both the Waterford 3 and Palo Verde Nuclear Plants in the area of Management Service. Recently, I have been l l i

contracted by Long Island Lighting Company to assist the < Shoreham Plant Staff. iO i i .l I' i i l i I l 4 i Y f i !O 4 1 i l f I i 1 1 l t I i 1 ( ( i, l i i. 1, 1 l

PROFESSIONAL QUALIFICATIONS PAUL W. RIEGELHAUPT Assistant Engineering Manager, General Division Stone & Webster Engineering Corporation My name is Paul Riegelhaupt. My business address is 245 Summer Street, Boston, Massachusetts 02107. I am employed by Stone & Webster Engineering Corporation (SWEC) as an Assistant Engineering Manager and have held this position since 1975. In this capacity, I am presently responsible for overall direction and guidance of advanced technology activities, and Marketing Department coordination. I was awarded a Bachelor of Chemical Engineering Degree, by the College of the City of New York (now City University of New York, CUNY). I successfully completed com-pany courses in cryogenics, radiological engineering, metal-lurgy, and instrumental analysis with former employers. Prior to joining SWEC in 1966, I gained a wide variety of experience in chemical research, commercial chemical engi- ! neering, and nuclear engineering. My career began as Shift ! 8 Supervisor (194%-1950) for the Kolker Chemical Works (now l Newark Division of Diamond Alkali) where I was responsible for i quality and quantity of product of manufacture of industrial and commercial chemicals. As a Research Engineer with the U.S. l l l l l t

r 7 9 Army Chemical Corps (1950-52), I performed studies and tests related to biological and radiological warfare. With Infilco, Inc., as a Field Engineer (1952-56), I g, was involved in process development, startup, test and con- .. . struction supervision of water, sewage, and industrial waste -, treatment plants. Special interests included development of e flotation purification of white water, automated process con- "y ' + trol of water treatment plants, and production of high purity water by the ion exchange process. From 1956-1958, I was associated with Air Products and Chemicals, Inc. As Project Engineer, I was responsible for the design and acceptance testing of the prototype of a mobile :, s () liquid oxygen and nitrogen production plant. These plants were . later used in support of the Redstone and Jupiter missile pro- '-

                                                                                           <        4 grams. As Senior Project Engineer with the Pillsbury Company (1958-1959), my efforts involved development of fermentation         J 2

processes and design of pilot plant production lines and equip-ment for bakery products. . i

                                                                                             ' s' '-

My experience with nuclear technology began with GE . p Knolls Atomic Power Laboratory (1959-1966) as an Engineer. Initially, I participated in the design of the D1G Prototype ~ nuclear reactor and nuclear propulsion plant for the USS .. Bainbridge DLGN-25. Specifically, I was responsible for the I ' design of off-hull support systems, core and primary plant spe- ' cial instrumentation, and evaluation of coolant technology :- ...

                                                                                                     - A
                                                                                      .,6 4

requirements. This engineering included design of a special reactor experimental facility to determine corrosion and cor-O rosion product behavior. After construction, I participated in the precritical and initial power testing of the D1G Prototype Reactor. Later, on the SSG Prototype Reactor Program as Project Engineer, my responsibilities included chemistry con-trol, evaluation of corrosion and corrosion product behavior, explosion haza:is, and materials selection in support of design, testing and operation. Subsequently, I was involved with and responsible for all chemistry control requirements, materials selection, evaluation of radiation buildup, purifica-tion system design, and crew chemistry training and qualifica-() tion for both the D1G and S5G Prototypes. I joined SWEC in 1966 and was appointed a Nuclear Engineer. My initial assignment was on the Virginia Electric and Power Company (VEPCO) Surry Power Station Project perform-ing support systems analysis. Later as Senior Nuclear Engineer, I was responsible for engineering and construction of two 800-MW pressurized water reactors (PWR) on the Surry Project. On VEPCO North Anna Station Project (1968-1972) as Assistant Project Engineer, then Project Engineer, my responsi-bilities included engineering and construction of two 900-MW PWR units. I was Project Manager on Wisconsin Electric Power Company's Koshkonong Nucleae Plant (1972-1976), Northeast Utilities Service Company's Connecticut Yankee Atomic Power i

s Station (1973-1976), and Wisconsin Electric Power Company's Point Beach Nuclear Plant (1976-1976). Earlier in October O# 1972, I had been appointed Assistant Chief Power Engineer. In February 1975, I had been appointed to the Nuclear Plant Security Committee. From December 1975 through August 1979, I was associated with SWEC New York Operations Center as Assistant Engineering Manager / Chief Power Engineer. I have accumulated 23 years of experience involving and Il nuclear technology ad engineering, in addition to }# years in chemical engineering. My nuclear experience includes design, development, hands-on operation, and backfit of nuclear power plants, both military and commercial. () I am a Registered Professional Engineer in Massachusetts, New York, Virginia, and Wisconsin. I am a mem-ber of the American Institute of Chemical Engineers, American Chemical Society, American Nuclear Society, and the Atomic Industrial Fortm Fusion Committee. In addition, I am a member of the Industrial Advisory Committee of the Department of Nuclear Engineerng of Rensselaer Polytechnic Institute. I am also the SWEC representative on the EPRI Steam Generator Owners Group A/E Advisory Committee. i l l l

PROFESSIONAL QUALIFICATIONS O/ DAVID J. ROBARE Manager, BWR 4/5 Projects Licensing General Electric Company My name is David J. Robare. My business address is General Electric company, 175 Curtner Avenue, San Jose, California 95125. I am currently Manager, BWR/4,5 Projects Licensing including the Shoreham project. As such, I am responsible for the management of all technical support work for GE's licensing interfaces with the NRC, Utilities and () Architect Engineers. I received a Bachelor of Science degree in Electrical Engineering from the University of Massachusetts in 1963. I worked for GE as a design engineer in the Large Generator Motor Operation from 1963 to 1967 and as project manager for the Rolling Mill Drive Systems Operation from 1967 to 1970. In 1970, I became a project application engineer in GE's Nuclear Instrumentation Department. In 1972, I was appointed project manager for control and instrumentation systems at reactor sites. From 1974 to 1975, I was lead application engi-neer in the BWR Projects Department. In 1975, I joined the i](- Safety and Licensing Operation assigned to the Shoreham pro-ject. i l i l [

During.my assignments in Safety and Licensing,-.I,have been responsible for'the licensibility'of Shoreham's nuclear safety systems, the LOCA and transient analyses, as well as the tech-O nical specifications and various generic licensing issues. I am a licensed professional nuclear engineer in the State. of California. O 4 i l !~ i i l l lO I i

O i Attachment 2 Information Relating to ANS-22

1. April 21, 1972 letter, Campbell and Moore to Koch, Submission of ANS PWR criteria.

(letter attachments 4 and 5 not included)

2. Foreward to ANSI /ANS-52.1-1978.

O O I . . . . . . . . . ____ _

D

                                                                      #v h3 man Westinghcusa E!actric Ccr;cratton-     ?cwerSystems                      w eev azza PmmePemervense2 0 April 21,1972 f

'O JCM-s83 Mr. L. J. Koch Chaiman, N-18 ANSI Comittee Illinois Power Company 500 South 27th Street Decatur, Illinois

Dear Mr. Koch:

SUBMISSION OF ANS PWR CRITERIA The standard " Nuclear Safety L-iteria for the Design of Stationary Pres-surized Water Reactor Plants" (PWR Criteria), January 1972 issue, is offered for promulgation as an American National Standards Institute (ANSI) standard. We understand that the number H18.2 has been set aside for this purpose. Enclosed is a copy of the document. We are femarding 350 copies to

 '1   W Roberta Brown of ANS Headquarters as we understand that, as Secretariat ANS will distribute copies to members of N-18, the Nuclear Technical Advisory Board, the Board of Standards Review, and the ANS Standards Comittee.

To supplement information contained in thiis letter, we offer the below listed attachments: Attachment Title

                                #1          Guidelines and Scope of the ANS PWR Criteria
                                #2          Historical Background
                                #3          System for Classifying Components
                                #4          Affiliations of Systems Engineering       -

Comittee Members

                                #5          PWR Subcomittee Members and Alternates.

Application of the PWR Criteria should help to fulfill objectives originally advanced, namely, simplifying and streamlining the licensing procedures, providing a means to obtain industry-wide agreement on objectives, providing a means to unifomly judge the protection afforded the public. O O

1

.           .:C;1-M3                              -E-                      1.o.m i I , :57   l Reasonable diligence was exercised by the cemittae to c: ordinate the            l content of the PWR Critaria with various c: des and standards so that            !

requirements would not overiao. In some instances, the cemittae included requirements that it fully axpects will be covered by other codes or stan-dards in the future, planning deletion of such requirements when they are covered in a more logical document. However, the cemittee knows of no significant conflicts presently existing with ANSI codes or standards. 'Je believe the proposed standard was developed in ac:ord with procedures established by the American Nuclear Society. Q If the PWR CMteMa are successful in fulfilling the objectives cited, tnen their issue within an ANSI standard will support the national interest 1 by reducing time lost in the licensing processes. This would contribute , to mitigation of the growing power shortage. j As a result of the makeup of the comittee and the guidelines followed in drafting the CMteHa, protection against unfair discrimination was auto-mati c. The rignts and opinions of each cemittee member were respected; also the positions of industry vis-a-vis the mgulatory agencies were care-fully considered. Technical quality is necessarily comensurate with the abilities, the applica-tion and the zeal of those who actively participated in producing the final product. From inception, high caliber individuals joined in, were interested in, and worked diligently on the creation of a standard for which all parties felt that a- great need existed. Some comittee members were key executives from industry, but all had some special contination of talent, experience, knowledge and ability, pennitting positive contributions by each. The prin-cipal criticism of the end result may be that of incompleteness in certain areas, since the scope was extremely broad, encompassing the entire nuclear plant. Throughout the period of endeavor, technical quality was an utmost concern. There was always recognition that the Criteria would never be static, but would require interpretation and change to fill in scope omissions, to further clarify intents, and to remain compatible with changing licensing requirements, n recognition of the procedures required to gain ANSI approval, the comittee. is ready to answer queries that arise, meet with N-18 for review purposes, and respond- to all timely, pertinent coments. Very truly yours, POWER REACTOR SYSTEMS ENGINEERING COMMITTEE

            /im 3D. A.ACampbell
                                                               < a4 d Attachments                             Secretary O

J S. No airman ec: Roberta Brown ANS Headquarters

   . . .   .   . _ _ _ .                                                       .                     i JCM-533

> April 21,1972 Attachment #1 GUIDELINES AND SCOPE OF THE ANS PWR CRITERIA As the original promoter of the PWR " supplementary criteria", the AEC recognized that joint AEC and industry participation would result in better identification of important safety problems of pressurized water reactor nuclear power plants, better recognition of each other's problems and, therefore, better ultimate definition of the AEC's General Design Criteria. The supplementary criteria would provide a basis to unifonnly judge the protection afforded, thus streamline the licensing processes that had, until that time, been handicapped by case-by-case evaluations of safety. By the time work of generating criteHa fell to a single comittee, the following ' guides had general concurrence and were being followed: a) The Criteria would deal with both systems and components. Codes and standards usually deal with one or the other. b) The prime concern would be nuclear safety, with conventional safety continuing under existing codes and standards. c) The application of the criteHa would be in the design stage but

             -                effects of deterioration in service (such as from fast neutron damage) would be considered.

d) The complexities and interactions in relating various components and systems with' respect to one, another, or combinations would be considered. e) .The criteMa would be primarily derived from "present practice" with minimum reliance on innovative approaches. O The major area in which the coninittee found it necessary to be innovative was in the definition of safety classes and their relationship to code h classes (as elaborated in Attachment #3).

2.1-533

   .                  April 21,1972 Attachment i/2 HISTORICAL BACXGROUND At the time that the Nuclear Safety Criteria for the Design of Stationary 0               Pressurized Water Reactor Plants was offered to ANSI's Nuclear Design Criteria Comittee, N-18, on Novemoer 4,1970, for thal use, a 5 page history and 3 page listing of major events occurring during creation of the Criteria were forwarded. This infonnation, and subsequent events,
                      &re sumarized below; greater detail, is available from D. A. Campbell, Secretary of ANS-20.

Histab! to Time of Trial Use Submittal Dr. Clifford Beck, of the AEC, went before the Reactor Safety Standards Comittee, N-6, of United States of America (now American National) Stan-dards Institute in October or early November,1965 with a request for industry help to develop nuclear safety standards. At the time, the AEC ms ready to publish the first version of their General Design Criteria, a framework of broad design objectives and perfomance goals required for public protection. Dr. Beck wanted industry, with AEC participation, to prepare supplementary cMteria to supplement and implement the general design criteria. . Objectives advanced included: simplifying and streamlining the licensing procedures, providing a means to obtain industry-wide agreement on objec-tives, providing a means to uniformly judge the protection afforded the public and avoiding the necessity of governmental articulation of such requirements in' the form of rulemakings. An ad hoc comittee, with C. R. McCullough as chairman,'was charged with developing a p.lan for im-plemnting Dr. Beck's proposal. A series of drafts were produced but a 1967 review by N-6 was highly critical of the msult. Responsibility for further work was transferred to a Systems EngineeMng Subcomittee structured under the ANS Standards Comittee but under the auspices of a new Nuclear Design' Criteria Com-mittee, N-18, of what is now ANSI. The comittee had representatives from utilities, manufacturers, architect engineers, and (two) from the AEC. Work on strengthening of the document was accomplished in large part by dividing the work into specific assignment areas, producing redrafts or new subdrafts for those areas, and using the full comittee to review the msult. Factortd in was the result of meetings with IEEE, ASME and the AEC. One meeting with the AEC was to discuss the relationship of AEC's General Design Criteria to the ANS criteria and subsequent meetings were specifically devoted to evolutionary differences between ANS and AEC approaches to classifying components. The AEC was highly influential in O'- establishing some of the basic objectives for the ANS classification system and its ultimate character. O L - .. . - . - - - - - -- --------_____ __-.____-.._________-.____ _____________________ ___ _ _________ J

JCf *M

l Attachment #2 (Ctd.) Draft Eight was fomarded to N-18 of ANSI where, after a review meeting and balloting, it was eventually approved, based on appropriate revision, and later accepted as revised 'or " Trial Use and Coment". About 56 comittee-days and untold time by subgroups and individuals had been expended to the time of the August 1970 draft, the revised version of O Draft Eight containing changes responsive to N-18 comments. History Followino Trial Use Submittal At least some of the infomation famarded to N-18 by the Powe Reactors d Systems Engineering Comittee (ANS-20, the comittee responsible for bringing the document to its submittal condition), was unfortunately i separated from the submittal to ANSI's Nuclear Technical Advisory Board (NTAB) and Board of Standards Review. Thus, the suggestion of issuing

the standard for " trial use and coment" was initially unknown to ANSI of ficials. Belatedly, the ANS Standards Comittee learned of this problem and ordered affixing of " trial use and coment" stickers, an action in-stituted on June 24, 1972. Because NTAB had earlier invoked a rule that such status could last only 12 mor.ths, the sticker specified that the trial use peMod would not continue beyond 12 r.cnths from the November,1970 date of publication of the August 1970 draft.

i Meanwhile, application trial use was made of the component classification system when utilities (e.g. , Duke Power, Alabama Power, Puerto Rico Power} , applying to construct new nuclear plants, identified in their safety report: - that they had adopted the ANS system. These actions led to another meeting (the third over a 15 month peMod) between ANS-20 and the AEC to identify l remaining differences of opinion on applicable quality levels to be applied to specific components. After some adjustments, the classifications proposed i by these utilities were found acceptable to the AEC. l In a September 20 ANS-20 meeting, to which the working group chaimen of ANS-21 were invited, it was resolved to proceed with necessary changes of the trial use version of the PWR CMteria to prepare the document for i i ANSI review and acceptance. ANS-21 is the PWR Subcomittee of ANS-20 charged with carrying on the work of updatino and revision to content of the document after it is accepted as a standard by ANSI. Acting on decisions reached in the September 20 meeting, a copy of the PWR CHteria was sent to each member of the ANS-20 and 21 comittees with a request for coments required to update and correct the document.

                  " Affiliations of Systems EngineeMng Comittee Members" and "ANS-21 Members and Their Alternates" are given in separate attachments. No increased scope was planned with the exception that fluid system inter-face criteMa were to be added ff ready in time. To expedite matters, a small' ad hoc group met October 13 to recomend changes to the full O                 co-4tt         aica === aa octa6 r 19-D
   .        J::t-U:

Attschment #2 (Ctd.) Scire changes, at first thcught to be minor, were necessary to the component classf fication system. These were refinements resulting from experience gained in actual applications. A new issue was opened, hcwever, this one trig-gered by extensive concurrent work being done by the ANS-22 BWR Subecmmittee n in preparing the ANS BWR Criteria (beginning Septemoer 15,1970). This was U whether a simple one-for-one relationship could exist for fluid system components between safety and code classes. Attempts to produce just such a relationship had been previously tried and abandoned. An ad hoc attempt at redefinitions to accomplish this was offered to a meeting of ANS-20 on November 16, 1971 to which members of ANS-21 and ANS-22 were invited. The result was a decision to make such a conversion, eliminating the old Safety Class 2a and 2b divisions of Safety Class 2. Rewrite of the ad hoc effort, followed by a series of refinements to deal with problems discovered by practical applications, led to the version of definitions and application information now found in the January 1972 issue. One effect of the updating effort was review and correction of all refer-ences. The need for change was exemplified by the many references to ASME Section III which had been issued in a new edition with a new scope and ti tl e. Additionally, new reference infonnation for the Probable Maximum Hurricane and a " List of References" were provided. The twelve current members of ANS-20 who actively participated in the preparation of the PWR Criteria were balloted on the January 1972 version and all ballots were affirmative; the last was dated March 30, 1972. Coments received were aporopriately considered and final changes made. Balloted members were: G. A. Arlotto V. Krecicki W. H. Owen D. A. Campbell J. V. A. Longcar J. W. Stacey R. L. Ferguson J. S. Moore R. A. Wiesemann H. E. Flora J. H. Noble W. K. Wilhelm. 8 4 0 J

~^
          '* 2. :-in                                                                       1 AcH 1 21, 1972

se Attachment #3 SYSTEM FOR C1.ASSIFYING COMPONENTS The need to have a rating sy: tem for classifying components in accordance with importance to safety was ;ecognized early during creation of the O " Nuclear Safety CHteHa for the Design of Stationary Pressurized Water Reactor Plants" (PWR CH teHa). Difficulties encountered in creating the system were attributable to uncertainties as to (a) how to make the dis-tinctions (b) how many distinctions to have, (c) how to apply the distinc-tions created, and (d) the scope of coverage. Early creative attests failed because objectives (see below) wi a neither clearly defined nor adequately understood. The objectives e,olved from the educational process l associated with many discussions: within the comittee, with ASME, and with the AEC.

Almost all participants to the licensing process had created one kind of classification system or another but there was neither a uniform system for what to classify nor agreement on definitions. There were systems classifying for quality level, aseismic design, functional requirements, and quality administration but nothing applied in some way to all of these considerations. It became clear that the ANS " safety classes", as they were (and are) called, should apply to all components incertant to nuclear safety and the classification system should permit correlation to aseismic design requirenants, to other environmental conditions, code classes, and quality assurance measures. The creation of the classification system was difficult; consequently, the definitions of class undenvent a series of iterations. Even so, the result was limited, for lack of time, to fluid system components. The latest system iteration, included in tne January 1972 issue of the PWR Criteria, incorporates the one-for-one equivalence between safety and code classes. Also, newest class definitions reflect ANS - AEC compromises. The conmittee recognized that consideration of classifiestions should be treated in the context of the total safety problem. The total safety problem is met, not only by the quality built into a component, but also by confonnance to m1ated criteria that, together, make up the total safety to the public. Recognizing that the class chosen for a component affects both the contained quality and the assurance of that quality, a body of related criteria mst dovetail with the classification system to suitably achhve appropriate total safety. That body of criteria, the PWR CriteHa, ripresents the only example where a classification system and related' criteria are fully integrated into a single docunent. The ANS definitions of safety classes are, as much as possible, functional to provide flexibility for future system realignments. The ANS functional O 9ar==ch wita t* rrvias =< r i t d crit r4 to ta d <4a4*4=as. =vo4ds a need for lengthy definitions. h

            .                      OCM-553 i

Attachment #3 (Ctd.) The ANS classification system fulfills these objectives:' i a) Defines safety classes having a broader connotation than the previously jO used s4esm4c cius , eus 11miting en groups of .cioso to wo: j safety and code b) Cmates simple, functional-type definitions, devoid of doses for easy application 1 c) Permits a direct cormlation between safety and code classes

d) Clearly recognizes that it is components, not systems, being classified even though it is the function of the system in which the component l 1s located. together with the importance of the conponent within the

!- system, that determines its classification e) Provides a framework upon which to expand the system beyond fluid , system components , i f) Rules out non-pertinent issues: (i) economics and (ii) conventional (versus nuclear) safety, as cases for definitions g) Avoids the vagueness included in previous definitions (e:g., adjectives

                                                         " substantial" and " excessive")

} h) Avoids tagging buildings and structures with clasi labels (exception:

. reactor' containment) as.being generally too complex to meaningfully 1 indic' ate the applicable safety criteria by this dans. .- .

During the evolutionary period of development, there were a whole series of formal and informal meetings in which drafts, mdrafts, proposals, and j counter-proposals of safety class definitions were offered. Doring this t time period, inputs obtained from ASME and the AEC were quite 1,nfluential j in shaping the final msult. ! To arrive at suitable definitions, the conurittee considered, among other j things,. the consequences of pressure boundary failure, the importance of the function mquired of the component, accessibility for service during safety functioning, whether backup means to perform the safety function 4 would be available should the classified cosponent fail, and practice with mgard to mdundancy. The msulting definitions establish class gradations having the effect of weighting, for each component, the importance of such considerations adjusted for the influence of applicable mlated criteria contained in the PWR Criteria.

 !O i'

D. A. Campbell

     . . .                                                                                                          4/21/72 0-
      - - ,         , - - - - -        , , . - _ - . - -         , _ - - - -  , . , . _ _ _ . , - . , , -. ~ - - -                  -

                              ~d " " ' ~ d ^-" a " *" ' S-d-d "-"- S" 'r caa"~ * "d Foreword =Stauonary Boilms Water Reactor Planta, ANSilANS-52.11978 IN212)

.q At the annual ANS meeting in June 1970, the ANS-20 Power Reactor Systems U Engineering Committee (now part of ANS-50, Nuclear Power Plant Systems Engineering Committee) accepted the responsibility to develop a set of Nuclear Safety o Criteria for Boiling Water Reactors. At the first organizing meeting in September 1970,

;")                 the ANS-52 (formerly ANS-22) Subcommittee reviewed the work of ANS 20 on the PWR criteria and considered in substantial detail an appropriate approach and format for the BWR criteria.

It was established that the PWR enteria (now N18.2) would be utilized to form basic structure of the BWR criteria. That is, it was agreed to adopt the. Safety Class concept, to establish Conditions for Design, and to adopt the same basic format regard-ing criteria for individual systems. The ANS 52 Subcommittee also agreed that the scope of the BWR criteria would be extended beyond that represented in the PWR criteria of August 1970, in three principal areas: a) Safety Classes The ANS BWR Criteria would be structured to include instrumentation, control, j fuel storage, electrical and other similar structures and equipment having plant safety functions. It would not be limited to pressure retaining components. b) Conditions for Design The ANS-52 Subcommittee adopted the philosophy that the definition of Con. ditions for Design should be the combination of plant process conditions and natural phenomena that could result in simultaneous effects on p' ant equipment. c) Design Requirements The ANS BWR Criteria would provide a set of Design Requirements for all ] categories of Safety Classes in terms of industry codes and standards for each of ! the Conditions of Design. The BWR Design Requirements would then be referen- ' ced to specific established standards and in this manner assure substantial in. terrelationship with the codes and standards developed by otter technical groups. In developing the format and technical content of the ANS BWR Criteria it has been the objective of the Subcommittee to achieve the following: a) To establish a disciplined, systematic method for defining safety requirements for a boiling water reactor power plant; b) To establish and delineate the basic functional safety requirements for the design of a boiling water reactor power plant; c) To be responsive to both the regulatory requirements of the Nuclear Regulatory Commission and the design and technical requirements of industry codes and stan-dards; d) To provide a framework for augmentation of the ANS BWR Criteria as additional standards are developed within the nuclear industry; e) To provide a uniform basis for design safety requirements which are to be reflected in regulatory licensing dccumentation. It is believed that these objectives have been achieved. To achieve a broad consensus forum, the ANS BWR Criteria was issued for the " Trial 9 Use and Comment" in 1974. The ANS BWR Criteria was updated after the " Trial Use and Comment" phase incorporating 1) input from the " Trial Use and Comment" reviewers; 2) the ANS-50 Glossary, des gnated CWG 1; and 3) the latest state-of the. art, e.g., recently issued ANSI Standards. The modification was made in accordance

a- 1 t i f with the ANS-50 format, designated CWG-4. The Standard as updated was approved ' by the ANS-52 Committee in April 1975, ANS 50 in January 1976; and by the ANSI N18 Committee in July 1976. ANS-52 recognizes that continuing efforts will be required to augment or modify the

criteria to implement changing licensing requirements, to achieve standardization among the various industry criteria and standards currently being developed, and to h provide additional clarification or interpretation as appropriate. The ANS-52 BWR Subcommittee meets regularly to consider revisions or modifications to these standards. Comments, suggestions and requests for in-terpretations should be addressed to the Chairman, ANS-52 BWR Criteria Sub-commitee, American Nuclear Society,555 North Keniegton Avenue, La Grange Park, IL 60525 i The ANS-52 membership at the time of approval of this standard was: W. H. D'Ardenne. Chairmats General Electne Com. A. T. Molin. United Eneneers & Constructors. Inc. > l pany H. Ostick. Eba.sco Seruces i G A Arlotto, US Nuclear Regulatory Commission D. R Patterson. Tennessac Valley Authorsty R Bner, US Nuclear Regulatory Commission G. E. Peterson, Commonuralth Edaon Company F. E. Ehrensperger. Southern Sensces, Inc. B. G. Schultz. Stone and n'ebster O. J. Foster. General Electne Company J. Steligowski, Consolsdated Edsson Company H. Frwnd, Bechtel Corporatson - G. F. Hoveke, Sargent and Lundy The membership of ANS-50 at the time of its approval of this Standard was: M. N. Bieldanea. Chantman, Black & Vratch T. J. Pashoe, Nuclear Sertices Corporatson J. F. Malley, Vice Chairman. Babmck & Hilcos D. R Patterson. Tennessee Valley Authonty G. A Arlotto, US Nuclear Regulatory Commasion C. H. Ponndester. Baltimore Gas & Electne Com. D. A. Campbell, n'estinghouse Electne Corporatson pany (Op) C. O. Cotter, Kaaer Eneneers C. Reed. Commonwealth Edison Company W. H. D'Ardenne, General Electne Company R F. Schreiber. %'estinghouse Electne Corporation F. A. Dougherty, EDS Nuclear Inc. W. C. Spencer, Virginia Electne Pbuer &s.

~'

J. Flavd. Metropolstan Eduon Company J W. Stacey, Yanker Atomse Electne Company C. J. Gill, Bechtel Power Corporation G. C. Vellender. Fluor.honeer E. W. Hewatt, Combustion Enenrenna. Inc. J. M. Waage General Atomic Company R W. Keauen. North Amencan Rockuell Cor. M. D. Weber, Amencan Nuclear Society paration G. L Wessman. General Atomic Company L L Newhart, Catalytse Inc. J. L Windhorst. Southern Company Services J. H. Noble, Stone & n'ebster Eneneenne Cor. F. Zapp, Oak Rutge Natsonal laboratory poration C. B. Zitek, Commonwealth Edaon Company

d 5

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NUCLEAR POWER E N GlH E E R IN G C OMMI T T E E C hierrasa Vice Chairenaa & Vice Chaierriani Standardo Coordinator Secretary O.E. ALLEN S. M. Aice J. T. Sauer C. M. Chiappetta tsauen E agiaeers & ovko Pe=or Compeay Geaeral Atomic Compear Seegent & Lwady E a glaser s C oa s te.c tees, tac. P.O. Ben 33509 P.o. Ses 8160s 55 E ast Mearee Street.F reer 34 0 Se n.17en Street Cheeleue. NC 20242 Sea olese. CA 92138 Chaceae, eL 60603

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Use of the Term, "Important to Safety" no e,.. e6A

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                   .          .         4 Tegt.ag                              a letter to Robert B. Minogue, Director of the Office H Ct. itoit                                                                    of Nuclear Regulatory Research, stating the NPEC posi-
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                ;c          .s. t e e n c c : te . ..'                             W. H. D'Ardenne (NUPPSCO) (w/1) isio re .... a s.

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                  . . . . . . 94 sus                                                  827 File (w/o)

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i Eucinicat ana D@ g )) ((h]{((@0[@ hh@0] ] 9> Eucinnics EnoN t t R s. INC. NUCLEAR POWER E N GIN F E R IN G COMMITTEE C hierman Vice Chairman & Vice Chairman Standardo Coordinator Secretary D.E. ALLEN S. M. Rice J. T. Sauer L aetse Eastaaers 4, C. M. Chiappetta Duke Pe.er Compear General Atom.c Covepaar Coa ste. ster s. lac. Sergent & L ndy Eas taeers P.o. Som 33 8 89 P.O. Ben 81608 6 54,th 17en Street Charlotte, NC 20242 55 East Monroe $treet. Fleer 34 Se.t Diese. CA 92138 Chicago. IL 60603

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    , u, c . e . . , , . . . . ,,,                                            Mr. Robert B. Minogue, Director 8 me n.=v se225

4i 68e.a500 E .t. 2io Office of Nuclear Regulatory Research s c : . o e .e.< .i .. U.S. Nuclear Regulatory Commission N. 5. Portir Washington, D.C. 20555 ai,*..a.u- P .: . .c P o ,* S o ci, S y s te =

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a. . . e, c . 4 4 9+352

Dear Mr. Minogue:

     . e a m e's c..      :3 .=i...-..

Subject:

Use of the Term "Important to Safetv"

            ,,, ...                ,, a r....,e g , g , 5,,,

References:

 <.             ........i- e c . . . a c ,, . , ,

1) Proposed Revision 1 to Regulatory

  . f; s e.                   3 7 6 r 4 * *'                                                              Guide 1.89. Environmental Oualifi-
 '~,, '. '.'

d* fy'l cation of Electric Equipment for Nuclear Power Plants, February, 1982.

                                          " ~" '

[D..,,(c,'[eci.cn6,.a.1 r a C ., . 2) Proposed Revision 2 to Regulatory

      -          ha' "                                                                                    Guide 1.105, Instrument Setpoints, 5 .J   ..,. o- e.:8.                                                                                     December, 1981.

e. , 6 ;; W.S E a t. 8 6 4 ? C,

  ., . r . se                o...,

J 4_ Pen a 4'id 3) Draft Regulatory Guide (Task IC 126-5),

              - -an e .u e s . . -                                                                        Instrument Sensing 1.ines, March,1982.

., 4 :.. . m r . .. s . e

     .,,,y,.                     -,,3,,3o
  ,. 3 . m - a                                                                                    4) Memorandum from Harold R. Denton to
          ... w....<cc is,sie..

All NRR Personnel, Standard Definitions J. E. Thomas for Commonly-Used Safety Classification

  >.. .. .. u..y.,,                                                                                      Terms November 20, 1981.

B. . .9+

                   ,.. . r_ ,o t3
      .,4         i   ..a.                                                                        5)     10CFR50, Propo.>ed Rule (47FRJ879,
              . . . . , . . . .. . s             c.....       i is.ia. s 1/20/82) Environmental Oual ification t         kg..a s                                                                                         of Electric Equipment for Nuclear 4-

. i . ., r

                           , a t     ..,...,y,,           r                                          Power Plants.
            ,       .c n                   p-                                                                                                                   m
*                 # ' ';                                                            A number of recent NRC documents have used the term
          '      4                                                          "important to safety" in describing the scope of systems
        $ ' "" ', '," ' '. "a .. , .                                       and equipment to which the document applies. Notable e.sanples are the proposed Revision 1 of Regulatory Guide
 ,Cr v.

r , c r. . . , V- ' ** 0' 1.89 (Reference 1), the proposed Revision 2 of Regulatory

               *' ' ' ' "                                                  Guide 1.105 (Reference 2), and the draft Regulatory Guide                            ,
          . s..                   .     .v-                            on Instrument Sensing Lines (Task IC 126-5) (Reference 3).

F Bates

          .      ', ~s..su.
        \ .*             %    4-    g uk : r e su

> .'s. 2C3

t Mr. Robert B. Minogur, Director May 10, 1982 82-C-015 l'.S. Nuclear Regulatory Commission Reference is made to 10CFR Part 50 as the source of the terminology. Over the years, the terminology of the General Design Criteria of Appendix A of.10CFR Part 50 has been understood through commen usage to equate systems important to safety to safety-related or safety systems. The repeated references within the General Design Criteria to preservation

      .of the safety function being performed by " structures, systems, and

[V ) components important to safety" reinforces this equivalence of terms. The current NRC intention in the use of/t he term "important to safety" appears to be to broaden the scope of equipment addressed to include more than safety-related or safety systems. In an internal NRC memorandum (Reference 4) Harold Lenton defined " safety-related" as a subset of "important to safety". Broadening the usage of the tem "important to safety" to encompass an undefined set of systems, in addition to safety-related or safety systems, increases confusion in the dialogue on current NRC requirements / guidance and creates an unworkable situation. A clear understanding of the principles for determining what is included and what is not included in " systems important to safety" is needed. For example, Regulatory Guide 1.105, Instrument Setpoints, recommends the substitution of " systems important to safety" for " nuclear-safety-related". This substitution of terms adds an unknown number of systems to the set of systems required to meet the draft ISA standard. The IEEE, through a Nuclear Power Engineering Committee (NEE _G) working group on standards project P827, is attempting to develop a methodology for assigning design criteria based on a system's level of importance to safety. Although considerable progress has been made on the subject in the last year and a half, the methodology has not been developed to the point of being easily understood and usable. Unfortunately, the complexity of the subject prevents the methodology, as currently developed, from being uniformly interpreted and applied by individual users. Work is continuing on over-coming this deficiency, so that application may be consistent from user to user and enforcement may be uniform from application to application. The difficulty in producing this methodology underscores the need for careful choice of teminology so that a basis is established to promote common understanding and not to introduce additional confusion. Until the P827 methodology has passed through the IEEE review (consensus) process and the tem " systems important to safety" has a commonly understood meaning, it is recommended that the ERC refrain from using this tem without also including a clear definition of which systems are addressed. This is exactly what was done in the development of the rule on equipment qualification (Reference 5). Alternatively, commonly understood terms, such as safety-related, and terms defined in voluntary standards, such as safety systems, should be employed. If it appears necessary to address systems beyond the scope of these terms, then the additional systems should be clearly identified. It is recommended that the tem " nuclear-safety-related" be retained in ] b the proposed Revision 2 to Regulatory Guide 1.105 and the proposed Regulatory Guide on Instrument Sensing Lines (Task IC 126-5) and that the tem "important to safety" not be used in these documentsa

Mr. Robert B. Minogue. Director May 10, 1952 L'.S. Nuclear Regulatory Commission 82-C-015 It is also recommended that the term " electric equipment important to sifety," be replaced by " Class IE electric equipment" in the second para-grcph of the introduction to the proposed Revision 1 to Regulatory Guide 1.89. As an alternative to these two recommendations, it is recoc. mended that the general terms he replaced with a specific tabulation of the systems; equipment to which the regulatory guide is applicable. Similar treatment should be accorded other NRC regulatory documents in the future, or until the P827 methodology reaches consensus in the nuclear power community. ,s Very truly yours,

p-

hC'5I. Allen Chairman, NPEC REA/mab cc: Paul C. Shevmon Chairman, Advisory Committee on Reactor Safeguards , Harold R. Denton, Director Office of Nuclear Reactor Regulation (g Edward C. k'enzinger, Chief , (_) Instrumentation and Control Branch j office of Nuclect Reactor Regulation j Letter File l 827 File j i t i i i I l i ! i !O i i 4

'e

l

Attachment 5 j i r Attachments to the Testimony , of Vojin Joksimovich i i i I I i i h i 4 h

i l

PRA PROCEDURES GUIDE, SECTION 3.7.2 DEFINITION OF DEPENDENT FAILURES Type 1. Common-cause initiating events: external and internal events that have the potential for initiating a plant transient and that increase the probability of failure in multiple sys-tems. These events usually, but not always, result in severe environmental stresses on components and structures. Examples () include fires, floods, earthquakes, loss of offsite power, air-craft crashes, and gas clouds. Type 2. Intersystem dependencies: events or failure causes that create interdependencies among the probabilities of fail-ure for multiple systems. Stated another way, intersystem dependencies cause conditional probability of failure of a . given system along an accident sequence to be dependent on the success or failure of systems that precede it in the sequence. There are several subtypes of interest in risk analysis. Type 2A. Functional dependencies: dependencies among systems that follow from the plant design phylosophy, system capabili-ties and limitation, and design bases. One example is a system that is not used or needed unless other systems have failed; another is a system that is designed to function only in con-junction with the successful operation of other systems. Type 2B. Shared-equipment dependencies: these are dependen-cies or multiple systems on the same components, subsystems, or

 /  auxillary equipment. Exampels are (1) a collection of pumps and valves that provide a coolant-injection and a coolant-recirculation function when the functions appear as different events in the event tree and (2) components in different sys-tems fed from the same electrica? bus.

