ML20199J247
| ML20199J247 | |
| Person / Time | |
|---|---|
| Issue date: | 07/13/1995 |
| From: | NRC OFFICE OF INVESTIGATIONS (OI) |
| To: | |
| Shared Package | |
| ML20199J167 | List: |
| References | |
| FOIA-97-325 NUDOCS 9711280084 | |
| Download: ML20199J247 (72) | |
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EXHI~ BIT 29 .
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,M Case No. 1 95 013 Exhibit 29
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1 UNITED STATES OF AMERICA 2 NUCLEAR REGULATORY COMMISSION .
3 + ++++
I 4 OFFICE OF INVESTIGATIONS l 5 INTERVIEW ,
6 -----------------------------o-x 7 IN THE MATTER OF:
i 8 INTERVIEW OF- : Docket.No. !
9 VIJAY CHANDRA : 195-013 10 : .
11 -------------------------------x 12 Thursday, July 13, 1995 13 14 Stan Labruna Conference Room.
15 PSE&G Administration Building 16 Buttonwood Road i 17 Hancock's Bridge, New Jersey 18 1-19 The above-entitled interview was conducted at [
20 3:15 p.m. .
21 BEFORE 22 KEITH LOGAN ~ Investigator ,
23 -
24 NEAL R. GROSS EXHIBIT '
CASENO. 1 - 9 a_ - 013 covat neartas Ano vaasscascas PAGE / OFMPAGEjS) i m a m i m uo m uuc.N w. -
<=3 eMun - **S""nON. D C M . y [4% . (202) eMMM .
2 1 APPEARANCES:
2 On behalf of Viiav Chandrat ,
3 MARK J. WETTERHAHN, ESQUIRE 4 MARCIA GELMAN, ESQUIRE 5 Winston and Strawn 6 1400 L Street, N.W.,
7 Washington, D.C. 20005-3502 8
9 10 11 12 13 14 15 16 17 18 19 ,
20 21 22 ,
23 24 1 25 l
l NEAL R. GROSS COURT REPORTERS AND TRANSCR$ERS 1323 RHODE ISLAND AVENUE, N W.
(202) 234 4433 WASHINGTON. O C 20005 (202) 2W433
- - . 3 1 PROCEEDINOS 2 (3 :15 p.m. ) -
3 Whereupon, 4 VIJAY CHANDRA 5 having been first duly sworn, was called as a witness 6 herein and was examined and testified as follows:
7 EXAMINATION ,
8 BY MR. LOGAN:
9 Q Mr. Chandra, my name -- or I guess it's Dr.
10 Chandra, my name is Keith Logan. I'm an investigator with 11 the U.S. Nuclear Regulatory Commission, King of Prussia, 12 Pennsylvania.
13 You're appearing today with counsel, is that 14 correct?
15 A Yes. Mr. Wetterhahn. Yes.
16 MR. LOGAN: Mr. Wetterhahn, could you identify 17 ,
yourself for the record, please? .
18 MR. WETTERHAHN: Yes. Good afternoon. My name 19 is Mark J. Wetterhahn. I'm with the firm of Winston &
20 Strawn, 1400 L Street Nerthwest, Washington, D.C. 20005.
21 With me is Marcia Gelman, also of the same firm.
22 We're here today representing Dr. Chandra. As 23 he is aware, we are also representing other individuals in 24 this investigation and also PSE&G as a company.
25 BY MR. LOGAN:
NEAL R. GROSS COURT REPORTER $ AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE, N W.
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. . 4 1 Q And, Dr. Chandra, with your knowledge that kr.
2 Wetterhahlt does represent other individuals in this case, -
3 is it still your desire to have him and Ms. Gelman here 4 today as your attorneys? -
5 A Yes.
6 Q Thank you.
7 Dr. Chandra, what is your current position at .
8 PSE&G?
9 A I'm an thermo-dynamics analyst, technical 10 support. Thermo-dynamics, hydraulics, technical 11 consultant.
12 Q And as a Thermo-dynamics, hydraulics 13 consultant, what are your responsibilities?
14 A My responsibilities are to as sist whoever asks 15 for help. If somebody has some difficulty understanding 16 some problem of a technical nature, they come and ask. So 17 my responsibilities are to assist whoever asks for it. .
18 0 Can you give me some examples of the kind of ,
19 assistance that you would provide, specifically on what 20 areas?
21 A Just to give an example, if somebody asks me, 22 say if I'm doing ventilation in this room, can you tell me 23 how the temperature would rise and what value in how much 24 time, and then I'll ask for his input, tell me what 25 equipment is operating in that room, what is the heat NEAL R. GROSS COURT REPORTER $ AND TRANSCR$ERS 1323 RHODE ISLAND AVENUE, N W, (202) 734 4433 WASHINGTON. O C. 2000$ (702n 2344431
5 1 lord, what 10 hect absorbing equipmsnt. Io there cny 2 ventilation path available?
3 I would collect all of the data, analyze the 4 problem using the computer and give him the answer.
5 Q Okay. And to whom do you report?
6 A Right now, the manager's position is vacant.
7 It is being filled by Andy Keogh. And he is the acting 8 manager.
9 Q And in the April 1994 time frame, who was your 10 supervisor?
11 A Jay Bailey.
12 O Can you spell the last name, please?
13 A B-A-I-L-E-Y.
14 O Calling your attention to the POPS issue, are 15 you familiar with the term POPS?
i 16 A Yes.
17 Q And when did you first become involved with i 18 the Westinghouse memorandum or memoranda and the POPS 19 issue with regard to Salem?
20 A I became aware of the issue in very early '94.
21 Either January or February.
22 Q And how did it come to your attention?
23 A I got a telephone call from Charlie Laskhari.
24 his real name is thunder Lashkari. And he told me that .
25 Westinghouse infermed us'that there is this problem and we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENVE. N W.
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6 I
, . . t I have an enginaaring evaluction in which the mechanical i 2 -engineers did it for A.S.M.E. cores which permits 10 f 3 percent additional pressurization above the limit. And he t f
4 said that we are not -- PSE&G doesn't have the NRC's i 5 permission specifically to clear for that code case. So !
i 1 6 can you do a calculation by taking crdit for the RS3 value 7 and tell me what the peak pressure would be.
i -
8 O Did you do that?
L i 9 A Yes.
10 0 And what did you conclude? -
. I I concluded the peak pressure would be much 11 .A. r i
less than what we are allowed.
12 13 O And what was it that you were allowed and what l 14 did you conclude, if you recall? -
15 A Well, the allowed pressure for Unit one is 450
, 16 and our pressure for Unit Two is 475 psi g. I would have 17 to consult my notes.
18 Q Please do.
19 A With one RCP operating -- now rememberr these 20 pressures are the absolute pressures so I have to subtract 21 15 from it to give you the psi G values. And the mass, it 22 is_447 minus 15, so it would be like 432. For the !
23 . temperature of"70 degreesin the pressurizer, and this is 24 with one RCP operation.
25 Q. 'Okay. Imd what code case did you use? I'm
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. . 7 1 sorry, what code did you use? Did you use the Gothic 2 code? -
3 A The Gothic computer program.
4 O Okay. And did you tell Mr. Lashkari your 5 findings?
c A Well, the calculation was officially sealed 7 later on, much later. But he was aware of this .
8 calculation.
9 0 When did'you tell Mr. Lashkari of your 10 calculntions?
11 A I know I personally gave him a copy of this 12 after it was issued, so it must be after June 13th. But he 13 was aware of the calculation before that, but I personally 14 gave him a copy that he asked for.
15 Q When you say this was issued -- what are you 16 referring to?
17 A This calculation, was calculated using RS3 .
18 valve.
19 Q What.is the document that you're referring to, 20 air?
21 A The document title is L Top Event with RSP 22 Valve. Calculation number is S-C-RC-MDC-1358 and Revision 23 0, 24 O And you provided Mr. Lashkari a copy of that, 25 you said, in June?
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. . B 1 A Somewhere after June 13th, because it was 2 issued on June 13th and I gave him a final version, copy. -
3 0 And prior to that time, you did share your 4 findings with him?
5 A Yes.
6 O And what did Mr. Lashkari say as a result of
? hearing your findings? .
8 A Well, I told him that the results are 9 acceptable using RS3 valves and it is also in this 10 document.
11 0 And what did he say?
12 A I don't know if he made any^special comment on 13 that.
14 0 Did he agree with you?
15 A Well, he did not say whether he agreed or 16 disagreed. Since this was an approved calculation, I 17 guess he would assume that it is correct.
18 MR. WETTERHAHN: Did he disagree?
19 THE WITNESS: No.
20 BY MR. LOGAN:
21 Q After you had this initial conversation in 22 January with Mr. Lashkari, did he send you anv documents 23 following up his call to you?
24 A No.
25 Q Did he ever show you the draft incident report NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE, N W (202) 2344433 WASHINGTON O C 20005 (202) 234 4433
9 1 which he preuared?
2 A No. I don't even know that he prepared an .
3 incident report.
4 0 Okay, you're not aware of any incident reports 5 that he may have prepared?
6 A I'm not aware of any incident reports.
7 Q Okay. And did you discuss the POPS issue with .
8 Mr. O'Gara?
9 A Many times.
10 0 Did you discuss the POPS issue with Mr. ,
11 Thomson?
12 A Not directly, nothing specific, but we 13 discussed this in a meeting with him. Which a lot of 14 people were present.
15 0 When did that meeting take place?
16 A I think sometime in April.
17 O April 1994?
18 A Possibly.
1 94 0 Is th,at the approximate time frame?
20 A Yes.
21 0 And do you recall who else attended that 22 meeting?
23 A Yes. Howard Berrick, Mahesh Danak, Frank 24 Thomson, Howard Onorati, Jerry Ranalli. There could be 25 some others, f
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I 10 t
< 1 Q Xen O'Gara?- * !
2 A I'm not sure about him. He may have been .
I l
3 there. I'm not sure.
l 4 0 Do you recall whether or not this issue was j 5 discussed'in conjunction with the issuance of an incident 6 report? !
. .r 7 A No. .
l t
8 -Q When you were doing your calculations, did you !
9 refer to any documents prepared by Westinghouse?.