Type 2C. Physical interactions: failure mechanisms, similar to those in common-cause initiators, that do notcause an ini-tiating event but nonetheless increase the probability of multiple-system failures occurring at the same time. Often they are associated with extreme environmental stresses created by the failure of one or more systems after an initiating event. For example, the failure of a set of sensors in one system can be caused by the excessive temperature resulting from the failure of a second system intended to cool the heat source. Type 2D. Human-interaction dependencies: dependencies intro-duced by human actions, including errors of omission and com-mission. The persons involved can be anyone associated with a plant-life-cycle activity, including designers, manufacturers, (- constructors, inspectors, operators and maintenance personnel. ( A dependent failure of this type occurs, for example, when an operator turns off a system after failing to correctly diagnose the condition of the plant -- an event that happened during the Three Mile Island accident when an operator turned off the emergency core-cooling system.

Type 3. Intercomponent dependencies: events or failure causes that result in a dependence among the probabilities of failure of multiple components or subsystems. The multiple failures of interest oin risk analysis are usually within the same system O , or the same minimal cut set that has been identified for a sys-tem or an entire accident sequence. Subtypes 3A, 3B, 3C, and 3D are defined to correspond with subtypes 2A, 2B, 2C, and 2D, respectively, except that the multiple failures occur at the subsystem and component level instead of at the system level. O O

        -                                                                   l System 1                                System 2 Initia ting Opt'81tb                                OPef 8ttl event 4
                                                                                                             #                                                       l Ye6                                     Yes 1

No f l No Yt& A .. NO . .- __ .. d f . _ _ . tua .,g s, it,- 3 g,. stem 2 J I* OPP'a1P1 OPP' ale) Yes Ob"~ 8. I i lO

'                               No                                          .t.

J

!                                                                         -                                                      3
!                                                                         No I

t t d i 1 4 1

                                 %                                      M                                                                                          j p rotes                             operates event
                                                             .Yes Yes j8 an.                                  _. <

l.l I l.

!                                                  ..                                             --{i!!3._f.                                        . .

T i I i i

           -                                                                                           -4 Syste m 2 System 1 O                          fails fails OR C
                      <m Ar.:                 F
                                                                 <>     AND A          F     G A        B                        D                                       E

(-o ' Compemen A Syste .1 ope ates ope'aies in.t;ating Cc-oc-en F Syrem 2

                                                                                         . operates                           operates e ent                                             s.                                               ,

7

                                                                                                                   ,,    j             3
                                                                                                       . . g g.. . g . .-
                                                                                                                         -e- r (O                                                                                                     i                   -e _ ,..
                                                                                                       - y . . g - s.-

i T

a . 2-1 A COMPAR SON OF PROBABILISTIC AND DETERMINISTIC SAFTY ASSESSMENT METHOD 0 LOG Probabilistic Deterministic Initiating events Complete spectrum Limited " credible" and p elected or deci on basis events Plant response A 1 relevant plant 6stricted to sequences re pense sequences sequences satisfying incl ing multiple " single failure" failur criterion Probability Median or an ,alues Usually not assessed evaluations and uncertal ries assessed Consequence Median r mean val s Upper bound values evaluations and uncertainties evaluated without ass /ssed. Assump- measure of uncertain-tons are realistic es. Assumptions are Ind always mechan- de d conservative istic. and e n nonmechan-some cases O- ' istic 1 Relative safety Established quan- Establishe uali-significance of titatively by means tatively sinc of focusing on probabilities a 5 cvents(e.g/.for guidance 1 risk or core melt margins on conseq nce recomenping dominant sequences are not assessed safety R&D) a-O T e

s. O Attachment 6 Tabic 3.1 of the Testimony of Dr. Edward T. Burns O O

Table 3.1 EXAMPLES OF EVENT TREE / FAULT TREE METHODS APPLIED TO THE ASSESSMENT OF POTENTIAL SYSTEMS INTERACTIONS SPECIFIC EXAMPLES TYPE OF SYSTEM GENERAL , TR INTERACTION APPROACH 5H R Functionally Coupled

Event Tree / Fault Tree Event Tree for sequences supported by depend- initiated by a Turbine Trip ency matrices Event Tree for sequences initiated by a Manual Shutdown Event Tree for sequences initiated by a MSIV Closure m Event ~ree for sequences (j initiated by a loss of Feedwater Event Tree for sequences initiated by a Loss of Condenser Vacuum Event Tree for sequences initiated by a Loss of Offsite Power Event Tree for sequences initiated by a Time Phased Loss of Offsite Power Event Tree for sequences initiated by a Inadvert'ent Open Relief Valve (IORV)

Event Tree for sequences g initiated by a Large LOCA

Shared systems are discussed separately. The definition used here is consistent with the PRA guide and is more explicit than those sometimes used in discussing SI.

  -                                              3-9

. Table 3.1 (Cont'd) EXAMPLES OF EVENT TREE / FAULT TREE METHODS APPLIED TO THE ASSESSMENT OF POTENTIAL SYSTEMS INTERACTIONS SPECIFIC EXAMPLES TYPE OF SYSTEM GENERAL INTERACTION APPROACH S E M R Functionally Coupled Event Tree for sequences (Cont'd) initiated by a Medium LOCA Event Tree for sequences initiated by a Small LOCA Event Tree for sequences initiated by a large LOCA Outside Containment Event Tree for sequences iditiated by a Reactor Pressure Vessel LOCA

                           .                          Event Tree for sequences initiated by a Turbine Trip ATWS Event Tree for sequence initiated by a Isolation Event ATWS Event Tree for sequence initiated by a Loss of Offsite Power                     ,

Event Tree for sequence initiated by a 10RV ATWS E.T. Event Tree for sequence initiated by a Flood of Elevation 8 3-10

Table 3.1 (Cont'd) EXAMPLES OF EVENT TREE / FAULT TREE METHODS APPLIED TO THE ASSESSMENT OF POTENTIAL SYSTEMS INTERACTIONS 5 L TYPE OF SYSTEM GENERAL APPROACH T E I E INTERACTION SHOREHAM PRA Functionally Coupled Event Tree for' sequences (Cont'd.) initiated by a Loss of a DC bus Containment Event Trees Giving impact of postulated degraded core on plant-system function-ability Shared Systems System Level Fault HPCI/RCIC/Feedwater F.T. Trees supported by O Dependency Matrices Dependency Matrix e minimum cutsets Dependency Matrix e quantitative evaluation Low Pressure Systems F.T. ADS - Dependency Matricc3 LPCI - Dependency Matrix CS - Dependency Matrix 3-11

Table 3.1 (Cont'd. ) EXAMPLES OF EVENT TREE / FAULT TREE METHODS APPLIED TO THE ASSESSMENT OF POTENTIAL SYSTEMS INTERACTIONS SPECIFIC EXAMPLES TYPE OF SYSTEM GENERAL D IN HE APPROACH INTERACTION fRA p Shared Systems System level Fault Electric Power and Instrument-Trees supported by ation F.T. and Dependency Matrix Dependency Matrices

O Service Water System F.T.

1 Dependency Matrix Chilled Water System F.T. < (ReactorBldg.RoomCooling) Dependency Matrix Scram System F.T. Standby Liquid Control System 1 p Instrument Air System F.T. 1U l- 3-12

Table 3.1 (Cont'd. ) j() EXAMPLES OF EVENT TREE / FAULT TREE METHODS APPLIED TO THE ASSESSMENT OF POTENTIAL SYSTEMS INTERACTIONS 5E TYPE OF SYSTEM GENERAL APPROACH DI E INTERACTION SHOREHAM PRA Shared Systems Fault Tree / Event Common Sources ~of Coolant (Cont'd.) Tree makeup minimum cutsets Common Discharge Points Intercomponeht Calculated Common Scram System Dependencies Mode or Dependent Failure Probability LPCI Pumps CS Pumps () . Diesels HPCI/RCIC Offsite Power Operator Action Human Coupling Fault Tree / Event Common-mode miscalibration Tree minimum cutsets Maintenance Errors Maintenance Unavailability Operator Manual Initiator O 3-13 k.. . . . . _ . . .

l Table 3.1 (Cont'd. )

 .O                                        EXAMPLES OF EVENT TREE / FAULT TREE METHODS APPLIED TO THE              .

ASSESSMENT OF POTENTIAL SYSTEMS INTERACTIONS CE LES TYPE OF SYSTEM GENERAL , INTERACTION APPROACH SHOREHAM PRA Spatial Dependencies Fault Tree / Event Postulated flodding of Tree Elevation 8 in the Reactor - Building (Adverse impact on ECCS) - Interfacing LOCA (Adverse impact on ECCS) Containment Leakage (Adverse impact on ECCS Operation) O Repair of RHR Following LOCA

    .                                                                            Initiators O                                            -

3-14

O Attachment 7 Selected Portions of the Standard Review Plan, NUREG-0300 (Revision 1, July 1981) O Appendix A to Standard Review Plan 3.2.2 Appendix B to Standard Review Plan 3.2.2 (Figure B-1) 0

APPENDIX A* ! CLASSIFICATION OF MAIN STEAM COMPONENTS OTHER THAN THE REACTOR COOLANT PRESSURE BOUNDARY FOR BWR PLANTS A. BACKGROUND p A pipe classification of "O + QA" fcr main steam line components of BWR plants () was proposed by the General Electric Company in 1971 as an alternative to Quality Group B and has been accepted by the staff in a number of licensing case reviews. However, we have recently identified a number of potential problems which are applicable to main steam lines of BWR plants. These problems relate to postu-lated breaks in high energy fluid-containing lines outside the containment. The criteria pertaining to protection required for structures, systems, and components outside containment from the effects of postulated pipe breaks, as contained in the Director of Licensing's letter to utilities dated July 12, 1973, reference ASME Section III, Class 2, which corresponds to NRC Quality Group B. The recent ASME Code Section AI revision contains in-service inspection requirements for Class 2 components. Steam lines classified as "O + QA" could be interpreted to be exempt from these inspection requirements. Such l interpretations would be contrary to the intent of the code and inconsistent with requirements of the NRC Codes and Standards rule, Section 50.55a of 10 CFR Part 50. Furthermore, the applicability of the following NRC Regulatory Guides. l U'N Standard Review Plan section, and Regulations, as they relate to ASME Section III, Class 2 components is not always clearly identified or implemented in case applications wherever "D + QA" classification is adopted:

1. SRP Section 3.9.3, "ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures."

2. Regulatory Guide 1.26, " Quality Group Classifications and Standards."

3. 10 CFR Part 50, s 50.55a, " Codes and Standards for Nuclear Power Plants."

4 10 CFR Part 50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants." In view of the foregoing, we find it necessary to clarify the quality group classification criteria for main steam components for BWR plants. B. BRANCH TECHNICAL POSITION The main steam line components of BwR plants should conform to the criteria listed in the attached Table A-1 of SRP Section 3.2.2. l Formally BTP RSB No. 3-1 l l 3.2.2-10 Rev. l'- July 1981

C. REFERENCES .

1. Letter of March 22, 1973, J. A. Hinds to J. M. Hendrie. -

2. Letters of August 13, 1973 and November 26, 1973, J. M. Hendrie to J. A.

Hinds. -

Table A-1 l CLASSIFICATION REQUIREMENTS FOR MAIN STEAM COMPONENTS OTHER THAN THE REACTOR COOLANT PRESSURE BOUNDARY Classification Item System or Component Quality Grcuo

1. Main Steam Line from 2nd Isolation B Valve to Turbine'Stop Valve.

2. Main Steam Line Branch Lines to E First Valve.

3. Main Turbine Bypass Line to S Bypass Valve.

4. First Valve in Branch Lines B Connected to Either Main Steam -

Lines or Turbine Bypass Lines.

5. a. Turbine Stop Valves, Turbine ' + QA1 Control Valves, and Turbine or

> Bypass Valves. Certification 2

b. Main Steam Leads fro:'.' Turbine 0 + QAl'3 Control Valves to Turbine. Casing. or Certification 2

 ^The following requirements shall be met in addition to the Quality Group D requirements:

1. All cost pressure-retaining parts of a size and configuration for' which volumetric examination methods are effective shall be examined

by radiographic methods by qualified personnel. Ultrasonic examination to equivalent standards may be used as as alternate to. radiographic methods. ,

2. Examination procedures and acceptance standards shall be at least equivalent to those specified as supplementary types of examination in ANSI B31.1-1973, Par. 136.4.

g 2The following qualification shall be met with respect to the certification requirements:

1. The manufacturer of the turbine stop valves, turbine control valves, turbine bypass valves, and main steam leads from turbine control 3.2.2-11 Rev. 1 - July 1981

I

                             ,/         Table A-1 (cont'd)                                _

valves to the turbine casing shall utilize quality control procedures eauivalent to tho'se defined in General Electric Publication GEZ-4982A, " General Electtic Large Steam Turbine - Generator Quality Control Program."

 <-        2. A certification shall be obtained from the manufacturer of these (3)            valves and steam leads that the quality control program so defined has been accomplished.

i , 3The following requirements shall be met in addition to the Quality Group D requirements:

              ~
          'l. All longitudinal and circumferential butt weld joints shall be radiographed (or ultrasonically tested to equivalent standards).

Where size or configuration does not permit effective volumetric examination, magnetic particle or liquid penetrant examination may be substituted. Examination procedures and acceptance standards shall be at least equivalent to those specified as supplementary types of examinationsi. Paragraph 136.4 in ANSI B31.1-1973.

2. All fillet and socket welds shall be examined by either magnetic particle or liquid penetrant methods. All structural attachment welds to pressure retaining materials shall be examined by either magnetic particle or liquid penetrant methods. Examination procedures and acceptance standards shall be at least equivalent to those specified as supplementary types of examinations, Paragraph

{} 3. 136.4 in ANSI B31.1-1973. All inspection records shall be maintained for the life of the plant. These records shall include data pertaining to qualification of inspection personnel, examination procedures, and examination results.

                        /
                      /

1 U .

3.2.2-12 Rev.1 - July 1981

CONTAINMENT QUALITY GROUP D OUALITY OUALITY OUALITY GROUP D GROUP A GROUP P R CERTIFICATION

                                                     *                  =                                      _o       --         '
                                                                                                                                            =

b h TURBINE BUILDING AUXILIARY BUILDING -BHANCH I LINE -TURBINE MAIN STOP VALVE ,

BRANCH LINE 6

                                            $                                 ,                                  A                              -TURBINE CONTROL VALVE          ,

D STEAM LINF *

       <a                                   5                  d4  -
                                                                                                         ;v:                                   [ MAIN STEAM LEADS b                                                                         FEEDWATER LINE                                                    TURBINE     -{F  G E'JER ATOR
      ~                                                    =    A  -
                                                                           ..-                          :>                                   N
                                                                ,g BRANCH LINE               <:1     {

a CONDENSER SilUTOFF VALVES k J TURBINE BY-PASS VALVE ISOLATION VALVES INTERFACEj RESTRAINTS

                                               - SEISMIC CATEGORY I                                                               NON-SEISMIC CATEGORY I STRUCTURES, SYSTEMS & COMPONENTS                                               STRUCTURE, SYSTEMS & COMPONENTS m
     ,                                                                                                            OUALITY GROUP D e

e E' q. G Figure B-1 NRC Quality Group and Seismic Category Classifications Applicable to Power Conversion

     $                                                                                System Components in BWR/6 Plants.

lO Attachment 8 Selected Portions of the Proposed Shoreham Technical Specifications O NOTE: The " pen and ink" markups shown in this attachment are as submitted to the NRC. Such markups are the method preferred by the NRC for submittal of proposed technical specifications, to readily show plant unique modifications from the NRC Standard Technical Speci-fications. O i

b l

                --    n.      ,--m.--
,                   7                    LONG ISLAND LIGHTING COM PANY 4
                      .d'3hh
                .: __.rse/ar.se:sf                  SHOREHAM NUCLEAR POWER STATION

_...z.:..:...._- P.O. 3OX 618, NORTH COUNTRY ROAD + WADING RIVER. N.Y. *1792 1 February 1, 1982 SNRC-665 i 1

i Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Cc= mission --

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

Dear Mr. Denton:

4 i ( Enclosed is one copy of proposed Technical Specification for l Shoreham Nuclear Power Station (SNPS). This submitcal consists j of a markup of SWR Standard Technical Specification pages to reflect Shcreham specifications. I Shoreham Technical Specifications were originally written to the then latest version of NUREG 0123, Rev. 3: however, to i facilitate staff review the document was upgraded to reflect a j draft BWR/4 STS provided to LILCO in May 1981. .i l

 ,           Discussions with the NRC staff have indicated that a Technical i             Specification submittal at this time will support an Augusc

1982 approval. Fuel Load is scheduled for Septe.ber, 19S2.

Should you require any additional informaticn, please centact i this office. i l Very truly yours, l s Zl S

                   , .WY_ _.h.--

J. L. Smith Manager, Special Projects Shoreham Nuclear Power Station RCW:=p cc: E. J. Weinkam J. Higgins

O

                     . ~ ~ . . . _ . . . .. . _.._. .      _u___      ,_=a,-

i TECHNICAL SPECIFICATICNS ; 1

               ;                    February 1, 1982 j

Shoreham Nuclear Power Station - Unit 1 t - - - - -- - - . . . . __ __ m i O' O

~*

we e e m e

REACTIVITY CONTROL SYSTEMS ROD BLOCK MONITOR LIMITING CCNDITION FOR OPERATION 3.1.4.3 Both rod block monitnr (RBM) channels shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to g of RATED THERMAL POWER. ACTION: *

a. With one RBM channel inoperable, restore the inoperable RBM channel to OPERABLE status within 24 hours and verify that the reactor is not operating on a LIMITING CONTROL ROD PATTERN; otherwise, place the inop'erble rod block monitor channel in the tripped condition within the next hour.

b. With both RBM channels inoperable, place at least one inoperable rod block monitor channel in the tripped condition within one hour.

SURVEILLANCE REQUIREMENTS 4.1.4.3 Each of the above required RBM channels shall be demonstrated OPERABLE by performance of a:

a. CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies and for the OPERATIONAL CONDITIONS specified in Table 4.3.5-1.

b. CHANNEL FUNCTIONAL TEST prior to control red withdrawal when the reactor is operating on a LIMITING CONTROL ROD PATTERN.

o- o e p O GE-STS (BWR/4) 3/4 1-18

INSTRUMENTATICN O . 3/4.3.6 CONTROL ROD WITH0RAWAL BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION O 3.3.6. The control rod withdrawal block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2. APPLICABILITY: As shown in Table 3.3.6-1. ACTION: ,,

a. With a control rod withdrawal block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, take the ACTION required by Table 3.3.6-1.

O

c. The provisions of Specification 3.0.3 are not applicable in OPERA-TIONAL CONDITION 5.

SURVEILLANCE REOUIREMENTS 4.3.6 Each of the above required control rod withdrawal block trip systems and instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION opera-tions for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1. O pe.edR'-I y OM 'uc.l.u d.e.S ve.H$c.o_Me9 -L4ai m u\ bPI'p tp] i N put S CM 6 b 6.5Dj O f coy M . -

e. g.

O - O 3/4 3-47 GE-STS (BWR/4)

TABLE 3.3.6-1 g CONTROL R0D WITilDRAWAL BLOCK INSTRUMENTATION vi HINIMUt1 APPLICABLE d OPERABLE CilANNELS OPERATIONAL g TRIP FUNCTION PER 1 RIP l' UNCTION CONDITIONS ACTION E 1. R0D DLOCK HONITOR(a) S

a. Upscale 2 1* 60

b. Inoperative 2 1* 60

c. Downscale 2 1* 60

2. APRM

a. Flow Blased C imulated7 Thermal Power - Upscale -

4 1 61

b. Inoperative 4 1, 2, 5 61

c. Downscal'e 4 1 61

d. Neutron Flux - Upscale, Startup 4 2, 5 61

3. SOURCE RANGE MONITORS w a. Detector not full in(b) 3 2 61 2

2 5 61 3 2 M Y

4

b. Upscale (c) 2 5 61 Inoperative (c) 3 2

c. ,

                                                                                                         )  _

3 2 6

d. Downscale(d)

4. INTERMEDIATE RANGE MONITORS

a. Detector not. full in (e) 6 2, 5 61

b. Upscale 6 .

2, 5 61

c. Inoperatig) 6 2, 5 61

d. Downscale 6 2, 5 61

5. SCRAM DISCllARGE VOLUME

a. Water Level-llinh f2Y 1, 2, 5** 62 (b. Scram Irip Bypassed (2) (1, 2.) 5"" 6

6. REAC10R COOLANT SYSTEM RECIRCULATION FLOW

a. Upscale 2 1 62

b. Inoperative 2 1 62

c. 4Comparator N Guwuxaiei' 2 1 62

k D g TABLE 3.3.6-1 (Cont'inued) CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION r O , ACT10N ACTION 60 - Declare the 9566 inoperable and take the ACTION required by Specification 3.1.4.3. ACTION 61 - With the number of OPERABLE Channels:

a. One less than required by the Minimum OPERABLE Channels

                         " per Trip Function requiremsnt, restore the inoperable channel to OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour.

b. Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.

ACTION 62 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour.

@                                              NOTES With THERMAL POWER > 420M of RATED THERMAL POWER.
     **    With more than one control rod withdrawn. Not       applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

a. The RBM shall be automatically bypassed when a peripheral control rod is selected.

b. This function shall be automatically bypassed'if detector count rate is

           > 100 cps or the IRM channels are on range 43P or higher.

c. This function shall be automatically bypassed when the associated IRM channels are on range 8 or higher.

d. This function shall be automatically bypassed when the IRM channels are on range 3 or higher.

e. This function shall be automatically bypassed when the IRM channels are g on range 1.

O GE-STS (BWR/4) 3/4 3-49

    %    .                                                                                _   ..___.d

TABLE 3.3.6-2 n CONTROL R0D WITilDRAWAL BLOCK INSTRUMENTATION SETPOINTS d TRIP FUNCTION TRIP SETPOINT All0WABLE VALUE

                   $  1. R00 BLOCK MONITOR                                       W4 14 l                                 d R      a. Upscale                                      < 0.66 $t f401%                 $ 0.66 W + W 3      b. Inoperative                                 SA                              NA

c. Downscale 1 45 W of RATED THERMAL POWER 1 (3y% of RATED TIIERMAL POWER

2. APRM-

a. Flow Biased 4Simulatedf Thermal w+

Power - Upscale ' 5 0.6610<42K $ 0.66 W + 445W

b. Inoperative NA NA

c. Downscale 1 (Sy% of RATED THERMAL POWER 5 (3y% of RATED TilERf1AL POWER

d. Neutron Flux - Upscale, Startup 1 412 % of RATED TilERMAL POWER $ f14K of RATED TilERMAL POWER

3. SOURCE RANGE MONITORS

a. Detector not full in NA 3 NA Ly 5

R

b. lipscale ( 42 x 10 F cps < 4& x 105( cps

c. Inoperative .

HA NA

d. Downscale 1 43? cps 1 42f cps

{ O .

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in NA NA

b. Upscale 5 fl08/125Y of full scale 5 fl10/125Y of full scale

c. Inoperative NA NA

d. Downscale 1 45/125for full scale 1 43/125T of full scale

5. SCRAM DISCilARGE VOLUME

a. Water Level-High 5-f18F ga11ons 'N 100) validiih Nb NA^

                                                                                               ~~
                         @. Scram irip uypassea                                                          NA U. REACTOR COOLANT SYSTEM RECIRCULATION FLOW

a. Upscale 5 (not/soor of full scale 5 Q/ oof' t of full scale

b. Inoperative NA NA

c. .(ComparatorY(Oe...me:c) $ 410y% flow deviation 5 (11)% flow deviation

                      *Ihe Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W). The trip setting of this function must be maintained in accordance with Specification 3.2.2.

4$-E00d MOP O k e.d- OM dCLECL Ohd Med Clarig %p.

 @                                                       TABLE 4.3.6-1 CONTROL R00 WITilDRAWAL BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS E                                                            CilANNEL                            OPERATIONAL E                                              CilANNEL     FUNCTIONAL         CllANNEL       CONDITIONS FOR WilICll 8  TRIP FUNCTION                                CllECK         TEST          CALIBRATION I *) SURVEILLANCE REQUIRED

1. ROD BLOCK HONITOR

a. Upscale NA S/U(b) g, q ja

b. Inoperative NA S/U ,H HA 1*

c. Downscale NA S/U ,H Q 1*

2. APRH

a. Flow Biased .MimulatedP Thermal Power - Upscale NA b) H -Q-Sk I

b. Inoperative NA S/U(b),H NA 1, 2, 5 w c. Downscale NA S/U(b),g S/U ,9_ g g j 2 d. Heutron Flux - Upscale, Startup HA S/UIU),H, -q- 5 A 2, 5 Y 3. SOURCE RANGE MONITORS l a. Detector not full in NA S/U(b) (c) NA 2, 5

. b. Upscale NA S/U , -q- E 2, 5 l c. Inoperative NA S/U(b), (c) NA 2, 5

d. Downscale NA S/U -Q-R 2, 5: 4. INTERHEDIATE RANGE MONITORS; a. Detector not full in NA S/U(b) (c) NA 2, 5; b. Upscale NA S/U , -Q- & 2, 5; c. Inoperative NA S/U(b), (c) NA 2, 5; d. Downscale NA S/U , Q S 2, 5; 5. SCRAM DISCllARGE VOLUME; a. Water Level-liigh NA Q R 1, 2, 5**

(D. 5 cram irip nypassed NA H NA (1, z.) 5""J

6. REAC10R C001ANI SYSTEM RECIRCULATION FLOW

a. Upscale NA S/U ,H Q 1

b. Inoperative NA S/U(b),H HA 1

c. .(Comparatory(0m...,c;L)- NA S/U ,H Q l

l TABLE 4.3.6-1 (Continued) CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SURVEILLANCE REOUIREMENTS NOTES:

a. Neutron detectors may be excluded from CHANNEL CALIBRATION.

b. Within 24 hours prior to startup, if not performed within the previous 7 days.

c. When making an unscheduled change from OPERATIONAL CONDITION 1 to OPERATIONAL CONDITION 2, perform the required surveillance within 12 hours after entering OPERATIONAL CONDITION 2.

With THERMAL POWER >.t20)% of RATED THERMAL POWER.

   **   With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

om O O GE-STS (BWR/4) 3/4 3-52 l

REACTIVITY CONTROL SYSTEMS a - 3/4.1.5 STANDBY LIOUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 5* ACTION:

a. In OPERATIONAL CONDITION 1 or 2:

1. With one pump and/or one explos'ive valve inoperable, restore the inoperable pump and/or explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours.

2. With the standby liquid control system inoperable, restore the system to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours.

b. In OPERATIONAL CONDITION 5*:

1. With one pump and/or one explosive valve inoperable, restore the inoperable pump and/or explosive valve to OPERABLE status within 30 days or insert all insertable control rods within the next hour.

2. With the standby liquid control system inoperable, insert all insertable control rods within one hour.

3. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:

a. At least once per 24 hours by verifying that;

1. The temoerature of the sodium mentaborate solution is(sithinj_

cthe limits of Figure 3.1.5-1.PgM.o te th04 Cx equoa. to (oS

F 2.

The than available or equal tovolume of sodium pentaborate 4 43 ga11onsrand, solution frtig. la'mit showc (p is greate[p3.1

3. The heat tracing circuit is OPERABLE by determining the temperature of the 4 pump suction pipin g :o be greater than or equal to 470FF.

G *With any control rod withdrawn. Not applicable to control rods removed per v' Specification 3.9:10.1 or 3.9.10.2. O GE-STS (BWR/4) 3/4 1-19

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) g b. At least once per 31 days by: V 1. Starting both pur..ps and recirculating demineralized water to the test tank.

2. Verifying the continuity of the explosive charge.

3. Determining that the available weight of sodium pentaborate is greater than or equal to fn%f lbs and the concentration of boron in solution is within the limits of Figure 3.1.5-1 by chemical analysis.*

4. Verifying that each valve, manual, power operated or automatic, in the flow path that is not~ locked, sealed, or otherwise secured in position, is in its correct position.

c. At least once per 18 months during shutdown by:

1. Initiating one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of that batch success-fully fired. Both injection loops shall be tested in 36 months.

2. Demonstrating that the minimum flow requirement of (41.27 gpm at a pressure of greater than or equal to Ofd& T psig is met.

lat) ti40

3. Demonstratingthathhepumpreliefvalvesetpointisgreater than or equal to (10001 psig and verifying that the relief valve does not actuate during recirculation to the test tank.

( 4. ** Demonstrating that all heat traced piping between the storage

tank and the reactor vessel is unblocked by @ umping from the storage tank to the test tankT and then draining and flushing l

the piping with demineralized water. l

5. Demonstrrting that the storage tank heaters are OPERABLE by verifyir; the expected temperature rise for the sodium penta-borate solution in the storage tank after the heaters are j energized.

  #    ^This test shall also be performed anytime water or boren H adMad to the solo ,

tion or when the solution temperature drops below 6he limit of Ficure 3.1.5-1J (o5 T

      ""This test shall also be performed whenever both heat tracing circuits have been found to be inoperable and may be performed by an'y series of sequential, O      overlapping or total flow path steps such that the entire flow path is included.

GE-STS (BWR/4) 3/4 1-20 l

U i. 0 l 5 3 0 5

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) VOL4HF 30 / - 2 2. _ _ _ _ m

INSTRUMENTATION O s 3/4.3.9 PLANT SYSTEMS ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION O 3l3.9 The plant systems actuation instrumentation channels shown in Teble 3.3.9-1 shall be OPERABLE with their trip setpoints set consistent witt. the values shown in the Trip Setpoint column of Table 3.3.9-2 c ~ -.. T'J::2

     - EU A5: 2l:T:" :::T::,:: T:.:: ' . . _ . ~ .   .. x ~' '^ ' '-t              '

APPLICABILITY: As shown in Table 3.3.9-1. ACTION:

a. With a plant system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.9-2, declare the channel inoperable and either place the inoperable channel in the tripped condition until the channel is restored to OPERABLE status with its trip setcoint adjusted consistent with the Trip Setpoint value, or declare the associated system inoperable.

im ..._ w a Luroine cypass system and suppression pool spray system:

1. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place at least one inoceraole c.kannel in the tripped condition within one hour or declare the associated system inoperabis.

2. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trio System requirement for both t d ry:t:::, m . ore Lne associatea system n, w x':.

              /h For the feedwater system / main turbine trip system:                    ,

1. With the riumber of OPERABLE channels one less than required by the Minimum OPERABLE Channels requirement, restore the inoperable channel to OPERABLE status within 7 days or be in at least STARTUP within the next 6 hours.

2. With the number of OPERABLE channels two less than required by the Minimum OPERABLE Channels per Trip System requirement, restore at least one of the inoperable channels to OPERABLE status within 72 hours or be in at least STARTUP within the next 6 hours.

O .

                          ,1
           ~

GE-STS (BWR/4) 3/4 3-Er 102. l

INSTRUMENTATION O % SURVEILLANCE REOUIREMENTS 4.3.9.1 Each plant system actuation instrumentation channel shall be (' demonstrated CPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.9.1-1. 4.3.9.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months and shall include calibration of time celay relays and timers necessary for proper functioning of the instrumentation system. - 4.3.9.3 The TURBINE BYPASS SYSTEM RESPONSE TIME shall be demonstrated to less than or equal to ( ) seconds at least once per 18 months. Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of reduncant caci...el: ia a soecific trip function.

n)y/ O e

-6

~J - O 3/4 3,34 103 GE-STS (BWR/4) l

OO CD o O TABLE 3.3.9-1 , PLANT SYSTEMS ACTUATION INSTRUMENTATION

     ^                                                                                      MINIMUM           APPLICABLE i     E                                                                                 OPERABLE CilANNELS     OPERATIONAL I

TRIP FUNCTION PER TRIP SYSTEM CONDil10NS f MAlH TURBINE BYPASS SYSTEM

a. : 1

2. SUPPRES SPl!AY SYSTEM

a. Urywell Pressure-Ii 1 1, 2, 3 coincident with .

b. Suppression Pool Pressure-liig 1 1, 2, 3 i s l u c. Suppression Pool Spray Line ow 1 1,2,3

[

d. Timer, 10 minutes 1,2,3 hl 1

      -                                                  ~

2 e. Timers System Ar 1 1, 2, 3

                     ?)   System B                                                            1                 1, 2, FEEDWATER SYSTEM / MAIN TURBINE TRIP SYSTEM                                .

a. Reactor Vessel Water level-lli0h level 8 3 1 4

O O O o O TABLE 3.3.9-2 M J, PLANT SYSTEMS ACTUATION INS 1RUMENTATION SETP0lNTS d

 -s                                                                                               ALLOWABLE E  TRIP FUNCTION                                                                 TRIP SETPOINT     VALUE R

3 --- na l H I U Ku l N t- U TI'855 JIaiih N

b. N .

2. SUPPRESSION POOL SPRAY SYSTEM

a. Drywell Pressure-liigh 69) psig

b. Suppression Pool P 1 ((1.89)

                                                                                                          ) psig        psig c.

d. Suppressi - T' igh Spray Line AP-Low _ (35 5

                                                                                 > (10) psid 3(10) minutes i(

y

                                                                                                          ) psid inutes imers t'

I) System A < (12) minutes < (13.2) minute

2) System B i (14) minutes 3(15.4) minutes w -

k FEEDWATER SYST[H/ MAIN TURBINE TRIP SYSTEM { a. Reactor Vessel Water Level-fligh Leve/ B $454.5 Finches

5 456.0} inches
A See Bases Figure B 3/4 3-1. ,

I l l I . .

$)

e TABLE 4.3.9.1-1 (Continued) PLANT SYSTEMS ACTUATION INSTRtlHENTATION SURVEILLANCE REQUIREMENTS n CllANNEL OPERATIONAL E CllANNEL FUNCTIONAL CllANNEL CONDlil0NS FOR WillCil IRIP FUNCTION CilLCK TEST CAllBRA110N SURVEILLANCE RLQlilRED {

   . MAIN 10                 WSTFM                                                     -

d. - I

b. - 1 J. SUPPRESSION POOL SPRAY SYSTEM

a. Drywell Pressure-Ill. I (NA) (M) '(Q) 1, 2, 3 f

b. Suppression Po essure-liigli (NA) (M)

(Q) 1, 2, 3

c. Si ssion Pool Spray Line R

AP-Low (llA) (M) (Q) 1 ,

d. Timer (llA) (M) (R) 1, 2, 3 Y e. Iimers -(NA) (M) (R) 1, 2, 3 k 3. FEEDWATER SYSTEH/ MAIN TURBINE TRIP SYSTEM g

e

a. Reactor Vessel Water Level-liigh (NA) (H) (R) 1 t

s CONTAINMENT SYSTEMS k PRIMAP.Y CONTAIN'1ENT AVEUGE AIR TE!!PERATL'?E LIMITING CONDITICN FOR OPERATION C 135 l 3.6.1.7 Primary containment average air temperature shall not exceed (1107 F. APPLICAEILITY: OPERATIONAL CONDITIONS 1, 2 and 3. ACTION: 136 With the primary containment average air temperature greater than F, reduce the average air temperature to within the limit within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REOUIREMENTS 4.S.1.7 Theori:arycontainmentaverageairtemperatureshallbetheAV Olumt.wt.h*d Gari t.metical Maverage of tne temoeratures at the following locations and snail ::e cetermined to be within the limit at least once per 24 hours: Elevation A:imuth

a. be'- o" B*,390*

e. 80'-o " tetO'Gn c.Rb cue.h

c. 8 3'-O " a 5 .*,135 > c155 -

d. i t o' - 0" HoS '* 35O'

e. 145'-O" 55 ' a30'

f. \N ke.t-liue cf the Ve.SSel m

v ) L GE-STS (BWR/4) 3/4 6-10

PLANT SYSTEMS O 3/4.7.4 REACTOR CORE ISCLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION

]     3.7.4 The reactor core isolation coolino-(RCIC-) system shall be OPERABLE with an OPERABLE flow path capable of((automatically)J taking suction from the sup ,

pression pool and transferring the water to the reactor pressure vessel. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome pressure greater than @ psig. ISD ACTION:

a. With a RCIC discharge line " keep filled" (pressure) (pump failure) alarm instrumentation channel inoperable, perform Surveillance Requirement g _4.7.4.a.1 at least once per 24 hours.

      @      With the RCIC system inoperable, operation may continue provided the @

system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to less than or equal to @ psig within the following 24 hours. ffo i . SURVEILLANCE REOUIREMENTS t 4.7.4 The RCIC system shall be demonstrated OPERABLE:

a. At least once per 31 days by:

by m.t.-, at & W,1, pe; t

1. VerifyingM hat tne system piping from the pump discharge valve to the system isolation valve is filled with water,

2. Performance of a CHANNEL FUNCTIONAL TEST of the discharge line Delete r

                         " keep filled" (pressure) (pump. failure) alarm instrumentation, a net 2@       Verifying that each valve, manual, power operated or automatic in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

b. When tested pursuant to Specification 4.0.5 by verifying that from the cold condition the RCIC pump develops a flow of greater than or

          'l00 ~ equal to dl 60 gpm in the test flow path with a system head correspond-ing to reactor vessel operating pressure when steam is Leing supplied q                   to the turbine at normal reactor vessel operating pressure,-(1000 +

'V 20, - 80F psig.*

      "The provisions of Specification 4.0.4 are not applicable provided the surveil-lance is performed within 12 hours after reactor steam pressure is adequate i    to perform the tests.

GE-STS (BWR/4) 3/4 7-10

PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)

c. At least once per 18 months by:

1. Performing a system functional test which includes simulated automatic actuation and verifying that each automatic valve in the flow path actuates to its correct position, but may exclude actual injection of coolant into the reactor vessel, 400

2. Verifying that/the system will develop a flow of greater than or equal to 600 gpm in the test flow path when steam is supplied to t e turbine at a pressure of 4150} +-(15) psig.*

(3. Verifying that the suction for the RCIC system is automatically , l transferred from the condensate storage tank to the suppression pool on a condensate storage tank water level-low signal and on a suppression pool water level - high signal.)

   "The provisions of Specification 4.0.4 are not applicable provided the surveil-lance is performed within 12 hours after reqctor steam pressure is adequate to perform the tests.

l l O - C GE-STS (BWR/4) 3/4 7-11

INSTRUMENTATION t 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.5 The reactor core isolation cooling (RCIC) system actuation instrumenta-tion channels shown in Table 3.3.5-1 shall be OPERABLE with tneir trip set-points set consistent with the values shown in the Trip Setpoint column of Table 3.3.5-2. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3 with reactor steam dome pressure greater than M psig.

                                                    \$O ACTION:

a. With a RCIC system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.5-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

b. With one or more RCIC system actuation instrument.ation channels inoperable, take the ACTION required by Table 3.3.5-1.

SURVEILLANCE REOUIREMENTS 4.3.5.1 Each RCIC system actuation instrumentation channel shall be demon-s,r. ated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TLST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.5.1-1. 4.3.5.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. lC - O 3/4 3-42 GE-STS (BWR/4)

7 c3 T TABLE 3.3.5-1 4

    "                                                                REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENT /, TION E

E HINIMUM 2

9PERABLE CllANNEL fRTRIPSYF(EH{,) FUNCTIONAL UNITS ACTION

a. Reactor Vessel Water Level . flow Low, Level 2r 2 50 -

b. Reactor Vessel Water Level - liigh, LP 9f I O 2(b) 51 h Condensate Storage Tank Water Level - Low' (2)lCl 52

                   . Suppression Pool Water Level - liigh.                                                                  (1)(d)            52 c..,t(     Hanual Initiation                                                                                    .(11/fsystem)I[)   M 5 A,.   -

g b (a) A channel may be placed in an inoperable status for up to 2 hours for required surveillance without i placing the trip system in the tripoed condition provided at least one other OPERABLE channel in the

      "            same trip system is monitoring that parameter.

(b) One trip system with two-out-of-two logic. , *

6) One trip system with one-out-of-two looic 3

            ,(tfT Single channel.

C I 9 9 0

TABLE 3.3. 5-1 (Contiriued) REACTOR CORE ISOLATION CCOLING SYSTEM ACTUATION INSTRUMENTATION

ACTION 50 - With the number of OPERABLE channels less thac required by the Minimum OPERABLE Channels per Trip System requis ement:

a. For one trip system, place the inoperable channel in the tripped condition within one hour or declare the RCIC system inoperable.

b. " For both trip systems, declare the RCIC system inoperable.

ACTION 51 - With the number of OPERABLE channels less than required by the minimum OPERABLE channels per Trip System requirement, declare the RCIC system inoperable. D2/steg , ACTION 52 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement, place at least one inoperable channel in the tripped condition within one hour or declare the RCIC system inoperable. ACTION M - With t' .r of OPERABLE channels one less than required by Q g' 52 the Min. pn OPERABLE Channels per Trip System requirement, restore the inoperable channel to OPERABLE status within

                         -f8). hours or declare the RCIC system inoperaole.

O me

O GE-STS (BWR/4) 3/4 3-44

i OO CD o 'O i f h TABLE 3.3.5'2 M m REACTOR CORE ISOLATION COOLING SYS'4EH ACTUATION INSTRUMENTATI0i1 SETPOINTS i Gx: ALLOWABLE i j FUNCTIONAL UNITS TRIP SETPOINT VALUE ! a. Reactor Vessel Water Levc1 - (Low Low, Level 2P > - t38f inches *

                                                                                                     > -4 46F inches *

! 54.5 5b l b. Reactor Vessel Water Level - liigh,LtVP.l8 1 f5+F inches * *1 (55.57 inches

f c. Condensate Storage Tank Level - Low ->( ) Inches >( ) inches

d. Suppression Pool Water Level - liigh 1( ) inches 1( ) inches jl

e. Manual Initiation NA NA

{ ^See Bases Figure B 3/4 3-1. , Y a S t

                                                                                                     ~.               .          _ _ - _ _ _ _ _

h , TABLE 4.3.5.1-1 M m REACTOR CORE ISOLATION COOLING SYSTFM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREHENTS G y CilANNEL CllANNEL FUNCTIONAL Cl!ANNEL 2

                      - FUNCTIONAL UNITS                             CilECK          IEST      CALIBRATION l

a. Reactor Vessel Water level -

M ow Low, level 27 S H R l l b. Reactor Vessel Water S , H R Level - lii0h a

c. Condensate Storage Tank Level - Low (S) H (R)

d. Suppression Pool Water Level -

lligh (S) H (R) {

e. Manual Initiation NA M(a) NA (a) Hanual initiation switches shall be tested at least once per 18 months duri'ng shetdown. All other circuitry associated with manual initiation shall receive a CHANNEL FUNCTIONAL 1EST at least once i

per 31 days as part of circuitry required to be tested for automatic system actuation. f L. _ _ . . _ _ _ . _ _

4 PLANT SYSTEMS 3/4.7.10 MAIN TURBINE BYPA59 SYSTEli LIMITING CONDITION FOR OPERATION 3.7.10 The main turbine bypass system shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITION 1. ACTION: With the main turbine bypass system inoperable,6e in at least STARTUD 6ii tnin 6 n'ou?T 3 w.ea,; e.,e he=, mte e cAs system t. oetmaste sta: s or were, / e nsePR to be eg"*I T* **~ l'** t tb*** Vke Atth ble MC!" " it **tb**t sypass, etko. se, tnke ti.e Acisou *epwored by steenfocatoea L t.S. SURVEILLANCE REOUIREMENTS

!         4.7.10 The main turbine bypass system shall be demonstrated OPERABLE at least
;         once per:

a. 7 days by cycling each turbine bypass valve through at least one complete cycle of full travel, and

b. ss 6s ays f.

P 16 months by), performing a system functional test which includes

d (b.

simulatec automatic actuation and verifying that each . automatic valve actuates to its correct position.

2. Pe 4vo i **~)os CHnxurt. CelsBRnism of the oa*** t*

G=* bit** **

sy ste e M tunto'eu om s tr ua. eu tafr***.

5. Dem et.et Tutteras 01 Pass sysrs"r Resrouse neue ta he less %a *r- egun/ 1'e Soo mill **secouds who*/s ~

vnlve PCsItfew opeutuy to n. , tg.eivnle t to 80 % o.n ted fles,o joc./<ds* d100 mill

seca4 s delay t]>,,e . y n

Nj GE-STS (BWR/4) 3/4 7-36

POWER DISTRIBUTION LIMITS P 5 3/4.2.3 MINIMUMCRITICALPOWERRATIO(6ptional-0DYNOption) LIMITING CONDITION FOR OPERATION bm V 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of average scram insertion time and core flow, shall be equal to or greater than MCPR deter-mined from Figure 3.2.3-1 times the K, shown in Figure 3.2.3-2,4provided that the end-of-cycle recirculation pump trip (EOC-RPT) system is OPERABLE per Specification 3.3.4.2(, with:

                                   ..; ,   (Tave       TB )           .

I I A B I A = @ time limit to notch 4395 per Specification 3.1.3.3,.86F seconds, co l t g=0.688+1.65[ ]h(0.052), [N$ i=1 n T ave = =1 N4 tj , n I N. I i=1 where: n = number of surveillance tests performed to date in cycle, th Ng = number of active control rods measured in the i surveillance test, tg = average scram time t'o notch 439P of all rods measured in the i tn surveillance test, and N total number of active rods measured in Specification y = 4.1.3.2.a APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to b 425)". of RATED THERMAL POWER. O GE-STS (BWR/4) 3/4 2-6a

POWER DISTRIBUTION LIMITS POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) ACTION eN

a. With the end-of-cycle recirculation pump trip system inoperable per Specification 3.3.4.2, operation may continue and the provisions of Specification 3.0.4 are not applicable provided that, within one hour, MCPR is determined to be greater than or equal to MCPR as a function of average scram time as shown in Figure 3 2.3-1 EOC-RPT inoperable curves times the Kf shown in Figure 3.2.3-2.)

0. 2. With MCPR, as a function of average scram insertion time and core flow, less than the applicable limit determined from Figures 3.2.3-1 and 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours or reduce THERMAL POWER to less than 4 5 7. of RATED THERMAL POWER within the next 4 hours.

SURVEILLANCE REOUIREMENTS 4.2.3 MCPR, as a function of average scram insertion time and core flow, with:

a. t = 1.0 prior to performance of the initial scram time measurements for the cycle in accordance with Specification 4.1.3.2, or

b. I as defined in Specification 3.2.3 used to determine the limit within 72 hours of the conclusion of each scram time surveillance test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable limits determined from Figures 3.2.3-1 and 3.2.3@.

                                                   ~                   2.

a. At least once per 24 hours,

b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MCPR.

O - O . GE-STS (BWR/4) 3/4 2-7a

O , 1 O I l , i l l

                                      #4eu L A , A,                u L h.                  t          l 2*~*& e4                h       l i.24                 ,

g,gg l Mnw ra we_Byp- s .u~ I /

                                                       ;. ope a n                                                 .

z. l 1.50 . lq

                                                                               . .(#7.5 l.18                           ,

I : d 1 Ils i g . g. u -l , I 1.14 -- r ! . .I i l i 1.11

                                                      !                 i I                 !

1.10 , i l.08 , 1 f.OV i e e e s e s o.0 c. l o.t 0. 5 ' O.4 'd. 9, 0.lo 0.T 0.6 o. ") l.O

     -                                                              T~                                            .

PIGU2.E S.'2. 5-l 5AidtSEUM ca.iricAi. PowE R 2ATio (u c.P R.) VERsus 't A T RATEt> F Lo W

O - e e 3/4 2-?6

l ll _ 0 _ 0

                                                                            ,                        8
            .                                                                                         0 f-N                 .

0 L O o R ! s I T N O C W O L

F C w.

I 0 T 1 . A ! 0'F 7 M O T

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    ,                                                        O      L. 0 0". 0                               1 27 27                                  .F IT     C01 1 A

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                                                                     == =*

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                                                                                                           .r ll    MMMM UUUU
                                                                                                           .O
                                                                                                           .C
                             ' '                         L A O C T MMMM I l l I R TA XXxX
                                               .         T    Nt AAAA i                                                 -

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W ES D LLLLO O'.OO WW 0 O E E FFFF f 4

L D N F L UIO A T.TI U P S 4 OO t OP A C , M S i 0 3 4 ,. 2 1 0 O 1 , 1 1, 1 K w W 7 . ll[l\lll\lll '

Attachment 9 Figures Regarding Reactor Water Level Instrumentation I

-~

a

CONOENSING CHAMSER 7 --_ ____.

                                                 .___y-a O                 POETRATIONS -         %  %
                                                       - u-REFERENCE 1
                                                                                                    )

(TYPICAL) REACTOR q.- k LEG 1

PRESSURE VESSEL y 1' - - 1,[ - , I g l . VARIABLE LEG 8 l

' - n--

VARIABLE LEG DRYWELL O # dp TYPICAL WIDE RANGE l TYPICAL NARROW RANGE g TYPICAL CONTROL ROOM LEVEL INDICATOR (LI) LEVEL RECORDER (LR) OR LEVEL TRIP UNIT (LIS) FIG.1 g g COLD REFERENCd LEG DESIGN SHOREHAM NUCLEAR POWER STATION-UNIT 1 FINAL SAFETY ANALYSIS REPORT REVISION 24 - DECEMBER 19 81

kk DRYERS ^$

                                                \           d
                        ) }                       }                     STEAM NOZZLE
        ~
                                       ~

7 SEPARATORS

w w - UPPER INSTRUMENT NOZZLE 1 (228 IN. ABOVE REF.) i DRYER SKIRT Y l i I ,, ek[

                                                         -s E                    MIDDLE INSTRUMENT N0ZZLE I      l                  !  5 @,_g                    (134 IN. ABOVE REF.)

I I E l l :

            ---....,NEf                        ~~~

LOWER INSTRUMENT NOZZLE (6 IN. BELOW REF.) 7z = d 1 I I I I I tiQ -$ 5 ---- TOP OF ACTIVE FUEL .

      )                                                9 {                  (REFERENCE) 1%

1 O , is t h BOTTOM OF ACTIVE FUEL 5_3_. ___ __ _ (14 4 IN. B ELOW R E F.) hNEaL- u h JET PUMP NOZZLE p 7 JET PUMP DIFFUSER TAP } [ (216 IN. BELOW REF.)

                                        ,                9!
                      !       _li l '       a      i 7

I _ l l l iii 1 i FIG. 2 y g LOCATION OF WATER LEVEL INSTRUMENT TAPS SHOREHAM NtlCLEAR POWER STATION-UNIT 1 FINAL SAFETY ANAi.YSIS REPORT l REVISION 24 DECE:ABER 1981

,                                                                        REFERENCE CONDENSATE CHAMBER m                                                                    ____
                                                                                                                                   \

UPSET RANGE l (+)180 in. - SHUTOOWN R ANGE STEAM SPACE REFERENCE l r- CONDENSATE CHAMBER R I TYPICAL OF 2 UPSET MARROW RANGE WIDE RANGE

                                                                                + 50 in.                 + 60 in.

yp egg I OF 2 TAPS~

                                                                    --7777-777--1 7

A SCRAM l______ HI LEVEL TRIP WATER RCIC HPCI fi,,, > LEVEL \ \ u __

                                                                     -0    7 T        T T
                                                                                                                        '0 1 1 O INSTRUMENT ZERO

_j l l 1-hiTJS$gNARROW RANGE LEVEL g,yic, - N."

                             !                                                                                                              g        I UPSET /SHUTOOWN LEVEL                            INITil.TE L._                                                     LPC' LPCS 19P_of_A1T11ELUE_yJAF{ -

160in- _ _ _ ___ TYPICAL

                                                #j                     _ _ __ _ _ _ _ _ _ _j                        /                                    FUEL ZONE i                                                                                                      /

INS UMENT

                                                    ,   ,                wtDE RANGE LEVEL                                                                 ZERO TAF BOTTOM OF ACTIVE FUEL (BAF)                                                                 (-)l50in 7

T l TYPICAL ELECTRIC SIGNAL l l g __ _ J TYPICAL OF 2 TAPS FUEL ZONE LEVEL TYPICAL OF 2D/P CELLS NOTES: 1. SEPARATE O/P CELLS ARE USED FOR NARROW RANGE INDICATION AND TRIP UNITS

2. INDICATION RECORD AND TRIP UNITS FOR W10E RANGE USE COMMON D/P CELLS FIG. 3 g TYPICAL REACTOR LEVEL INDICATORS ON REACTOR CONTROL PANALS SHOREHAM NUCLEAR POWER STATION-UNIT 1 FINAL SAFETY ANALYSIS REPORT REVISION 24 -DECEMBER 1981

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NO TE: FIG. 4 ALL OIMENSIONS IN INCHES. LEV EL - ELEVATIO N ! CORREL ATION CHART SHOREHAM NUCLEAR POWER STATION-UNIT 1 FINAL S AFETY ANALYSIS REPORT 1 i REVISION 24 -OECEMBER 19 81 J

4347 ll 1 MR. ELLIS: Thank you, Judge. 2 At this point, with the panel's permission, we 3 would like to have -- I think it would be a ppropriate 4 for Dr. Burns and Dr. Joksimovich to sit in the audience 5 and proceed with the cross-examination rith the o remaining witnesses. 7 JUDGE BRENNER: Off the record. 8 (Discussion off the record.) 9 JUDGE BRENNER: All right, let's go back on 10 the record. We can continue. 11 MR. ELLIS: With the panel's permission, we 12 also have the opening statement that the Board has 13 permitted to these contentions, if the Board would like 14 that at this time. 15 JUDGE BRENNER: Right. Now, recall, it will 16 not be referenced as testimony upon which findings may 17 be based. But we appreciate it as a summary. 18 MR. ELLIS: This will be given by Mr. Dave. 19 WITNESS DAWE The testimony being presented 20 by and on behalf of the Long Island Lighting Company on 21 these contentions is directed to demonstrating that an 22 adequate, well established and well accepted methodology 23 was employed for the classification of systems at 24 Shoreham. Moreover, the testimony will demonstrate and i 25 demonstrates that, contrary to the implication of the l ALDERSON REPORTING COMPANY,INC, l 400 VIRGINIA AVE., S.W., WASHINGTON, D C. 20024 (202) 554-2345

4348 () 1 Intervenors' contention, systems interactions have been 2 taken into account in connection with the classification () 3 4 and design of the systems at Shoreham. First, section 2 of our testimony describes 5 the General Electric Company and Stone E Webster 6 Engineering Corporation experience and procedures that 7 were brought to bear on the Shoreham project. It is 8 very important to bear in mind that Shoreham was not 9 designed and built in a vacuum. It was designed and 10 built by organizations with a substantial amount of 11 design and/or operating experience. 12 Stone & Webster and General Electric have both 13 been involved with the nuclear power industry since its 14 very beginnings. Each has designed and built a number 15 of nuclear power plants now licensed and operating, some 16 of which are essentially similar to the Shoreham plant. 17 The third section of our testimony describes 18 the methodology used at Shoreham for classification of 19 systems. The various subsections of this testimony 20 identify and describe the component parts of that 21 methodology, these being the design basis analysis, 22 industry standards, more specifically ANS-22, the 23 General Electric nuclear safety operational analysis, ( 24 Regulatory Guides 1.26 and 1.29, applicable Nuclear 25 Regulatory document, formerly Atomic Energy Commission l l ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE.,3.W., WASHINGTON. D.C. 20024 (202) 554 2345

4349 () 1 r eg ul a tion s , principally 10 CFR Part 100, and Appendix A 2 to 10 CFR Part 100, 10 CFR Part 50, Appendices A and B 1 () 3 4 specifically or most specifically, and 10 CFR'Part 50.55(a). 1 5 And finally, included in this section is a 6 reference to the experience of the organizations 7 involved in applying these component parts to the 8 methodology. Contrary to the Intervenors' contention, 9 this methodology is not inadequate; it is comprehensive 10 and systematic, and it constitutes essentially the same 11 methodology that has been successfully used to design 12 and license other boiling water reactors. 13 Design bases as described in the regulations 0 14 means that information which identifies the specific 15 f unctions of a structure, system or component, and the 16 values of parameters chosen for the reference design. 17 This information may be restraints derived from 18 generally accepted sta te of the art practices for 19 achieving functional coals or it may be derived from 20 analyses of the effects of postulated accidents for 21 which those structures, systems and components must meet 22 functional goals. 23 The design bases of structures, systems and 24 components for Shoreham were established using the 25 methodology described in this section of the testimony, O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4350 () 1 leading to their classification. As has been the 2 practice throughout the indust ry , the application of 3 this methodology is founded on the principle of defense 4' in depth. 5 This leads to a plant design which is 6 engineered for highly reliable operation. The plant is 7 designed in a sound, conservative manner, so that it can

 ,     8  be built, tested , opera ted and maintained in accordance 9  with accepted quality standards and engineering 10  practices.

11 This reliability goal minimizes the chance of 12 malfunctions occurring, and despite this first level of 13 assurance the plant is then protected against 14 operational transients due to equipment failures or 15 malfunctions, so that these are safely and properly 16 accommodated in the plant design. 17 The third level of defense in this defense in 18 depth concept provides multiple backups such that no 19 undue risk is presented to the public as a result of 20 postulated but highly unlikely accidents. The analysis 21 -- analyses of these hypothetical events establishes the 22 design basis accidents for which the saf ety-related 23 equipmen t is designed. 24 Section 4 of the testimony relates to the 25 treatment of non-safety-related structures, systems and ( ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

l 4351 () 1 components at Shoreham. An indication of Intervenors' 2 prepared testimony and responses to cross-examination is 3 that structures, systems and components at Shoreham that 4 are not safety-related received little or no care in 5 their design and structure. In this section of the 6 testimony it reflects that this is simply not the case. 7 Structures, systems and components that are 8 not safety-related at Shoreham do receive design and 9 quality assurance attention commensurate with their to functions, whatever those functions may be. They are 11 not designated safety-related because they are not 12 relied upon to perform a sa fety-related function. 13 Section 5 of the testimony is a discussion of 14 the consideration of systems interactions at Shoreham. 15 The section includes both Shoreham-specific studies and 16 generic studies applicable to Shoreham, all of which 17 have been conducted in support of the design process. 18 These studies include both deterministic and 19 probablistic type studies. 20 This section f urther demonstrates that systems 21 interactions have not been ignored in the design and 22 construction of the Shoreham plant. 23 24 i

25 O

ALDERSON REPORTING COMPANY,INC,

4352 () 1 The sixth section of the testimony is 2 presented in three parts. The first part is the 3 testimony of Dr. Joksimovich, a member of the peer 4 review group for the Shoreham PRA. Dr. Joksimovich in 5 his testimony reviews PRA methodology in general, and 6 the PRA methodology specifically employed in connection 7 with the Shoreham PRA. His testimony clearly reflects

p 8 that the Shoreham PRA cor.siders systems interactions at 9 Shoreham and has used such techniques as event tree and 10 fault tree analyses and dependency matrices.

11 The second portion of this section of the 12 testimony is authored by Dr. Burns, one of the 13 individuals actually involved in the conduct of the PBA O 14 for Shoreham. His testimony also makes clear that the 15 Shoreham PBA considers systems interactions and uses 16 such state-of-the-art techniques as fault and event tree 17 methodology and dependency matrices. 18 The third part of this section of the 19 testimony is the testimony of Mr. Kascsak, an employee 20 of LILCO. Mr. Kascsak's testimony is direct 1,y 21 responsive to the Board's request for information f rom 22 LILCO cencerning its intended use of the Shcreham PRA, 23 its current status and its schedule for completion. 24 The seventh section of the testimony is also 25 presented in several parts. The first part is authored O ALDERSON REPORTING CCMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4353 () I by Mr. Paul McGuire, a consultant to long Island 2 Lighting Company, having extensive operating experience 3 on boiling water reactors. Mr. McGuire's testimony 4 responds to the implications in intervenor's testimony 5 that the'use of non-safety related equipment and 6 emergency operating procedures implies an improper 7 classification of that equipment. 8 His testimony shows that the use of the 9 equipment referred to by intervenors is a sensible and 10 logical approach to be taken in emergency Operating 11 procedures. It shows that their use as described in 12 these emergency operating procedures is not a basis for 13 changing their classification. O 14 The inclusion of non safety-related systems in 15 emergency operating procedures is based on the sensible 16 principles that operators should be directed to use the 17 full capabilities of the plant in dealing with l 18 transients and other events, because these normal non 19 safety-related systems are required, by design, to be 20 highly reliable, and their use in these circumstances 21 will often make an unnecessary cha11ence or call upon 22 the safety-related systems of the plant. 23 The second part of Section 7 responds to ( 24 improper cla ssifica tions alleged by the intervenors for 25 four specific systems or portions of systems at the O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON D.C. 20024 (202) 554 2345

l 4354 () 1 Shoreham plant. These are the rod block function, the 2 RCIC systes, the high water level trip, and turbine 3 bypass system. This part of the testimony addresses 4 each of there examples in turn and demonstrates not only 5 the correctness of the classification, but also, the 6 adequacy of design and construction of these portions of 7 the plant for their intended function. 8 The third part of this section addresses the 9 consideration of systems interaction in the reactor 10 water level instrumentation. This part of the testimony 11 demonstrates that this matter has been thoroughly 12 studied at Shoreham and presents no problem- In 13 addition to demonstrating the inherent capability of 14 water level instrumentation to withstand the interactive l 15 effects of drywell environment, the testimony further 16 demonstrates the existence of operating procedures and 17 technical specifications in place at Shoreham to avoid 18 or mitigate these effects. i 19 The fourth part of this section a ddresses the 20 classification of the standby liquid control system, 21 demonstrating its correctness for its intended functions 22 in the design basis of the plant. And the final two 23 portions of this section specifically address alleged () 24 inconsistencies, alleged inadequate level of detail of 25 FSAR Table 3.2.1-1. Specific responses to the O ALDERSON REPORllNG COMPANY,INC, 400 VIRGINIA An'E., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4355 O i intervenor s =octentions in these ere s are provieed in 2 these final two sections of the testimony. 3 MR. ELLISa So that completes our statement, 4 Judge, and we are now -- the panel is now ready for 5 cross examination. 6 JUDGE BRENNER s Just before we do that, as I 7 look at the documents I have received -- I suggested 8 might be helpful to also bind in the listing of the 9 additional FSAR sites for LILCO's testimony, since

   ~10 normally we have just given them orally.                     But since 11 there is a list of four, I think we should just bind it 12 in at this point. And in the absence of any objection, 13 let's do that.

O 14 (The listing of additional FSAR sites for 15 LILCO testimony follovss) 16 17 18 19 20 21 22 23 24 25 O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE.. S.W., WASHINGT" N, D.C. 20024 (202) 554-2345

o a 3 t/G i' UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board

    \, J In the Matter of                      )
                                                      )

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

(Shoreham Nuclear Power Station, ) Unit 1) ) ADDITIONAL FSAR CITES FOR LILCO'S TESTIMONY ON SC/ SOC 7B AND SOC 19(b) Because of the broad nature of these contentions, much of the testimony relates to some part of the FSAR. Where a por-tion of the FSAR was particularly relevant, the section was

   , s          cited by the authors. In reviewing the testimony, LILCO believes several additional references may be helpful.

(1) The results of the reactor vessel water level instrumentation study referred to in Section V(p) and VII.C of the testimony are found in the PSAR as a response to NRC Staff question 223.91. (2) The Rod Block Monitor discussed in Section VII.B.1 of the testimony is discussed in FSAR Section 7.7.1.1. (3) The RCIC System discussed in Secti.on VII.B.2 of the testimony is described in FSAR Section 5.5.6. (4) The Standby Liquid Control System discussed in l

/3

() Section VII.D of the testimony is discussed in FSAR Section 4.2.3.4. l l

l , Respectfully submitted. LONG ISLAND LIGl! TING COMPANY 4

                                          /              .       n   _          ,

{( r i.s[ ( ((f _ '(g T. S. Ellin[ II'I J/j/' Anthony F.Carley, Jr. f g/ I llunton & Williams Post Office Box 1535 l Richmond, Virginia 23212 DATED: Juno 15, 1932 l 1 l l l 1 O

4356 (]) 1 JUDGE BRENNER: In addition, one thing I meant 2 to do person ally to help me keep track of the testimony 3 up here was to run additional copies of a few of the

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NJ 4 pages, a nd I am wondering if that could be done in the 5 next day or two for the four 'of us up here. That would 6 be the attachments list, the authorship attribution to 7 page list, and then the table of contents, because I am 8 afraid if I pull them out I will loose them in the long 9 run. I can keep track better with that. 10 MR. ELLIS: We vill do that, Judge. 11 JUDGE BRENNERa Thank you. I meant to do it 12 and forgot this week. 13 Nr. Lanpher, I only have the written cross 14 examination plan that you gave me on Friday. Am I 15 missing another one? You promised you would revise it. 16 I do not want to necessarily hold you to that. 17 MR. LANPHER: I do have a revision -- I did 18 not copy it this morning -- of a portion of it, and part 19 of it -- all of it is revised, but not in a form to give 20 to the Board. Actually, the portion that has been , 21 retyped and revised I will give to you af ter lunch. I l l 22 will not be addressing that portion initially, so I do 23 not think I will put you at that d isad va n ta ge . () 24 JUDGE BRENNER: All right. I was just 25 wondering if you gave me something I was missing, was i (:) ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4357 () 1 the main reason for the inquiry. Why don't you proceed? 2 MR. LANPHER: Judge Brenner, just so that 3 people are somewhat -- know where we are going a bit, 4 Ms. Letsche is going to be handling th e PR A section. We 5 will consult with Mr. Ellis and see if there is a means l l 6 of accommodating schedules or whatever. And she may be i 7 handling another section also. We have not decided that 8 at this time. 9 I as intending to also ask some preliminary to questions relating to professional qualifications, but I 11 intended to do that rather quickly at this point. And 12 given the multiple authorship, I may come back to , 13 professional qualifications as to particular portions. l (Z) 14 JUDGE BRENNER: All right. 15 CROSS EIAMI'? ATION 16 BY MR. LANPHER: 17 0 Mr. Dave, when did you first start work on the 18 Shoreham nuclear project? 19 A (WITNESS DAWE) I first started work on the l 20 Shorham nuclear project very shortly after starting l 21 working at Stone C Webster in 1973. 22 0 Is it f air to say that by that time, the 23 project had essentially been designed, at least with ( 24 respect to its basic structures and components and 25 equipment? O ALDERSON REPORTING COMPANY,INC, ! dm VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4358 () 1 A (WITNESS DAWE) The basic conceptual designs 2 had been in place for the construction permit 3 app lica tian . That is true. But a great deal of the 4 engineering and design of the plant for construction 5 occurs af ter that bssic design is done. So there was a 6 great deal of the engineering and design going on in 7 1973 when I was first associated with the project. 8 Q You stated, sir, tha t the basic conceptual 9 design was essentially in place by the time you started 10 on the project. Would that include basic decisions as 11 to classification of systems, structures and components? 12 A (WITNESS DAYS) There, in fact, were changes in 13 the method of classifying systems, structures and O 14 components which you can see if you compared the final 15 safety analysis report from -- with the preliminary 16 safety analysis report. The early classification 17 methodologies were demonstrated in the preliminary 18 safety analysis report in the timefrase when I first 19 sta rted working on this project. Additional work was, 20 in fact, going on relative to the classification of 21 systems, structures and components. 22 A (WITNESS GARABEDIAN) I would like to add to 23 tha t a little bit. What was existing at that time was ( 24 far different from what you see now. For the CP 25 application, the requirements were basically to identify O ALDERSoN REPORTING COMPANY,INC, 400 VIRGIN!A AVE., S.W., WASHINGTON, D.C. W24 (202) 554-2345

4359 () 1 seismic category 1 systems and structures. 2 (Panel of witnesses conferring.) 3 0 Did that complete your supplementation? 4 A (WITNESS GARABEDIAN) Yes. 5 Q Let me just follow up on that. You said at 6 that time -- we are talking about 1973? In that time 7 period? 8 A (WITNESS GARABEDIAN) No, I as talking about -- 9 well, 1972 to 1973 -- prior to 1973. 10 0 So prior to 1973, what had been done was to 11 identify those systems, structures and components that 12 needed to be designed to withstand the effect of the ( 13 safe shutdown or design basis earthquake, is that ( 14 correct? 15 A (WITNESS GARABEDIAN) Yes. 16 0 Had a decision been made at that time -- 17 strike that. 18 Then is it also fair to state that as of that 19 time, decisions had not been made regarding the quality 20 assurance requirements that would be applied to 21 particular systems, structures and components? l l l 22 (Panel of witnesses conferring.) 23 A (WITNESS GARABEDIAN) The Q A requirements for 24 Appendix B I believe came out -- when was that? In l l 25 1970. The basis of the PSAR was pre-1970.

ALDERSON REPORTING COMPANY,INC, l 400 VIRGINTA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4360 O i o We11, I thought we were ta,.xing, sir, about 2 the 1972 to 1973 time period, and I am trying to get a 3 grip on what the classification system was -- 4 classification, quality contro1 and tha t . I believe you 5 testified that as of that time, the seismic categories 6 had been decided. Correct? 7 A (WITNESS GARABEDIAN) Yes. 8 Q Is it fair to say that the qua11ty ascurance 9 categories had not been decided? 1C A (WITNESS GARABEDIAN) No. I was talking about, 11 when I said the seismic categories had been determined, 12 pub'ished in terms of what was then the PSAR. _ When the 13 project restarted, in order to accommodate, you know, 10 0 14 CFR Appendix B and Appendix A, we reclassified the 15 systams to delines te Q A requirements, seismic 16 requirements, code requirements and the like. 17 0 Now, had that been done by late 1972, early 18 1973? 19 A (WITNESS GARABEDIAN) It had been done before 20 we got the CP. I mean it had been initiated before we 21 got the CP, in order to take into account the new codes 22 and regulations that were issued in 1970, 71, 72 23 timeframe. 24 0 What new codes are you referring to, sir? 25 A (WITNESS GARABEDIAN) The ASME-3 code, as well O ALDERSoN REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

4361 () 1 as the regulations. 2 Q Are the regulations that you are referring to [') v 3 Appendix B to Part 50? 4 A (WITNESS GARABEDIAN) Yes. And 10 CFR 50.55. 5 (Counsel for Suffolk County conferring.) 6 0 Was another thing that you were having to 7 change for 10 CFR 50.55(a)? 8 A (WITNESS GARABEDIAN) Yes, sir. 9 0 Maybe you just said that and I just had not 10 heard you. So it would be fair to say that by the time 11 the construction permit was issued, the seismic category l 12 1 structures and components had been identified as well 13 as those systems, structures and components as to which O 14 quality assurance category 1 requirements would be 15 applied. 16 A (WITNESS GARABEDIAN) Yes, sir. 17 0 Mr. Dawe, turning your attention to your 18 Statement of Prof essional Qualifications, -- l 19 MR. ELLIS: Could I have a clarification? 20 Could I have that last question and answer read back? I 21 am not sure I understood it. 22 JUDGE BRENNER: All righ t. 23 (The reporter read the record as requested.)