10 A I looked at the PSE&G calculations that vere 11 done using the Westinghouse algorithm that were done in 12 1977.
f 13 O Did you, at any point during your analysis of l i
14 this problem, refer to a memo prepared by Westinghouse 15 that was dated in March of 1993, or one in September of 16 1993?
r 17 A I remember one document of Westinghouse sent ,
18 to all utilities that didn't account for the effect of the *
- 5 19 RCP operation. Is that the one you are.refering to? ,
20 0 You are aware of one of them. Do you have.a 21 copy of that one with you? ,
22 A No , not here. These'are all these 23 calculations.
24 MR. WETTERHANN: Would you like me to proffer-25 a copy of the March 15th letter to the witness?
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i U-r,- , -,.c ,.ow%--r, -w-.-,mer.~ wennww w - , - * - - - = - +
. . 11 1 BY MR. LOGAN:
2 Q Would you~like to take a look at that document .
3 dated March 15th, 1993, and tell me if you recognize it? l l
1 4 A Yes, I have seen this., ;
I 5 Q Do you recall seeing any subsequent documents ;
1 ,
i 6 prepared by Westinghouse on that same subject, that might !
! 7 have been more plant specific than that one is? ,
8 A I don't recall. This ic the only one I 9 remember. If it was referred to me, I might recognize it, . ,
1 1
- 10 but without --
11 Q Do you recall seeing a letter from Mr. Howard 12 Berrick to Frank Schnarr dated December 30, 1993?
13 A Yes. This is the one which took credit for l 14 the code case?
15 MR. WETTERHAHN: I have a copy. '
26 THE WITNESS: Yes, I have seen that one.
17 BY MR. LOGAN: ,
18 Q Did you ever discuss this memo with Mr.
19 Berrick? ,
20 A I discussed it with Mahesh Danak. He is the 21 one who actually prepared it but Howard is his supervisor.
22 0- So you didn't discuss this with Mr. Berrick, :
l 23 though?
~
24 A .No.
25 Q- Did you discuss.the substance of this memo i
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12 1 with cnyone else?
2 A No, only Mahesh Danak, because he is the one ,
3 who wrote it.
4 0 Did this letter come up as a matter for 5 discussion in the April 1994 meeting with Mr. Thomson and 6 others?
7 A I'm not sure the letter specifically, but the 8 subject of the RCP pressure should be added to the old 9 calculation.
10 0 What was your opinion, as of thr* meeting, on 11 whether or noc salem was operating outside its design 12 basis? -
13 A My opinion was this, that the Salem plant does 14 not operate with RCPs when the reactor and the pressurizer 15 are water solid. And if they're water solid and the 16 safety ignition pump comes on, the only ti..ng that should 17 be added would be a mass and contransient should be the 18 hydrostatic effect, not the RCP effect, because the RCP 19 would not be operating.
20 And I also called Westinghouse and I asked 21 them, since the Salem plant doesn't operate its RCPs in 22 the water solid condition, why are you telling us to add 23 RCP effect, and they said this letter was written to 24 encompass all plants in the country. You use this letter 25 as it is applicable to your plant. .
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. . 1 1 1 Q And when did you discuss this with F i
2 Westinghouse? .-
+ t 3 A I discussed this -- ;
i 4 O Before or after you spoke with Mr. Lashkari? !
i 5 A Oh, after. Much after.
6 O So you're not aware of a second letter in 7 September which addressed Salem in particular, that was I
8 prepared by Westinghouse? ;
1 9 A I'm not aware of a letter but if you showed it l 10 to me, maybe it will jog my memory. -
11 Q If you're not aware of it, you're not aware of l i
12 .it. That's okay.
13 MR. WETTERHAHN: I think the witness asked 14 that his recollection possibly be refreshed, that he may 15 be aware of it.
16 MR. LOGAN: Okay. ,
17 BY MR. LOGAN: ,
18 Q S'd in April, again you concluded that Salem 19 was or was not op,erating outside its design basis? !
20 A Well, that conclusion was not made at that 21 time, ,
22 Q It was not made. What was your belief at that 6
23 time?
1 24 'A My belief was that the -- if you follow the 25 design basis as=it is written in the tech spec, and you NEAL R. GROSS -
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I get verification, you can look at the. calculations that l 2 were made to support that, the design basis is that the !
3 safety injection pump will come on and work in a water 4 solid condition and the pressurizer would be water solid i
5 and its temperature would be the same as the reactor f 6 temperature, and the reactor pressure under that condition ,
7 should not exceed what is permitted by the tech spec. .
8 Q Does that mean you believed you were not 9 operating -- Salem was not operating outside its design I s
r 10 basia?- What I'm asking you to do is draw a conclusion.
11 A My conclusion was that this letter, as it is 12 written here -- ,
I 13 0 When you say "this letter," you're referring 14 to? ,
15 A I'm referring to this NEC 93-917.
16 Q From whom?
From Howard Berrick to Frank Schnarr.
17 A -
18 Q okay.
19 A What*it did is it took the pressures 20 calculated in 19't? and they added the effect of -the RCPs.
21 My conclusion was that Salem RCP's would not operate for 22 the design conditions as described in the tech spec.
t 23 So only pressure for the massive transient 24 should be added to the nydrr> static *effect --for heat 25 transient you should add one-RCP factor.
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15 1 Q Which means what? You are or are not outside 2 your design basis? .
3 A But that conclusion can only be made after we 4 looked at, add the hydrostatic facts.
5 0 Which you have not done yet?
6 A No. We have not, no.
? O Have you done it to date? ,
8 A I didn't do it. I asked Mahesh or the people 9 in the room, why this memorandum improperly Lccount for 10 the pressures.
11 Q Was that done?
12 A Yes.
13 0 And what was the conclusion?
14 A The conclusion that of we add the hydrostatic 15 effect, the massive transient to the cid calculations that 16 were done in 1977, the pressure will be 450.7 psi g, which 17 is .7 psi higher than the Salein One limit. ,
18 0 Which would then put Salem outside its design 19 basis by .7? ,
20 A If you take the old calculation as the right 21 calculation or if you assume there is no margin in your 22 calculation.
23 0 Are the old calculations the proper ones to 24 use in making this determination?
25 A No, old calculations have some great NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1373 RHODE ISLAND AVENUE N W.
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16 l 1 inconsistancies in them. ,
l 2 Q Shouldn't they have that H: tit in
{
3 conservatively? l 4 A Not necessarily. !
5 O It's wrong?
i 6 A It's not wrong. I mean, it is just i 7 conservatism in there. It doesn't have to be there. And 8 I repeated the calculation in Gothic computer code and the 9 same input that was made in the 1977 calculations. I came 10 up with a : ressure of 438 psi G, instead of 446. Now, ;
11 keep in mind that the old calculation was done using a lL2 algorithm Westinghouse, and that algorithm is not in !
13 Salem's specific algcrithm. It's designed.so that 14 anybody can use it. So, by nature it would be 15 conservative. If Westinghouse did the analysis of Salem 16 specific'using the old algorithm, they would come up with 17 less than 446. But since we are now using a method which 18 covers all the plants in the country, by measuring, it l
19 would add some conservatism to that. When you 20 interpolate, you - although it is a special plant 21 calculation, it is not a plant specific algorithm. It 22 doesn't use a plant specific algorithm.
23 My feeling was that the .7 psi exceedance, 24 although we are exceeding by .7 psi, but this 446 is a 25 conservative limit, so actually we should be adding 4.7 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBER $
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17 i 1 psi to 438 which weuld keep us below the 45_0 limit. !
'l 2 Q At-the time of licensing, a safety analysis is .
l 3 done on a plant, is that correct?
1 4 A I'm sorry, what is the question again? ;
. 5 0 When a plant is licensed, the N.R.C. does a 6 safety analysis on the plar*., is that correct? f 7 A They issue a safety evaluation report, I know ,
b 8 that.
9 Q What code is used for that report? Is it the 10 Gothic Code?
11 A Oh, no, the othic code actually is very 12 recent. It didn't exist then, so it was not used.
13 Q If the Gothic Code wasn't used when the safety ;
14 analysis was done, shouldn't we use the same code that we i 15 used at that time and not the Gothic code?
16 A See, this transient is so simple that it's not 17 really a core dependent type problem, it's a very simple ,
18 problem, mass and mass going out and really I could have
-19 written the ( , ) program by hand, and it 20 would give all the answers. Since the Gothic was there, I 21 used it.
22 O I see.
23 -A And then I look at the Gothic answers, I did 24 an eneigy balance and a mass balance and it matched. So 25 Gothic is right-and the answers are right. I have done an NEAL R. GROSS COURT REPORTERS AND TRANSCRt9ERS 1323 RHODE ISLAND AVENUE. N W (200 2MM33 WASHtNGTON. O C. 20005 (202) 2M*33
18 1 . independent: hand calculation..
2 0 ~Okay. I'm not.taking11ssue with_the accuracy .
3 of your calculations,-just the use of the Gothic Code, 4 MR. WETTERRAKV; Mr. Chandra, do you know-5 whether the N.R.C.-did an independent calculation back 6 when the plants were licensed? Do you know that, whether 7 they used any code whatever? 7 8 THE WITNESS: I'm not aware of it. I was not 9 here in 1977, but I have looked as they are now, and I 10 became aware later on that -- specifically looked at the l 11 calculations that were submitted at that time and approved
- 12 it. .
, 13 BY MR LOGAN:
l'4 O And when you finished your calculation,s, who 15 did you send the results of the calculations to? .
16 A I have done many calculations. I don't know 17 which one you are referencing to, ,
2 18 0 I was referring to the one where you F 19 concluded, the one you mentioned-earlier, the last one in 20 your book there.
7
. L21 Q Oh, that=one that-Charlie Lashkari requested?
22 A Okay.
23 Q Who was that sent to, after you finished those 24 calculations?
f 25 A It was sent to the doeur ,at distribution room. .
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19
- 1) 12 You'didn't sent it to.Charldt Lashkari, you 2 must have sent'them to somebody else first, right? You ,
3 said you sent a_ courtesy copy to Mr. Lashkari.