   ) 24             MR. ELLISs     I think there was an ambiguity in 25  the question as I understood the question, and maybe I O

ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4362 1 should have clarified the ambiguity before the answer, 2 rather than stating the ambiguity. 3 JUDGE BRENNERa Well, I think you are in the 4 realm of fixing it on redirect if you think you need to, 5 unless Iou think a lonq+line is going to go in the wrong 6 direction. 7 MR. ELLISa I do, and that is why I think the

8 witnesses should have an opportunity to clarify. 9 WITNESS DAWEa I think the original question 10 -- 11 JUDGE BRENNER Well, wait. 12 MR. LANPHER: Judge Brenner, this is a strange 13 p ro ced o re . 0 0 14 JUDGE BRENNER: All right, hold it. I did not 15 hear an objection. The witness felt he was able to 16 answer the question. I have discouraged objections as 17 to ambiguity of questions throughout the proceeding 18 believing that it is the witness who can best determine 19 it within reason. Admittedly, sometimes a question is 20 very bad in the sense of ambiguity, and the witness 21 might not perceive it, and we could save the next three 22 hours if we clear it up. 23 But speaking for myself, I did not hear 24 anything so terribly ambiguous in that question that it 25 could not be answered yes or no. O ALDERSoN REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON D.C. 20024 (202) 554-2345

i 4363 () 1 MR. ELLIS: Well, then, rather than state it, 2 I would lite to have the witnesses have an opportunity, l 3 if one of the other witnesses wants to add something, 4 which has been a typical practice of the panels. 5 JUDGE BRENNER: Yes, but not typical with all 6 the prompting we have now had. Let's just proceed, and 7 you can fix it on redirect in this instance. 8 MR. ELLIS4 But I think -- 9 WITNESS DAWE I would like a chance to -- 10 JUDGE BRENNERs Hold it, Mr. Dave. Let's 11 proceed. 12 BY MR. LANPHER (Resuming)s 13 0 Mr. Dawe, on your resume in the first page O 14 about halfway down you say that your duties include 15 assuring project awareness of regulatory developments. 16 Can you explain what that duty or those duties entail? 17 A (WITNESS DAWE) Yes, sir. As the supervisor of 18 project licensing, I am responsible for the supervision, 19 both technically and administratively, of licensing 20 personnel from the Licensing Division assigned to the 21 various projects in the Boston Stone C Webster office. 22 That statement that you referred to is the 23 portion of my duty which is to make sete that the lead ( 24 engineers, and in turn, the support engineers on the 25 project are aware of the current licensing status in the O ALDERSON REPORTING COMPANY,'NC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4364 () 1 industry and in the eyes of the NBC. 2 In other words, I assure that changes to 3 regulation, appropriate Federal Register notices and so 4 forth are made available to the people on the projects. 5 0 How do you do this, Mr. Dave? Are you sort of 6 a clearinghouse on all developments from industry and 7 the NRC? 8 A (WITNESS DAWE) Sometimes I feel like a 9 clearinghouse, but in fact, I have a large 10 organization. It is not my organization; it is the 11 Stone & Webster organization of which I an a part. 12 There are people who receive these things. There are 13 people who make the distributions for us based on O 14 pre-determined lists which I have an input to and so 15 forth. I review many of these things myself and discuss 16 many of them myself with the specific lead engineers. 17 0 You have used that term twice, lead 18 engineers. What is a lead engineer? ! 19 A (WITNESS DAWE) A lead engineer -- 20 0 Let me finish. Generally, what would lead 21 engineer's duties be? And if we can tie it to the 22 Shoreham project, that would be helpful. l l 23 A (WITNESS DAWE) A lead engineer is a term ( 24 describing a person's position in a project 25 organization. The lead engineer for a pa rticula r ; O ALDERSoN REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W WASHINGTON, D C. 20024 (202) 554 2345

4365 O ' disciatine o stone c vehster orosect is the senior 2 representative f rom a particular technical discipline 3 assigned to that project who is responsible for the 4 proper conduct of that discipline's activities on the 5 project. 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 l 23 24 25 1 I O l l ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

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

4366 () 1 Q On the Shoreham project what are the different 2 disciplines of lead engineers that Stone and Webster has? 3 A (WITNESS DAWE) Well, there are a large number 4 of them. Just to list several, there would be a lead 5 power engineer, a lead structural engineer, a lead 6 instrumentation and controls engineer, a lead electrical 7 engineer, a lead materials engineer, a lead 8 environmental engineer, a lead geotechnical engineer. 9 These are all components of the Stone and Webster 10 organization represented on the project. 11 Q Do they all report to you, Mr. Dave? 12 A (WITNESS DAWE) No, sir, they do not.

13 0 Who do they report to?

t {/ S 14 A (WITNESS DAWE) Each lead engineer reports on 15 the project to the project engineer, and additionally he 16 reports to his division to somebody comparable to myself. 17 Q You are a project engineer with overall j 18 responsibility for Stone and Webster's activities, say 19 for Shoreham, is that correct? l 20 A (WITNESS DAWE) Yes, sir, for Stone and 21 Webster's engineering activities for Shoreham. 22 0 Mr. Dawe, you said you started with Stone and l 23 Webster on the Shoreham project sometime in 1973. What () 24 work did you perform starting in 1973 on the Shoreham 25 project? Os/ ALDEP. SON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

1 1 4367 1 () 1 A (WITNESS DAWE) In 1973 without being directly 2 assigned responsibility to the Shoreham project which ; 3 occurred in January 1974 when that project was assigned 4 to me for licensing, I was assisting another licensing 5 engineer in doing review of the Shoreham licensing 6 sta tus because the project had experienced a long CP 7 construction stage, starting in 1970 and not completed 8 until the CP was issued in 1973. 9 At that point in time I was reviewing the 10 initial decision from the ASLB at the construction 11 permit stage. I was reviewing and comparing general 12 design criteria which were then in existence in Appendix 13 A to 10 CFB 50, but which had not been in existence at O 14 the time the Shoreham PSAR was completed and filed. It 15 was a draft version that existed in the 1967 time 16 frame. So in the first few months in 1973 I was 17 starting to do that work for Shoreham. 18 0 You said you compared the th en-e xisti ng GDC to 19 the PSAR which had not specifically addressed certain of 20 those. i 21 A (WITNESS DAWE) No, sir, I did not say th a t . i 22 0 Well, was it your testimony, Mr. Dawe, that 23 certain of the GDC had not been specifically addressed 24 in the PSAR? 25 A (WITNESS DAWE) No. Let me resta te what I O ALDERSON REPORTING COMPANY,INC, 400 VIRGIN!A AVE., S.W., WASHINGTON. D.C. 20024 (202) M54 2345

4368 () 1 have said and try to clarify for you. 2 Ihe general design criteria which appear in 3 Appendix A to 10 CFR Part 50 were not placed in the code 4 of federal regulations until 1971, I believe. The 5 Shoreham PSAR was filed first in 1968. The Shoreham 6 PSAR did address general design criteria but not the 7 version that appeared in 10 CFR Part 50 Appendix A 8 because that did not exist. However, at that time, in 9 the late 1960s, the AEC did have available draft 10 versions of what became 10 CFR 50 Appendix A, and those 11 were addressed in the PSAR. 12 0 So the PSAR addressed the draft GDC, and was 13 it your purpose in your initial work on the Shoreham O 14 project to review how well the PSAR would apply to the 15 GDC which were finally adopted? 16 A (WITNESb DAWE) Initially I think it could be 17 more clearly stated as to compare the two versions of 18 what could be called the general design criteria for how 19 they had changed. 20 0 You performed this comparison. What was done 21 with this comparison? 22 A (WITNESS DAWE) It was used as a piece of 23 working information along with a lot of other things () 24 tha t were being done at that time to assess the 25 licensing status of Shoreham relative to current O At.DFRSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345 1.

4369 l () 1 regulations in the 1973-74 time frame. 2 0 Did this review result in any changes in the 3 classication of any structures, systems or components? 4 A (WITNESS DAWE) I would say yes, it did. Not 5 the conceptual design of the plant changed, but again as 6 I mentioned earlier when you first asked about 7 classifications of systems, the classification 8 terminology has changed substantially, if not tae basic 9 subset of safety-related versus not safety-related 10 equipment, structures and components. 11 I think you can best characterize the review 12 of current regulation at that time when the project -- 13 as being something that we normally do throughout a 0 14 project just to confirm constantly that we are meeting 15 our regulatory requirements. 16 0 In this comparison did you compare the PSAR or 17 portions of it with Regulatory Guide 1.26? 18 A (WITNESS D AWE) No, I would not. No. The 19 work I was talking about was not a comparison to 20 Regulatory G uide 1.26. 21 0 Did you compare it to ANS-22, as I guess it 22 was c?lled back then? 23 A (WITNESS DAWE) I think that yes, I have 24 performed comparisons to ANS-22, but the subject we are 25 talking about is a comparison of the general design O e ALDERSON REPORTING COMPANY,INC, 400 ViRG!NIA AVE., S.W., WASH'NGTON. D.C. 20024 (202) 554 2345

4370 l () 1 criteria at two points in time. And the particular 2 activity that I was referring to at the sta rt did not I 3 involve the comparison with ANS-22. 4 0 If I understood your testimony earlier, this 5 was sort of an initial project for you on Shoreham 6 before you were formally assigned to the Shoreham 7 project, and then you went on to other jobs with respect 8 to Shoreham, is that correct? 9 A (WITNESS DAWE) Yes, that is correct. One of 10 my first tasks was this comparison we have been talking 11 about. I was formally assigned to the Shoreham project 12 as the licensing engineer for Shoreham in 1974 in 13 January. We are only talking a matter of about three 14 months because I was only with Stone and Webster af ter 15 August of 1973, and from that time forward I have had 16 involvement with the Shoreham project. 17 0 Now, you were a licensing engineer at the 18 Shoreham project from 1974 to sometime in 1976. 19 A (WITNESS DAWE) Yes. 20 0 What were your responsibilities during that 21 time period? 22 A (WITNESS DAWE) During tha t time period I was 23 responsible for advising the project and consulting with ( 24 people on the project relative to licensing commitments 25 that had been made, licensing commitments that were O ALDERSON REPORTING COMPANY. INC, 400 VIRGINIA AVE., S.W., WASHINCTON, D C. 20024 (202) 554-2345

4371 () 1 being made, and the proper attention to regulation by 2 the project. 3 Also included in that period of time was the 4 preparation of the final safety analysis report and 5 environmental report, the preparation of the operating 6 license application for the Shoreham project. 7 0 During this time as a licensing engineer were 8 you involved personally in the classification of 9 systems, structures or components? 10 A (WITNESS DAWE) Yes, sir. 11 Q In what manner? 12 A (WITNESS DAWE) Well, as a licensing engineer 13 on the project daily duties would be conferring, 14 consulting with all members of the project. During that 15 period of timo questions of classification would be 16 raised, and I would be involved in meetings or 17 researching regulation or helping people interpret the 18 requirements of regulatory guides, and that is part of 19 the function of assisting in the classification of 20 com ponents in the plant. 21 Q Was it part of your responsibility +o, for . 22 instance, perform analyses to determine whether a 23 particular system would be called upon to perform a 24 safety function? 25 A (WITNESS DAWE) No, sir. I was not O ALDERSON REPORTING COMPANY,INC. l 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4372 () 1 particularly responsible for those - performance of 2 those analyses. 3 Q But in your job as licensing engineer you 4 would consult with the people that performed such 5 analyses and to ensure -- well, I'll leave the question 6 at that. 7 A (WITNESS DAWE) Yes, sir, I consulted with 8 people who performed those analyses, primarily with 9 respect to the final safety analysis report. 10 Q And is it true that you participated actively 11 in preparntion of the final safety analysis report, 12 particularly Section 3.27 13 A (WITNESS DAWE) Yes, sir. I

   )   14            From 1976 to 1980 you were the lead licensing 0

15 engineer, correct? 16 A (WITNESS DAWE) Yes, sir. 17 0 Did that change your responsibilities? 18 A (WITNESS DAWE) It increased my 19 responsibilities without really changing them. Prior to 20 1976 the licensing engineer on the project reported to 21 the lead power engineer, licensing being a portion of 22 the Power Division. In 1976 the Licensing Division was 23 formed, and I maintained my role and responsibilities on () 24 the project but then became the lead li cansing engineer. 25 After 1976 there would be very little O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W WASHINGTON, D.C. 20024 (202) 554-2345

4373 () 1 difference in my duties and responsibilities except in 2 an administrative nature.

/"

V) 3 0 You worked on preparation of the FSAR both as 4 a licensing engineer and as the lead licensing engineer. 5 A (WITNESS DAWE) The enttre preparation period 6 of time my title would have been licensing engineer. 7 The FSAR was filed for acceptance review in 1975. After 8 1976 we were continuing to maintain the FSAR. The plant 9 was then going through the NRC's review period, and I 10 had the same responsibilities toward the FSAR. It is 11 only that my titles changed. 12 0 Mr. Robare, I would like to direct several 13 questions to you. O 14 Have you ever personally classified systems, 15 structures or components? I 16 A (WITNESS ROBARE) I have not personally 17 classified them in an engineering sense. I have been 18 involved in the classification process, however. l i 19 0 Have you been involved in the classification 20 process in the sense of reviewing the classification 21 work that others have performed? 22 A (WITNESS ROLARE) That is correct. I am also 23 involved in the establishment of regulatory requirements l () 24 for our engineering departments. 25 Q Did you personally assist or participate in .O i j ALDERSON REPORTING COMPANY,INC, I

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4374 () 1 the design of any systems, structures or components for 2 the Shoreham project? 3 A (WITNESS ROBARE) I am not sure what you mean 4 by assist in the design. 5 Q Well, then, let's delete that. Did you 6 participate in the design? 7 A (WITNESS ROBARE) In my position as the l l 8 licensing engineer for Shoreham I have participated in 9 the design, yes. l 10 0 When did you first participate in design 11 activities relating to Shoreham? 12 A (WITNESS ROBARE) April of 1976. j 13 Q Is it fair to state that by that point in time I i 14 the basic design with respect to the GE scope of desiva 15 was pretty well completed? 16 A (WITNESS B0 BARE) The design had been 17 escablished in terms of the basic number of systems, 18 yes. I do not know how to quantify; the basic design 19 was completed. 20 0 Well, for instance, by that point in time had 21 GE determined what structures, systems and compenents 22 which it was supplying should be classified as seismic 23 category 1? () 24 A (WITNESS B0 BARE) Yes. 25 0 And is it true that within the GE scope of O ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D C. 20024 (202) 554 2345

4375 () I supply, when GE classifies an item as seismic category 1 2 that also means that the full Part 50 Appendix B quality 3 assurance requirements apply in GE's view to those items? 4 MR. ELLIS: May I have that question read back 5 to me, please? I think the same aabiguity is creeping 6 in again. 7 JUDGE BRENNER: It is so ambiguous that I do 8 not see it creeping in. 9 Can you repeat the question, Mr. Lanpher, or to would you prefer to have it read back? 11 MR. LANPHER: Let me rephrase it. I cannot 12 repeat it word for word. I do not mind rephrasing it. 13 BY HR. LANPHER: (Resuming) 14 0 When GE classifies in item as seismic category 15 1 does that also mean that GE applies the full Appendix 16 B to Part 50 QA requirements to that item? 17 JUDGE BRENNER: Hold the answer. 18 Do you object to that question? 19 MR. ELLIS: No, I do not object to that 20 question. That was not the one -- was the previous one. 21 JUDGE BRENNER: All right. You may answer the 22 question. 23 WITNESS ROBARE: That is generally true. () 24 BY MR. LANPHER: (Resuming) 25 0 When is it not true? O ALDERSON PEPoRTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON D C. 20024 (202) 554 2345 m

4376 ( ,) 1 A (WITNESS ROBARE) There are certain exceptions 2 to that rule. I would have to take some time to

  ,/~l  3  evaluate that to give you the exact exceptions.

4 0 How long will that -- I do not want to delay 5 this, but I mean can you generally explain what those 6 exceptions would be? And by the way, Mr. Robare, I know 7 you have other people from GE on the panel. If someone 8 else -- I am asking you, but I am not trying to single 9 you out. 10 (Panel of witnesses conferring.) 11 A (WITNESS ROBARE) For componen ts that were 12 purchased or designed significantly prior to the 13 issuance of Appendix B where General Electric selected 14 them to be safety-related components it is possible that 15 to literally say they meet Appendix B is not possible. 16 However, the quality assurance that was applied could be 17 said to meet generally all the requirements of Appendix l 18 B, but maybe not literally the Appendix B label. So i 19 those components would be seismic category 1 but not 20 necessarily full Appendix B quality assurance.f i l 21 0 Is that the only exception to my prior l 22 question that you can recall at this time? l 23 A (WITNESS ROBARE) Yes. (/ 24 0 Has GE performed an analysis or compiled a 25 list or something of that nature as to what components ) l ALDERSON REPORTING COMPANY. INC, l 400 VIRGINIA AVE , S.W,, WASHINGTON. O C. 20024 (202) 554 2345 i

                                                                                                                                                     '4377

() 1 purchased prior to enactment of Appendix B which may not 2 literally meet the full Appendix B requirements?

 /'                                     3             A (WITNESS ROBARE)                                 Not that I am aware of.

4 (Counsel f or Suf f olk County confe rring. ) 5 0 Mr. Ga rabedian -- is that the right 6 pronunciation? 7 A (WITNESS GARABEDIAN) Yes, sir. 8 0 I would like to direct some questions to you. 9 What are your current responsibilities at 10 Stone and Webster, sir? 11 A ( W IT N ES.: GARABEDIAN) I am a project manager 12 at Stone and Weaster for 1MFBR activities, specifically 13 that relate to the effort that we have going on with the () 14 Department of Energy on their large fast breeder design i 15 program, and the support effort that we have will be 16 National Nuclear Corpcration, NNC, on their large 17 breeder design. 18 Q Anide from presenting the instant testimony l 19 and helpin; to prepa re it, obviously, do you have any 20 present responsibilities with respect to the Shoreham 21 project? 22 A (WITNESS GARABEDIAN) No. 23 0 From your resume you have had responsibilities l ( 24 with respect to Shoreham in the past. When was your l 25 last -- when were yo ur last activities with respect to O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4378 O 1 Shoreham except for this testi:nony? 2 A (WITNESS GARABEDIAN) Oh, 1977, 1978 I last 3 reviewed the Shorehsm project. 4 5 6 7 8 9 10 11 12 13 14 15 1 16 17 18

19 20 21 22 23 24 25 O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. O C. 20024 (202) 554-2345

4379 () 1 MR. LANPHER: Judge Brenner, if I could ask, 2 if you could move that a little closer. I am having a 3 hard time hearing you, sir. 4 WITN ESS G AR ABEDI AN: I apologize. BY MR. LANPHERs (Resuming) 6 0 In 19,77 or 1978, what were these activities I that you were involved in with respect to Shoreham? 8 A (WITNESS GARABEDIAN) Well, it was trying to 9 understand, you know, what was happening to Shoreham 10 with all these new regulations and requirements, and 11 trying to, you know, come up with an analysis or coming 12 up with an analysis on which guides, you know, were 13 issued and when they were issued, and how Shoreham had 14 really done a cuper job in accommodating issues that, 15 you know, had occurred in the 1970's. 16 It is an amazing story. 17 0 Why was this review performed? 18 A (WITNESS GARABEDIAN) To assist Stone C 19 Webster management in examining Shoreham from a broad 20 perspective. 21 Q What do you mean, "from a broad perspective," 22 sir? 23 A (WITNESS GARABEDIAN) Well, as you know, you () 24 know, thera are many issues that have been raised by 25 Shoreham these past few years, and why has it taken so () ^ ALDERSON REPORTING COMPANY,INC, 400 VIRGiNI A AVE., S W., WASHtNGTON, D.C. 20024 (202) 554-2345

4380 l l l () 1 long to complete construction or why, you know, had the l 2 costs escalated so much. And those are the broad l 3 perspectives I am talking about. 4 0 What were the conclusions that you reached in 5 performing tha t review or assisting in that review? 6 MR. ELLISs I think, Judge, we would object to 7 tha t as being outside the scope of this conten tion. And 8 this is not discovery. 9 JUDGE BRENNER: Well, first of all, your 10 discovery objection carries no weight. You can ask the 11 same questions here as you can ask on discovery. 12 As te the first point, now yet I do not know 13 where it is going. If it is clear that the review he O 14 performed is outside of anything pertinent to this 15 contention, you may renew the objection or I will cut it 16 off on my own. But I certainly cannot tell from the 17 exchange so far that it is not pertinent. So we will 18 proceed for now. But we vill see where it goes. 19 BY MR. LANPHER: (Resuming) 20 0 Do you recall the question, sir? 21 A (WITNESS GARABEDIAN) Could you repeat it, 22 please. 23 0 If Ijcan remember it. What were the -- () 24 -JUDGE BRENNERs He wanted to know what you 25 found out. O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, O C. 20024 (202) 554 2345

4381 () 1 BY MR. LANPHER (Resuming) 2 0 Yes, what you found out, what were the l 3 conclusions? 4 A (WITNESS GARABEDIAN) Well, basically that 5 Shoreham was not unique, you know, in what had been 6 going on with Shoreham. It was really symptomatic of 7 what is going on in the nuclear industry. 8 0 Did this work involve in any manner the 3 classification of systems, structures and components? 10 A (WITNESS GARABEDIAN) No. 11 0 Did it involve -- 12 A (WITNESS GARABEDIAN) Not this particular 13 effort. O 14 0 I as talking about this effort we have been 15 talking about, I think you said in 1977 or '78. 16 A (WITNESS GARABEDIAN) Yes. 17 0 Did it involve in any manner identification or 18 p-- 19 A (WITNESS GARABEDIAN) Could you repeat one of l 20 the earlier questions? I may not fully have answered l 21 it. ( l 22 (Panel of witnesses conferring.) l 23 0 Sure. ( 24 HR. ELLIS: My problem again with the question 25 of the effort in 1977 or 1978, again there is an O l i ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, O C. 20024 (202) 554 2345

4382 r (_)/ 1 ambiguity. That was not the only thing he was doing in 2 1977 1r 1978. 3 JUDGE BRENNER: Well, I do not know about that 4 particular ambiguity, but the witness has indicated he 1 5 has a problem and we will respect that problem. And I 6 think Mr. Lanpher was about to rephrase it in an attempt 1 7 to clarify it. So we will see where it goes. l 8 MR. LANPHER: Judge Brenner, I think I have 9 been interrupted an awful lot of times with, well, I 10 think there is an ambiguity. As you said before, he has 11 an opportunity for redirect and that is the time for it, 12 it seems to me. 13 JUDGE BRENNER: Why don't you proceed. I 14 ruled on this. 15 WITNESS GARABEDIAN Well, there was one -- 16 JUDGE BRENNER: M r. Garabedian, he is going to 17 a ttempt to clarify it for you, and then if you have a 18 problem understanding what he is trying to ask you can 19 state that, and if possible indicate what the difficulty 20 in understanding is. 21 HR. ELLIS: But if I am not mictaken, Judge, I 22 believe before I broke in Mr. Garabedian said he wanted 23 to say something. I am not sure whether that is true. (O _) 24 MR. LANPHER: He asked me to repeat the 25 question. O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE , S W . WASHINGTON, O.C. 20024 (202) 554 2345

4383 () 1 JUDGE BRENNER Hold it, hold it. That is 2 right, Mr. Ellis, and Mr. Lanpher was about to 3 accommodate the witness when you broke in. 4 Back to M r. Lanpher. 5 BY MR. LANPHER: (Resuming) 6 0 Mr. Garabedian. 7 A (WITNESS GARABEDIAN) Yes, sir. 8 0 This effort in 1977 and 1978, which I 9 understand was the last work you did on the Shoreham 10 project up until this testimony, did that involve in any 11 manner the classification of systems, structures or 12 components at Shoreham? 13 A (WITNESS GARABEDIAN) Yes, it did. You see, 14 when I answered the previous question in terms of 15 safety, you know, I should have been a little bit more 16 careful in saying that, you know, one of the major 17 issues in the nuclear industry has been the promulgation 18 of new standards, new requirements. You know, ASME-3 in 19 particular. 20 And it did involve review of, you know, the 21 more stringent quality requirements than plants being 22 built in the 1970's. In that respect, it did include 23 that as part of the review. (,,) 24 0 The more stringent quality requirements you 25 just referred to, are those requirements like ASME which () I ALDERSON REPORTING COMPANY,INC, i 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4384 ( 1 came in in the early 1970's, I believe someone 2 testified, or Appendix B which came in in 1970 or '71? 3 v) [ 4 Are those the quality requirements you are referring to, sir? 5 A (WITNESS GARABEDIAN) ASME was part, you know, 6 during the design process, and ASME as implemented 7 during the construction process. And there are, you 8 know, a substantial amount of codes and standards issued 9 in the 1970's. 10 0 Did you personally review during this time i 11 period, 1977 to 1978, the adequacy of the classification 12 of systems, structures and components at Shoreham? 13 A (WITNESS GARABEDIAN) I reviewed that earlier O 14 when I was part of the project team. 15 0 During what time period was that, earlier 16 review? 17 A (WITNESS GARABEDIAN) Oh, that was back in 18 1972 to 1974, when I was assistant project engineer. 19 Q But in the 1977-78 period you did not perform 20 such a review? 21 A (WIINESS GARABEDIAN) No. That was not the 22 purpose of the review. 23 0 Prior to this 1977-78 review which we have () 24 been talking about, when was the next most recent work 25 which you performed on the Shoreham project, sir? O ALDERSON REPORTING COMPANY, INC. 400 VIRGINIA AVE., S.W., WASHINGTON, D C. 20024 (202; 554 2345

4385 () 1 A (WITNESS G AR ABEDI AN) Prior to that? 2 0 Yes. 3 A (WITNESS GARABEDIAN) Over the period I just 4 mentioned, 1972 to 1974, when I was assigned to the 5 Shoreham project ss assista nt project engineer. 6 0 Let me go back to one thing. This recent work 7 which you described relating to the Shoreham project in 8 1977-78, is that referred to on page 4 of your resume, 9 the first full -- the paragraph starting approximately 10 in the middle of the page? 11 MR. ELLIS: What page? 12 MR. LANPHER: That is page 4 of Mr. 13 Garabedian's resume. 14 HITNESS GARABEDIANs That is referred to, you 15 know, at the bottom of page 4. 16 BY MR. LANPHER: (Resuming) 17 0 The sentence, "I headed another task force 18 effort for the engineering manager to analyze and 19 evaluate cost and schedule impacts," et cetera? 20 A (HITNESS GARABEDIAN) Yes, yes, sir. 21 0 Do you believe that sentence is a fair summary ! 22 of the work which you performed in 1977-1978 with 23 respect to the Shoreham project? ( 24 A (WITNESS GARABEDIAN) It is a very terse 25 summary of it. ALDERSON REPORTING COMPANY. INC, ( 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4386 () 1 Q Do you think it is a fair terse summary of 2 that work, sir? 3 (WITNESS GARABEDIAN) In my opinion, yes. ( A 4 Q From 197 2 to 1974, you were assistant project 5 engineer; is that correct? 6 A (WITNESS GARABEDIAN) Yes, sir. 7 Q Was Shoreham your only responsibility at that 8 time? 9 A (WITNESS GARABEDI AN) Yes, sir. 10 Q And what were your responsibilities generally 11 during that time period? 12 A (WITNESS GARABEDIAN) Well, generally I was 13 responsible for overseeing the power -- assisting the O 14 project engineer in coordinating the power division 15 efforts in the engineering mechanics group, and that was 16 assisting the project engineer in'those respects. 17 Q When you talk about power division efforts, is i 18 that the mechanical engineering group? 19 A (WITNESS GARABEDIAN) Yes. . l 20 Q I thought in an earlier answer you had said 21 that during that time period, '72 to 1974, you had 22 reviewed the classification of systems, structures and 23 components; is that correct? 24 A (WITNESS GARABEDIAN) Yes, sir. 25 0 Ihat was part of your work in assisting in the O ALDERSON REPORTING COMPANY, INC.

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C387 ( 1 power division efforts? 2 A (WITNESS GARABEDIAN) That was part of the (f 3 4 work that was dons under -- under my responsibility. 0 Okay. With respect to the review, for want of 5 a better word, that you conducted -- 6 A (WITNESS GARABEDIAN) Sure. 7 Q -- what did that entail? 8 A (WITNESS GARABEDIAN) Well, what it entailed 9' was taking a look at, you know, the new standard -- I to mean, the new code requirements that were issued and 11 preparing a table in much more detail than what we had 12 before. You have to remember, you know, that during the 13 conceptual design period, you know, we classified, you O 14 know, the systems and structures f or seismic category 15 one and it's associated QA and we did that from the J 16 conception of the project. 17 After, you know, 1970 there were new codes 18 issued, Reg Guide 1.26 got issued, 1.29 got issued, and 19 the requirements now were that the NRC was looking for a 20 more detailed breakdown in terms of the gradation of 21 quality requirements for the systems and structures. 22 And so one of the immediate concerns of LILCO was, you 23 know, let's, you know, take a look at, you know, 1.26 24 and 1.29 and determine, you know, what are the 25 requirements, you know, on the design that these O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. 0.C. 20024 (202) 554-2345

4388 ( 1 regulations require. 2 JUDGE BRENNER4 M r. Lanpher, we are going to () 3 4 break whenever it is convenient. If you are going to be finished with Mr. Garabedian soon we can stay with it. 5 O th erwise , we can break whenever you want. 6 WITNESS GARABEDIAN This -- 7 JUDGE BRENNER: Did I interrupt a question? 8 MR. LANPHER4 No. I was going to follow up. 9 JUDGE BRENNER: All right. 10 MR. LANPHER: We may have interrupted an 11 answer. 12 JUDGE BRENNER: I as sorry. 13 WITNESS GARABEDIAN: That is fine. O 14 JUDGE BRENNER: If you would like to add 15 something, Mr. Carabedian, to your previous answer, feel 16 free to do so. 17 WITNESS GARABEDIAN I was just going to say, 18 this is part and parcel of the design process. 19 MR. LANPHER: Let me just follow up with one 20 question. 21 BY MR. LANPHER: (Resuming) 22 0 You said a much more detailed table. This is 23 a Q List table or Table 3.2.1-17 24 A (WITNESS GARABEDIAN) No, I do not refer to it 25 as a Q List table. O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4389 ( 1 0 Do you know what I mean when I refer to FSAR 2 Table 3.2.1-1? 3 A (WITNESS GARABEDIAN) Yes, sir. 4 0 Is that the so.me table that you are referring 5 to? 6 A (WITNESS GARABEDIAN) Yes, sir. ! 7 MR. LANPHER: Judge Brenner, this probably is 8 as convenient as any time. 9 JUDGE BRENNER: All right. 10 Before we break, I want to make a comment 11 about the objections with respect to ambiguity and so 12 on. One reason I have been discouraging th ose 13 objections is I just have not seen any abuse on the part l () 14 of cross-examiners in general, or Mr. Lanpher in l 15 particular in any examination he has conducted, and 16 specifically this one, in attempting to cut the witness 17 off or asking confusing questions which would coniuse 18 the questions. 19 In fact, quite the contrary. Sometimes it 20 seemed to me overly lengthy in spots in making sure the 21 witness is clear on whtt he is responding. And I think l l 22 the particular exchcage here was an example of that. We 23 went over not once, but about three times, which studi*es 24 the witness was talking about and what those studies 25 were. l l ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D C. 20024 (202) 554-2345

l 4390 () 1 And as long as that is the pattern, I am just 2 not going to permit interjections by counsel that, I do 3 not think the question is clear. The caliber, the 4 sophistication of the witnesses we deal with in these 5 hearings, makes it clear, to me at least, that the 6 witness is very capable in general. And I will 7 certainly admit there are particular times where the 8 witness may have 31ssed something in the question, but 9 in general the witness is able well to explain where he to is not clear. And I have seen that on the part of Mr. 11 Garabedian also. 12 Part and parcel of the discouraging of such 13 objections has been this Board's willingness, on the O 14 oth er hand, to give the witness a lot of leeway whenever 15 the witness wishes to add something. , And I think we 16 have been very quick, when counsel was about to cut off 17 the witness, to let the witness finish the answer. So I 18 think that is all part of the picture. 19 In addition, when you made the objection 20 stating that the particular study was irrelevant, I 21 trust that you yourself did not know what the study 22 was. If you thou7ht you did, you were wrong as it 23 turned out, as we heard from the answers. If you did 24 not know what the study was, then my point th a t the 25 objection was prema ture was correct. O ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

I 4391 () 1 As f ar as the question that you f elt was so 2 ambiguous, Mr. Ellis, that we are going to spend a lot 3 of time wasting our time because of it, I stated at the 4 time and I will repeat, I just did not see that there 5 was that potential in that question. However, if you l 6 fee] strongly about it, I suggest you might want to l 7 disr uss it with Mr. Lanpher. 8 If he agrees that everybody would be wasting 9 their time unless it is clarified, perhaps we can 10 interject the clarification. If you get nowhere with 11 your discussion with him and still feel strongly about 12 it, I will entertain a bench conference after lunch to l 13 consider it. ( 14 Let's break until 1:20. 15 (Whereupon, at 12:17 p.m., the hearing in the 16 above-entitled matter was recessed, to reconvene at 1:20 17 p.m. the same day.) 18 19 20 21 ! 22 23 24 25 O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4392 () 1 AFTERNOON SESSION 2 (1:20 p.m.) 3 JUDGE BRENNERa All right, we are ready to 4 proceed. Do we need to have a bench conference? 5 MR. LANPHERa We spoke during lunch. I do not 6 think we do. 7 JUDGE BRENNERa Mr. Ellis, Mr. Ea rley? 8 (No response.) 9 JUDGE BRENNER: All right. Proceed. 10 Whereupon, 11 EDWARD T. BURNS 12 GEORGE F. DAWE 13 GEORGE GARABEDIAN O 14 PIO W. IANNI 15 VOJIN J0KSIMOVICH 16 RDBFRT M. KASCSAK 17 PAUL J. McGUIRE 18 PAUL W. RIGELHAUPT 19 DAVID J. ROBARE, 20 the witnesses on the stand at the time of recess, 21 resumed the stand and, having been previously duly 22 sworn, were examined and testified further as follows: 23 CROSS-EXAMINATION -- RESUMED () 24 BY MR. LANPHER: 25 0 Mr. Garabedian, I would like to turn your , O)

\.

ALDERSON REPORTING COMPANY,INC, 400 VIRGINI A AVE, S.W., WASHINGTON, D.C. 20024 (202) 554-2345 l

4393

   )   1 attention to page 3 of your resume,, sir.

2 A (WITNESS GARABEDIAN) Yes, sir. 3 Q The sacond sentence, you are spesking of the 4 March '67 to August '71 time ,9tiod. You say you were 5 responsible for preparation of portions of the PSAR, 6 established basic design criteria. When you say you 7 established basic design criteria, what do you mean by 8 that, sir? 9 A (WITNESS GARABEDIAN) What I mean by that is, 10 establish for the rest of the project what the basic 11 criteria were for integrating the NSS as supplied by 12 General Elactric into the balance of plant systems and 13 structures. l (2) 14 0 How did you go about developing those basic 15 design criteria? 16 A (WITNESS GARABEDIAN) Well, that is a pretty 17 general question. Basically, it was developing a 18 knowledge of what the BWR as of f ered by GE to LILCO, 19 what it consisted of, what its interface requirements 20 were -- and what its interface requirements were. 21 0 Was this the first BWR that Stone C Webster 22 worked on, sir? 23 A (WITNESS GARABEDIAN) No, sir. ( 24 0 Was it the first one you functioned as the 25 architect-engineer for? O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON D.C. 20024 (202) 554 2345

4394 O i A m TNtSS GAaABED nN) 1 do net seueve it wee 2 the first one tha t we functioned as an 3 architect-engineer on. 4 0 Which one do you think you may have done prior 5 to this time? 6 A (WITNESS GARABEDIAN) Well, around the same 7 time -- I am not sure of the da tes. You know, we 8 functioned as an architect-engineer for Fitzpatrick 9 project. We also functioned from the construction 10 standpoint for the Nine Mile Point 1 project. We also 11 were the architect engineer for the Eastern project. 12 0 In the context of establishing these basic 13 design criteria, I believe your testimony was that you O 14 examined the BWR to understand where your interfaces 15 would be? 16 A (WITNESS GARABEDIAN) Yes, sir. 17 0 Did you make determinations regarding how to 18 classify various structures, systems and components, 19 wha t quality to require for basic structures, systems 20 and components? 21 A (WITNESS G AR ABEDI AN) I assisted in tha t, yes, 22 sir. 23 0 What criteria did you utilize in doing that 24 work? 25 A (WITNESS GARABEDIAN) 10 CFR 100 was the O ALDERSON REPORTING COMPANY,INC, 400 VIRGINTA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

4395

      )  1 principal criterion, as well as 10 CFR 20.

2 0 As well as what? 3 A (WITNESS GARABEDIAN) 10 CFR 20. 4 You have to remember that Shoreham, you know, t 5 was not designed in a vacuum. You know, there were a 6 great deal of precedents prior to Shoreham. We had 7 access to key licensing documents for other BWR's going 8 on at the time, and we had access to requirements in 9 terms of licensing requirements for PWR's. And so there 10 was a great deal of information available, you know, for 11 use by the Shoreham project. 12 0 Did you establish functional design criteria 13 for determining how to classify structures, systems and 14 components? 15 A (WITNES5 GARABEDIAN) What do you mean,

16 " functional"?

17 0 Let me come at it a different way. Did Stone 18 C Webster in classifying items first determine what the i 19 functions would be of those systems, structures or 20 components, as basically a first step in determining 21 wha t quality would have to be applied? j 22 A (WITNESS GABABEDIAN) We had to understand i

23 wha t the whole basic project design was, yes.

24 Q Did you perform analyses to -- as to specific l i 25 components'or systems to determine what functions they O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE S.W, WASHINGTON, D.C. 20024 (202) 554 23,6

4396 1 would be called upon to perform? 2 A (WITNESS GARABEDIAN) As part of the PSAR, 3 yes, sir. 4 0 So that is all documented in the PSAR? 5 A (WITNESS GARABEDIAN) Yes, sir. 6 0 Did you participate in the preparation, I 7 guess, of the original table, the counterpart of the 8 FSAB Table 3.2.1-1? 9 A (WITNESS GARABEDIAN) Yes, sir. 10 (Counsel for Suffolk County conferring.) 11 Q Mr. Ianni. I would like to ask a few questions 12 to you. In reviewing your resume, sir, I may have 13 missed it, but I do not identif y any Shoreham-specific 14 work which you have highlightad; is that correct? 15 A (WITNESS IANNI) I did not highlight 16 Shoreham-specific work because we are a functional 17 organization and I work on all the BWR's, including 18 Shoreham, and I thought tha t was understood. When I 1 l 19 would say I worked on ECCS, I work on all the ECCS's for i 20 all the BWR 's, including Shoreham of course. 1 1 21 Q As an example, what ECCS work did you do for 22 Shoreham? 23 A (WITNESS IANNI) Well, back when I was doing 24 ECCS work, at the time we first conceived of the BWR-4 25 design, I was a technical leader in the group that did O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON D.C. 20024 (202) 554 2345

4397 O 1 the ECCS networks calculations. We set the systems s_/ 2 c ri te ria , we determined the pump sizes, we determined 3 f the flow rates that were needed, and we did the 4 reliability studies on them. 5 So design, basically, would mean my group. We 6 designed the network as you see it today pretty much. 7 That was for the BWR-4. And then when we sold the 8 product to the customer that was the design we sold. 9 0 So would it be fair to say that you worked on 10 the ge ne ri: BWR-4 and -- 11 A (WITNESS IANNI) But we also did specific 12 plant analyses for Shoreham. As the design proceeds, we 13 have so go back and redo calculations, because you buy a 14 pump, it is not e:Jactly the exact pump size, it may be a 15 little bit higher or lower, and you redo all the 16 calculations. So we have a complete set of calculations 17 for each pump. 18 Q Did you work -- 19 A (WITNESS IANNI) So we did Shoreham-specific 20 calculations, for example, for the PSAR. We did those 21 in my group. 22 0 Now, you referenced the PSAR. That was back 23 in the late sixties, early seventies, correct? 24 A (WITNESS IANNI) A long time ago, yes, sir. 25 0 Have you done Shoreham-specific work O ALDERSON REPORTING CCMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4398 k) 1 subsequent to that PSAR work? 2 A (WITNESS IANNI) My people have, yes, sir. In 3 fact, as recent as 1977 or '78 they were redone again. 4 Now, you have to understand that these plant 5 -- the plant design, even though we did it in 1960, 6 whatewer it'was, we keep going back over and over the 7 design process because as a change or an issue comes up 8 we have to, by regulation and by procedure, go back and 9 examine every plant for that specific item. And as a 10 result of that, I guess, you know, we have redesigned 11 this plant two or three times maybe. 12 In other words, it is a continuous process 13 that we go through, like we were discussing this morning O 14 about the classification. You know, we reviewed all the 15 classification in 1979, I remember, and every time 16 something comes up we review it. So it is not as though 17 ve design a plant and suddenly everything stops. It is 18 a continuous process. And even today there are several 19 things we still have to complete. 20 It is more of a continuous process until the 21 plant is finally turned over. So that is my involvement 22 in the Shoreham project in that time period. 23 0 I understood you until you last statement, "in 24 that time period."? 25 A (FITNESS IANNI) You asked me about the time O ALDERSON REPORTING COMPANY,INC. 400 VIRGINIA AVE., S.W WASHINGTON, D.C. 20024 (202) 554-2345 , l

4399 () 1 frame 1960 -- the PSAR time frame, o ka y ? 2 Q But your answer carried me up to 1979, I 3 thought. (' 4 A (WITNESS IANNI) Right, yes, it did, because 6 after that except for a couple of years on Mark III, 6 then I spent five or six intensive years on the Mark II 7 containment, which was Shoreham -- Shoreham was part of 8 tha t design also. 9 0 Mr. Ianni, you said tha t the latest GE review 10 -- I assume you are speaking from GE's point of view -- 11 was in 1979? 12 A (WITNESS IANNI) That is one that I recall 13 being involved in. My people and I, we were asked O 14 questions as the classification reviewers were going 15 through their work. Ihey asked us questions with 16 respect to systems, functions, and things of this sort,

17 so they can determine classification. And we get 18 involved in that way in my group.

10 That is the last time that I personally was 20 involved, that I recall being involved, was 1979, on 21 Shoreham-specific review which I recall. In fact, I 22 t h i n .'- it was in that time frame that I can remember 23 being s>9cifically involved. 24 (Counsel for Suf f olk County conf e rring. ) 25 0 Do you recall what the systems were being ALDERSON REPORTING COMPANY. INC. 400 VIRGINIA AVE., S W., WASHINGTON. D.C. 20024 (202) 554 2345

4400 () 1 classified or reclassified in 1979, sir? 2 A (WITNESS IANNI) As I recall, again, we are 3 involved because the people who classified the systems 4 1sk us questions so that they can determine how to 5 classify them. And as I recall, it had to do with that 6 table in the FSAR, 3.2.1. And as I recall, there were a 7 few things that were chanced as a result of the work. 8 And I think Mr. Robare could maybe tell you more about 9 it, since he had more to do with that than I do. 10 But we get involved because of the questions 11 and analyses. 12 0 Now, were you personally involved in the 1979 13 review, or was that your group? O 14 A (WITNESS IANNI) I think my group was, and I 15 remember seeing paper come across my desk, work orders 16 and things of this sort. 17 0 So the record is clear, what is the title or 18 name of your group that you had? 19 A (WITNESS IANNI) The name of my group is the 20 Nuclear Systems Performance Group. 21 Q What is the basic responsibility of that 22 group, sir? l 23 A (WITNESS IANNI) Well, it is to do the () 24 calculations and analyses and the systems design th a t 25 leads to the steady-state performance of the plant. O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON D.C. 20024 (202) 554 2345

l 1 4401 (/ 1 That is one parts the generation of steam; the control 2 system relative to its performing the transient 3 functions that it is supposed to perform; the emergency 4 core cooling system so it can handle the accidents, 5 postulated accidents; and then the containment 6 performance so it can handle this postulated accident. 7 And it is called performance because we deal in how it 8 performs. 9 0 Does your group prepare the chapter 15 10 analyses for GE's scope of the plant? 11 A (WITNESS IANNI) Indeed it does. 12 (Counsel for Suff olk County conferring.) I 13 Q Mr. Ianni or Mr. Robare, I do not know who O 14 this should go to, but you said that this review was in 15 terms cf the table in the FSAR, Table 3.2.1-1. You are i 16 f amiliar that in the course of this proceeding in the ( 17 last couple of months there were a number of changes l 18 made in that table, as documented in an exhibit to the 19 Suffolk County prefiled testimony, correct? 20 A (WITNESS ROBABE) Yes, I am. 21 22 l l 23 ( 24 25 O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W, WASHINGTON, D.C. 20024 (202) 554-2345

4402 1 0 Now, in your review did you review to 2 determine -- your review in 1979, did you review to 3 determine whether that table in terms of the GE scope of 4 supply was accurate and correct? 5 A (WITNESS ROBARE) No, we did not. That was 6 not a licensing review per se. It was an engineering 7 review. 8 (Counsel for Suffolk County conferring.) 9 0 Do you have a copy of that table in front of 10 you, Mr. Robare, as it was attached to Suffolk County 11 testimony? 12 (Panel of witnesses conferring.) 13 A (WITNESS ROBARE) I will in a minute. O 14 JUDGE BRENNER a Y ou are talking about 15 Attachment 2, Mr. lanpher? 16 ER. LANPHER: Yes, I am, Judge Brenner. 17 WITNESS ROBARE: I do now. 18 BY MR. LANPHER: (Resuming) 19 0 If you turn your attention to page 14 of that 20 table, toward the bottom of the page, the main steam 21 isolation valve laakage control system, when we deposed 22 you on March 31, Mr. Robare, I think you will recall 23 that this table had a lot of blank spaces with respect 24 to the main steam line, correct? 25 A (WITNESS ROBARE) That is correct. O ALDERSON 9EPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4403 () 1 0 And subsequently and prior to the hearing it 2 was filled in with these notations pursuant to a letter 3 tha t we received, correct? 4 A (WITNESS ROBARE) Correct. 5 0 Do you know why this was not filled in 0 correctly prior to that time? 7 A (WITNESS ROBARE) That particular system is a 8 recent addition to the plant, and as such there were 9 some editorial errora made in this FSAR table at the 10 time. 11 O What do you mean by " editorial errors," Mr. 12 Robare? 13 A (WITNESS ROBARE) It was not correctly 14 assembled. That is not a reflection on the design of 15 the system, however. I am trying to separate the 16 table 's ac:uratanass with the engineering design of the 17 equipment. 18 Q Is preparation of this table or accurate 19 preparation of this table a GE responsibility or a Stone 20 and Webster responsibility or whose? 21 A (WITNESS ROBARE) GE is responsible for the 22 items within the GE scope of supply, to provide that 23 information to Stone and Webster who assembles it for () 24 the whole plant. 25 (Counsel for Suffolk County conferring.) O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE. S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4404 () 1 JUDGE BRENNERs M r. Robare, I am not sure I 2 understand what you mean by editorial error either in 3 this context. Stone and Webster receives the 4 compilation, including the printing process, of the 5 FSAR, is that correct? 6 WITNESS BOSARE: Correct. 7 JUDGE BRENNER: As agent for LILCO, of 8 course. Was the error in the material that was supplied 9 by GE or was the error in the use of the material by 10 Stone snd Webster? 11 WITNESS DAWE: Judge Brenner, I think I can 12 respond to your question. The main steam isolation 13 valve leakage control system was added to this table O 14 just prior to this October revision in 1981 as part of 15 the SER open item resolution process. In Washington the 16 ' table was marked up by a team of personnel who were 17 dealing with the NRC, and this is a system they 18 requested be added to this table, so it was. 19 The version that was given to the NBC, 20 handwritten in Washington, as part of the SER process 21 did not include the quality assurance category on the 22 table. Subsequently, Stone and Webster was directed to 23 add that part in, and when we did that we did not () 24 realize that the information was missing as part of the 25 amendment which put the SER discussions onto the table. O ALDERSON REPORTING COMPANY,INC. 400 VIRGINI A AVE., S.W., WASHINGTON. D.C. 20024 (702) 554-2345

4405 O i The srstem 1s correct 1r c1esstried in the ted1e es trtco 2 QA Category 1. 3 JUDGE BRENNEPs Did the team that assembled 4 the handwritten tab 1e consist of GE and SEW personne1? 5 WITNESS ROBARE: GE was there. 6 WITNESS DAWEs Judge Brenner, I am not sure 7 whether Stone and Webster personne1 vare there that day 8 or not. 9 JUDGE BRENNER: A11 right. 10 BY MR. LANPHER: (Resuming) 11 0 Mr. Dawe or Mr. Robare, I would like to turn 12 your attention to page 1 of that tab 1e, item 7A, the 13 shroud head and separator assemblies. O 14 A (WITNESS ROBARE) Yes. 15 0 And the table as of October '81 it listed 16 seismic category 1 and quality assurance category 1, and 17 subsequently it was changed to qua11ty assurance 18 category 2 and seismic N/A, is that correct? 19 A (WITNESS ROBARE) That is correct. 20 0 This was not an item that was recently 21 inc1uded in the FSAR, was it? 22 A (WITNESS ROBARE) No, it is not. 23 0 Do you know whether this -- this is an item O 24 ithin the GE ecooe or supp11, correct 2 25 A (WITNESS ROBARE) Yes. O ALDERSON REPORTING COMPANY,INC, 400 VIRGINTA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4436 () 1 Q Do you know whether the original or the 2 apparent error in the original table reflects an error hj 3 in the information provided by GE to Stone and Webster

  \~J 4 for incorporation in the FSAR?

5 A (WITNESS ROBARE) I believe that was the case, 6 that it was a General Electric error. The Shoreham 7 plant was the only plant that had that error. All the 8 other BWRs were QA 2, seismic N/A. 9 Q Right under that, the dryers have a seismic 10 N/A category now. It used to be seismic category 1. Is 11 this again a GE error to the best of your knowledge? 12 A (WITNESS ROBARE) Yes, it is. And it is also l 13 -- it makes those components consistent with the other

    ) 14 BWRs.

15 (Counsel for Suffolk County conferring.) 16 Q Have you determined, Mr. Robare, how those two 17 problems as you just described managed to slip through? 18 Is it just a typo or something more serious? 19 A (WITNESS ROBARE) It is certainly not serious 20 in the sense that the categorization is now consistent 21 with all BWRs. The evaluation of those components, 22 their failure modes and consequences have been 23 thoroughly investiga ted, and the seismic N/A category ( 24 and QA 2 category are quite appropriate considering 25 there is no safety function of those components. l ALDERSON REPORTING COMPANY,INC, 400 vtRGINIA AVE., S.W., WASHINGTON D.C. 20024 (202) 554 2345

4407 () 1 Q When GE supplies data for inclusion in the 2 FSAR is that subject to the GE quality assurance program? 3 A (WITNESS ROBARE) It is subject to licensing 4 procedures. Certainly it is not subject to the full 5 Appendix B quality assurance program. 6 Q I do not mean every single criterion, but I 7 sean do you have a quality control program to ensure -- 8 A (WITNESS ROBARE) Yes. 9 Q Let me finish the question. Do you have a 10 quality control program to assure that the da ta you 11 supply for inclusion in the FSAR are accurate? 12 A (WITNESS ROBARE) We have procedures in 13 licensing requiring signoffs and verifications by our O 14 engineers to assure the quality of that documentation, 15 yes. 16 0 Were those procedures followed in this case? i 17 A (WITNESS ROBARE) I was not personally 18 involved when that information was transmitted. I 19 sim ply don ' t know in that case. 20 0 You have procedures, however, that are 21 designed to prevent this kind of an error. 22 A (WITNESS ROBARE) That is the intent of the 23 procedure. ( 24 (Counsel for Suf f olk County conferring. )

25 0 Mr. Rigelhaupt, I would like to direct some l

O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W.. WASHINGTON, D.C. 20024 (202) 554-2345

4408 () 1 questions to you now, sir, similar to my question to Mr. 2 Ianni. When I reviewed your resume I could find no 3 specific reference to Shoreham specific work. Is that 4 accurate? 5 A (WITNESS RIGELHAUPT) The question of Shoreham 6 specific work as it applies to Stone and Webster 7 supervisors and managers has the same dilemma that we 8 had in Mr. Ianni's testimony. During the period of time 9 of approximately 1972 to '75 I was assistant chief power 10 engineer in our Boston office, and going back to Mr. -- 11 I believe it was Mr. Garabedian's answer to your 12 question concerning the Power Division -- 13 (Discussion off the record.) () 14 The answer -- 15 0 I lost the train of your answer. If you could 16 start over. Do you recall the question, sir? l 17 A (WITNESS RIGELHAUPT) Yes. If I may repeat my 1 18 answer, as I understand, the question was raised whether 19 I had Shoreham specific experience. My answer to the l 20 question was we get into the same dilemma ve got into 21 with Mr. Ianni's testimony that supervisors and l 22 managers have to relate to these projects in a somewhat 23 different way, and I want -- I proceeded to give you ( 24 some specifics of my tangency in relationship to the 25 Shoreham project. O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D C. 20024 (202) 554-2345

4409 During the period 1972 to 1975 I was assistant (]') 1 2 chief power engineer in our Boston office. And I then 3 referred you to Mr. Carabedian's answer to one of your 4 questions as to whether the Power Division was the 5 mechanical division. That answer really needed 6 amplification. 7 Our Power Division was a consolidation of our 8 Mechanical Division, which had to do with the design of 9 fossil plants, and our Nuclear Division in which our 10 nuclear technology resides. We combined those two 11 divisions in the late '60s, and so as a consequence 12 under my specific direction was the nuclear technologies 13 such as licensing, safeguards, rad waste processing, 14 shielding, and nuclear technologies of that sort. 15 And so Mr. Dawe in his testimony concerning 16 his assignment as licensing engineer, those assignments 17 were received from my supervisor of licensing. So in an l 18 overall sense of engineering, management of 19 classification of systems and other project concerns 20 eventually crossed my desk as f ar as my management 21 review and specific approvals on certain aspects of this. I 22 Similarly, as assistant engineering manager j 23 and chief power engineer in our New York office, at that () 24 time I had the responsibility for a group of piping 25 design engineers which were physically loca ted at the i O l ALDERSON REPORTING COMPANY,INC, ( 400 VIRGINI A AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345 l

4410 () 1 Shoreham plant and were part of the Power Division. 2 Again, this was in an overall management context rather 3 than a specific individual contributor relationship to 4 Shoreham. 5 0 Would it be fair to state, sir, that your work 6 with respect to Shoreham has been as a manager and 7 supervision of personnel who might be doing day-to-day 8 work on the Shoreham project? 9 A (WITNESS RIGELHAUPT) I would say that in most 10 cases it was supervision of the supervisors. 11 Q So -- 12 A (WITNESS RIGELHAUPT) To clarify that issue, 13 it was a level above that. O 14 Q So sort of one step further removed perhaps. 15 A (WITNESS RIGELHAUPT) To be absolutely cl e ar , 16 yes. 17 Q In the first paragraph of your resume, sir, 18 you state that you are now responsible for overall 19 direction and guidance of advanced technology 20 activities. What does that involve? 21 A (WIINESS RIGELHAUPT) This involves a series 22 of technical activities that range all the way from 23 micronized coal applications to fossil plants through 24 solar work, power work of all sorts into fusion 25 technology and finally way out on the outer fringes of O ALDERSON REPORTING COMPANY,INC, 400 VIRGINTA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4411 () 1 our advanced technology efforts a're things like 2 consulting with HIT on the design of a gravitational 3 vsve antenna for the National Science Foundation. So we 4 cover the b roadest possible range of new technologies as 5 they relate to the power field. 6 0 Ihis current work does not relate specifically 7 to Shoreham? 8 A (WITNESS RISELHAUPT) It has no relationship 9 to Shoreham in that sense, but again, taking one step 10 backward, in my management responsibility as an 11 assistant engineer in management of our engineering 12 department there are general administrative duties and 13 even technical duties that again are tangent to all of l ' 14 our projects. So in that very ill-defined area, if you 15 will, non-specific area, there is some tangency with 16 Shoreham, but clearly not a close link. 17 0 Not a day-to-day kind of responsibility. 18 A (WITNESS RIGELHAUPT) No, certainly not a l l 19 d ay -to -d ay responsibility. 20 0 Have you ever classified any systems, 21 structures or components at Shoreham? l l 22 A (WITNESS RIGELHAUPT) Again, going back to my l l 23 previous testimony, at the time period that Shoreham was 24 going through these processes we had over 20 nuclear 25 units going th ro ugh these pregesses, and my assumption 1 ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

l l 4412 () 1 was to develop overall management guidelines and 2 philosophies as to how this kind of activity relates to 3 the regulatory process. So in that sense, which is a 4 very key sense, I have played a role in this. 5 (Counsel for Suf folk County conferring. ) 6 0 Mr. McGuire, I would like to address certain 7 questions to you. 8 You corrected your resume on the first page to 9 add the sentence, "In 1970 I was sleighted to be one of 10 the chief engineers at Shoreham and worked on Shoreham 11 for eight sonths," correct? i 12 A (WITNESS MC GUIRE) That is correct.

;     13      0     What were your responsibilities during those

( 14 eight months, sir? 15 A (WITNESS MC GUIRE) The specific 16 responsibilities were to review the systems that were 17 being generated by GE and Stone and Webster and to look l 18 at them from an operational standpoint whether or not 19 they would function. 4 20 0 What do you mean whether or not they would 21 function? 22 A (WITNESS MC GUIRE) Well, my previous 23 axperience was working in fossil plants, and we were ( 24 supposed to look at this to make sure that the systems 25 were maintainable specifically, and that they are laid ( ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4413 () 1 out in such a manner that there was esse of operation. 2 0 You say you were sleighted to be one of the 3 chief engineers. Does that mean you in fact did not 4 become one of them? 5 A (WITNESS MC GUIRE) That is correct. 6 g prior to the testimony you are providing in 7 this proceeding was that your only Shoreham specific 8 work? 9 A (WITNESS MC GUIRE) I have been on contract 10 with the plant staff since April. This is one of the 11 duties. l 12 0 Okay. I mean since then other than this work 13 since April, maybe there is other work as well, this O 14 testimony, this plus that work back in 1970 was the only 15 Shoreham work. 16 A (WITNESS MC GUIRE) Well, Shoreham was 17 included as a plant that was reviewed in 1979 with 18 General Elactric and the BWR owners group subsequent to 19 the THI incident. 20 0 Those are the owners group activities which 21 you describe in your actual prefiled testimony, sir? 22 A (WITNESS MC GUIRE) That is correct. 23 (Counsel for Suf f olk County conferring. ) ( 24 25 O ALDERSON REPORTING COMPANY,INC, 400 VIRG'NIA AVE., S.W., WASHINGTON. O.C. 20024 (202) 554 2345

4414 () 1 Q The last sentence of your resume say, 2 " Ra cen t ly , I have been contracted by Long Island Light 3 Company to assist the Shoreham plant staff." That was 4 in April 1982? 5 A (WITNESS McGUIRE) Yes, sir. We have been 6 discussing this over the last year, based on my 7 availability. 8 (Pause.) 9 Q Mr. Kascsak, in your work for LILCO, have you 10 been involved in the classification of systems, 11 structures and components? 12 A (WITNESS KASCSAK) My involvement has been one 13 of review and personnel that have worked under me in our 14 review of design documents and how those design 15 documents relate to classification. 16 0 What is the purpose of your review, or what 17 has been the purpose of these reviews? 18 A (WITNESS KASCSAK) Well, my reviews have been 19 in different capacities, so you are going to have to 20 clarify as to in what capacity you were referring. 21 Q Sure. Let me -- let*c start chronologically. 22 I guess your first work on the Shoreham project was in 23 J uly 1974, is that correct, sir? () 24 A (WITNESS K ASCSAK) That is correct. 25 0 And that was in the position of lead O ALDERSON REPORTING COMPANY,INC. 400 VIRGINIA AVE., S.W., WASHINGTON D.C. 20024 (202) 554 2345

4315 () 1 mechanical engineer. 2 A (WITNESS KASCSAK) That is correct. 3 0 What were your responsibilities in that 4 position? 5 A (WITNESS KASCSAK) My responsibilities in that 6 position dealt with the review of design documents 7 sssociated with mechanical equipment that was to be used 8 within the Shoreham plant. 9 Q And when you say you reviewed design 10 documents, were these documents generated by GE or Stone 11 C Webster? 12 A (WITNESS KASCSAK) Yes, these documents are 13 documents that we defined -- LILCO has defined

 \J 14 procedurally within our project manual as documents that 15 require review by LILCO engineers.                  And we assign lead 16 engineers by discipline to review those specific design 17 documents.

18 0 What was the purpose of the design review? 19 A (WITNESS KASCSAK) The purpose of the design i l 20 review is to perform a technical overview to ensure that 1 21 the design basis of the plant was being maintained to 22 evaluate the functional aspects of the documents to see 23 if we had any concerns or questions that we could -- you 1 () 24 know, that might come in to pla y in producing those 25 documents. l l ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4416 () 1 0 I believe you stated tha t this review, at 2 least in part, vss to assure that the technical design 3 basis was main tained. Did you review these design 4 documents against some design basis criteria? 5 A (WITNESS KASCSAK) Well, we have not got into 6 the whole design evolution process, but the process is 7 sn evolution of design documents, and these design 8 documents evolve one to another and there needs to be a 9 consistency in those design documents relative to the to specifications, the flow diagrams, elementary diagrams, 11 wha tever, the system f unctions and how those system 12 functions are fulfilled through those design documents. 13 0 What I do not understand, Mr. Kascsak, is that 7_ (_) 14 you stated that you -- this review was to assure that 15 .the design basis was maintained. How did you know 16 whether the design basis was maintained or not when you 17 conducted the review? 18 A (WITNESS KASCSAK) Well, I think those are your 19 words. I do not believe I used the words " design basis 20 was maintained." I said we did a technical review to 21 evalute the design, and to ensure that th e design -- th e l l 22 design was consistent with the function and requirements 23 of that particular system or document that was to be 0 Q 24 reflected in that review. 25 (Counsel for Suffolk County conferring.) O(~% l ALDERSON REPORTING COMPANY,INC. 400 VtRGINIA AVE., S W., WASHINGTON. O C. 20024 (202) 554 2345

4417 l () 1 0 Did your review include determining whether 2 the classificaticn of a particular system or structure 3 or component was proper? 4 A (WITNES3 KASCSAK) Not specifically. In most 5 cases, the design documents are reflective of that, or 6 many of the design documents produce results that could 7 maybe be used to determine system classification. But 8 LILCO per se was not involved in directly classifying 9 systems. 10 0 And so you were not, either, correct? 11 A (WITNESS KASCSAK) No. 12 0 Have you ever been part of the LILCO quality 13 assurance staff? 14 A (WITNESS KASCSAK) No, I have not. 15 (Counsel for Suffolk County conferring.) 16 0 M r. Dave and Mr. Robare, I would like to turn 17 your attention to page 50 of the prefiled testimony. 18 JUDGE BRENNER: M r. Lanpher, I take it for now 19 you are leaving the area of qualifications. 20 MR. LANPHER: Yes, I am. 21 JUDGE BRENNER: I wonder if I could jump in. 22 MR. LANPHER: Of course. 23 JUDGE BRENNER: Mr. McGuire, could you tell me ("% () 24 what else you have been contracted to do since April 25 besides obviously, this testimony. O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W WASHINGTON, D.C. 20024 (202) 554 2345

4418 () 1 WITNESS McGUIRE: Yes. At this point in time, 2 many of the plant staff are going through a training 3 program as part of licensing. I have been asked to come 4 in and take a look a t each f unctional area to see what 5 things have to be completed so that they will be 6 operationally re s ?.y at the time of fuel load. 7 JUDGE BRENNER: You mean in terms of hardware 8 or procedures or some combination of the two? 9 WITNESS McGUIRE4 Tha t is correct. Systems 10 right now are being turned over from the startup 11 organization to plant staff, and the plant staff has 12 written many procedures based on design documents. It 13 is now to a point where a systematic approach is taken U 14 where we look at the operating procedures anJ the 15 surveillance procedures to make sure that they will work 16 and function as the system is actually built. 17 JUDGE BRENNER: Do you think you were brought 18 on because of your connection with Shoreham , or does the 19 firm you are presently will do this type of thing? 20 WITNESS McGUIRE: I have been doing this type 21 of thing at the Louisiana Power & Light plant, 22 Waterford-3, for the last year. The fact is I do know 23 most of the plant staff from my previous employment with ( 24 Long Island Lighting Compan y. I am sure that has some 25 influence on it. O ALDERSON REPORTING COMPANY,INC. 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4419 O 1 3 UDGE eREa*Ea. So rou are en emo1eree of UES 2 based in Atlanta, but you are assigned to Waterford and 3 that is why you are in Louisiana. 4 WITNESS McGUIRE: That is correct. 5 JUDGE BRENNER: All righ t, thank you. 6 BY MR. LANPHER (Resuming): 7 0 Mr. Dawe and Mr. Robare, in this section of 8 your testimony you address the so-called Denton meno, 9 which was A ttachment 1 to the Suffolk County prefiled 10 testimony. Correct? 11 A (WITNESS BOBARE) That is correct. 12 0 Do you disagree with the Denton memo? 13 (Panel of witnesses conferring.) O 14 A (WITNESS D AWE) In the context of the Denton 15 memo and the definitions that are contained in 16 enclosures to that Denton memo, we disagree that those 17 were the historical things used as definitions on 18 Shoreham. And in fact, on many, if not all, other BWRs 19 of this vintage. 20 0 Did the Denton memo state that these l 21 definitions were used on Shoreham or on other plants? 22 A ( WITNESS D AWE) I would like a moment just to 23 quickly review the memorandum. 24 (Witness reviewing document.) 25 Mr. Lanpher, could you repeat the question, O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4t20 () I please? 2 0 Does the Denton memo state that Shoreham or 3 other plants have used this or' failed to use this 4 terminology? That was not exactly the question. 5 A (WITNESS DAWE) The Denton memorandum that we 6 are referring to does not specifically mention Shoreham 7 and whether it used or did not use the terminology. 8 However, it does speak to the way the terms have been 9 used in the past, and it does state that this is the 10 consistent way to be used in the future. It, I believe, 11 states in language that I understand that this is the 12 way it was applied in the past, and that all he is 13 attempting to do is clarify for his staff the way it is

 ! (   14  to be used in the future.

15 I disagree with the statement that that is the 16 way it has characteristically been used in the past. I 17 agree with the Denton memorandum in its earlier 18 paragraphs that talk about the fact that they have been 19 used interchangeably in the past. And to that extent, I 20 think I have answered your question. 21 Q Has LILCO used the terms "important to safety" 22 and " safety-related" interchangeably in its FSAR? 23 A (WITNESS DAWE) Yes, sir, they have. And I () 24 think consistently with the statements of the 25 regulations tha t were being addressed in the FSAR. l ALDERSON REPORTING COMPANY,INC. 400 VIRGINIA AVE., S W., WASHINGTON, D C. 20024 (202) 554 2345

4421

 )  1 0        There are a lot of regulations.                                                                                                                   Do you have 2      an7 particularly in mind, Mr. Dave, when you make that 3      statement?