- 4 A He specifically asked for it'. :He said he-wants a copy of my calculations, so I made a copy of my
~
5 6 copy and gave it to him. -
7 Q And did-you send them also to your supervisor?
8 A No.
9 Q To Mr. Thomson?
10 A No.
11 Q To Mr. Berrick?
12 A- Berrick, yes. In fact, he is the one who 13 requested -- well, no, he didn't request it but since he 14 is in charge of this whole L top, I made him aware that it 15 had been issued.
16 Q Who else got a copy of that?
17 A My feeling is I kept a copy for myself and I 18 gave one to document distribution list.
19 Q And who was on the distribution list?
,, 20 A _Nobody else.
21 Q So-just Mr. Berrick?
-22 A' Mahesh probably had a copy of this. I think I 23 probably gave-him one,-too. I'm not really sure. I don't 24 have a distribution list here. But_I know it went to_the_
25 document distribution group, and'I kept.a copy for myself NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE, N W (202) 234 4433 WASHINGTON, D C. 20005 (202) 2344433
g .._
20 ,
1 and I'think I prcbably-gave one to Mahesh and probably the 2 peer reviewer also has a copy. .
3 0 Okay. May I just~see the cover sheet for a 4 minute?
5 A This sheet?
6 Q Yes.
7 MR. LOGAN: Can we go off the record for a ~
f 8 minute?. .
9 (Whereupon, a brief recess was taks .
10 MR. LOGAN: Back on.
11 BY MR. LOGAN:
12 Q Again, this was approved by.Mr. Bailey as your 13 supervisor, is that correct?
14 A- Yes.
15 Q It's titled "L Top Event with RH3 Valve"?
16 A Yes.
17 MR. WETTERHAHN: The unique number, that would 18 identify it is just what?
19 THE }iITNESS: 1358.
.20 MR. WETTERHAHN: Any revision on it?
21 THE WITNESS: I didn't revise it, so --it-is 22 only -- revision zero you can call it. .
23 MR. WETTERHAHN: Okay.
24 BY MR. LOGAN:
25 Q Did you discuss this with Gita Narasimhan?
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21 1 A Narasimhan. Yes.
2 Q' .
Did you?= ,
I 3 A- She wasn't here, I think, in those days.
-4 0 .But have you discuss it with her? t 5 A- Later on, I told her that I did a calculation 6 in the past, yes, s 7 Q When was that, "later on"?
8 A Sometime in December time, November or 9 December.
10 Q Of 19S4?
11 A '94.
12 O Why -- what caused you to discuss this with 13 her?
14 A She was the reviewer of other calculations I 15 .did.
16 O So she was also the reviewer of this one?
4 17 A No, she is not the reviewer of this one. No.
18 Q I guess the question is, why did you discuss 19 it with her then, if she was not a reviewer of that one?
20 A I told her that the model which I'm using here
.- 21 -was.the earlier -- it was not really discussing, but 22 making her awar:t of the calculation.
23 MR. WETTERHAHN: You felt -- let me just --
24 MR LOGAN: All right.
25 MR. WETTERHAHN: You felt that she needed that
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22 1 information in' order to adequately do a peer review of a 2 later calculation? ,
3 THE WITNESS: Right, because some of the 4 numbers that were used here were double op here.
5 BY MR. LOGAN:
6 Q When you say "here" and "here," what are you ,
7 referring to?
8 A I'm sorry, numbers left.in the calculations, 9 14:;3 were double opped in 1538.
10 0 And what is the other document for?-
11 A Now, this document in which I calculated _the 12 reactor pressure without taking credit for RS3 valve and 23 documents with 1PURV, Calculation No. S-C-RC-MDC-1413 Rev.
14 0.
15 0 And why did you complete that calculation?
16 A Okay, thi$calculationwasdoubleoppedatthe 17 request of Licensing. ,
18 0 Who in Licensing requested it of you?
19 A Ken O'Gara and Dave Smith.
20 Q Why.did Mr. O'Gara request that?
21 A He said that we aren't allowed to take credit 22 for RH3 valve, so we have to do the calculation without i 23 it.
24 _
Q Why was he asking you to do those calculations l' 25 withcut;it? Had you done some-prior calculations with.it?
l
- l. - NEAL R. GROSS l . COURT REPORTERS AND TRANSCRIBERS 1323 RHOOE ISLAND AVENUE, N W.
(-
.-. ' (202) 2344433 WASHINGTON. D.C 20005 (202) 2344433 lL .. , ,- . - - . . - . . - - . -
, _ _ _ _ = .
23 1 MR. WETTERHAHN: Taking credit for the RH3 2 valve. ,
3 THE WITNESS: Yes, he was aware of this 4 calculation, yes. ,
5 BY MR. LOGAN:
6 Q What's the date on that, on that second one 7 that you were talking about?
, 8 A It was issued December 16th, '94..
9 Q So that was the -- the December 16th, '94, was 10 after the June report that we just discussed,'is that
'11 correct?
12 A Yes.
13 -Q Now, were you ever hsked to review an incident 14 report that concerned the POP system?
15 A Incident report?
16 Q Incident report.
17 A I have never seen an incident report. ,
18 Q On the POP system?
19 A No. ,
20 Q I'm going to show you a letter that was sent 21 from Mr. Hagan to the Nuclear-Regulatory Commission 22 concerning Licensing Event Report No. 94-017-00. -It's 23 dated December 14th, 1994. Have you ever seen that
-24 document, Dr. Chandra?-_ '
25 A Yes, I believe I've seen this.
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHOOE ISLAND AVENUE, N W. -
(202) 734 4433 WASHINGTON, D C. 20005 (202) 234-4433
24 1 Q -You have acen that document?.
2 A -- Yes.
-- 3 Q When did you see'it?
-; 4 A I don't remember exactly.
l 5 Q Let me ask you this. Were you involved in the i 6 preparation and/or review of that document?
7 A No.
8 O So you saw the document after it had been sent 9 to the N.R.C.?
10 A Yes.
11 Q All right. Do you agree with the conclusions 12 in that document?- ,
t 13 Why don't you take a moment,to read it? Let's l'4 go off the record.
15 (Whereupon, there was a br'.ef pause off the 16 record) 17 MR. LOGAN: Back on the record.
18 BY MR. LOGAN:
19 Q Dr. Chandra, after having reviewed that LER,
-20 you indicated that you-did or did not agree with the 21 conclusions?
22 A- That's right.
23 Q You didn't?
c 24 A I do not agree with the conclusion that we
-25 .were outside design basis. ,
' NEAL R. GROSS' COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE. N W.
(202) 234 4433 WASHINGTON. D C. 20005 932) 2344433
7-. , ,
25 1 Q Why -- 1 2 A The design = basis, _ as stated in the tech spec.. -
3 0 And why is that?
~
4 A Because my feeling is that here-th=y are using 5 the -- they are again adding this RCP effect. This should 6 not-be added. That's my feeling.
7 Here they are saying that they will exceed-the- .
8 allowable reactor pressure but my feeling is that the 9 pressure would not be exceeded.
10 MR. LOGAN: Okay.
11 THE WITNESS: For the transient that is 12 described in the tech spec.
13 MR. LOGAN: That's all the questions that I 14 have.
15 Mr. Wetterhahn, is there anything that you 16 would like to ask any questions about?
17 MR. WETTERHAHN: Let's go off the record for -
-18 just a minute.
19 MR. LOGAN: Okay.
20 (Whereupon, a discussion was held off-the 21 record) 22 MR. LOGAN: Back on the record.
23 MR. WETTERHAHN: Dr. Chandra, I only have one 24 question.
25 .During our preparation for this interview, did NEAL R. GROSS COURT REPORTERS AND TRANSCRIB'.RS 1323 RHOOE ISLAND AVENUE, N.W.
{ Qon 2344433 - WASWNGTON O C PontM im SMJW
26
-1 I ask you to read a March 30, 1995 Salem Engineering 2 Follow-up. Inspection 94-32?
3 THE WITNESS: Yes.
4 MR. WETTERHAHN: Did you so review that 5 document, the inspection report?
6 THE WITNESS:- Yes.
7 MR. WETTERHAHN: On Page 4 of :he Detail 8 section of that inspection report, under the heading, 9 " Revised Design Basis, Transients," is a sentence that 10 ' reads as follows:
11 "On September 27, 1994, Problem Report (PR),
12 No. 940927126 was initiated after the licensee determined 13 that they could not rely on the establishment of a 14 pressurizer (bubble) to resolve the problem."
25 Could you state whether you agree or disagree 16 with that statement?
17 THE WITNESS: We never relied on the bubble
.L 8 formation. All the calculations were done on the water 19 solid pressurized.
20 However, what we did consider, we considered 21 other transients than what was considered in 1977. In the 22 1977 transients the pressurizer was considered water solid 23 and was at the same temperature as the reactor. My
.24 transients, what I considered that if the bubble is 25 present, that means RCP will be operating. Then and only NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISt.AND AVENUE. N W.
(202) 234-4433 WASHINGTON, D.C. 20005 (202) 234 4433
.-- . - -. -. = . . .- . . . . . . . - .. - .
27.
'l then wo should add the RCP offect, and if the RCP is 2 operating, this bubble is present. oow the safety' ignition .
3 pump comes on and it stays on and th- bubble will 4 . disappear and then we will have water polid pressurizer 5 and the pressure would be at a higher temperature.
6 MR. LOGAN: Okay.
7 THE WITNESS: So once it becomes water solid, 8 then the peak pressure occurs. So we didn't rely on the 9 bubble cushion to reduce the pressure.
10 MR. WETTERHAHN: And that's the method of 11 calculation and analysis that would tend to increase the 12 pressure?
. -13 THE WITNESS: Yes, that's a more severe
~
14 transient and I personally feel that the transients should 15 .be a part of the design basis and it was not. It is not.
16 MR. WETTERHAHN: Thank you.
17 MR. LOGAN: Ms. Gelman?
18 MS. GELMAN: No questions.
19 MR. LOGAN: Dr. Chandra, is there anything 20 that you would like to state for the record, before we 21 conclude this interview?
22 THE WITNESS: Yes. What-I would like to say is 23 that when I look at the Westinghouse Algorithm, which was 24 approved by N.R.C. and it says use this and this method to 25 come up with the peak pressure, I realize that that NEAL R. GROSS COURT REPORTERS AND TRANSCRGERS 1323 RHCOc ISLAND AVENUE, N W.
- (202) 2344 433 ~ WASHINGTON. D.C. 20005 (202) 2344433
28 ,
- 1 methodology is not encompassing all.possible situations..
2 that can occur. A more realistic-transient that should >
3 have been made part of this design. basis, or should be 4 part of this design basis is if there is a bubble in the 5 pressurizer, RCP are operated. Now, if by mistake a 6 safety ignition pump comes on and it stays on for as long 7 as it takes for the bubble to disap' pear, and then the 8 pressure becomes water solid and we.are to relieve
. 9 circulated water, that condition would be more severe.and 10 it was never considered as part of this design basis. And 11 that should be in the plans, design. basis, should be, it
. 12 was not.
13 And in th"e calculations which I considered 14 later on, I considered hot pressurizer, which is a more -
15 severe condition.
16 MR. LOGAN *: Okay.
17 MR. WETTERHAHN: So you were even more 18 conservative than the original Westinghouse calculation 4 19 which kicked this issue off?
20 THE WITNESS: That's right. ,
21 MR. LOGAN: Thank you.
22 (whereupon, the proceedings were concluded at 23 3 :58 p.m.)
24 25 NEAL R. GROSS
. COURT REPORTERS AND TRANSCRIBERS 1323 RHOOE ISLAND AVENUE, N W.
(202) 234 4433 - WASHINGTON. D.C. 20005 (202) 2344433
-C E R'T-I F I.C A T'E
~
. ,This-is to certify that the attached proceedings before the United' States Nuclear-Regulatory .
commission in the matter of:
Name of' Proceeding: Interview of-Vijay Chandra .
Docket Number: 195-013 Place of Proceeding: Hancock's Bridge, New Jersey, were held as herein appears, and that this is the. original .
transcript thereof for the file of the United States Nuclear Regulatory Commission taken by me and, thereafter
. reduced to typewriting by me or under the direction of the ,
court reporting company, and that the transcript is a true
. .and accurate record of the foregoing proceedings.
R , , \
- . QQ LWQ _
John Burke,
- Official Reporter Neal R. Gross & Co., Inc. ,
3, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS '
1323 RHODE ISt.AND AVENUE. N.W.
, CNE) 234-4433 WASHINGTON. D.C. 20005 (202) 234-4433-
,c , - _ - . . ,. ,, . , _ . _ _ ~
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i EXHIBIT 31 l
1
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\D Case Nn 1.QS.f)19 c ,. . ., 4 . ,,
RN
, ATAcle INT "2. / /s R !
L .. - j PISIIESennce Electnc and Gas Company P.O. Box 236 Hancocks Bndgo. New Jersey 08038 Nuclear Department . .
MEC-93-917 3
4'- '
,'._l TO: F. Schnarr Reliability & Assessment Group FROM: Howard Berrick Salem Mechanical Engineering Supervisor
SUBJECT:
NONCONSERVATION IN POPS SETPOINT ~
ATS OPEN ITEM - WESTINGHOUSE NUCLEAR SAFETY ADVISORY LETTER PSE-93-204 (NSAL-93-005B)
DATE: October'29, 1993 ,
Background
Westinghouse NSAL-93-005B transmitted via PSE-93-204 identified a potential issue regarding a nonconservatists in the POPS setpoint development. The pressure difference from the wide range pressure transmitter to the reactor vessel midplane (where the Tech. Spec. heatup and cooldown pressure / temperature limits are defined) was not considered in Westinghouse analysis. This pressure difference effectively results in the pressure in the reactor vessel midplane being greater than that seen by the wide range pressure transmitters used to actuate the PORVs, potentially resulting in violation of the Tech. Spec, heatup and cooldown pressure /temperatur'e limit curves.
The Salem POPS analysis . (SGS/M-DM-042 and 062) used methodology provided by Westinghouse in their report " Pressure Mitigating .
Systems Transient Analysis" (July 1977). The methodology in this report did not consider the pressure difference of concern and therefore the subje'ct NSAL applies to Salem 1 & 2.
Discussion The Tech. Spec. heatup and cooldown P/T limit curves (attached) are determined in accordance with the requirements of Appendix G of 10CFR50 and ensure maintenance reactor vessel integrity. The pressurizer overpressure protection system (POPS) protect.s the RCS from exceeding the Tech. Spec. P/T limit curves by opening the PORVs during cold overpressure transients (RCS below 312* F) .
_ .,o c.-* ' ' , _ .
p'- -
)
Cf , -
h/
The pwe a m uw iunds. 8"._._ ~
EXHIBIT 5 ._T .:. s.; - ,u, .e. ,, ,,
CAsm o7-oya PAGE / OF 'g't PAGE(S) e ha
t F. Schnorr 2 10/29/93 ' hj The POPS uses the two wide-range RCS pressure sensors PT403 and
- PT405.to actuate the PORVs. These sensors sense hot leg pressure. The pressure at the vessel midplane will be higher than the pressure at the hot leg due to the dynamic and static pressure difference between the locations. The dynamic pressure difference depends on the number of reactor coolant pumps (RCPs) in operation at the time. Westinghouse did not consider the delta-P associated with the difference in location of the wide-range transmitter relative to the vessel midplane, in their analysis; this leads to a concern that the POPS setpoint may be nonconservative.
To resolve this issue POPS setpoint for actuating the PORVs must be shown to provide adequate protection, with the additional delta-P incorporated in the setpoint analysis. The Salem POPS analysis calculated the maximum pressure attained during a cold overpressure transient to be 446 psig with the PORV set at 375 psig. Therefore, it must be shown that 446 psig plus the delta-P of concern does not exceed the Tech. Spec. P/T limits, in order for the POPS PORV setpoint to be adequate.
The Tech. Spec. P/T limit curves define tne allowable temperature and pressure combinations for heatup rates up to 60' F/hr and for cooldown rates ranging from O'F/hr to 100'F/hr. For the POPS analysis a composite curve made up of the heatup curve and the 20*F/hr cooldown curve is used. The use of cooldown rate of
. 20* F/hr is justi11ad beemune me rh. lau *==neratures when POPS is c.,3 i TrmTdT' higher cooldown rates are not achievabih A review of the
' - Tech. Spec. v/T 11mic curves snows tnat the 20*F/hr cooldown
"" curve is more limiting at lov temperatures, on both Units. The
- " "' pressure that must not be exceeded is 450 psig on Unit 1 and 475 psig on Unit 2.
Additional margin in the Tech Spec. curves can be-gained for the POPS application by taking credit for ASME Code Case N514. This code case states "LTOP systems shall limit the maximum pressure in the vessel to 110% of the pressure determined to satisfy Appendix G of Section XI', Article G-2215". (LTOP - Low Tempressure Overpressure Protection is the same as POPS) , By taking credit for this code case the allowable pressure can be increased b p'ns= _In this case the lowest be exceeded 65 psiQ Unit 1 ancLS22.% pressure that
)on Unit 2. must not
( ' , C ' ,$ $
'~
EXHPIT_ h PAGE
._OF1 PAGE(S)
' , 2 F. Schnasr 3 10/29/93 ' M,
- Evaluation .
Table 1. summarizes the results of the evaluation. Comparing the POPS analysis maximum pressure of 446 psig to the Tech. Spec. P/T limits shows that the margin availa.ble to accommodate the delta-P is 4 psig on Unit 1 and 29 psi on Unit 2. An additional 45 psi (Unit 1) and 47.5 (Unit 2) can be gained by taking credit for the code case. Westinghouse indicated that based on generic analyses the delta-P is 74 psi with four RCPs operating. It is clear that Unit 1 does not currently have the margin to accommodate the expected delta-P with four pumps operating."
To quantify the-Salem spacific delta-P and assess benefit of fewer operating pumps, Westinghouse was requested to calculate the delta-P for one, two and four RCPs operating. The results of the calculation provided delta-P values of 31 psi, 39 psi and 73 psi for one, two and four RCPs respectively (PSE-93-707).
Westinghouse assumed the transmitters are zeroed out to the RHR suction line at 92.4 ft. The transmitters are zerced to the hot-leg (97 ft). To correct for this dif ference_2][sh was added to the Westinghouse results. The maximum pressure including the above delta-P values is presented in Table 1. '
- . - gp p , ..
'r.. vW Table 1 shows that Unit 1 Tech. Spec. minimum of 450 psig is exceeded by the two RCPs and four RCP casts. Taking credit for the code case the Unit 1 Tech. Spec. minimum is 495 psig which can be met by the-two RCPs operating case. The Tech. Spec.
pressure limit increases with increasing temperature and exceeds the four RCPs maximum pressure of 511 psig at 200* F. Therefore, int restricting the number of operating RCPs to two RCPs below 200* F the POPS PORV setpoint will provide adequate protection.
The temperature of 200* F was selected to coincide with cold shutdown Tech. Spec. operational mode (Mode 5).
Table 1 shows that Unit 2 Tech. Spec minimum pressure of 522 psig (taking credit for the code case) can be met with four RCPs '
operating. However, to maintain similarity in the operation of the units, and to providt margin for future evaluations of the Tech. Spec P/T curves, the same restriction on RCP operation is recommended on Unit 2.
.It should be.noted that 1) restricting the number of RCPs is one of the recommendations in the subject NSAL and 2) taking credit for the 10% margin in the limits as afforded by-the ASME code case was discussed with Westinghouse and this margin has been credited by other utilities (eg FP&L), to address the subject issue.
.,,, * , ,:~~
PAGE OF Y PAGE(W
F. Schnars 4 10/29/93 h Recommendation .
In summary to address the POPS setpoint nonconservation identified in Westinghouse Nuclear Safety Advisory letter PSE 204, we recommend restricting the number of RCPs in operation while in mode 5 to no more than two RCPs.. Procedure change request has been issued to incorporate this change into IOP-2 (Cold Shutdown to Hot Standby) and IOP-6 (Hot Standby to Cold Shutdown). The ATS open item NSAL-PSE-93-204 is considered closed by this letter.