4 A (WITNESS DAWE) Yes, sir. Cur prefiled 5 testimony states the regulations that used the te rm , or 6 approach the term or interchange the term, if you will. 7 0 The term "important to safety" is defined in 8 the introductory material to 10 CFR Part 50, Appendix A. 9 Correct? 10 (Witness reviewing document.) 11 A (WITNESS DAWE) Yes, sir, that is correct. 12 0 You are familiar with the sort of generic 13 definition of the term " safety related" performing the O 14 functions which come from Part 100, Appendix A -- 15 correct? -- preserving the integrity of the reactor 16 pressure coolant boundary and the other items. 17 A (WITNESS DAWE) Yes, I am familiar with those 18 definitions from Part 100, Appendix A. 19 0 Do you agree with Part 100, Appendix A 20 definitions, with the definition of "important to 21 safety" from 10 CFR Part 50, Appendix A? 22 A (WITNESS DAWE) The definitions f rom Part 100, 23 Appendix A and the terminology using Pa rt 100, Appendix 24 A defin the safety-related functions to be performed by V 25 certain, structures, systems and components in the plant, O V ,' ALDERSON REPORTING COMPANY,INC, 400 VIRGIN! A AVE., S.W , WASHINGTON, D.C. 20024 (202) 554 2345

4422 () 1 and those definitions of " safety-related functions", 2 using that terminology, are those to which you alluded. 3 The integrity of the reactor coolant pressure boundary, 4 the achieving and main taining safe shutdown, and the 5 prevention and mitigation of accidents, whose 6 consequences could a pproach the guideline values of Part 7 100. 8 It is our interpretation of the way we used it 9 in classifying our systems and in using the terminology 10 in the FSAR, that the prevention of undue risk to the 11 health and safety of the public was, in fact, the proper 12 performanca of those safety-related functions. The 13 general design criteria themselves, when they use the O 14 term "im por ta n t to safety" talk about performing a 15 safety function. 16 I think that is a consistent application that 17 has been used throughout the industry, not just on this 18 particular plant or in its licensing documents, but in 19 all that I have looked at. I think it is consistent 20 with the industry standards such as the American Nuclear 21 Society which defines safety function and the 22 performance of safety function by safety-related systems. 23 I think that yes, we use in the FSAR the 24 terminology "important to safety" to be synonomous with 25 " safety related." As we have explained in our O ALDERSON REPORTING COMPANY,INC, 400 VIRGlNIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4423 ( 1 testimony, we do not believe in the end result that that 2 results in anything different in terms of the plant 3 design. But we do believe that that was the 4 intepretation throughout the time that Shoreham was 5 being developed, and that is the way we have used it. 6 0 From that answer -- we will come back to the. 7 level of safety -- your position that you do not think 8 it makes a difference. In fact, your plant -- but from 9 your answer, I interpreted that to mean that you'do 10 disagree with the Denton memo which says that important 11 to safety is really bigger than safety-related. 12 A (WITNESS DAWE) Yes, sir, I personally disagree 13 with that interpretation, but I think it is a 0 14 distinction that may come to be in the future. But I do 15 not think it is a distinction that has been made in the 16 past. 17 Q Certainly that was not the intepretation of 18 Part 50 Appendix A which LILCO applied in its FSAR. 19 A (WITNESS DAWE) Which interpretation, Mr. 20 Lanpher? 21 0 LILCO did not distinguish between "important 22 to safety" and " safety related" in the FSAR. 23 A (WITNESS DAWE) In the use of the terminology, 24 no. In -- I think that answers your question. 25 Q Now, I want to understand your position O ALDERSON REPORT;NG COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON,1C. 20024 (202) 554 2345

4424 () 1 clearly, Mr. Dawe. Yc u believe nevertheless that even 2 _u items which LILCO would call non-safety related or 3 not important to safety, you believe that LILCO, GE, 4 Stone C Webster, whoever had the responsibility applied 5 a level of quality assuranc7 c7mmensurate with the 6 function of that equipment or component or whatever? 7 A (WITNESS DAWE) In specific response, yes, I 8 do. I do nnt think you need limit it just to say a 0 level of quality assurance or total engineering 10 judgments that were made. Our design, our construction 11 has checked that quality assurance, if you will. Yes, I 12 think that has been done commensurate with its f unction 13 in the plant. O 14 0 Now, in determining what level of engineering 15 quality and the other things tha t go into producing a 16 product to determine what degree of ef fort is 17 commensurate with a function to be performed, how is 18 that determination made? 19 A (WITNESS DAWE) I am sorry, M r. Lanpher , could l 20 you repeat the question for me, please? l 21 A (WITNESS D AW E ) To determine what level of l 22 quality and engineering and the other things that go 23 into producing a product that is commensurate with the n/ (- 24 function to be performed, how do you determine what l 25 degree of effort to put into tha t product? What is the l 1 ALDERSON REPORTING COMPANY,INC, 400 VIRGINTA AVE , S W . WASHINGTON, D.C. 20024 (202) 554-2345

4425 () 1 process by which you reach that conclusion? 2 A ( WIT!!ESS D AWE) I have difficulty with your 3 applica tion of the term " level of effort." Effort is 4 required to engineer and design a plant. The effort for 5 engineering is a function of how difficult that portion 6 of the engineering is. 7 With respect to -- excuse me just a second. 8 (Panel of witnesses conferring.) 9 I am having some difficulty answering your 10 question because the question is so broad and perhaps 11 not in a terminology that I can relate to. The level of 12 eff ort a pplied to a rarticular portion of the 13 engineering is obviously directly that level that needs

l f'# 14 to be applied to come out with a sound engineering 15 product. 16 Likewise, the level of quality assurance that 17 is applied to it is the level that is needed to ensure 18 that the engineering product that we want to put in 19 place, when it is finally constructed, is what we intend 20 to be there. 21 For our safety-related components, there are 22 very clear and definci additional regulatory 23 requirements determined on it. Our non-safety 7 24 requirements are more -- are a judgment of what is 25 required, but even within that judgment there are O v A DERSON REPORTING COMPANY,INC, 100 VRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4426 O i re2uire. eats g1eced on us sr the .ay .e eo our work. 2 And in fact, our -- speaking for Stone C Webster, that 3 .ork is procedurally controlled and directed. And all 4 these factors go into determining . hat I think you mean 5 by the 1eial of effort. 6 That level of effort is just not a term I can 7 relate to in such a broad question. 8 9 10 11 12 13 14 15 16 17 18 19 20 l 21 22 23 24 25 1 ALDERSON REPORTING COMPANY,INC, l 400 vtRGlNIA AVE, S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4427 l () 1 0 Let me direct your attention to page 54. You 2 state thats " Normal non-safety equipment, while not 3 subject to the full requirements of Appendix B, is 4 nonetheless designed and constructed to a range of high 5 str ndards conmensurate with its function." With that 6 st:itement in mind, I understand with respect to items 7 which you classify as safety-related, you have set up or 8 attempted to set up a quality assurance progran 9 consistent with the very specific regulatory 10 requirements of Part 50, Appendix B, correct? 11 A (WITNESS DAWE) Stone C Webster has a quality 12 assurance progrsa which meets the requirements of 13 Aptendix B, yes. O 14 0 And that will in your opinion ensure, with 15 recpect to safety-related equipment, that the high 16 standards, the necessary high standards are met? This 17 is for safety-related equipment. 18 A (WITNESS DAWE) Yes, sir. The quality 19 assurance that we perform for safety-related components 20 will assure that the standards we have established with 21 respect to saf ety-related components is met. 22 Q Now, for items of equipment which are not 23 saf e ty-rela ted , or systems, how do you determine what is 24 com mens ura te with the system's function and thus what 25 standard to construct it to or to design it to? And O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

l 4428 l l () 1 using these righ t h e re , wha t a re your criteria? How do 2 you do it? 3 ( Panel of witnesses conf erring. ) 4 A (WITNESS DAWE) For the non-saf ety-rela ted 5 components, the determination of standards commensurate 6 with its function is a normal engineering decision. 7 There is really nothing magic to the fact they are a 8 non-safety component. If we are designing, for example, 9 1 piping system to contain fluids at a certain pressure, 10 we would apply certain industrial codos to that piping 11 system, and that is what we mean by -- an example of 12 what we mean by standards commensurate with its 13 f unc tion. ' 14 These are all highly reliable components 15 designed in the plant to perform a function. 16 Q Well, what functions are we talking about 17 here? Are these the normal functions or the whole range 18 of functions that a system migh t be called upon to 19 perform, even in say an accident situation? 20 A (WITNESS DAWE) It is the functions that are 21 within the design basis of the system. Now, it could be 22 called upon or requested to perform, if you vill, those 23 functions during any mode of operation. All right, it () 24 is specifically designed for the service conditions in 25 which its function was intended to be performed. That O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

4429 p) q, I would not necessarily preclude a function from being 2 performed under some other service condition, but it is 3 a service condition that is defined, and it is designed 4 to perform that function under those service 5 conditions. 6 A (WITNESS GARABEDIAN) May I add? 7 0 Certainly. 8 A (WITNESS GARABEDIAN) The balance of plant, 9 what is called non-safety-related equipment, systems, 10 structures, are extremely important, you know, for the 11 power generation of the plant and for putting together a 12 well-integrated plant design. Ever since time zero, 13 when we started in conceptual engineering, we have 14 representatives, you know, on the project who look at 15 the mechanical aspects, environmental aspects of the 16 plant design, alongside with the nuclear aspects. Both 17 of them are important. 18 Now, our goal is, you know, to have a well 19 designed and well operated plant. And so when you start 20 getting into some of the turbine generator systems, you 21 know, they are extremely complex. They require many 22 calculations and s tremendous amount of coordination 23 with the vendors, wi th the client. r-(_)\ 24 But you know, all of that has to be well 25 engineered and well designed. And the amounts of ('T v ALDERSON REPOR3NG COMPANY,INC, 400 VIRGINI A AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

4430 () 1 engineering design required of course varies as you go 2 through the process. 3 So that is all I wanted to add. 4 JUDGE BRENNER: Let me see if I could help. 5 Mr. Carabedian, I thought that was all well and good, 6 but where Mr. Lanpher is going, stated differently, is 7 he wants to know, when it is stated in the testimony 8 that the non-safety system is designed to high standards 9 commensurate with its function, how that function is 10 designed. Are you talking about just a normal function 11 or do you take a look at the emergency situations in 12 which admittedly a non-safety system might be called l 13 upon as one of the initial lines of defense, so to 1 CE) 14 speak, before you call on a safety-related system, and 15 say, well, we night want to use it for that, so we 16 should design it to more than just the normal 17 operational functions of startup and operation and 18 shutdown and so on. 19 (Panel of witnesses conferring. ) 20 JUDGE BRENNER: Is that a fair paraphrase of 21 where you were headed, in part, Mr. Lanpher? 22 MR. LANPHER: Yes. Thank you, Judge Brenner. 23 JUDGE BRENNER: So we accept your general b q,/ 24 sc oladas, but it does not help zero in on that avenue 25 of questioning. ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE, S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4431 () 1 WITNESS DAWE Judge Brenner, I think I can 2 answer your question in the way you have framed it. A 3 systen would be designed, one of these non-safety 4 systems, would be designed for its normal service 5 condition. We would not be postulating more than its 6 normal service condition. 7 Howevar, there is no reason to believe that 8 necessarily, even in another type of situation such as 9 an abnormal operational transient or an accident, that 10 its service condition would be exceeded. And even 11 within its service condition the high standards, the 12 industry standards that are applied, do design margin 13 into the systems. O 14 BY MR. LANPHER. (Resuming) 15 0 I just want to follow up on that, Mr. Dawe. 16 When you say " normal service condition," I understand 17 tha t to mean that you design it not for those unusual 18 conditions, for instance in an accident, where it might 19 be called upon to operate; is that correct? 20 A (WITNESS GARABEDIAN) It is considered. You 21 know, the sa f e ty-rela ted , non-safety-related interf ace 22 is an extremely important interface that you have to 23 consider in the nuclear power plant design. You know, ( 24 that is, you know, one of the reasons why we have to 25 describe the whole plant in the FSAR and not just the O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

4432 1 safety-related systems. 2 0 I believe your answer was that normal -- that 3 in design of non-safety-related equipment, you design 4 for normal service conditions but you consider abnormal 5 conditions; is that right? 6 (Panel of witnesses conferring.) 7 0 I am trying to paraphrase. If I am wrong 8 correct me. 9 A (WITNESS DAWE) The last part of your 10 question, in considering beyond the normal service 11 conditions, the answer to that is yes, to the extent 12 that we would want to be sure that that type of a 13 component does not experience a failure mechanism that O V 14 would affect another system that is designed for this 15 different service condition. I think that is '- 16 0 Would it be fair to say, then, it is 17 considered in the sense that you attempt to ensure that i 18 this non-safety-related system will not have, say, an l 19 adverse interactions with a saf ety system,

   ._. ) safety-related system, that you would be relying on in 21    an accident or non-normal situation?

22 A (WITNESS DAWE) Yes, sir. 23 (Counsel for Suffolk County conferring.) p Q 24 JUDGE BRENNER: Mr. Lanpher, while there is a l 25 pause, I have been trying to follow you a little bit and , O ALDERSON REPORTING COMPANY,INC. 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4433

  )  1 I ended up on page 12 of your plan and I guess I should 2 not infer that we have gone through the first 11 pages.

3 MR. LANPHER That is correct. 4 JUDGE BRENNER: Are you going to be jumping 5 around? 6 ER. LANPHER: I will -- I should have told 7 you, I am not going to be dealing wi th the testimony in 8 the order that it is presented. 9 JUDGE BRENNER4 All right. That answers the 10 question. 11 MR. LANPHERs I will, knowing tha t there are a 12 lot of different elements, when I finish an area I will l 13 inform the Board, so if you, for instance, have O 14 questions on the Denton meno and want to tie that aspect 15 of the record up, you will have that chance. 16 JUDGE JORD AN : Does that mean you have now 17 finished? 18 MR. LANPHER: No, it does not, Judge Jordan. 19 (Laughter.) 20 JUDGE BRENNER: We tried that trick, but it 21 never worked. 22 (Laughter.) 23 MR. LANPHERs It worked, but later in the 24 week. l 25 (Laughter.) ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4434 () 1 BY MR. LANPHER: (Resuming) 2 0 .ir. Robare, you heard Mr. Dave's answer and 3 M r. Carabedian's answers to that with respect to the GE 4 scope of sapply. You also do not use an important to 5 safety category, correct? 6 A (WITNESS ROBARE) That is correct. 7 0 For itess which are not important to safety, 8 are those designed just for their normal service 9 conditions? 10 A (WITNESS ROBARE) We consider more than the 11 normal service condition in the design of those 12 systems. 13 0 What more do you consider, Mr. Robare? O 14 A (WITNESS ROBARE) We consider all credible 15 modes of operation that the equipment may be needed l 16 for. 17 0 You say needed for. The testimony also talks 18 about something that is not required to perform a safety 19 function. When you use the word "need" do you equate 20 that with " require"? 21 A (WITNESS ROBARE) No, that was a bad choice of 22 words. There are examples in the County's testimony, 23 and explainec .T our testimony, of systems and 24 components that have not been categorized full safety 25 grade, but have been given special design and quality O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4435 () 1 assurance attention. And I believe those are the types 2 of items that you are referring to. 3 I also believe we have demonstrated the extra 4 attention that they received. 5 0 I think the testimony indica tes -- your 6 testimony identifies four or'five such systemsa rod 7 block monitor, level A trip. Are those the items that 8 you are referring to, Mr. Robare? 9 A (WITNESS RGBARE) Yes. 10 0 Let's go back to page 54 of the testimony, 11 because this is -- I was asking Mr. Dawe about this 12 statement in the middle of the page where, with respect 13 to Stone & Webster 's scope of supply, he tastified that O 14 non-safety-related equipment is designed and constructed 15 to a range of high standards commensurate with its 16 function. 17 Now, how does General Electric determine 18 whether -- wha t standa rd to design and construct an item 19 to if it is not safety-related? 20 A (WITNESS ROBARE) Tha t would be a judgment 21 made by a combination of our design engineers and our 22 quality assurance engineers, dependent upon the use of 23 that system for plant reliability and its potential use ( 24 in transient situations. 25 (Counsel for Suffolk County conferring.) O ALDERSON REPORTING COMPANY,INC, 400 VIRGINTA AVE. S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4436 () 1 C So would it be fair to state, Mr. Robare, that 2 at GE you look beyond the normal service condition to 3 the potential use in mitigation or prevention of 4 transient conditions? 5 A (WITNESS ROBARE) Only for those systems or 6 components that are -- have been -- have been 7 demonstrated to be of some use in those situations. 8 0 How do you determine what systems, structures 9 or components might be of some use in transient -- in 10 preventing or mitigating accidents or transients? 11 A (WITNESS ROBARE) When we performed the 12 transient analysis, as in chapter 15, we assume the 13 mitigation is obtained by systems that would . ( 14 mechanistically perform that function, and from that l l 15 analysis we can determine what systems and components 18 are required and desion for them accordingly. 17 0 Now, this is safety-related systems, those 18 that are identified in chapter 15, aren 't they, sir, 19 excr at for a few minor exceptions? 20 A (WITNESS ROBARE) They are mostly safety 21 systems, yes, but not totally. 22 Q Well, there are those that are discussed in 23 the latter portions of your testimony which are not. O (/ 24 But except for the rod block monitor and those others 25 tha t are identified, those are the only non-safety b U l l l ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. O C. 20024 (202) 554-2345

4437 O i syste = in chepter ,s,ere they not, u=ed for prevention 2 or mitiga tion? 3 MR. ELLISS Hay I inquire whether prevention 4 or mitigation is of accidents or transients? 5 MR. LANPHER: Either one. 6 (panel of witnesses conferring.) 7 JUDGE BRENNER: I guess the answer is yes, you 8 might inquire. 9 (Laughtar.) 10 WITNESS ROBAEE: I want to be sure we are 11 differentiating between transients snd accidents. The 12 accidents utilize only safety-grade equipment for 13 mitigation. The transients generally use safety-grade O 14 equipment. There are a few exceptions that are noted in 15 our testimony, and those exceptions are the only what I 16 would call active mitigators of those transients. In 17 other words, those systems that are required to operate 18 in order to turn the event around. 19 - 20 21 22 23 bi V 24 25 O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (2C2) 554-2345 l

4438 () 1 JUDGE JORDAN Do you mind if I ask for just a 2 little clarifica tion on tha t? You say the eq uipm en t 3 involved with transients also may be required for 4 safety. 5 Now, I was thinking in terms, for example, of 6 feedwater. Feedwater transients I presume are 7 moderately frequently in-plant power failures of the 8 offsite power. These are all transients. Failure of 9 the control system, are these not transients and are 10 these not transients in nonsafety-related equipment? 11 WITNESS ROBARE: The equipment you mention was 12 use as -- we postulate its f ailure to be the event 13 itself, and we do not use that equipment to mitigate. O 14 JUDGE JORDANS All right. But you are saying 15 then that transients which can lead to accidents can 16 orijinate in the nonsafety equipment. 17 (Panel of witnesses conferring.) 18 WITNESS ROBARE: I would not state that the 19 way you did. There are no transients that can lead to 20 unacceptable accident consequences that require 21 nonsafety grade equipment f or mitiga tion. 22 JUDGE JORDANS Well, challenges to the safety 23 equipment frequently come from the nonsafety equipment, ( 24 failures in the nonsafety equipment, is that not correct? 25 WITNESS ROBARE: That is possible -- that is O ALCERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D C. 20024 (202) 554 2345

4439 () I true. I do not know about frequently, but -- 2 JUDGE JORDAN 4 Yes. And so that equipment 3 therefore is -- well, I am saying I think the obvious 4 then, that -- I wanted to make sure I was understanding 5 you. This equipment such as control systems, feedwater 6 systems and so forth you are surely not including in 7 your definition of important to safety. 8 WITNESS ROBARE: Definitely not. 9 JUDGE JORDANS Very well. All right. Go 10 ahead. I will have other questions later on that. 11 MR. LANPHER: Thank you, Judge Jordan. 12 BY MR. LANPHERs (Resuming ) 13 0 In that last answer, Mr. Robare, Judge Jordan O G 14 asked you whether you include such things as feedwater 15 systems in your definition of important to safety. You 16 said definitely not, meaning, I believe, you definitely 17 do not classify those as sa fety-rela ted. 16 A (WITNESS ROBARE) Plus I definitely do not use 19 the term "important to safety." 20 (Laughter.) 21 0 Yes. You answered his question, though. Do 22 you agree, Mr. Robare, that it is desirable to avoid 23 challenges to safety systems such as the challenges () 24 which Judge Jordan was postulating? 25 A (WITNESS ROBARE) I agree with that. l () ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D C. 20024 (202) 554-2345

4440 1 A (WITNESS DAWE) If I might add to that answer, a 2 Mr. lanpher, I think that is what the very essence of 3 our testimony states. It is important to avoid those 4 challenges. 5 Earlier when we talked about the design fe: 6 normal service conditions, I think that you should 7 understand that in my terminology of normal service 8 conditions the same types of considerations go into it 9 that Mr. Robare was referring to. The transient to operation of a nonsafety system would have been included i 11 in my definition of its cervice condition for which we 12 have defined it. And I think that it is the very pcint 13 that our nonsafety equipment is designed to be highly 7S V 14 It is designed to rvoid or prevent the reliable. j 15 failures of that equipment which will cause transients, 16 but it is also true that the plant is designed to be 17 able to withstand those transients without those 18 transients progressing to the hypothetical severe 19 accidents for which the safety systems are designed. 20 And, in fact, within the transients there are in fact l 21 saf ety-rela ted equipment that does mitigate the l 22 trsnsients. The reactor protection system, for example,

23 is one of the main mitigators to turn those transients, i: 24 and GE models those in their transient analysis.

25 But that does not say that. Everything that t !O ALDERSON REPoATING COMPANY,INC, 400 VIRGINIA AVE. S.W WASHINGTON. 0 C. 20024 (202) 554 2345

4441 () 1 defines tha t transient or to prevent the occurrence of 2 that transient in the first place needs to be a 3 safety-related component. 4 It is true that we designed the plant to be 5 reliable to minimize the frequency of expec ted 6 transients. And I say expected only in the sense that 7 we have to recognize what components of the plant can 8 cause transients and design to preclude er minimize 9 those transients from occurring. 10 0 It is my understanding that both GE and Stone 11 and Webster evaluate, at least on a judgmental basis, 12 system structures and components which are not 13 classified as safety-related to determine what degree or O 14 what range of high s ta nda rd commensurate wi th f unction 15 needs to be applied, correct? 16 A (WITliESS ROBARE) That is correct. 17 0 Are these -- 18 Mr. Dawe? 19 A (WITNESS DAWE) That is correct f or Stone end 2L Webster, and it is at least a judgmental basis more 21 often than not. It will be based on an evaluation of 22 the component's function and the service conditions it 23 is going to see. 24 0 Are these -- did you finish your answer, Mr. 25 Dawe? I did not want to interrupt you. O ALDERSON REPORTING COMPANY,INC. 400 VIRGINIA AVE., S.W., WASHtNGTON, D C. 20024 (202) 554 2345

4442 () 1 A (WITNESS DAWE) Well, I would just like to add 2 that where it is judgmental, that judgment is based on 3 experience. 4 0 Where this judgment is based on analysis or at j 5 least in part are these analyses documented in the FSAR? 6 A (WITNESS DAWE) In the context that we are 7 talking about, the nonsafety-related systems, structures 8 and components of the plant, the design bases and the 9 functions to be performed and the descriptions of the

  . 10  systems and se on are stated in the FSAR.                        The 11  information in the FSAR is in fact backed up by the 12  encineering documentation of the project.                        To that 13  extent it is documented in the FSAR.

O 14 You will not find at the end of every line a 15 statement that refers you to a specific analysis. 16 (Counsel for Suffolk County conferring.) 17 JUDGE CARPENTER: Mr. Lanpher, while you are 18 conferring, if I might ask a couple of questions. 19 You keep using the term " highly reliable." I 20 am curious as to what that means. You certainly do not 21 order a piace of aquipment specified as being highly 22 reliable. Can you give me some feel for what that means 23 engineering-wise? 24 (Panel of witnesses conferring.) 25 WITNESS ROBARE: Reliable in terms of quality O ALDERSON REPORTING COMPANY, INC, 400 VIRGINI A AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

I _ h443 () 1 assurance. Speaking for GE now, on pages 42 a:td 3 in 2 our prefile testimony I have shown the degree of quality 3 assurance that is typically applied to nonsafety grade 4 equipment. This degree of QA is very close to that

                                                                   - i 5  applied to our fuli Appetlix B safety pride equipment, 6 and this degree is applied to approximately 90 percent 7 of the nonsafety grade equipment in the General Electric 8 scope of supply.

9 JUDGE CARPENTER: The other context I was 10 trying to understand highly reliable in, earlier in' the s 11 testimony it was said design is a process that goes on s 12 for a number of years. I think I heard the commept that 13 in fact you might say in one sense the plant has been , O 14 redesigned several times.- And' then I am listening fto 15 this testimony about essentially an ad hoc analysis l 16 about the reliability that should be specified for some 1 l 17 particular piece of equipment at some point lin time, and 18 then downstream from that if there is this evolving , 19 design you have to go back and redo all of these ad- hoc , 20 evaluations. 21 WITNESS IANNI: Yes. Judge Carpenter, I think , i 22 I can help you with that. I think someone mentioned 1 23 experience here a few minutes ago, and the thi~ng you

24 have to remember is that none of these plants were- - 25 started with a clean sheet of paper. The facts are th'at-

O - ALDERSON REPORTING COMPANY,INC. 400 V.'RGINIA AVE., S.W., WASHINGTON, D C. 20024 (202) 554-2345

4444 1 () I these specifications and reliability has largely come 2 from experience on these nonsafety systems. We found 3 out the hard way that pressure regulators are very 4 important on a BWR, and as a result of that, why, as the 5 designs evolved from the early designs -- we insist on 6 highly reliible pressure regulators -- designs were 7 changed. As we got new information from the field we

                       '8  fed it back into the design.             As a result of that, we
                       '9  have a highly reliable pressure regulator.                        And 10  experience now tells us in the operating plants that 11- that is a good piece of equipment; the same way on, for n                                                                                          .

l 12 example, the bypass valve. The same way on the r ' 13 feedwater system. 14 But a first of a kind design one worries about i 15 these things. We have changed a lot of things. You 16 - look for it and if you find something wrong you feed it 17- back. So if you plant this on a Duane plot, you find 18 that characteristically it comes downward in a negative 3 19 slope with time as you feed back the changes. 20 So when we say high reliability we mean nature 21 has told us whether we have high reliability or not. If 22 we do not, our plants have high downtime. So when I use 20 the term "high reliability" in terms of these systems,

        .)             24  that is what I mean.       It is in the context of the 25  opera tion of the plant, do we achieve high reliability a

ALDERSON REPORTING COMPANY,INC. 400 VIRG!NIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

4445 t () 1 or don't we, and tha t is really the bottom line of 2 reliability, in my opinion.

     /    3                       JUDGE C ARPENTER:         Thank you.           I think that 4     helps explain why some of these things are not 5     documer.ted in the FS AP , because they are sort of 6     buried. I do not know how you would ever dig them out 7     in a summary document as to what is going on in terms of                                        1 8     judgments over so many years.                     I have a lot of trouble 9     with how you document that other than, as you say,                                              I l

10 building the plant, seeing that it does not have much 11 downtime. 12 WITNESS IANNI Let me tell you, Judge 13 Carpenter, if you are responsible for the pressure O 14 regulator -- and we have two or three plants that go

15 down bocause of the pressure regulator -- that 16 engineering group knows it and knows it quickly, and 17 they get the heat put on them, and they have to resolve 18 the issue. So the feedback to the person who is 19 responsible for the equipment comes very quickly and , 1 20 very fast, and it is there. 21 And in addition to that our reliability people 22 do keep records of all the pieces of the plant and LERs 23 and things of this sort. It is computerized, and our 24 design engineers have access to it. But it is difficult , 25 to summarize, I admit, but the point I'm trying to make C:) ; ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4446 , () 1 is that when you are responsible for a piece of 2 equipment and it has trouble out of the field for 3 reliability reasons, that group knows, and they are made 4 to as quickly as possible rectify the situation. 5 JUDGE CARPENTER: Thank you. 6 (Counsel for Suffolk County conferring.) 7 BY MR. LANPHER4 (Resuming) 8 0 Going back to the statement we have been 9 focusing on on page 54, in view of that statement is.it 10 fair to say that you agree that within the class of 11 nonsafety-related there is a range of desired for 12 quality and reliability, some things are more important 13 than others. Is that true? 14 A (WITNESS DAWE) I would not state it quite 15 that way. I would say that some things are more complex 16 or more demanding than others and therefore require more 17 a tte n tion . 18 Q Let's take an example. We will come back to 19 it later in the testimony. The turbine bypass valve, 20 that is not fully safety-related, correct? 21 A (WITNESS DAWE) That is correct. That is not 22 a saf ety-talated component. I 23 Q It helps to prevent challenges to the safety 24 relief valves, correct? 25 A (WITNESS DAWE) It is not my understanding O ALDERSON REPORTING COMPANY. INC, 400 VIRGINTA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

r 4447

   )  1 that tha t is quite a correct statement.                        If you had a 2 turbine trip, for example, at a low enough power level 3 within the full capacity of the bypass valve, it would 4 prevent operation of the safety relief valves but at 5 high power level it only relieves a portion of the steam 6 demand, and the safety relief valves are still going to 7 operate.

8 0 The bypss valves can only take about 25 9 percent of the steam, correct? 10 A (WITNESS DAWE) Twenty-five percent of the 11 full power. ( 12 Q Mr. McGuire, is it your understanding that the l , 13 turbine bypass valve in certain situations can be ( 14 utilized to prevent challenges to the safety relief 15 valves? 16 A (WITNESS MC GUIRE) My experience basically is 17 that these plants operate at 100 percent power. Now, if 18 you get a reactor, a scram, they will open, so the l l 19 bypass valves are utilized, you know, looking at certain i 20 transients. But in the real world I have never really 21 seen it utilized to stop the lif ting of a relief valve. 22 JUDGE BRENNER: Let me make sure I understand 23 tha t, M r. McGuire. You mean that given your postulation 1 24 of normal, one hundred percent or close to one hundred 25 percent operation even if the bypass valves work fully l l l ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4448 1 there vill still be a challenge to the safety relief 2 valves; they will lift anyway? 3 A (WITNESS MC GUIRE) In certain conditions, 4 yes, that is correct. It all depends on the transient. 5 JUDGE JORDAN Perhaps I need a little 6 education at this point, and I have been thinking in 7 terms of PWRs for a long time, and I am having a little 8 trouble getting back to BWRs. 9 In the case of the loss of load turbine 10 overspeed and the closing of the turbine stop valve -- 11 is that the correct sequence -- would there not also be 12 a scram at the same time? 13 WITNESS MC GUIRE: That is, if you are above 14 30 percent power for this plant. 15 JUDGE JORDAN That is what I am talking 16 about, a plant operating at a hundred percent power, say. 17 WITNESS MC GUIRE: That would occur. When you 18 mentioned the load rejection, the first thing, you know, 19 you lo ' offsite power. Basically that is that 20 situation. That is dif ferent than the other. 21 JUDGE J3RDAN: That is not the situation you 22 were talking about. 23 WITNESS MC GUIRE: Well, if you had a load 24 rejection, your main steam isolation valves would close 25 and you would lift your relief valve definitely. O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE, S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4449 () 1 JUDGE JORDANS And as the main steam line 2 valve closed, then do you not also at the same time get 3 a reactor scram? 4 WITNESS MC GUIREs That is affirmative. . 5 JUDGE JORDAN And does that not therefore 6 reduce the power quickly enough that you do not 7 necessarily hit the relief valve? 8 WITNESS MC GUIRE: In certain cases that would 9 be true. It depends on the transient, that is the 10 thing. But my point here was at a hundred percent power 11 under transient conditions where the reactor does scram, 12 the bypass system in many cases does not aid that much 13 in mitigating the transient. I would not look at it as l l

   ) 14 a system as the words important to safety.                       I would not l     15 need it as an operator.

l l 16 1 l 17 18 19 20 21 22

23 24 25 l (2) i ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W WASHINGTON, D.C. 20024 (202) 554 2345

4450 () 1 JUDGE JORDANS Are you saying then, in the 2 majority of the cases where the turbine stop valve 3 closes, tha t a majority of those transients, there will [)) 4 be a pressure rise to the point where the pressure 5 relief valves open? 6 WITNESS McGUIRE: Again, it depends on your 7 power level and what transient you have that causes the 8 scram. 9 JUDGE JORDANS Yes, but this is more or less 10 the usual situation. That is what I gather from what 11 you are saying. Am I right? 12 WITNESS McGUIREa That is correct. 13 JUDGE JORDANS All right. O 14 (Counsel for Suffolk County conf erring.) l j 15 JUDGE BRENNER: Mr. Lanpher, when you come to l 16 convenient pause, we will take the mid-afternoon break. 17 3R. LANPHER: Well, I was going to -- I have a 18 bit more questioning on this section of the tastimony l 19 and then I was going to go into another section. So why 20 don't I try to finish this line up? l I l 21 BY MR. LANPHER (Resuming) l l 22 0 I would like to turn your attention, Mr. 23 Robare and Mr. Dawe, to pace 53 of the testimony at the ( 24 center of the page. "The absence of a distinct ... 25 that sentence, "The absence of a distinct separate l O ALDERSON REPORTING COMPANY,INC. 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4451 1 category of structures, systems and components that are 2 'important to safety' but not safety related leads to 3 essentially the same result as that suggested in the 4 Denton memorandum." 5 You use the word "e ss e n tia lly" there. That 6 implies it does not lead to exactly the same result. 7 Wha t distinction were your trying to draw here? 8 A (WITNESS DAWE) I think to me, Mr. Lanpher, 9 tha t means the only difference in results is that I do 10 not have a list called "important to safety." It does 11 not mean a difference on the results. 12 For those things that are in the plant, there 13 is no specific regulatory criteria that would be applied O 14 in terms of design requirements or QA requirements or 15 otherwise, except as maybe in regulatory guidance 16 documents, as opposed to regulation. I mean things like 17 a regulatory guide or a NUREG document. 18 Those are very specific in terms of their 19 application to specific components of the plant or 20 specific features of the plant, and this plant has 21 addressed those, has documented the way it has addressed 22 those in its safety analysis ceport, and those things 23 are the way they are supposed to be in terms of 24 regulatory guidance. I think the only difference is 25 that there is not a list called "important to safety." O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4452 ( 1 There is a list called " safety related" and the rest is 2 non-safety related. 3 But its functionability is quality. Its 4 design criteria are commensurate with what it is there 5 to do. 6 Q I understand that there is not a separate list 7 "important to safety", but do you compile lists of other 8 components or equipment that you think are relatively 9 important in your range of priorities? And thus, 10 attempt to ensure that special inspection or quality 11 controls are applied for those groups of items? 12 A (WITNESS DAWE) The compilation -- I am not 13 sure I understand. The work is performed under -- in ( 14 the Stone & Webster scope -- under procedural control. f 15 Non-safety related equipment, for example, is purchased 16 under specification in almost all cases. The 17 specifications will clearly define the requirements that 18 are placed on a piece of equipment which is being 19 purchased. 20 And when I say the requirements being placed 21 on it, I mean any particular special needs of the 22 component, the codes and standards, the industry quality 23 standards, if you will, to which that is supposed to be 24 made based on our engineering judgment. The l 25 specification will include quality assurance () l l ALCERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4453 I () 1 requirements on it. The development of that type of a # 2 document which procures a component; in this case, a 3 non-safety component, is based on the f ull judgment of 4 our company. It is based on the engineering skills of 5 the project. 6 It is f ully reviewed before it is provided to 7 our client. It is approved by our client before it goes

8 out to purchase. All of those things happen. That is 9 part of the engineering process and cycle. Each one is 10 designed as it needs to be designed for the function it 11 is being put into the plant to perform. 12 0 I believe you stated before that in most 13 cases, or a t least in some cases, actual evaluations of O 14 the function to be performed are conducted before a 15 classification decision is made. Is that correct? 16 A (WITNESS DAWE) I think my statement before was 17 that evaluation of its function and service condition 18 was performed before a particular requirc=:nc was placed 19 on it for its design or its fabrication for precurement. 20 Certainly, as our testimony shows, analyses 21 and evaluations are performed in the total scope to 22 determine our required safety-related set. The area I 23 was talking about earlier was dealing with the specific 24 parameters of a system, the pressures it would see 25 internally, the flows it would have to produce and so O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

             - - _ _ - _ _ _ _ - _ _ - _ _ - _ _ _ _ _ _ _ _ - _ - _ _ _ - _ _ _ _ _ . _ - _ _ _ _ - _ _ _ _ - _ _ _                                                       _ . _     ____--______________0

4454 1 on. That is what I meant by analysis for the function 2 of the system to determine the quality standards applied 3 -- the codes and standards applied. 4 MR. LANPHER: Judge Brenner, I think this is a 5 convenient time. 6 JUDGE BRENNER: All right. Why don't we take i l 7 a little over 15 minutes and break until 3:25. 8 (A short recess was taken.) 9 10 11 12 13 O 14 15 16 17 { 18 ( 19 20 21 22 23 24 25 O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W WASHINGTON, D.C. 20024 (202) 554-2345

4455 A During (/ 1 JUDGE BRENNER: Back on the record. 2 the break I was informed by the staff and Suffolk County

     )  3 that they have each have some miscellaneous matters they 4 want to take up at the end.             I know Suffolk County's 4

5 will be quick. I am not sure of the staff's, and what 6 we would like to do, therefore, is stop the testimony at 7 around 10 to 5:00 to have time to go into the 8 miscellaneous matters. And I have one miscellaneous 9 question about SOC that I will ask you now and we will 10 come back to it. 11 You indicated that you would follow up on 12 SOC's cross examination plan on their Contention 9, i 13 rather than follow your own.