GN:
C J. Serwin MEC File Standards Records Coorindator e
e 4
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PAGE_______Opg PAGE(S)
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EXHIBIT 32 i
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I Case No. 1 95 013 Exhibit 32
c.:
FORM EC.MA-AP.55-0006-1 M
(
NUCLEAR CEPARTMENT INCIDENT REPORT FORM I
COMMITMENT NUMBER INCIDENT REPORT No.
UgJ CONTINUATION SHEETS IF NECESSARY SECTION I REPORT
SUBJECT:
(Initiator)PS
_ Pn Llan?J Alom - /b u-sm /:< m UNIT S3,HC\: DATE OF INCIDENT: 4 / 20 /R TIME:
SUMMARY
OF EVENT (IF ESF ACTUATION, INCLUDE SOE PRINTOUT):
NMv. Su Obxhr REPORTED BY: KW6e DEPT: 44.A PHONE EXT: 13 7 o SECTION II (SNbO/ OPS MG)
RX PWR AT TIME OF EVENT: % UNIT LOAD: MWe Op Con / Mode REPORT MADE PER ECG? (Y/N): (IF YES, ATTACH ECG CuPY)
LCO #: A/S #: DATE IN: TIME IN: _
W.R.#:
INITIAL CAUSE DETERMINATION: EQUIP DESIGN PERSONNEL PROCEDURAL OTHER:
REPORTABLE: YES/NO, REASON SNSS/NSS SIGNATURE: DATE: / /
COMMENTS:
OPERATIONS MANAGER REVIEW : DATE: / /_
Nuclear Common Page 1 of 4 R47q 5 EXHIBlT J7 CASE N1 1-D~-013 f fduc PAGE / OFdPAGE(S)
l Westinghouse identified in letter PSE-93-204 dated March 15, 1993 (NSAL-93-005B) a potential non-conservatism in the calculation of the pressurizer overpressure protection system (POPS) setpoint (375 Psig) that effects Salem Units 1 and 2. The pressure ,
difference from the wide range pressure transmitters (PT403 and PT405) which sense hot leg pressure to the reactor vessel midplane (where the Tech. Spec. heatup and cooldown pressure / temperature (P/T)_ limits are defined) was not considered in the Westinghouse analysis, pg The Tech. Spec heatup and cooldo curves are determined in accordance with the requirements f 10CFR50, Appendix G and ensure reactor vessel integrity. The POPS protects the RCS from exceeding the Tech. Spec iiI vu ece by opening the PORVs during cold overpressure transients (RCS Temperature below 312'F) . The current heatup and cooldoun curves (Tech. Spec. Figures 3.4-2 and 3.4-3) POPS limits are approximately 450 and 475 psig for Salem Units 1 and 2, respectively. The Salem POPS analysis calculated a maximum peak pressure during an overpressure transient of 446 -
psig with the PORV set at a pressure of 375 psig. Further analysis was required to show that the pressure difference between the RCS hot leg and the midplane of the vessel plus the maximum calculated pressur,e (446 psig) did not exceed the Tech.
Spec. P/T curves.
The results of this evaluation by engineering are provided in letter MEC-93-917 dated December 30, 1993 (Attached). In .
summary, the calculated maximum pressures assuming 1 or 2 RCPs in operation compared to the Tech Spec P/T curves are as follows:
Unit RCPs in Serving Max. Press. Tech Snec Limit 1 2 485 450 1 1 477 450 2 2 485 475 2 1 477 475 Based on the above, when the non-conservatism is removed by adding the pressure difference calculated, the limits of both Salem Unit's P/T curves are exceeded. DEF # 94-0060 dated 4/19/94 was issued to evaluate this issue.
As identified in MEC-93-917 additional margin on the Tec . pec.
curves can be gained when operating with POPS (RCS less an 312*F)by taking credit for ASME Code Case N514. This Code Case states that the LTOP systems shall limit the maximum pressure in the vessel to 110% of the pressure determined to satisfy Appendix G of Section XI, Article G-2215. Crediting the Code Case will allow the maximum pressure for POPS to be increased to 495 psig and 522 psig fo. Salem Units 1 and 2, respectively. However, utilization of the Code Case will require NRC approval prior to implementation. This Code Case has been relied upon by another utility under these circum pst n s- _
. +LJW LM Also, Procedure revisions [ato limit the number of RCPs in 6/y&*,
J' operation to 2 while in Mode 5 will ensure that the maximum pressure will not be exceeded when credit is taken for Code Case
%k N514.
Engineering is in the process of completing plant specific analysis of the POPS utilizing the RHR relief valves (RH3).
Westinghouse WCAP 11640 allows plants to credit the RH3 valves for LTOP applications if the Autoclosure Interlocks for valves RH1 and RH2 tied to PT403 and 405 (> 375 psig) have been removed to prevent the inadvertent isolation of RH3. These interlocks have deleted from both Salem Units. The RH3 setpoint along with the valve capacity were generically evaluated by Westinghouse to provide the Appendix G protection during low temperature overpressurization events without relying on the PORVs. Although the' plant specific analyses for Salem 1 and 2 have not been comple ed at the present time, the results are expected to produce acceptable results.that the present Tech. Spec. P/T limits would be satisfied assuming only one RCP in operation.
These analyses do not credit Code Case N514.cwck CLu\ f.NCfdMO\ h.
Based on the above, reasonable assurance exists that the current Tech.' Spec. P/T limits would be met when considering the pressure difference between the midplane of the Reactor Vessel and the location of PT403 and 405. Therefore, it is judged that this issue-is not an immediate operability or safety concern.
F0ib CWh? d ve*% % %%
t:_Rif deM9 Ws %d % creda % EH 3, wave.
l EXHBIT PAGE d 0FJ PAGE(S) i
k I 9
EXHIBIT 33 e
V Case No. 1-95 013 E
s f(anyThoonSdn
)px E I f
)B038 J S 0 1 ,j ~
U T -'~7d Pubhc Service Electnc and Gas Company P.O. Box7.y. 236 Hancocks 3,, 3_ . Bndge, ,
Ne MEC-94-630 Nuclear Department - , , -v. t l
~~
..-n _
"'S
.1 -
J. Wiedemann ',_,,,'".--
TO: Technical Engineer J_;/f */*
,, { ~_ Z _- _.
H. Berrick Sup rvisor
( lJ-FROM: .i._ .s.. ..
Salem Mechanical En inee Y
SUBJECT:
NONCONSERVATISM IN POPS SETPOI (NSAL-93-005B)
ADVISOKY LETTER PSE-93-204 from Berrick to Schnarr
REFERENCE:
MEC-93-917 Dated 12/30/93 DATE:
May 26, 1994
Background
3 204 identified a Westinghouse NSAL-93-005B transmitted via PSE-9 h POPS - setpoint potential issue regarding a nonconsc.vatism in t eMEC-9 development. This memorandum (MEC-94-630]
resolution. transmitter The pressure d3fference from the wide range pressure (where the Tech.
was not to the reactor vess;l mid planecooldown pressure / temperature considered in Westinghouse analysis. sel mid effectively results in the pressure in the reactor ves essure plane being greater than that seen by the potential wideldown range to pr transmitters used to actuate the PORVs, with aresul pressure / temperature limit curves. used methodology The Salem POPS enalysis (SGS/M-DM-042 and 062) Mitigating provided by Westinghouse (July in their report "PressureThe methodology 1977). and i Systems Transient Analysis" report did 2.not consider the pres therefore the subject NSAL applies to Salem 1 &
EXHSIT N l-9~-013 n - mua Blu/e'g m as / a= m g ,as, L -
~
5/26/94 2
h, Wiedemann Discussion /T limit curves (attached)
Tho Tech. Spec. heatup and cooldown integrity.
Pwith The the requirements of Appe pressurizer cro determined in accordance of 10CFR50 and ensure reactor protects vesselthe RCS fromby opening the P ovorpressure protection system (POPS)cxeceding th during cold overpressure transients bolow 312*F). ure sensors PT 403 and Tha POPS uses the two wide-range These sensorsRCSsense presshot leg PT 405 to actuate the PORVs.The vesselpressure mid plane at thewill be higherto the dy prcssure. The dynamic pressure than the pressure at the hot leg duei s coolant pumps (RCPs) prossure difference between the Westinghouse locat did not on . thein consider difference location in operation at the time. l mid plane leading to a dolta-P associated with the difference conservative.
rcngo transmitter relative to the calculated vesseconcern the that the (SGS/M-DM-042 and 062)ld overpressure transient to The Salem POPS analysis m ximum pressure attained during a cod 418Therefore, psig for the heat be it must bo 446 psig for the mass input case anl s the additive pressure input case with the PORV set at 375 psig.
shown that the above peak pressures, p udoes not exceed the Tec.h based on the Westinghouse in order to notification comply with the Appendix G Spec. P/T limits, -
requirement. the allowable temperature The Tech. Spec. P/T limit curves definetes up to 60*F/hr and for For the POPS -
and pressure combinations for heatup 100*F/hr.
cooldown rates ranging from 0*F/hr to f theraheatup curve and the analysis a composite curve made up The o use of cocidown rate oftempera 20*F/hr 20*F/hr iscooldown curve isatused.
justified because the lowhievable.
A review of the armed, higher cooldown rates are not ac the 20*F/hr cooldown Tech. Spec. P/T limit curves shows thatt low temperatures, d is 450 psig on curve is more limiting than heatup curve aThe pressure tha both Units.on Unit 1 and 475 psig on Unit 2.
9 EXH IT M PAGE . OF/d_ pAGE(S)
h*
5/26/94
! 4 3
- 7. Wiedemann Spec, curves can be' gained.This for the .
Additional margin in the Tech Code Case NS14. m pressure i
POPS application by taking credit for ASMEtates "LTOP sy
. in codthe 3 case s vessel to'110% of the pressure {LTOP - Low 5". determ This App 3ndix G of Section XI, Article G-221 h same as POPS).
Tcaperature Overpressure Protectionh is_t 93 e Winter Addenda h of 1993 Coda Case has been now incorporated ed by 10%. in t eBy In thistaking credit for th ASME Codo,Section XI.
the allowable pressure can be increasbe. exceeded
,However, is 495_(vs ccso the lowest pressure that must not i Unit 2. ,
450] psig on Unit 1 and 522.5 [vs 475) psig onof the Code can be noither the Code Case or the updated vers h onfor Salem.
cpplied at this time pending NRC approval of t em Comparing Evnluation .
h esults ssure of of the 446evaluation.
psig to the Tech. Spec.
.Tcbles 1 & 2 summarize ths POPS analysis maximum pre t e r ble to accommodate the P/T limits shows that the margin availa Unit 2. d on dalta-P is 4 psid on Unit 1 and 29 psi d assess the benefit of To quantify the Salem specific delta-P an ted toThe'results calculate of Jcwer operating pumps, Westinghouse was requestwo and the delta-P for one, l
93-707).
the calculation provided delta-P va uetwo and four RCPs respectiv d to the psi for one,WastinghouseThe assumed transmitters are zerce are dded the transmitters hot-leg (97 f t) . To correct for this difference 2 psi was aThe m suction line at 92.4 ft.
to the Westinghouse results. i ressureas difference applicable, is e
above delta o values or just the stat c p l ,
between the transmitter and the core mid p an ,
presented in Tables 1 & 2. i of 450 psig is t t start of one SI Table 1 shows that Unitd1anTech. Spec, min mum infringement of the pump for the mass input case. ns. [a] Recent up and.cooldown curve is not considere timated peak pressure Appendix G concern ~
38 psig,forandthe[b] has following Recent reasoinform calculation calculated peak
.for mass input case to be 4 i using RH 3 valve to provide LTOP mitigat on[ Salem-has removed i
pressure for mass input case as 420 ps g
~Autoclose' Interlock).
EhyflTM E b,g2PAGE(S)
0 t
J. Wiedemann 4 5/26/94 Recommendation In summary to address the POPS setpoint non conservatisms identified in Westinghouse Nuclear Safety Advisory letter PSE 204, we recommend restricting the number of RCPs in operation while in Mode 5 to no more than one RCP. This has been already incorporated into station procedure IOP-2 (Cold Shutdown to Hot Standby) and IOP-6 (Hot Standby to Cold Shutdown). Future changes to the Station procedures should not rescind the .
restriction of Mode 5 operation with no more than one RCP. The next capsule on Salem Unit 1 is scheduled for removal during the Spring 1995 refueling outage. To address any nonconservative shift of the Appendix G curves at that time, (1) Licensing should pursue approval of Code Case 514 for Salem and (2) Initiate a License-Change Request to take credit for RH3 safety relief valve for LTOP. This ATS open item NSAL-PSE-93-204 is considered closed by this letter.
GD:
Attachments V, p C M. Danak V. Chandra C. Lashkari K. O'Gara K. Pike J. Ranalli -
J. Serwin F. Schnarr D. Smith ,
ATS File MEC File Standards Records ~ Coordinator e
EXHIBIT PAGE Y OF_/O_PAGE(S)
A TABLE 1 i SALEM ESTIMATED INITIATING EVENT DELTA P [ VESSEL STATIC HEAD CORRECTED LTOP ALLOWABLE
' UNIT PEAK PRESSURE FOR LTOP MID PLANE TO FROM PEAK PRESSURE' PRESSURE FOR
' NUMBER IN PSIG FOR TRANSMITTER IF TRANSMITTER TO AETER LTOP CO*iDITION' MASS [M] & ~[ BASED ON' TECH APPLICABLE VESSEL MID ADDRESSING TO MEET APPENDIX-HEAT [H] SPEC BASIS' '[ BASED IN CORE PLANE IF WESTINGHOUSE G CRITERIA !
INPUT CASES. 3/4.4.9 FOR UNIT DELTA P APPLICABLE NSAC-93-05B-IN
[FROM 1& 3/4.4.10 FOR CALCULATED BY PSIG [ BASED ON TECH ;
SGS/M/DM 42 UNIT 2] WESTINGHOUSE; & [NOT TO BE SPEC HEAT UP:AND -!
AND 62 BASED ON ADDED WHERE [ SUM OF COLUMN COOL DOWN CURVES RESPECTIVELY] OPERATION OF DELTA P FROM #2 AND COLUMN PROVIDED IN T.S.
ONE RCP AS PREVIOUS COLUMN #4 OR #5 AS FIGURES 3.4-3 PROCEDURALLY IS ADDED APPLICABLE] AND 3.4-4]
RESTRICTED IN ALREADY]
MODE 5) ;
UNIT 1 446 PSIG THE START OF A N/A 97'MINUS 450.7 PSIG 450 PSIG ,r MASS INPUT SI PUMP AND ITS 86'-2" = 10.8' f INJECTION INTO A .
[NO RCP RUNNING) 418 START OF AN IDLE 29+2=31 PSIG N/A 449 PSIG 450 PSIG UNIT 1 i HEAT INPUT RCP WITH THE [29 IS FOR 1 SECONDPRY WATER RCP FROM PSE-TEMP OF S/G LESS93-707 AND 2 THAN OR EQUAL TO PSI WAS ADDED i 50 F ABOVE RCS TO CORRECT r COLD LEG TEMP TRASMITTER EL OF 92.4' TO HOT LEG EL OF 97'] l t
Pn m t UM= _
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TABLI: '
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CORRECTED LTOP ALLOWAQLE-DELTA P (VESSEL STATIC HEAD- PRESSURE FOR INITIATING EVENT FROM PEAK PRESSURE SALEM ESTIMATED MID PLANE TO AFTER LTOP CONDITION PEAK PRESSURE FOR~LTOP TRANSMITTER IF TRANSMITTER TO TO MEL7' APPENDIX
- UNIT ADDRESSING IN PSIG FOR APPLICABLE VESSEL MID G CRITERIA NUMQER PLANE IF WESTINGHOUSE MASS [M] & [ BASED ON TECH SPEC BASIS { BASED IN CORE APPLICAELE NSAC-93-05B IN HEAT '[H] DELTA P PSIG [ BASED ON TECH INPUT CASES 3/4.4.9 FOR UNIT CALCULATED BY SPEC HEAT UP AND 1 &~3/4.4.10 FOR WESTINGHOUSE; & [NOT TO BE COOL DOWN CURVES UNIT 2] ADDED WHERE [ SUM OF COLUMN
[FROM BASED ON #2 AND COLUMN PROVIDED IN T.S.
SGS/M/DM 42 OPERATION OF DELTA P FROM FIGURES 3.4-3
- 4 OR #5 AS AND 62 ONE RCP AS PREVIOUS COLUMN APPLICABLE] ANb 3.4-41l RESPECTIVELY) PROCEDURALLY IS ADDED RESTRICTED IN ALREADY)_ ,
14QDE 5) 450.7 PSIG 475 PSIG:
97'MINUS THE STARf OF A 'N/A 86'-2" = 10.8' UNIT 2 446 I MASS INPUT SI PUMP AND ITS = 4.7 PSIG ;
INJECTION INTO A MATER SOLID RCS 475 PSIG 449 PSIG
[NO RCP RUNNING]
29+2=31 PSIG N/A 418 START OF AN IDLE ;
UNIT 2 RCP WITH THE (29 IS FOR 1 >
HEAT INPUT RCP FROM PSE- -
SECONDARY WATER -
TEMP OF S/G LESS93-707 AND 2 PSI WAS ADDED !
THAN OR EQUAL TO '
TO CORRECT 50 F ABOVE RCS '
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EXIfBIT PAGE Y OF/f1PAGE(S) !
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Case No. 1 95 013 Exhibit .M
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O PSEG .
Pubhc Service Decinc and Gas Company P O Boa 236 HanCecks Bndge. New Jersey OB038 Nuclear Depanment Tot J. Ranalli Nuclear Mechanical Engineering Manager J. Bailey Nuclear En f3 neer ipg Scier es[ Manager !
FROM: D. A. Sm th Principa ngine r - N lear % Licensing
SUBJECT:
NONCONSERVATISM IN POPS SETPOINT (DEF 94-0060)
DATE: SEP 2 81994 REF: HLR-I94400 The purpose of this memo is to request further evaluation by Muclear Mechanical Engineering (NME) and Nuclear Engineering Sciences (NES) to support the prepe. ration of a License Change l
Request (LCR) to credit RHR Relief Valve RH3 as part of the Pressurizer overpressure Protection System (POPS) Technical Specifications. This LCR is required to address an issue ,
Westinghouse identified in letter PSE-93-204 dated March-15, 1993 '
(NSAL-93-005B) regarding a potential non-conservatism in the calculation of the POPS setpoint (375 Psig) that effects Salem Units 1 and 2. The pressure difference from the wide range pressure transmitters (PT403 and PT405) which sense hot leg pressure to the reactor vessel midplane (where the Tech. Spec.
heatup and cooldown pressure / temperature (P/T) limits are defined) was not considered in the Wentinghouse analysis.
The Tech. Spec, heatup and cooldown curves are determined in accordance with the requirements of 10CFR50, Appendix G and ensure reactor vessel. integrity. The POPS protects the RCS from exceeding the Tech. Spec, limits by opening the PORVs during cold overpressure transients (RCS cold leg temperature below 312*F).
The current heatup and cooldown curves (Tech. Spec. Figures 3.4-2 and 3.4-3) POPS limits are 450 and 475 psig for Salem Units 1 and 2, respectively. The Salem POPS analysis calculated a maximum peak pressure during an overpressure transient of 446 psig with the PORV set at a pressure of 375 psig. Further analysis was required to show that the pressure difference between the RCS hot leg and the midplane of the vessel plus the maximum calculated pressure (446 psig) did not exceed the Tech. Spec. P/T curves.
CASE ND. 1 - 9 .5 - 0 1 3 .
>> EXHIBIT The Energy Peopk g PAGE / OF PAGE(S)
SEP 2 81994 J. Ranalli/J. Bailey 2 .
NLR-194400 The resulta of this avaluation by NME are provided in letter MEC-94-630 dated May 26, 1994 (Attached). To quantify the Salem specific delta-P and assess the benefits of fewer operating RCPs, Westinghouse was requested to calculate the delta-P with one, two and four RCPs operating. The results of the calculation provided delta-P values of 29 pai, 37 psi, and 71 psi with one, two and four RCPs, respectively. In addition, the static head between the core mid-plane and the transmitter was also added for the mass addition transient (4.7 psi). However, one assumption that NME considered in the mass addition transient was that an RCP would not be running with the RCS in a water solid condition.