-(:)

14 MR. LANPHER: Since it is out on the record, 15 tha t is righ t. 16 JUDGE BRENNER: Well, that is -- I do not 17 think -- if I thought I was disclosing anything I would 18 not have said it. Be aware that as of now they have no 19 plan for your to follow up on that I have received. ! 20 MR. LANPHERs I am aware that they were 21 working on a plan which I have, in fact, discussed with 22 them some and I think it will be here shortly. That is 23 my understanding. 24 JUDGE BRENNER: Today? You don't know? Well, 25 today is the due date, you know. I just wanted you to O l a ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4456 1 know that I have not received one, and with all the 2 people floating around I wanted to make sure -- 3 MR. LANPHERa I do not have it. 4 , JUDGE BRENNER: All right. That is the 5 purpose of my inquiry. You have answered my question. s 6 We will proceed with the examination. 7 BY MR. LANPHER (Resuming): 8 0 Mr. Robare, I asked Mr. Dave before the break 9 whether he agreed or disagreed really with the Denton 10 memo, and I believe he basically said that he 11 disagreed. Do you agree or disagree? 12 MR. ELLIS: I would object to that 13 characterization. The record will speak for itself. He 14 can ask the question without characterizing Mr. Dawe's 15 testimony. 16 JUDGE BRENNER: Well, you know, you usually 17 get that object when a questioner characterizes it on i 18 the way to asking another question. But in this case, 19 the questioner askeda is my characterization correct, 20 in essence. l 21 MR. ELLIS: But he is asking a different 22 witness. He said Mr. Dave said I, and then he went on 23 to say, do you agree with the Denton memo. And all I am 24 saying is that I am simply inserting for the record that 25 I do not acquiesce in his characterization of Mr. Dave's O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON D.C. 20024 (202) 554-2345

4457 rs (_) 1 answer, and the record will speak for itself. 2 MR. LANPHERs Fine. I obviously cannot 3 testify in this proceeding. I would like to go on with 4 the question. 5 JUDGE BRENNER: Why don't you rephrase it 6 without the characterization and just ask him directly. 7 BY MR. LANPHER (Resuming): 8 Q Mr. Robare, do you agree or disagree with the 9 Denton memorandua? And could you please explain why? 10 (WITNESS HOBARE) I agree with portions of the 11 Denton memorandum that imply there has been confusion 12 about safety-related, important to safety and safety 13 grade. In the application of safety-related at General 7-(_)/ 14 Electric, we simply do not use systematically a third 15 category. We use safety-related and we use non-safety 16 related, and we design the non-safety related 17 commensurate with its impor tance to plant reliability 18 and, in some cases, transient mitigation. That is the 19 answer. 20 Q L'ith respect to non-safety grade equipment, 21 you mentioned plant reliability and then in some cases, 22 transient witigation. Would it be f air to state that 23 the quality controls and engineering controls which you

   ^
  /  )

) (_/ 24 apply to non-saf ety related equipment prima rily relate 25 to those quality controls you deem to be necessary to O ALDERSON REPORTING COMPANY,INC, 400 VIRGINI A AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4458 () 1 ensure reliable power generation equipment? Really, 2 that was a little convoluted. In terms of reliable, you 3 are talking about reliable in terms of generating power. 4 (WITNESS ROBARE) That is correct. 5 0 Yes? 6 (WITNESS ROBARE) The majority of the 7 additional quality assurance requirements are for 8 purposes of plant reliability. 9 0 Now, Mr. Robare -- 10 (WITNESS ROBARE) Since the safety systems are 11 designed -- the systems that are needed f or saf ety are 12 designed to the full design and quality assurance of 13 safety type equipment. Q 14 0 Now, I believe you said that you agreed with 15 the Denton memorandum in its statement that there has 16 been some confusion. Am I to infer from th at that you 17 disagree with the remainder? That is, Mr. Denton's 18 belief that there is an important to safety category 19 implicit in the regulations, and tha t as a subset there 20 is a safety-related category? 21 (WITNESS ROBARE) I cannot disagree with how he 22 feels about that. I can only tell you how we apply 23 saf ety-rela ted and non-safety related in our designs. 24 And I think I have done that. 25 Q I think you have, too. O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON D.C. 20024 (202) 554-2345

4459 O i ., R . uNeaER, audge aorden, I e - ng te go 2 to another area, or -- Judge Morris? 3 (Board conferring.) 4 JUDGE CARPENTER: A little while ago I asked a 5 question about highly reliable and you suggested I look 6 at testimony on page 42 and 43. And in doing that, I 7 had previously noted I had some difficulty in resolving B the statements in the first paragraph on page 42 that 9 describe a series of quality assurance metals, with the 10 second paragraph which originally read, " General 11 Electric requires an identical degree." This morning, 12 it was modified by inserting the word " essentially 13 identical", which I have not any idea what that means in 9 14 a quantitative sense, to put that modifier in there. 15 So I am very confused as to why -- what is the 16 thrust of the second paragraph, as a modifier of the 17 first paragraph? The first paragraph says there are 18 different levels, and the second paragraph says they are 19 essentially the same. It would be helpful if you could 20 elaborate a little bit. 21 WITNESS ROBARE: The first paragraph describes 22 the quality assurance that is applied to General 23 Electric-provided hardware. The second paragraph talks 24 about quality assurance of General Electric 25 engineering. The reason the term " essentially" was O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. O C. 20074 (202) 554-2345

4460 1 inserted in the second paragraph is as follows. For 2 non -sa f e ty grade equipment, there is a judgment made 1 3 about the degree of design verification, design review, 4 and the supporting analysis for design adequacy'. 5 Depending upon the use of that equipment, 6 those functions could be somewhat less than the full 7 safety grade equipment. Is that a sufficient 8 explanation? 9 JUDGE CARPENTER: Yes, thank you. I was not , 10 sensitive to the emphasis on engineering in the second' , 11 parag ra ph. Thank you. l l 12 (Board conferring.) 13 JUDGE MORRISs I hate to be the dead horse, 14 but I wouli like to come back to the infamous sentence l 15 on page 54 which ends up saying, "...a range of high l i 16 standards commensurate with its function." I am just ! 17 having some semantic problems. If they are high ( l 18 standards, what does the range mean? l l 19 (Panel of witnesses conferring.) 20 WITNESS DAWE: Judge Morris, a range of high 21 standards commensurate witn its function there, as 22 discussed in other portions of our testimony, for 23 example, a quality assurance treatment of non-safety 24 related equipment means that there -- it is a graded 25 approach, but within all of the standards, for example, O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON D.C. 20024 (202) 554-2345

4461

/G

(_) 1 or codes, industry codes and standards. We have a large 2 number to select from, and this selection process, for 3 us at Stone C Webster, is based on a corporate knowledge 4 and procedures and guidelines of a technical nature that 5 sre put out to our enginaars parforming the work. 6 We can look at a mechanical code, ASME-3 7 versus ASME-8 versus a B-31-1 code and so forth. In 8 fact, in our plant, you will find that some of our 9 components which are non-safety related are built to the 10 ASME-3 code because it was our judgment that 'you would 11 use that code on that component not because it had a 12 saf ety-related function but because we wished to apply 13 that code. b;l 14 There are ranges within the codes, depending 15 upon their application to the type of component ther 16 have. It is not an absolute, so it implies a range. 17 JUDGE MORRIS: The codes are in so you can 18 design to a range, meeting the code depending on service l 19 c on di ti o n s. Isn't tha t true? 20 WITNESS DAWE: Yes, sir. l 21 JUDGE MORRIS: And as you point out, you might i 22 use Section 3 or you might use Section 8, depending on 23 function. l l N. 24 WITNESS DAWE: Yes, sir. l 25 JUDGE MORRIS: And these are mechanical l l ALDERSON REPCrlTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WAS iiNGTON, D.C. 20024 (202) 554-2345

4462 1 codes. Are they equivalent to electrical codes or 2 environmental standards which you would include in this 3 concept? 4 WITNESS DAWEs Yes, sir. 5 JUDGE MORRIS: Is this, in fact, what you mean 6 by this phrases "s range of high standards"? 7 WITNESS DAWEa Yes, sir. Tha t along with the 8 graded approach to quality assurance, which is discussed 9 elsewhere in our testimony and in GE testimony. 10 JUDGE MORRIS: When you re' late the function, 11 do you look at a system by itself and decide what its 12 function is for a range of service conditions, or do you 13 somehow characterize functions as pressurized fluid 14 containers or things of tha t kind? What kind of 15 categorization or systemization do you do when you are 16 looking at functions? 17 (Panel of witnesses conferring.) 18 WITNESS DAWE: I think your description is 19 accurate. Examples might help. For example, the 20 function of the system may be a high pressure system or 21 it may be a low pressure system, which could have some 22 bearing on its design. An example would be the turbine 23 building. 24 If we get out of systems and into structures, 25 turbine building has no safety-related function, but it O ALDERSON REPORTING COMPANY,INC, 400 VIRG!NIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

                           .                                                                 4463

( 1 has design requirements that we have placed on it to 2 assure it has no seismic failure mode that can impact an 3 adjacent building, which is the safety-related control 4 building. 5 So by function, we look not only at its own 6 function, but also its relationship in the plant to 7 other things around it. 8 JUDGE MORRISa You see,.one of our problems I 9 think is in this one sentence you have described a -- 10 pardon the expression -- a ' broad range of activities, 11 and we are having trouble relating to specifics. So 12 could you give us a few more? 13 WITNESS DAWEa Well, ano ther specific would l ( j 14 be, for example, you could have a non-safety system or a ! 15 portion of a system which, in fact, interfaces with the 16 safety-related portion of the system. At that interface 17 -- actually, I suppose this is a bad example because the 18 interface itself would be safety related -- but at the 19 interface going from the non-safety related portion into I 20 the safety-related portion, the upgrade would be made 21 right there, but that interface is, in fact, safety 22 related. 23 (Panel of witnesses conferring.) s/ 24 In evalua tion of systems, for example, in a 25 high energy line break situation, the high energy line C:) ALDERSON REPORTING COMPANY,INC, 400 VIRGINI A AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4464 1 break would look at the line break, whether it was a 2 safety-related line or a non-safety related line, which 3 means taht the design attention to the existence of that 4 line and its location would be considered in the design 3 of the system as an application of the range of 6 standards. 7 I agree that it is a broad sta tement. The 8 engineering discipline in a plant as complex as Shoreham 9 or any other nuclear plant is a very, very broad process 10 to condensa to the size of this testimony. 11 JUDGE MORRIS: Let me ask you a question which 12 you may not be able to answer. I am receiving your 13 testimony to imply that you look system by system or O 14 component by component. I have not heard you say yes or 15 no to whether you have some systematic approach to doing 16 this in deciding what level of high standards to apply 17 in design and construction. Sort of analogous to the 18 graded a pproach to 0 A. 39 WITNESS DAWE: Yes, sir. I think that 20 systematic approach for Stone & Webster is described in 21 our testimony. It is described in the section where we 22 discuss the Stone & Webster organization. The existence 23 of pr cedures and technical guidelines for our people is () 24 discussed in the preparation of specifications and use 25 f master specifications, which represent the O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE, S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4465 O 4 =aect 11=== ae ta a v 1on at or our xao 1 do over 2 period of time.

    )  3             I think in that context, it is truly
  .)

4 systematic. 5 JUDGE MORRIS: So you say it is not a single 6 discipline that is systematic, but it is the aggregate 7 of the things you have just described, and perhaps 8 thers which provide the systematic approach. 9 * ' * * *

reviews of the work as it is being done and the inter-disciplinary reviews that have been done to ensure that all people who need the knowledge are aware of the 3

way the plant is being engineered and designed and constructed. I 14 5 (Pause.) JUDGE MORRIS: Mr. Ianni, you talked about 6 reliability and the importance of feedback from operations; whether it is a safety-related system or 8 g not. And, of course, there is a goal to have reliable generation of electricity, and it is very costly if you g do not. Do you wait for something to go wrong before consideration of improvements to the system, or is there l some kind of active program that attempts to measure l 23 reliability prior to failure? WITNESS IANNI No, we do not wait for 25 l O ALDERSON REPORTING COMPANY. INC. 400 VIRGINTA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4466 1 something to happen. That example is by way of plant 2 experience. The only way you find out by plant 3 experience is if something happens or does not happen. 4 That tells you how well you are doing. 5 We do have an active program as part of the 6 design process, which looks at all the reactor 7 occurrences, the LERs and our own plant performance, and 8 computerizes all this data and disseminates this 9 information to the lead systems engineer and the 10 ' managers that are responsible for the various systems 11 and the various components. And based on that 12 information, they make a decision as to whether 13 something should be done or not done. And clearly, O 14 there is the matter of threshold, and based on that past 15 experience they either take action or they do not. 16 In addition to that, this group looks ahead, 17 and specifically, if you have a new design, for example, 18 they project ahead and set certain goals for the various 19 groups te achieva in terms of reliability. 20 JUDGE MORRISs These are perf ormance 21 objectives? 22 WITNESS IANNIs Performance objectives in 23 terms of unavailability for the particular component. 24 And you try to achieve that in your design by using a 25 lot of common sense and analyses where you think they O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4467 k 1 help. But the idea here is that everyone has a goal he 2 is trying to achieve, and there are certain overall 3 goals for the plant availability, say, for example, for 4 a new product line. 5 And we try and forecast what this is and where 6 we think we have weaknesses in the design. Then the 7 design reviews are -- either we have design reviews to 8 address that specific issue, and where we think 9 something needs to be done we attempt to do it. So, 10 where we have new areas of work or we change the design 11 significantly, we do have design reviews and we do try 12 to set what the availability objectives are to be. 13 JUDGE MORRIS 4 These are quantitative? O 14 WITNESS IANNI Yes, sir. And everyone is, 15 you know - quality is the key thing that we have to , 16 offer the customer; how much steam he gets out in a 17 year's time. And so, quality has been, you know, one of 18 the foremost things in our mind, along with safety, of 19 course. Tha t we have to have a reliable product, and 20 throughout the company, quality is a real key thing. 21 JUDGE MORRIS: I think both of you have 22 objected to the term "important to safety" and what that 23 apparently means to Mr. Donton, as being a class larger 24 than safety-related. But from the kind of activities 25 that you have described in design, design review, do you O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4468 () 1 think -- did I understand you correctly that you both 2 believe in achieving the objectives, although you do not 3 use that term? 4 WITNESS IANNIs I think that is correct, that 5 we are achieving the objectives with the real hardware 6 out in a real power plant. I think it is essentially 7 the same as if whatever Denton has in his mind would be 8 done in his manner. 9 Another point is that when we do make changes 10 and we have to redo analyses, we redo analyses. We keep 11 going back, then we make a change across the board and 12 examine all the plants. So I think tha t the end 13 objective is the same; essentially the same. b J 14 JUDGE MORRIS: Mr. Dawe, do you want to add 15 something? 16 WITNESS DAWEa Judge Morris, I was just going 17 to provide essentially the same answer that Mr. Ianni 18 did. He just started first. I agree with what he said. 19 JUDGE MORRIS: I have nothing further now. 20 JUDGE JORDANS I also wish to explore the term 21 "important to safety" and perhaps some of the questions 22 would be better to have first asked the staff and then 23 to have asked Mr. Dave, or Mr. Ianni, but we are 24 probably not going to get that chance. 25 Now, as Judge Morris has just pointed out, O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4469 1 O 1 stef f uses the tum 1mpettent to eafety es e much 2 broader term, which includes the terms that you have , 3 used that are well defined; namely, safety related or 4 safety grade. 5 What I am concerned about, however, are we 6 really trying -- are we sure we are aiming at the same 7 objective, both by the applicant and the staff? And so 8 let me ask a few questions which will show that my 9 understanding of the regula tion is not as f ull as it 10 ought to be. 11 First of all, when the staff, according to the 12 Denton memo, applies the term "important to safety", 13 does that mean that it has to -- that those items that O 14 are important to safety -- have the criteria been 15 defined, for example, do they have to meet the general 16 design criteria? What does " meet the general design 1 17 criteria" -- is it just the " safety grade" or the 18 "important to saf e ty"? Mr. Dawe, could you help me on 19 tha t? 20 21 22 23 0 v 24 25 O ALDERSON REPORTING COMPANY,INC. 400 VIRGINIA AVE., S.W., WASrilNGTON. D.C. 20024 (202) 554 2345

1 4470 () 1 WITNESS DAWE4 As we view the general design 2 criteria, we view it to be a design criteria, a 3 principal design criteria to be applied plantwide. The 4 term "important to safety," as I stated before, we at 5 Stone and Webster, and as GE has stated and as I believe 6 as many organizations in the industry would state, has 7 equated that to be the safety-related set of structures, 8 systems and components providing or being relied upon to 9 perform the safety-related functions described in 10 regulation. 11 I think that the general design criteria -- I 12 know that the general design criteria have been applied 13 to this plant, but within our licensing documents the 14 commitments we make for the term "important to safety" 15 are commitments, made f or safety-related. 16 JUDGE JORDANS And so are you saying, 17 therefore -- and I am just only trying to understand -- 18 that those systems which we will now all agree are 19 important to safety are the ones -- those systems which 20 will prevent an accident once on its way, these systems 21 have to meet the general design criteria, is that 22 correct? Or do all the plant systems have to meet the 23 general design criteria? 24 WITNESS DAWE I think that is a very 25 difficult question to ask the way you have phrased it, O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WA3HINGTON, D.C. 20024 (202) 554-2345

N > k4?1' () 1 Judge Jordan. m 2 JUDGE JORDANS could you phrase it better.for 3 me? 4 WITNESS DAWEa The general design criteria are 5 of different types. Some are very general in nature. I ' 6 think I would characterize them all as very general in 7 nature, which is probably appropriate to a general . 8 design criteria. But some apply to specific systems or 9 system functions; others apply to a philosophy of 10 design. 11 In some of the general design criterla the 12 term "important to safety" is used and in others it is / 13 not. - ( Well, I noticed in the ! 14 JUDGE JORDANS l 15 introduction of Appendix A they do use the term 16 "important to safety." 17 WITNESS DAWE: Yes, sir. 18 JUDGE JORDAN And, therefore, I wondered if l 19 that was really the same term that Denton was using -- 20 do you think it is -- when he uses the term "important 21 to safety?" 22 (Panel of witnesses conferring.) t 23 WITNESS DAWE: I think that the actual answer l /~'s I b/ 24 to that question would have to come from the man who 25 vrote the memorandum. l O l ALDERSON t.EPORTING COMPANY, thC, l 1 400 VIRGINI A AVE., S.W , WASHINGTON, O C. 20024 (202) 554-2345

4472 h 1 JUDGE JORDANS Yes. 2 WITNESS DAWE: If you will excuse me just one 3 iosent, I would ILke to -- h k 4 (Panel of witnesses conferring.) 5 WITNESS DAWEa Judge Jordan, looking at the 6 attachment to the memorandum in question, I could b 7 certainly infer'that he is looking at that same section E 8 to which you were referring, the first paragraph or so V E 9 of the appendix. He makes reference to it in the words E 10 quoted from it. g 11 I think what I tried to explain earlier 12 relative to our use of these terms was our definition of E "r And our definition of these terms, 13 these terms. c lll 14 particularly to the extent that the words "important to h 15 safety" are frequently tied, I would say in almost all 16 cases tied to the statement safety function, that leads 17 us to believe in our design philosophy that the tera 18 "important to safety" and " safety-related," at least as 19 we use it, is tied to the safety-related functions. 20 But as we have also stated, the plant design 21 has been designed and reviewed by the NRC. It does meet

22 regulation. We are convinced it is a safe plant, 23 conservatively designed plant, and I believe that the g d 24 difference in terminology that exists since -- or the 25 question in terminology that has existed since the

~ I ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

0473 () 1 memorandum is not substantive to this plant that nov 2 exists. 3 I think the questions may be somewhat 4 substantive to the industry in future licensing actions 5 in commitments that have been made using certain 6 terminologies. But I do not think those are what is at 7 issue here today. 8 JUDGE JORDAN The Denton memo came out, of 9 course, after the Three Mile Island accident, and I to presumed was in part in response to that because among 11 the Lessons Learned documents f rom Three Mile Island wa s 12 the statement that systems and equipment which were not 13 graded safety grade were nevertheless very important , 14 from the standpoint of safe operation of the plant. I 15 believe this is one of the conclusions of the Three Nile , 16 Island Task Force. And so, therefore, apparently the 17 staff is attempting to upgrade the reliability, perhaps 18 is the right word, of equipment that has not been 19 classified safety grade previously. And therefore, the 20 Denton memo was aimed at that. 21 But now then, did the Den ton -- in using the 22 term "important to safety" and saying that these systems 23 were broader in terms -- that you in fact would say were 24 nonsafety systems, has it been reflected in an increase 25 in requirements by the NRC? Hac there been anything to O ALDERSON AEPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. O C. 20024 (202) 554 2345

4474 () I go along with the Denton memo that says now these 2 systems are going to have to perform according to a 3 higher standard -- dif f erent criteria? 4 Do you know of any change in the regulations 5 that has gone along implementing this broadened term? 6 AITNESS DAWE: Judge Jordsn, I personally am 7 not aware of such changes since the Denton memorandum, 8 and in fact, the Denton memorandum itself as opposed to 9 the enclosure clesrly states tha t it was not Mr. 10 Denton's direct intention to dictate new technical 11' requirements, modify existing technical requirements or 12 broaden tre existing scope. 13 I think to the extent that the Commission as O 14 well as the industry in the Commission guidance 15 documents and the industry stsndards has interpreted 16 these terms in each case by each party, industry in its 17 standards, NRC in its guidance documents, our individual 18 companies by our own internal documents established 19 specific criteria and specific conservatisms in specific 20 areas of the plant. Those are known to us. Those were 21 used. I believe there is some confusion in the 22 terminology, but I think the essence of the guidance 23 documents from which we worked and from which the 24 Commission worked were well known to all of us, and in 25 fact have been the subject of long technical review on O ALDERSON REPORTING COMPANY INC, 400 VIRGINI A AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

447S

   )  1 this plant.

2 And I think that the safety analysis report 3 that we have submitted, and this is the evaluation 4 report that the NGC has issued , demonstrates that those 5 were known, were addressed, were done in this plant and 6 have been satisfied. 7 I would also like to add to that that beyond 8 the specific items that might be found in the Nuclear 9 Regulatory Commission guidance documents, as I said 10 before, we have areas of our plant that we have 11 recognized things as well and have done and continue to 12 do not only in this plant but in all of our plants. 13 As we have stated in our testimony, the term O l 14 "important to safety" as a regulatory term may be in 15 question. The importance to safety of careful design of 16 the entire plant is not in question, and I do not 17 believe that the difference in terminology has resulted 18 in the difference in this plant or any comparable plant 19 coming up for license toda;. 20 I think that the problems that may result from 21 the terminology will be in documents that have to be 22 changed, methods tha t have to be developed and so on, 23 and I think that is a long-range process. I think the 24 reason why we need to address that here only is because 25 of the issue raised about the term "important to safety" O G ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4476

( ) 1 and the lack of such a term in our classification 2 scheme. That is why it has been addressed. It does not 3 result in a change in the level of safety of this plant.

4 JUDGE JORDANa This is a term that I probably 5 should ask Judge Morris, because he undoubtedly can 6 answer it very well. 7 WITNT,SS IANNIa Judge Jordan, you have touched 8 on something uhich I would like to add a comment or two 9 because I am sensitive to it. And that is that this to letter you said was written after the THI situation, and 11 it arose because in that particular plant there were 12 heavy interactions between safety systems and nonsafety 7, 13 systems. And you also pointed out earlier that maybe tj 14 not everyone is familiar with the differences between a 15 PWR and a BWR. 16 I personally submit that had we had a BWR on 17 that site it would have been back on the line in three 18 weeks because our safety systems are so deep and so 19 redundant, and the nature of the BWR is such that it 20 would have flown through that situation without any 21 trouble. We do not have to depend on nonsafety systems 22 to meet the safety criteria to the extent that that type 23 of plant did, and I think this runs all through this (~s

'~) 24 whole thing with FMI and the BWR.                  There are lessons to 25 be learned and we tried to apply them, but I think we O

ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4477 ( 1 are sort of moving in the direction of neatly concluding 2 that therefore we have similar situations without 3 recognizing the true differences between the plants. 4 And I st mit that is a very important factor to keep in 5 mind. 6 Also, when we get to systems directions, the 7 BWR is a much simpler machine. It does not get confused 8 with an accident and normal operation because you do not 9 go through the supercritical situation that the other 10 type of reactors do. The system is simple. It is a pot 11 With some pipes connectai to it. It is not a little pot 12 with big pipes connected to it and with huge steam 13 generators sticking out of it at different elevations. 14 And the safety systems are redundant, and they are the 15 best in the industry. And it is pretty well-recognized 16 and I think for these reasons. You know, had there been 17 really truly adequa te safety systems, I wonder if this 18 question would ever have even come up, and I thiuk that 19 ve have to keep this distinction in mind in this. 20 JUDGE JORDAN: Well, thank you. I think it is 21 refreshing to . hear your feelings. 22 JUDOE BRENNER Okay. Wait a minute. We are 23 getting a little bit far afield. Let me interject -- I 24 will go to you in a minute, Mr. McGuire -- with this 25 guidance. O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4478 (3 (J 1 While I can certainly see, Mr. Ianni, how your 2 comment was stimulated by Judge Jordan's comment, and I 3 can also well understand why you wanted to say it, we 4 are trying to dall in the decisionmaking process with 5 greater specifics than that kind of comment can give 6 us. And we hopefully would have gotten to those 7 specifics with or without your comment. 8 What I am trying to point out is Judge Jordan 9 made the comment leading to his question, and your 10 response did not directly assist, as I heard it, in 11 answering the question; tha t is, how do you go about 12 determining which systems would fall in which catego y: 13 and he is exploring the possible ambiguities in the 14 semantic titles of things and also exploring also -- you 15 all, of course, are not the staff end not the authors of 16 the memo; your perceptions of those uses of the term as 17 they affect you or might affect you in the future in 18 order to better ascertain the distinctions that you drew 19 in your testimony and in response to questions. 20 So what I am trying to gently state is given 21 the length of your comment, it was not directly focused 22 on where we need to go; that is, to deal with 23 specifics. I do not think it was unfair. I do see how 24 it was related to Judge Jordan's lead-in, but you might 25 keep in mind the fact that we have been getting quite a O N/ ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4479 () 1 bit of generalities today in the answers, sometimes only 2 by way of introduction before you get to the specifics, 3 I will admit, but we are not going to deal with those [v) i 4 generalitias in our findings without the specifics. 5 They may nicely cap off the specifics if ycu had made 6 your case, but in and of themselves they are not going 7 to be that helpful. 8 I only make the comment in terms of thinking 9 of the future and the length of time we are going to be to here on this subject, and not worrying about a criticism 11 of the past now th a t the comment has been made. 12 With that in mind, Mr. McGuire, to you have I 13 some specifics or just a general comment, because I () 14 think Judge Jordan was ready to follow up. But I do not 15 vant to cut you off if you have something specific in 16 response to his question. 17 WITNESS MC GUIRE: Well, specifically I was l 18 going to interject my experience, because I have been 19 involved in transients, total loss of offsite power and 20 everything else, and I wanted to -- I have also been 21 trained on a pressurized wa ter reactor. And I would 22 like to just interject that I agree with Mr. Ianni that 23 the systems interaction on a boiler under the full range ( 24 of transients that I have experienced in over eight ! 25 years just does not exist like a pressurized water l l ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON O C. 2C024 (202) 554-2345

4480 ( 1 resctor because it is simpler. 2 One example would be the cleanup system. When O 3 you initiate standby liquid control, that system, / / V 4 cleanup system, could interact by removing boron from 5 the reactor wa ter if it was left to operate. But the 6 design is such that that system isolates when you start 7 the system, so interactions are looked at and are trying 8 to be negated, I think , in the boiler. And that is why 9 I just want to interject my experience. 10 JUDGE JORDANS I take part of the 11 responsibility for setting up the long answer from Mr. 12 Ianni, although I had not anticipated it, and I do not 13 have far to go now. Nevertheless, as a consequence of O 14 the TMI accident, there is a document, the 0737, some of 15 which I believe applied to BWRs. 16 Now, were there any requirements -- and I 17 cannot think of any offhand -- were there any 18 requirements from 0737 which changed the amount of 19 either quality assurance or redundancy or anything else 20 of the nonsafety grade system? In other words, was that 21 observation that nonssfety equipment led to the accident 22 as one of the Lessons Learned, was that ever followed up 23 in any of the NUREC requirements of 0737? 24 Mr. Dawe, do you know? 25 WITNESS DAWE4 Just one moment, Judge Jordan. O ALDERSON REPORTING COMPANY,INC, 400 VIRG ANIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345 l

4481 () 1 JUDGE JORDANS All right. 2 (Panel of witnesses conferring.) 3 WITNESS DAWEs Judge Jordan, in response to 4 your question, va are not aware of any changes in 5 classification that came out of NUREG-0737. There are 6 certainly many, many items in there that the boiling 7 water reactors are addressing with respect to 8 re-analysis or administrative items, modes of operation 9 and so on, but not that change the classification of to components to our knowledge. 11 J0DGE JORDANS Okay. I think that is all I 12 have. Thanks. 13 JUDGF MORRIS: I just wanted to be sure that 7_ V 14 that answer included GE. 15 WITNESS ROBARE: Yes, that is correct. 16 WITNESS XASCSAK Excuse me, Judge Brenner. I 17 wonder if you could allow me to just add a little 18 footnote to this discussion we have had on quality 19 standards, and although I have been somewhat -- LILCO 20 has been somewhat silent in these discus,11ons. I wanted 21 to point out that. 22 As I mentioned earlier, in our design review 23 process one of the things we fervently try to interject (3 w/ 24 is our experience, and although our experience is 25 somewhat limited in the nuclear field, we do have five ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W , WASHINGTON, D.C. 20024 (202) 554-2345

4482 () 1 sites where we operate fossil plants. We have been 2 operating them for many years, and much of the equipment 3 that is used in those facilities are the same types of 4 equipment that are used in a nuclear facility, 5 particularly in the balance of plant area. And is that 6 type of feedback, that experience and the quality 7 standards that our engineers are very familiar with. 8 And I am familiar with the problems we have had at those 9 facilities and in reviewing the design documents that 10 are prepared by GE and Stone and Webster, that type of 11 information is fed back, and we attempt to apply as high 12 a quality standards in terms of the engineering 13 standards to assure high reliability for these O 14 facilities. 15 So I just wanted to interject that little 16 footnote. 17 JUDGE BRENNER: Mr. Lanpher, did you want to 18 move to that cther area now, or do you want to come back 19 to this? Okay. 20 ER. LANPHER: The Board's questioning was a 21 little broader than I originally had thought it would be. 22 JUDGE BRENNER: We get carried away sometimes. 23 MR. LANPHERs Judge Brenner, I was passed a 24 message which I would like to just bring to your 25 attention now. It is from Mr. Latham concerning the O ALDERSON REPORTING COMPANY. INC, 400 VIRGINIA AVE S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4483

    )           1 satter you brought up I think off the record about his 2 cross plan. And it says that he was planning to deliver

(} 3 4 it tomorrow morning, and he wanted me to interrupt the proceeding to see if that would be an accaptable 5 schedule. 6 JUDGE BRENNER: Yes. Mainly because he will 7 not be here today and he will be here tomorrow? 8 MR. LANPHERa I did not speak to him. That is 1 9 the sum of my knowledge. 10 JUDGE BRENNER: That is acceptable. 11 HR. LANPHER: Thank you. 12 JUDGE BRENNERa My main reason for raising it 13 other than the schedule was to make sure that you did O 14 not willy-nilly up te the date of the contention 15 believing somebody else had a cross plan that you 18 thought you were going to follow up on and find out 17 thare was none. And I did not know the extent of your ) 18 knowledge as to their plans, and that is one reason I 19 raised it. 1 20 MR. LANPHER: I appreciate that. Thank you 21 very much. l 22 BY MR. LANPHER: (Resuming) 23 0 Mr. Dawe, to follow up on one question that p)rs 24 Judge Jordan asked, I think it was to you, it concerned 25 whether the general design criteria have been applied to O l ALDERSON REPORTING COMPANY. INC. 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4484 10 1 etructure the erste=s and co oonente other then those 2 classified as saf ety-related. I just want to understand 3 your answer. I believe it was that you have applied it 4 only to -- you applied the GDC only to those items which 5 you hava Olsssified as safety-related at Shoreham, is 6 that correct? 7 A (WITNESS DAWE) No, sir, that is not correct, 8 and I do not think I implied that. As we have stated, 9 quality standards and quality assurance are applied 10 scross the pisnt whether we are talking saf ety-rela ted 11 or nonsaf ety-rela ted. And in the nonsafety-related 12 areas we feel that we have applied those commensurate 13 with the function of the systems in the plant. O 14 I do not think I tried to imply that the 15 split, as you described it -- 16 17 18 19 20 21 22 23 24 25 O ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4485 () 1 0 I understood you to equate "important to 2 safety" with " safety-related." , 3 A (WITNESS DAWE) Yes, sir. The terms are 4 synonomous the way we use them. 5 0 Right. In equating the two, why is it not 6 true, then, that you have applied the GDC only in 7 connection with safety-related systems, structures or 8 components? 9 A (WITNESS DAWE) I think the reason that is true 10 is, as I said, we view the general design criteria as a 11 general design criteria for the plant, and we have 12 applied the philosophy of the general design criteria to 13 all of our work. And in fact, in the general design O 14 criteria, there are specific criteria that mention items 15 that are definitely not safety related such as the 16 offsite power systema. But our offsite power systems 17 have been addressed as they are stated in the general 18 design criteria. 19 In response to Regulatory Guide 1.29, we have 20 addressed seismic events and their impacts on the 21 non-safety equipment to assure that the non-safety 22 equipment will not damage the safety-related equipment. 23 That is a specific requirement, but that is an 24 implementation of a general design criteria, and that is 2e the way va have designed this plant. And I think that O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4486 1 type of a philosophy is found throughout the general 2 design criteria. 3 (Counsel'for Suffolk County conferring.) 4 MR. LANPHER: Okay. I am going to turn to a 5 different area. 6 BY MR. LANPHER (Resuming): 7 Q Mr. McGuire, I would like to turn to your 8 testimony. I guess it is page 129, sir. 9 MR. ELLIS Judge, may I make a suggestion 10 that might make things easier? Perhaps Mr. McGuire can 11 change places with Mr. Robare. Perhaps that would make 12 things easier. 13 JUDGE BRENNERa Do you think Mr. McGuire is 14 going to ha ve a stiff neck if we leave him there too 15 long? Off the record. 16 (Discussion off the record. ) 17 JUDGE BRENNER: Let's go back on the record. 18 BY MR. LANPHER (Resuming): 19 Q Mr. McGuire, in your prior nuclear experience 20 it has been primarily in operations, I believe you said. 21 A (WITNESS McGUIRE) That is correct. 22 0 Have you ever classified systems, structures 23 or components for inclusion in a table such as Table 24 3.2.1-1, which we have been referring to? 25 A (WITNESS McGUIRE) I have reviewed systems that O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHING TON, D.C. 20024 (202) 554-2345

4487 () 1 have been classified to see if th e y were correct, from 2 an operational standpoint. For example, every time they 3 make a plant modification, we have to review the impact 4 of that modification on safety-related equipment, even 5 if it is a non-safety related component. 6 This is usually done by the plant operating 7 review committee, which is a multi-discipline group that - 8 looks at the modification from the standpoint of its > 9 impact on safety. 10 0 Okay. So you have reviewed classifications. 11 You have never performed them initially yourself? 12 A (WITNESS McGUIRE) That is a design function 13 and I am an operator, so the answer is no. 14 0 The answer to my next question -- you have 15 never designed systems? 16 A (WITNESS McGUIRE) That is correct. 17 Q Now, the first paragraph of the long paragraph 18 about seven lines down or so, you say, "I am of the 19 opinion, as a reactor operator and nuclear plant 20 manager, that the classification of the structures, 21 systems and components used in the Shoreham EOPs is 22 correct and consistent with other BWRs." This goes on. 23 Now, you have the caveat "as a reactor 24 operator." What are you conveying by that caveat? 25 A (WITNESS McGUIRE) Well, you know there is a O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

4488

  /3 V   1 design criteria which you were talking about before.