The RCPs would not be running unless a bubble has been established in the pressurizer in accordance with plant Tech.
Specs. Therefore, the pressure difference calculated by H between the core mid-plane and the pressure transmitters was not added to the analysis results.
Procedure revisions have been implemented to limit the number of RCPs in operation to 1 while in Mode 5 to ensure that the P'T limits will not be exceeded.
As recommended in MEC-94-630, additional margin on the Tech.
Spec. curves can be gained when operating wit'h POPS (RCS loss than 312*F) by taking credit for ASME Code case N514. This Code case states that the LTOP systems shall limit the maximum pressure in the vessel to 110% of the pressure determined to satisfy Appendix G of Section XI, Article G-2215. Crediting the Code case will allow the maximum allowable pressure (Tech. Spec.
P/T limits) for POPS to be . increased to 495 psig and 522 psig for Salem Units 1 and 2, respectively. However, utilization of the Code case will require NRC approval prior to implementation.
This Code case has been relied upon by another utility under these circumstances. It is NLR's intention to submit this -
request to the NRC by October 30, 1994.
Engineering has also completed a plant specitic analysis of the POPS utilizing the RHR relief valve (RH3) . Westinghouse WCAP 11640 allows plants to credit the RH3 valves for LTOP applications if the Autoclosure Interlocks for valves RH1 and RH2 tied to PT403 and 405 (> 375 psig) have been removed to prevent the inadvertent isolation of RH3. These interlocks have been deleted from both Salem Units.
The additional plant specific analyses for Salem 1 and 2 completed by NES have determined that the present Tech. Spec. P/T limits would be satisfied assuming either 1 or 2 RCP(s) in operation utilizing RH3. These analyses do not credit code case N514. The results of the analyses for both the mass addition and b
PAGE OF PAGE(S)
SEP 2 81994 J. Ranalli/J. Bailey 3 N LR-I94 4 00 heat addition transients is documented in Calculation S-C-RC-MDC-1358 dated June 13, 1994. This calculation is to be utilized by NLR as the basis for a License Change Request to credit RH3 as part of the LTOP ao recommended in memo MEC-94-630. NLR has reviewed this calculation, and requests NME and HES provide the following additional information to support this license change:
- 1) Calculation S-C-RC-MDC-1358 presently only supports the use of 2 PORVs and RH3 for overpressure protection assuming a single failure of a PORV or RH3 with up to 2 RCPs running. It was NLRs belief that the analysis would support a Tech. Spec. change that would allow either one PORV or RH3 to be available to provide sufficient relieving capacity to ensura the P-T limits are not exceeded, and still meet the Tech. Spec. LCo. NES should perform additional calculations using the methodology and assumptions contained in calculation S-C-RC-MDC-1358 to justify the use of either 2 PORVs, RI 1 PORV and RH3 assuming a single active failure to mitigate overpressure transients.
- 2) Calculation S-C-RC-MDC-1358 conservatively assumes as the worst case that the pressurizer will be water solid following the inadvertent start of a SI pump, and donsiders the delta-P k associated with one or more RCPs operating. The evaluation documented in memo MEC-94-630 does not assume any RCPs operating with the inadvertent start of a SI pump into a water solid 4 pressurizer as discussed -bove. NME should assess the results of the additional calculations performed in response to Item 1 to g determine if the current Tech. Spec. Bases 3/4.4.9 for POPS (1 PORV provides sufficient relieving capacity to ensure the P-T limits are not exceeded) are still met under the conservative t assumptions of Calculation S-C-RC-MDC-1358 (eee attached Bases Section 3/4.4.9 for Unit 1). The current Tech. Specs. Bases for I POPS do not all redit to be taken for RH3.
- 3) ECCS Te . Spec. Surveillance requirement 4.5.2 for the SI pump and ifugal charging pump identifies a total pump flow rate of $ 64 gpm and 560 gpm, respectively. These flow rate are currently nded by the worst case LTOP flow rate of 780 gpm used in the analyses. Additional margin can be gained by using the more realistic flow rate currently governed by the plant Tech. Specs. Since 10P-6 requires the tagging of the electrical supplies to both SI pumps upon entry into Mode 4, the Centrifugal charging Pump maximum flow rate of 560 gpm per Tech. Spec. 4.5.2 could be used as the worst case LTOP flow rate since the SI pumps power supplies have been removed.
A meeting was held between NME, NES and NLR on September 16, 1994 to discuss the approach for resolution of the 3 items discussed above. It was agreed during this meeting that the following actions would be taken to address the POPS setpoint issues:
k PAGE OFh PAGE(S)
SEP 2 81994 J. Ranalli/J. Bailey 4 NLR-I94 4 00
- 1) NME vould issue a problem report documenting that sufficient relieving capacity of the POPS would be available to ensure the P-T limits 'tre not exceeded assuming a single failure of one PORV when considering the conservat:.ve assumptions of Calculation S-C-RC-MDC-1358 (See attached Basen Section 3/4.4.9 for Unit 1) . -
- 2) NES would perform additional analyses to address the ;
comments provide in questions 1 through 3 above. These '
i additional analyses will be utilised in support of a License !
Change Request (LCR) to credit RHR Relief Valve RH3.
- 3) NLR will prepare a letter to the NRC requesting that application of Code Case N-514 for Salem Units 1 and 2.
This will allow the maximum allowable pressure (Tech. Spec. 'i P/T limits) for POPS to be increased to 455 psig and 522 psig for Salem Units 1 and 2, respectively.
)
- 4) Upon receipt of the additional analyses from NES, NLR will
- prepare a LCR to credit RH3.
i Should you have any questions, please contact D. Smith at extension 1282. !
l KOG/ I C Manager - Licensing & Regulation (N21)
Operations Manager - Salem (Sol)
Technical Manager - Salem operations (502)
H. Berrick (N50)
M. Danak (N50)
V. Chandra (N32) ,
J. Curham (S02)
P. Steinhauer (N31)
Manager - Station QA Salem (S16)
C. Bersak (S16)
Station Licensing Engineer - Salem (N21)
Filet 2.3.1 EXHiiT er T
'PAGE q OFhPAGE(S)
L_ . . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _- _ _ _ ___ .. . - _ .
RDCTOP C00lMT SYSTDI '
nAsr.s Finally, the new 10CFR50 rule which addresses the metal temperature of the closure head flange regione is considered. This 10CFR50 rule states that the metalteeperatugeoftheclosureflangeregionsmustescoedthemateria by at least 120 F for normal operation when the pressure escoeds 20 percent the preservice hydrostatic test pressure (621 pegg for Sales). Table 83/4.41 indicates that the limiting RT of 24 7 occurs in the closure head flange,ofSalesUnit1,andtheminimunT1ovabletemperatureofthisregion is tal F at pressures greater than 621 peig. These limits do not affect Figures 3.4 2 and 3.4 3.
Although the pressuriser operates in temperature ranges above those for which there is reason for concern of non ductile failure,*operatim limits are provided to assure compatibility ot o ration with the fatigue analysis
,poog - ... e __
The OPERASILITY of two POPSs or an RCS vent opening of greater than 3.14 square inches ensures that the RCS will be protectd from pressure transients )
which could onceed the lietts of Appendia G to 10 gFR Part 50 when one or more of the RCS cold less are less than or equal to 312 F. Either POPS has adequate relieving capability to protect the RCS free overpressurisation when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50'r above the RCS cold les temperatures, or (2) the start of a safety injection )
pump and its injection into a water solid RCS.
/
A e
SALDi WIT 1 B 3/4 4 11 Amendment No.108 EXHIIT PAGE-
_OFh PAGE(S)
4 3/6.5 D'DGENCY CORE C001.2NG SYSTEMS RASES 3/4.5.1 ACCWULATORS .
The OPERABILITY of each RCS accumulator ensures that a sufficient volume of borated water will be immediat61y forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the cors provides the initial cooling mechanise during large RCS pipe ruptures.
The limits on accumulator volume, boron cor.contration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.
The accumulator power operated isolation valves are considerd to be
" operating bypasses" in the contest of IEEE Std. 279 1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valvss is required.
The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimisen the time esposure of the plant to a LOCA l event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.
l 3/4.5.2 and 3/4.S.3' ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergancy core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.
Either subsystem operating in . conjunction with the accumlators is capable of supplying suf ficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold les pipe downward. In addition, each.ECCS subsystem provides long term core cooling capability in the ci : 1 & n Q during the accident recovery period. '
W _]
The limitation for a maximum of one safety injection pump or entrifugal I charging pusp to be OPERABLE and the Surveillance Requirement to verify all safety injection pumps except the allowed OpIRA312 pump to bc inoperable below 312'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single F0Fs relief valve}
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SA12M - WIT 1 8 3/4 5 1 gtp[
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EXHIBIT 35 L '
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Case No. 1 95 013 Exhibit 35
/$
FORM NC.WA-AP.EE-0006-1 ,
NUCLEAR DEPARTMENT INCIDENT REPORT FORM I
\ VI- H n s COMMITMENT NUMBER INCIDENT REPORT NO.
USE CONTINUATION SHEFTE IF NECESEARY SECTION I (Initiator)
REPORT SURJECT: Pops A ,l a h n n , m / ,,, aa ,dm ,J #de UNIT S3,HC): 51,5 2 DATE OF INCIDENT /! / t'1/ 9 TIME: / 730
SUMMARY
OF EVENT (IF EST ACTUATION, INCLUDE SOE PRINTOUT):
hi MLJ,2 Sa ,,,in. ~ J nd-0' /
b REPORT 7.D BY: Nes O'[m DEPT: A/4A PHONE EXT: /3 70 SECTION II (SNSS/ OPS MG)
RX PWR AT TIME OF EVENT: 100 % UNIT LOAD: Il6D MW. Op Con / Mode I REPORT MADE PER ECG? (Y/N): _k (IF YES, ATTACH ECG COPY)
LCO #2 A/S #: DATE IN: _ TIME IN: W.R.#:
INITIAL CAUSE DETERMINATION: EQUIP DESIGN vPf.RSONNEL PROCEDURAL OTHER: __
REPORTABLE:
REA ON See Attaele) Ger.um Edb - On H-4 bM SNSS/NSS SIGNATUREt sN C DATE: il /O /N COMMENTS:
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OPERATIONS MANAGER REVIEW : M gyi g,- @TE: ///M/ff CASENi 1-95-013 PAGE / 0@ PAGE@
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e EX fH IT b PAGE 7 _ OF/Y _ PAGE(S)
_ _ _ _ _ _ _ _ _ --_ _. _ _ _ _ _ . _ _ m.____ _ _ _ _ . _ _ . _ _ _ _ ___
i ASME Code Case N-514 allows exceedance of the Pressure / Temperature limits calculated in acecrdance with 10CFR50, Appendix G by 10%. During development of a letter to the NRC requesting application of this Code Case for Salem Units 1 and 2, the following concern was identified regarding the mass addition transient assumptions that may result in exceeding the current P/T limits contained in Technical Specification Figures 3.4-2 and 3.4 3 based on recent POPS analyses completed by .