2 Saf ety-rela ted systems are redundant; they are 3 independent, they are testable, they have a programmatic 4 quality assurance program applied to them. As a 5 sideline from that, then you go to operate them you have 6 so many systems to rely on, you know, that even 7 considering multiple failures, I feel very safe in 8 operating a boiling water reactor. 9 Q I am not sure you answered -- ! 10 A (WITNESS McGUIRE) I think there is a 11 difference between a designer and an operator I think is 12 what I am trying to put across. 13 2 That is clear from your testimony, sir, and I O 14 am trying to find out wha t the operator perspective is 15 in terms of classification of structures. You say that 16 you are of the opinion as an operator that the 17 classification is correct at Shoreham. 18 A (WITNESS McGUIRE) That means tha t I feel as an l 19 operator, if I was to operate the plant and we had a 20 loss of offsite power which left me with just the 21 safety-related systems, that I would he.,e more than one 22 way to successfully put that plant in cold shutdown. 23 That is what I meant. f~h d 24 0 So this testimony really is, or this sentence 25 says that you believe, after reviewing the l l 1 l ALDERSON REPORTING COMPANY,INC, l 400 VIRGINIA AVE., S.W., W ASHINGTON, D C. 20024 (202) 554 2345

4489 (- (,/ 1 safety-related systems, that there are sufficient 2 safety-rela ted systems to bring the plant to cold 3 shutdown. 4 A (WITNESS McGUIRE) That is correct. And also, 5 if there da.- failure of the safety-related systems, 6 there a re ways of also bringing the plant, you know, to 7 cold c".utdo sn . Because I mentioned before the design 8 requires redundancy, independency and everything else, 9 so that enables the operator, if he does not have a core 10 spray system, the A-1, the B-1 is always available. If 11 he does not have the B core spray, he has the LPCI 12 injection mode. So there are many ways of getting water

  -  13 into the reactor, is what I am alluding to.

14 0 In that last answer, you referred each time to 15 "other safety-related systems." Correct? 16 A (WITNESS McGUIPE) That is correct. 17 Q Wha t review have you performed of the 18 classification of systems, structures and components 19 used at Shoreham? 20 A (WITNESS McGUIRE) As I mentioned before, in 21 1979, I was involved with the short-term Lessons Learned 22 Task Force for the boiling vster reactor owners group. l 23 Our job basically was to look at the systems for the BWR 24 1 through 4, and to answer specific questions posed by 25 the NRC. Many of those questions were way outside the O ALDERSON REPORTING COMPANY,INC. 400 VIRGINIA AVE.. S.W., WASHINGTON. O C. 20024 (202) 554-2345

4490 g) (_ 1 general design criteria, including multiple failures, 2 loss of all AC power on the site, just to try to get a 3 feel for whether or not these plants did require any I 4 modifications. 5 And what we found when we looked at each 6 individual boiling water reactor was that when the 7 designers applied the design criteria, tha t there was I 8 enough redundancy, enough back-ups, so that these plants 9 could achieve the objective of cold shutdown many 10 different ways with the safety-related systems that were 11 available. 12 0 Let se make my question more narrow. Aside 13 from your work on the short-term Lessons Learned phase, O 14 Lessons Learned Task Force, I guess you used that term, 15 what review of the classification of structures, systems 16 and components at Shoreham have you performed? 17 A (WITNESS McGUIRE) That is all. 18 (Panel of witnesses conferring.) 19 Do you mean outside of these proceedings or 20 including these proceedings? In these proceedings, I 21 have reviewed them with the designers in preparation f or 22 the hea rings. 23 0 Okay. You said you reviewed the systems I 24 think with the designers. What did that review entail, 25 sir? O ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4491 ( ') 1 A (WITNESS McGUIRE) Basically, looking at the 2 table in question here, 3.2.1-1, and taking a look at 3 the emergency operating procedures and how the emergency 4 operating procedures were within them to utilize both 5 systems. And very nicaly, this is a boiling water 6 reactor number 4, which is the same that I got af 7 initial training on, so the sistems are basically 8 essentially identical in function. 9 0 When you reviewed Table 3.2.1-1, was that 10 reviewed for the purpose of identifying in your own mind 11 what systems were classified as safety related and what 12 were not, and components within them? I 13 A (WITNESS McGUIRE) What that does to me, it U~3 14 indicates to me when I look at a component that is on a 15 list like that I know the designers have applied the 16 criteria: redundancy, independence and everything else 17 to those items, and that is what I am familiar with. 18 That is what the review consisted of. 19 0 The review was not for the purpose of 20 analyzing whether in fact the correct classifications 21 had been imposed on a particular component? 22 A (WITNESS McGUIRE) I would say the review was 23 more or less to, in my own mind, to figure out the s s/ 24 availability of that component under various adverse l 25 situations. l ALDERSON REPORTING COMPANY. INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4492

  )  1      0     And did you -- did you finish your answer, sir?

2 A (WITNESS McGUIRE) Well, also considering it is 3 consistent with other plan ts, the review, you know, that 4 is the type of review I did, to see that it was 5 consistent. 6 0 But you did not independently determine that a 7 particular classification or classifications were, in 8 fact, correct? 9 A (WITNESS McGUIRE) Tha t is a design f unction, 10 as I mentioned before. 11 JUDGE BRENNER: Mr. Lanpher, I wonder if I 12 could jump in for a moment. 13 MR. LANPHER: Of course. O 14 JUDGE BRENNER: As I understand the sequence, 15 Mr. McGuire, you basically focused on the procedures to 16 respond to the allegations in the county's testimony, 17 but looked back to that FSAR table to make sure you 18 understood how the different equipment called for in 19 different sequences of the procedure had been classified 20 by the designers. 21 WITNESS McGUIRE: Yes, sir, that is correct. 22 JUC3E BRENNER: In the course of this review, 23 did you make any attempt in your own mind to close the k 24 loop, so to speak; that is, when you saw what equipment 25 was being called upon in the emergency procedures, to O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON D C. 20024 (202) 554 2345

4493 () 1 determine in your own mind how that equipment should be 2 classified given its use in the procedure, and then take 3 a look at the table to see if it was appropriately 4 classified f rom an operator's point of view, given the 5 use in the procedure, in the emergency procedure. 6 WIINESS McGUIRE: The emergency procadure -- 7 again, there is a difference between designers and 8 operators, Judge. I will try to show you how we look at 9 things differently. 10 When you have a transient, there is nobody-11 that I know of that can tell you what transient you are 12 experiencing until that transient is completed, or 13 whether you have a transient or an accident. What the 0 14 energency operating procedures attempt to do, after TMI 15 we looked at this and said that we had not given the 16 operator enough guidance in considering more than a 17 single failure. So again, the single failure is a 18 design criterion, and what we wanted to do is write 19 something that encompassed everything from the smallest 20 transient to a large accident, and do this in a 21 systema tic approach. 22 And we actually started out with a flow 23 diagram considering components that were available, 24 whether safety rela ted or non-saf ety related, because as 25 I mentioned, if the reactor water level is going down, O ALDERSoN REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

4494 () 1 it could be a minor thing which feedwater could take 2 care of. So with that in mind, we wanted to guide the 3 operator and to improve his confidence tha t the actions 4 that he was taking in systematically trying to restore 5 water level, which is the primary objective for an 6 operator in any transient initially, that he had proper 7 guidance so that he would react properly. 8 The feedwater is running like -- if you 9 plotted your own design criteria, loss of offsite power 10 vill stop you from having those non-safety related 11 systems. But if you have a feedwater system running and 12 water level is decreasing for a different reason than 13 feedwater, the feedwater system itself is going to 0 14 automatically try to compensate for that. 15 So what we tried to do is there is no cutoffs 16 there is a range I guess you might want to call it of 17 operator actions depending on what he is faced with. We 18 are more concerned with achieving an objective and how 19 we achieve that objective -- we do what we feel is > 20 taking a component that is operating if it is available 21 and utilize it. If that is not available, the next step 22 there is always a safety-related system behind these 23 non-safety related systems tha t the operator knows he 24 can fall back on. 25 JUDGE BRENNER: Well, what I was wondering in O ALDERSON REPORTING COMPAN'i,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

4495 ( 1 part of your review was whether you looked for any 2 sequences from an operator's point of view that would be 3 difficult for the operator to perform; that is, move to 4 the next step if a particular piece of equipment failed, 5 either because ha might have trouble diagnosing its 6 failure promptly enough or for some other reason, And 7 therefore, go bsek to ascertain whethat or not tha t 8 piece of equipment -- which failure would make the next 9 step quite dif ficult for the operator -- whether that 10 was classified appropriately given the difficulty of 11 operation. And tha t is what I am asking you from the 12 standpoint of an operator. 13 WITNESS McGUIRE: Yes, I did that and I did O 14 not see any problems. 15 JUDGE BRENNER: 111 right. I guess we later 16 on will gat to some particular systems which might be 17 talked about.

18 JUDGE JORDANS Just one item of l

19 clarification. You said we did this in the way of l 20 operating -- getting these improved operating procedures l 21 for emergency operations. Who is "we"? l 22 WITNESS McGUIRE: The BWR owners group in 23 conjunction with General Electric. General Electric 24 Em e rgen cy Procedures Guidelines, which basically I think 25 address systems sequentially in the order in which they ( l l ALDERSoN REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4496 78 () I should be used. The plant then specifically wrote a 2 procedura from that guideline, the Shoreham procedures 3 -- in this whole process we used the old type of 4 emergency procedures as well as the new type at 5 simulators to try to get a feel for the operator's 6 responsiveness to the different types of reactions that 7 were exparted of him. 8 And we used old operators, experienced 9 operators and those with very little experience. The 10 performance using the new emergency operating procedures 11 far outweighed the other ones. In fact, with the 12 emergency operating procedures, the new ones, all 13 operators were in a position to place the plant in a 0 14 safe condition. 15 With the older procedures which were 16 event-oriented or which required an operator to diagnose 17 basically whether he had a small break, a big break, an 18 intermediate break, the operator stalled at a certain 19 point. 20 JUDGE JORDANS Yes, I understand that. But 21 what I did not understand was the part tha t you pla yed . l l 22 Did you issue the operating guidelines from GE? Did you 23 write the emergency operating procedures for Shoreham, () 24 or did you just review them is what I thought you had 25 done. O ALDERSON REPORTING COMPANY,INC, 400 VIRGINI A AVE., S.W., WASHINGTON, D C. 20024 (202) 554-2345

4497 () 1 WITNESS McGUIRE: I was involved initially in 2 getting the thing started, and I backed out of the 3 owners group because of my job as a plant manager. I 4 could not spend the time necessary. 5 But I have been involved from the standpoint 6 with INPO when I worked for them in taking a look at the 7 approach, so intermittently over the past couple of 8 years I have been involved, and specifically, I did 9 raview Shoreham's. 10 JUDGE JORDAN: Thank you. 11 JUDGE BRENNER: Let me pick up on your last 12 answer because it also follows up on what I asked you 13 during the qualifications round in what seems like guite O 14 a while ago today. 15 Did you review only those procedures necessary 16 to respond to the county's testimony, which are 17 understandably the ones focused on in your testimony, or 18 are you also being engaged to perform that review of all 19 Shoreham emergency procedures? 20 WITNESS McGUIRE: I have been asked by the 21 plant basically to look at what has to be done. As I 22 mentioned before, many design changes have occurred at 23 Shoreham in the construction process. And what we want () 24 to do now is systematically put together a program for 25 them, state an objective such that when the systems are O ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D C. 20024 (202) 554-2345

4498 () 1 turned over to plant staff , we can take the operating 2 procedures and surveillance procedures and all 3 procedures having to do with operating the plant and 4 basically validate them, that they are true technically, 5 and run those where we can run those for performance. 6 You know, make sure they will work. 7 JUDGE BRENNER: Are you going to be involved 8 in the training at all also to see if the curricula 9 properly protrays what should be done by the operator in 10 the procedures that you have reviewed? 11 WITNESS McGUIRE: Part of the systematic -- 12 well, the operators themselves right now have to perform 13 at the simulator, and this is going to be performance f\ \'! 14 tests which basically utilize the emergency operating 15 procedures as part of that phase, showing that they are 16 competent and can handle various situations. 17 Personally, the only thing I am going to do is 18 make sure the lesson plan they use to train operators as 19 part of the systematic approach are correct. 20 JUDGE BRENNER That is what I was asking. 21 You answered it in much better terms than I asked it. 22 Mr. Lanpher. 23 BY MR. LANPHER (Resuming): () 24 0 Mr. McGuire, you referred to new emergency 25 operating procedures and the old one. You equated the O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4499 l l l O v 1 old ones with event-oriented, I believe. Are the new 2 ones you are referring to the symptom-oriented? i 3 A (WITNESS McGUIRE) Yes, sir. 4 0 Turning to page 136, I understand you did not ! 5 choose these emergency procedures to review. But can we 6 ca tego rize these among event-oriented or 7 symptom-oriented? 8 A (WITNESS McGUIRE) These are event-oriented. s 9 0 Are they all event-oriented? to A (WITNESS McGUIRE) Except for maybe the , 11 shutdown procedure. . 12 13 14 i 15 16 17 18 19 20 21 22 23 24 25 O ALDERSON REPORTING COM.PANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

r 4500 e (]) 1 Q Now, see thase -- these are emergency 2 o pe ra ting procedures at Shoreham that will probably be 3 revised in the future, but these are the ones in effect 4 or have been in effect. I think we just got some new 5 ones just in the last couple of days. 6 A (WITNESS MC GUIRE) The plant itself has been 7 going through the process of validating these procedures 8 at the simulator, and that process I think was 9 completed, procedures were written with comments from to the NRC. This is what I have been told. And they have h 11 been reviewed by the Review of Operatione Committee, and 12 they have been approved towards the beginning of May, 13 the majority of them. So these are superseding the se 14 procedures that were in the testimony. y 15 One of the reasons we did this was that in L 16 tryino to look at every event possible we had probably I ~ 17 on the order of 30 to 35 emergency procedures which were 18 event-oriented. If we tried to cover every event, we 19 probably would end up on the order of another 20 after k 20 post-TMI. 21 The process that was looked it was the 2 22 response to each event was looked at, and if you overlap 23 them, they are very common in nature, the response of D s (f 24 the operator. In fact, the commonalities were so 25 gla r2ng that we were able to do basically the same thing r [ ALDERSON REPORTING COMPANY,INC, g_ 400 VIRGINI A AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

4501 () 1 in nine procedures instead of I would say 50. So this 2 is why we went to a symptom base. This is easier for 3 the operator to understand. 4 0 Hr. McGuire, from reading four testimony and 5 hesring your earlier comments, you do not have -- you do 6 not believe in event-oriented operating procedures, is 7 that correct? 8 A (WIINESS MC GUIRE) That is correct. I can 9 clarify tha t, expand on that if you like. The reason I 10 do not or nost operators do not is that we are there for 11 a purpose. We have a specific objective, and that is to 12 make sure that that plant is running safely. 13 Now, we will use everything that you can 14 possibly think of to achieve that objective and sort of 15 figure out whether you have a small break, you know, 16 with a loss of feedwater. As f ar as an operator is 17 concerned, his water level is going down, and he will 18 verify that safety systems are starting, that, you know, 19 level decreases fast enough, or he will turn the systems 20 on himself. 21 What we try to do is make it simple for him, 22 and they were being confused by trying to memorize all 23 these different events and trying to correlate ( 24 anunciators to identify an event where it does not make 25 any difference. The idea is to get the water level back O ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4502 h 1 up. 2 0 I am confused then. At page 132 you said that 3 the current -- toward the bottom of the page, sir -- the 4 current Shoreham EDPs are consistent with the 5 recommendation of the BWR owners subcommittee. Those 6 recommendations were to go to symptom-oriented E0Ps, 7 correct? 8 A (WITNESS MC GUIRE) Yes, sir. 9 0 And all or all except one of these E0Ps that 10 se are talking about in this specific testimony are 11 event-oriented, so please tell me how thes; E0Ps are 12 consistent with the subgroup recommendations. 13 A (WITNESS HC GUIRE) What I have done is I have 14 answered your testimony. I have taken each one of the 15 procedures which are now outdated and I reviewed them 16 from the way you had testified on those. And though you 17 had mentioned, for example, if a nonsaf ety-related 18 system was mentioned, the reason I did this is because I 19 do not think it makes a difference whether it is 20 event-orianted or symptom-based. We were going to 21 mention nonsafety-related components in the emergency 22 procedures. That is the point. What I wanted to show 23 here is that most of those are just for equipment 24 protection. 25 0 I want you to go back and look at your O f ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4503 () 1 testimony on page 132 where I just quoted at the bottom 2 of the page. When you say the current EOPs you mean 3 some EOPs other than the ones you later testify about, 4 or is it these E0Ps that were referenced in the Suffolk 5 County testimony that you are referring to? 6 A (WITNESS MC GUIEE) These are the E0Ps that 7 were referanced in the Suffolk County testimony. 8 Q Haven't you testified orally today that those 9 E0Ps are not consistent with the subgroup 10 recommendations? 11 A (WITNESS MC GUIRE) Yes, sir. And in those 12 cases where they were -- where the other procedure would 13 substitute for the ever.t procedure, they had been done, 14 and that process took place about a month ago. , 15 JUDGE BRENNER: Let me make sure I 16 understand. I think you asked a double question, the 17 question before, and that is why you got the answer 18 yes. These EDPs, M r. McGuire, let me see if I can 19 straighten it out. Maybe I won't. 20 The E0Ps that you discuss in your testimony, 21 which are the ones raised by Suffolk County's testimony, 22 are not those to which you apply the term " current," are 23 they? h 24 WITNESS MC GUIRE: Some of these, that's 25 true. Some are still current, but the majo rity of them O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE.. S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4504 I have been ceplaced. ([ ) 2 JUDGE BRENNER: So in the sentence that is 3 troubling Mr. Lanpher, when you refer to current E0Ps, 4 you are thinking of the EOPs as th ey have been updated 5 recently. 6 WITNESS MC GUIRE: That is correct. 7 BY MR. LANPHER: (Resuming) 8 0 Aren't these still event-oriented E0Ps, these 9 updates tha t we are talking about? 10 A (WITNESS MC GUIRE) No. I would classify them 11 more as a symptoms f or example, one is classified at 12 level restoration which does not tell you why you are _ 13 losing water, but it just tells you how to get water

\'   14 back into the reactor.        So that is more reacting to a 15 symptom which is a decrease in water level than to try 16 to analyze what is causing that.

17 (Counsel for Suffolk County conferring.) 18 JUDGE BRENNER: Mr. Lanpher, whenever it is 19 convenient you can come to a stopping point for the day. 20 (Pause.) 21 JUDGE JORDANS Just one matter of 22 clarification. I am familiar with the ATOG program. Is 23 this simila r or identical to the ATOG program? () 24 WITN ESS MC GUIRE: Yes, sir. It is similar in 25 nature. It is trying to achieve the same objective.

O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4505 () 1 JUDGE JORDANS Just a different term. 2 WITNESS MC GUIRE: That is the BCW terminology 3 if I am not mistaken. 4 JUDGE JORDAN: I sea. Now, this was in 5 response to the Lessons Learned, I believa, and is 6 included in 0737, is that not correct, as a requirement? 7 WITNESS MC GUIRE: It is in response to 8 Lessc7s Learned, but I am not sure whether it is covered 9 in the regulation right now. 10 JUDGE JORDAN: Okay. I think there is an 0737 11 item, but it does not matter. 12 MR. LANPHER: Judge Brenner, this is probably 13 as convenient a place as any. O 14 JUDGE BRENNER: Maybe we should note what ATOG 15 is for the record. I forget. 16 JUDGE JORDAN: I believe it stands for 17 anticipated transient operating guidelines, but I am not 18 positive about that. 19 WITNESS MC GUIRE: Yes, that is what it is. 20 (Board confarring.) 21 JUDGE BRENNER: We can excuse the panel for 22 the day if you woald like to leave. I imagina you 23 might. And although we are not r.ecessing for the day -- () 24 we have some business, procedural business -- we will be 25 reconvening at 9:00 in the morning in this room. I hope O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

I 4506 (]) 1 I said 9:00. If I did not, that is what I meant to say. 2 JUDGE JORDAN: You did. 3 JUDGE BRENNER: All right, Mr. Lanpher. We 4 vill take your business first because I think it will be 5 quick. 6 MR. LANPHER: The first item, Mr. Latham would 7 prefer to respond to the motion to strike at the start 8 of business next Tuesday if -- 9 JUDGE BRENNER: I really -- is he going to

10 respond orally or in writing? It is a simple motion, 11 and frankly, it is fresh in my mind this week. Why 12 don 't you tell him to get in sometime this week? If.he 13 insists on a written response I will understand why he 14 might want to wait until next Tuesday to file it, but I d 15 am perf ectly happy with an oral response. It is a very 16 simple motion to strike, and I would just like it behind 17 me this week while it is fresh in my mind if possible.

18 I am not berating you. I am jost -- basically 19 -- 20 MR. LANPHER: I think it was my fault. I had 21 understood that you had said either you would take it up 22 either some morning this week or at the very beginning i 23 of next week. I think that is what I conveyed, and that llh 24 is what he jumped on. 25 JUDGE BRENNER: This is the difficulty of his O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

4507 () 1 not being here. What I meant to convey is if he 2 insisted with a good reason, I would carry it over; but 3 somebody has to be here tomorrow to deliver the cross 4 plan. 5 MR. LANPHER: My message to him got there too 6 late. There is a messenger delivering it now. 7 (Laughter.) 8 JUDGE BRENNER Well, we have come here much 9 further than he has to come. 10 MR. LANPHER: I will get back to him. I am 11 just conveying information at this point. 12 JUDGE BRENNER Tell him unless it is 13 impossible to come in at the beginning of one day this O 14 week, I would like to do it. And the reason I wanted to 15 know the day in aivance, we are ready in any time, but 16 if LILCD wanted to take a position, I wanted them to do 17 it at the same time, and I did not know if they needed a 18 particular counsel available. And I think we can do it, 19 and it would assist the staff to get a ruling one way or 20 the other sooner rather than later presumably. 21 MR. LANPHER: I will convey that message. 22 The second item I would like to bring up is 23 that with respect to emergency planning, discovery and (]) 24 testimony I think are the two items you are mentioning, 25 my colleague, Mr. Brown, is going to be coming to Long h ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554 2345

i 4508 () 1 Irland tomorrow, and we would suggest that we take up 2 that matter at the opening of business on Thursday 3 morning, if that is agreeable to the Board. 4 JUDGE BRENNER: That is fine. 5 MR. LANPHER: Thank you. 6 JUDGE BRENNER: Since we are holding it over 7 until Thursday, it would be useful if the parties could 8 mutually discuss a discovery schedule before then, and I 9 will tell you what we had in mind in broad parameters. 10 I am not seeking a response. I understand this is Mr. 11 Brown's bailiwick. 12 But consistent with what we think will be the 13 overall schedule, potential schedule of the proceeding b ' 14 and what we have contemplated in some earlier rulings in 15 forecasting where we would go, I think what we had in 16 mind, the contentions are going to be filed June 22, and 17 we will talk on Thurstay, but they will be discussed 18 prior to the filing date so we can get the positions

19 down also on that same day to the other parties and get, 20 you know, refined contentions that were the subject of 21 discussion. 22 MR. LANPHER: I think those are being 23 delivered tomorrow to LILCO -- I know there has been A (_/ 24 discussion going on -- either tomorrow or Thursday for 25 discussion pucposes. ALDERSON REPORTING COMPANY. INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2145

4509 s () 1 JUDGE BRENNER: All right. As I said, I was ! 2 just giving you a forecast. I wanted to talk about it 3 on Thursday rather than putting you on the spot right 4 now, but I am glad to hear you report on it. We have in 5 mind a discovery schedule that ended around the end of 6 July and the testimony filing schedula around the end of 7 August. And I think that is reasonable and consistent 8 with what va hai :ontemplated. And also to require tha t 9 discovery documents on emergency planning began to take 10 place sometime ago, and we have seen evidence that that 11 has been taking place, as well as extensive informal 12 discussions and so on, all the tools we used in the past 13 to expedite discovery. 14 MR. LANPHER: Judge Brenner, I was wondering 15 if I could ask that when we have this session on 16 Thursday morning, I know that one of the factors before 17 was the staff review of the onsite emergency 18 preparedness and where that stood in discussing the 19 schedule; so it might be good if we could get an update 20 on shall we say the best estimate. 21 MR. REPKA: I can tell you that right now.

22 JUDGE BRENNEP: All right. 23 MR. REPKA The onsite appraisal is scheduled () 24 for July 19 right now. 25 I just want to also indicate that we have not O ALDERSON REPORTING COMPANY,INC, 400 VIRGINI A AVE., S.W., WASHINGTON. D.C. 20024 (202) 554 2345

4510 t () 1 seen any emergency planning contentions yet, so if there 2 is -- 3 JUDGE BRENNER: He said tomorrow or the next 4 day. 5 MR. REPKA: Okay. 6 JUDGE BRENNER: Why did you add the "right 7 now" in a rather tentative tone of voice in terms of the 8 sch edule ? Is that going to be as firm as the last one? 9 NR. REPKA: That is as firm as any schedule. 10 I have no reason to believe that that schedule will not 11 be met, and that date was just given to me on Tuesday of 12 last week, so -- 13 JUDGE BRENNER: All right. It is the Board's O 14 view that a date for something like that should be less 15 tentative than an estimated date for the completion of 16 review; that i,s , they are going onsite to do something, 17 however it turns out, and that should be subject to less 18 uncertainty than the completion of a review where you 19 run into a stumbling block in the middle of the review. 20 MR. REPKA: The date is not tentative as far 21 as the staff member is concerned. It might be a little 22 tentative as far as I am concerned, so I do not want to 23 -- () 24 JUDGE BRENNER: Okay. All right. 25 MR. REPKA: -- Read that into anything. O ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) 554-2345

1 l 4511 i (v) 1 JUDGE BRENNER: I have a delivery from Mr. 2 Latham's office. It is the cross examination plan on 3 SOC Contention 9. I guess unfortunately that tomorrow 4 would be okay got there too late, but we have it now. 5 That completes your items, I take it, Mr. 6 Lanpher. 7 MR. LANPHER4 Well, I can speak to Mr. -- I am 8 just curious about the appraisal is scheduled for July 9 19, when will the results be known in that; but I will 10 just speak to him. 11 JUDGE BRENNER: Why don't you talk to him, and 12 then we would be willing to hear more about it on 13 Thursday morning. 14 All right. I guess we can hear from the Staff 15 now on the matter it wanted to take up. 16 MR. REPKA: Thank you, Judge Brenner. 17 On May 4 Judge Morris requested that the Staff 18 respond to a series of questions on, first, safety 19 questions raised at limited appearances, and second, the 20 Staff's unresolved safety issues, and, third, the status 21 of the SER open items. The Board set June 22 as the 22 S taff's due date to respond to those questions. 23 foday we would like to ask that the Board () 24 grant the Staff an extension of ten days in which to 25 respond to those questions. Our reason essentially is O m/ ALDERSON REPORTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4512 l 1 () 1 manpower. The people we need to respond to those 2 questions have been involved in other cases, and we feel 3 that the brief extension will give them an opportunity 4 to provide more quality responses, and it is not 5 anything that will delay this proceeding. 6 Given this Board's desire for quality, we 7 think it would be only appropriate. 8 JUDGE BRENNER: No matter what you say it gets 9 used against you some day. 10 (Laughter.) 11 I tell you, we have a manpower concern, too, 12 and it is our manpower. And we took the schedule of 13 that week off in contemplation of the large number of O 14 things that were due on the 22nd and scheduled the 15 following week off. We need that week to look at it. 16 And I guess I would like to ratchet you back down a 17 little bit to June 29; that is, give you one week. And 18 the reason, as I indicated, is that the few days in that 19 instance do matter very much to the Board. We will be 20 together in our offices. We have not been together as 21 often as we would like on this case, and we have been 22 busy with the day-to-day business. 23 Recall also your request was for the entire () 24 package, and as to the week's extension, that is 25 granted. I am not going to worry about do you need the O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON 0.C. 20024 (202) 554 2345

4513 () 1 extension for all of it or not. , But when you get to the 2 29th, I think somewhere in the transcript you will see 3 tha t we indicated that if you were not giving us the 4 answer at that time, we would accept a status report as 5 to why not, and when we might receive it, it could 6 affect the proceeding. You know, if we see something 7 that we want to raise as an issue in the proceeding all 8 of a sudden, a lot of things are going to flow from that 9 decision. 10 And in the normal course of events had there 11 been a more orderly transition to this hearing, we would 12 have requested that information well before starting the 13 hearing or at least those on which you could give us a 14 definitive report. So we have all these other 15 considerations in mind, and that is why the few days -- 16 every day might count on this report, and hopefully the 17 week would be a reasonable compromise so that we could 18 focus on it, and also giving you the out, if you will, 19 that any that are not ready, you know, hopefully we will 20 get it a few days thereaf ter, and you can indicate which 21 portions you have not included. If we are going to get 22 it in a day or two, no big explanation would be needed. 23 If it is going to be delayed beyond that as to () 24 particular portions, a status report of what the problem 25 is. So, you know, we are reasonable, but we need O ALDERSON REPORTING COMPANY,INC. 400 VIRGINIA AVE., S.W., WASHINGTON, D C. 20024 (202) 554 2345

4514 O ' o etataa to er* oa ta t eex-2 HE. REPKA We appreciate that, Judge Brenner, 3 and at this stage any grant of extension looks good and 4 helpful. 5 JUDGE BRENNER Okay. Let's carry it over 6 until the 29tn, and we would want it received in our 7 offices by the close of business that day. "It" being 8 e ve ry thing hopefully, you know -- when I say if portion 9 of it are not ready, okay, we are not going to get one 10 small part on the 29th and the bulk of it the other way 11 around. I will depend on your good faith, and I think 12 you know what we mean. 13 MR. REPKA: I assure you our best efforts. 14 JUDGE BRENNER: A lot could flow from our 15 assessment of that information, so it could affect the 16 pace of this proceeding very dramatically depending on 17 what we think of it after we receive it. 18 HR. REIS: Let me say the Staff is going to 19 make every effort, and we expect to give you the vast 20 bulk of the items presently on the 29th. We are a 21 little concerned with asking for the items in the 22 context of searching for items that you might raise 23 under 2.7608A. O 24 1 de act we=t to ao 1ato it in ore t detei1, 25 but I just wanted to point to recent Commission and O ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE.. S.W., WASHINGTON, D C. 20024 (202) 554-2345

4515 __-- m

                                                                                                           -s h              1 Appeal Board cases which indicated the scope of review.

E And we call the Board's attention to the scope of review 2 4 3 in operating license proceedingsa the San Onofre case, -f 4 the Summer case, the Comanche Peak cases. And I just - 5 wanted to mention those.

6 We will attempt to give you the information. - 7 I do not think that calls for any ruling right now, but 8 we are aware of those cases. We are concerned. 9 JUDGE BRENNER: Well, I am aware of them, too, to and I cou?,d let Judge Morris respond, but I will be a 11 lot more restrained than he will be, notwithstanding our 3 12 normal differences in personality. 13 If you take a look at the SER and the items we - O 14 asked questions on, we could have jumped directly into

15 saying the way the SER had those items written up, that 16 raises a question as to the quality and completeness of 17 the staff's review on those items sufficient to inject 18 it directly into the case. And we could have done that 19 and in fact we considered doing that as to many of them, 20 perhaps all of them, because you just cannot tell which . 21 end is up on some of those items. - 22 And I will not go into it as to my views and

23 other people's views on the writing of SERs and their ' 24 amenability to scrutability in a hearing. But we did 25 not take that step. We are taking this intermediate -: O ALDERSON REPORTING COMPANY,INC, 400 VIRGINTA AVE., S.W., WASHINGTON, D.C. 20024 (202) 554-2345

4516 () 1 step. And I think we have some pretty direct words at 2 the time we took that step to the writers of the 3 response that we were seeking should consider us as the 4 audience. And in the sense that if we did not 5 understand from the furtt.er response what the basis for 6 the staff's conclusions were, either tha t everything is 7 okay or they are still looking at things, but it is okay 8 to go ahead in the interim or alike, then we are going 9 to inject it into the hearing because we have a safety 10 concern, because we do not know what the status is. 11 And I would leave it at that. Subsequent to 12 the time we made the request we discovered in the 13 con text of a t least one contention tha t when the Staff 7._ 14 writes " resolved subject to confirma tion," my view of 15 the English lar.guage and the Staff 's view of the English 16 language are at great variance in terms of what 17 " resolved subject to confirmation" means. So that just 18 reinforced our view. 19 Well, I do not have to go much further. The 20 Staff's position to take no position on the response to 21 the motion to strike 7(b), the Staff in passing alluded 22 to matters similar to that, and the Board thought it was 23 out of place there, although we did not comment on it, (~s (_) 24 because the d3 cision on the proper scope of the 25 contention is only very tangentially related to sua ALDERSoN REPORTING COMPANY,INC, j 400 VIRGINIA AVE., S.W., WASHINGTON, D.C. 20024 (?O2) 554 2345

4517 l () 1 sponte matters. I did not bring it up, because the 2 Staff itself seemed to raise it most tentatively; that 3 is, by the way, attached to the filing, so I did not 4 pick it up in that context. 5 Maybe it would have been better off if I had 6 let Judge Morris speak at that, but this was the 7 considered position by the three of us as to just great 8 uncertainty with respect to the items we asked about. 9 In addition, unless I read the Appeal Board 10 precedent, with which I am quite familiar, on generic 11 unresolved safety issues incorrectly, we are charged 12 with the obligation to look at those matters, and that, 13 therefore, would not be in the sua sponte category, and 14 we have some questions about those matters also. 15 And you would think by now that the writers of 16 SERs af ter all these years since 1975 or 1976 when River 17 Bend first came down, followed by North Anna, and made 18 it applicable to in operating license case, would l 19 understand that Boards have that obligation and would l 20 vrite up these matters so that we could ascertain for 21 ourselves just what the status of those items are in the 22 S ta ff 's view in the context of a particular plant. And 23 in some instances which we raised, in our view at least, () 24 which may be erroneous, but in our view that SERs did

25 not do that.

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4518 (} 1 But if we find a sua sponte question, we vill 2 let you know. And in all seriousness, if we do decide j , 3 to pursue an issue in the hearing given the response, we 4 are trying to do it as soon as possible. I will not i 5 belabor anothat item, but in one of our orders we 6 commented on what we see as a discrepancy between the 7 status of the review by the Staff, and in some cases 8 LILCO, and the readiness of the schedule for hearing 9 also advocated by the same parties. And while we are to happy to grant the week, the fact that there are many 11 matters still being considered by the Staff is 12 inconsistent with that posture. We agree it is prudent, 13 and it is consistent with due process. It makes sense 14 to proceed as far as we can on all issues on which we 15 can proceed, but as long as this hearing seems to be 16 taking, the pacing item may well turn out to be late 17 completed reviews which we then first have to pursue. I 18 will leave it at that. 19 (Board conferring.) 20 Well, you have your week, and more than you . 21 bargained for, I am afraid, but it is late in the day. 22 Are there any other matters? 23 (No response.) () 24 All right. We will resume at 9:00. 25 (Whereupon, at 5:04 p.m., the hearing was O ALDERSON REPCRTING COMPANY, INC, 400 VIRGINIA AVE., S.W., WASHINGTON. 0.C. 20024 (202) 554-2345

f4519 1 recessed, to be reconvened at 9400 a.m., the following 2 day, Wednesday, June 16, 1982.) 4 5 6 7 8

9 10 11 12 13 14 4

15 16 17 18 19 20 21 22 23 24 25 0 ALDERSON REPORTING COMPANY,INC, 400 VIRGINIA AVE., S.W., WASHINGTON. D.C. 20024 (202) S54 2345

   )                       NUCLEAR REw1* CRY COMMISSIC*T O

7 This is Oc. certify that the attached proceedings before the ATOMIC SAFETY AND LICENSING BOARD j in the matter cf: , LONG ISLAND LIGHTING COMPANY (Shoreham Nuclear Power Station) , Date of Proceeding: June 15, 1982 ] Oceket Number: 50-322-oL i i PlaC6 Of Proceeding: Haupeauce, New' York i were held as herein appears, and that this is the criginal transcript thereof for the file of the Commission. i 1 i David'S. Parker Official Reporter (Typed) .I

_ _M . _ _ _ (SL-u"URE OF RE7CE) l l l 1 4 1 i lO .I !O, I _ __._.}}

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