Nuclear Engineering Sciences. '
BACKGROUND The Tech. Spec. P/T limits for plant heatup and cooldown, which are determined in accordance with the requirements of 10CFR50, Appendix G, ensure reactor vessel integrity. The Pressurizer Overpressure Protection System (POPS) protects the Reactor l Coolant System (RCS) from exceeding the Tech. Spec. limits by ,
opening the Power Operated Relief Valves (PORV) during cold l overpressure transients (RCS cold leg temperature below 312*F) . .
Either PORV has adequate relieving capacity to protect the RCS l from overpressurization when the transient is limited to either j (1) the start of an idle RCP with the secondary water temperature i less than or equal to 50*F above the RCS cold leg temperature j
.(heat addition), or (2) the start of a safety Injection (SI) pump and resultant injection into a water solid RCS (mass addition).
5 The pressure limits at the low temperature end are 450 and 475 psig fo'r Salem Units 1 and 2, respectively, as read from the current heatup and tooldown (20*F/hr) curves (Tech. Spec. Figures 3.4-2 and 3.4-3) The original Salem POPS analysis calculated a maximum peak pressure for the most limiting mass addition transient of 446 psig with the PORV set at a pressure of 375 psig. In this analysis, the RCS pressure due to injection of 780 gpm SI fl,ow into an initially cold water solid RCS was .
considered. In the limiting heat addition transient, a maximum peak pressure of 418 psig was calculated with the PORV set at a pressure of 375 psig.
Westinghouse identified in letter PSE-93-204 dated March 15, 1993 (NSAL-93-005B) a potential non-conservatism in the calculation of the peak pressure for the heat input and mass addition transients that affects both Salem Units 1 and 2. The pressure difference between the wide range pressure transmitters (PT403 and PT405) locations, which sense hot leg pressure and the reactor vessel midplane (where the Tech. Spec. heatup and cooldown pressure / temperature (P/T) limits are defined) with the reactor coolant pumps (RCP) operating was not considered in the Westinghouse analysis.
l l Page 1 EXHIIT by L PAGE OF /Y PAGE(S)_
L. _ __ _. - --- _ _ _ . .- _ -- - - - -
t To quantify the Salem specific pressure difference associated with RCP operation, and assess the benefits of fewer operating RCPs, the delta-P was calculated for one, two and four RCPs operating. The results of the calculation provided delta-P 4
values of 29, 37, and 71 psig with one, two and four RCPs operating, respectively. In addition, a correction pressure of 2.0 psig was added to these delta P values to account for transmitter elevation differences assumed in the calculations.
Procedure revisions have been implemented to limit the number of RCPs in operation to 1 while in Mode 5 (less 'than 200*F) .
However, even with the administrative controls in place to limit only 1 RCP in operation, the calculated peak pressure for the mass addition transient is 477.0 psig when considering the calculated delta-P value for i RCP in operation.
Therefore, this would renuit in the pressure limits at the low temperature end of the Technical Specification heatup and cooldown curves for Salem Units 1 and 2 (450 psig and 475 psig, ;
respectively) being exceeded during the mass addition transient when considering the pressure differential due to the operating RCP.
Further engineering evaluation.and anrlyses were required to determine if conservatisms that formeo che bases for the original POPS analysis could be removed to determine if the current Technical Specification P/T limits could be met. It was determined that the SI pump flow rate of 780 gpm used in the original POPS mass addition transient analysis was extremely conservative. Technical Specification Surveillance Requirement 4.5.2 identifies a maximum floy rate for a SI pump and a Centrifugal Charging pump as 675 gpm and 560 gpm, respectively.
Current plant operating procedure requires that the power supplies be removed f rom both SI pumps upon entry into Mode 4.
Therefore, the most limiting mass addition transient that could .
occur would be the inadvertent start of a centrifugal charging pump.
The mass addition transient was analyzed using the GOTHIC computer code realistically assuming a maximum Centrifugal Charging pump flow rate of 560 gpm. The resulting peak pressure was 407 psig, When considering the pressure differential due to the operation of one RCP and the 2 psig elevation correction (31 psig), the new POPS analysis peak pressure of 438 psig is below the POPS limits 450 and 475 psig for Salem Units 1 and 2, respectively. The esults of the new POPS analysis based on the Centrifugal Chargi g Pump flow rate only for the mass addition transient is .marized in the following-table:
0{
Page 2 EX BIT PAGE _OF/[ PAGE(S)
UNIT POPS CALC. PEAK aP W/ 1 CORRECTED T/S SETPOINT PRESS. RCP PEAK RCS PRESS.
(PSIG) (PSIG) RUNNING PRESSURE LIMITS +
SASED ON (pslo) (PSIG) NEATUP OR MASS INPUT 20*F/NR CASE W/ 560 CooLDOWN GPM FLOW (PsIG)
S1 375 407 31 438 450 S2 375 467 31 438 475
~
At the present time based on the reanalysis assuming the Centrifugal Charging pump flow rate, this issue does not represent an immediate operability issue since the current (
Technical Specification P/T limits are not exceeded. "
Administrative controls arn currently in place to limit one RCP in operation during Mode 5 to ensure that the current P/T limits for heatup and cooldown contained in Technical Specification Figures 3.4-2 and 3.4-3 are not exceeded. In addition, current operating procedures require that power be removed from the SI pumps during Mode 4 to preclude the possibility of an inadvertent pump start. These actions were taken as documented in Problem Repo.rt (PR) 940927126.
CURRENT. ISSUE Recently, during resolution of an unrelated issue regarding the Inadvertent Safety Injection at Power transient, it was determined that the Positive Displacement Charging Pump (PDP), if already in operation, would continue to operate upon initiation of a SI signal if offsite power is available. Letdown would be automatically isolated as part of the SI actuation. The POPS -
analysis provided above had not considered the additional flow associated with the PDP should a safety Injection actuation occur when the POPS is required to be operable. The combined flow of 665 GPM from the PDP (105 GPM) and the CCP (560 GPM) would result in exceeding the P/T limits for Salem Unit 1 based on analysis performed by Nuclear Engineering Sciences assuming the SI pump runout flow of 675 GPM. The calculated peak pressure for a mass addition transient assuming a total flow of 675 GPM is 474 psig.
This analysis was performed using the GOTHIC Code. Based on the results, the P/T limit of 450 psig for Unit 1 would be exceeded, The P/T limit of 475 psig for Unit 2 would continue to be met.
The peak pressure includes the delta-P of 31 psig from the operation of one (1) RCP as discussed above, s
Page 3 gyg) 7h PAGE 7 _OF/Y PAGE(S) .
I I
The additional flow f rom the PDP is only a concern for f.he period of time when the RCS is less than 200 degrees F und the PDP is an l operation removed per and One (1) CCP has not had associated power cupply procedure.
1 This scenario is beyond the present accident analyses basas for t.he POPS which assumes only the inadvertent Centrifugal Charging Pump as discussed above.o t a r t. ot a Baned on the above, it is recommended that when RCS temperature '
is less than 200 degrees F and the PDP is in operation, ensure relief valve isvalve RH3 is available and the associated RHR locked open with power removed. Analyses for isolation a mass addition transient of 780 GPM, assuming either 2 PORVs, or 1 PORV and RH3 are available, have determined that this combination of valven provide sufficient relieving capacity to ensure that the ;
current requirement P/T limits will not be exceeded. This procedure ;
and the POPisisnecessary while power is available to the one CCP in operation, overpressure protection Tech. Spoco. and POPS is relied on to meet the This requirement would net he requ!, red if a vent path has been establis.hed inntead of POPS i in accordance with Tech. Specs. Also, the SI pumps shall have power removed in accordance with procedure as discussed in Problam F.eport (PR) 940927126.
t cannd on tn* above, this concern doeu not pene ar, immed: ate.
uperabiitty concern. )
However, the current den:cn basin ter iha 1"l PS ic tnat one valve the event of an overpressure provides adequate rejiev:ng capacity in transient. Based on the above, one POPS is not outficient for Unit 1, and credit. muut be taken tv:
!<10 1soue to mitigate the mass addition transient . rharetore, this is reportable in accordance with 10CFR50.72 tb) (1) f a il .
Nuc; ear Licensing and Regulation is currently an the process 01 preparing to S14 a letter to the NRC requesting app 12eatic.n of Code Case limits wnach allows exceedance of Lhe Preusure/Tamperat .:re (P/1) calculated Arp*:atix C by 10% . in accordance with the requirements ni 10CFR50, u/T limits operating This additional margin for the POPS 10%durang marcin will increasa tna 1..ow overpressure Protection (1. TOP ) conditicos, and allowTemperature the npet at int restrictions discuaued aoove rn na temnved A !.t ase
- ?
4 P@Shang.: Request is also under considerat.an to c:enit R!D au par.
Pw .1 E BIT N PAGE OF/ik PAGE(S)
I i
ECG ATT 12
- Pg. 1 of $
ATTACEMENT 12
- ONE EOUR REPORT -
Mr AS t A. Refer to Attachme- 1, Non-Emergency Notifications .
Reference, for t' udrrent lit. ting of individuals and phone numbon .
B. Each step shall be 'nitialed by the responsible individual when completed.
/
I. NOTIFICATIONS
'a 1. Notify the operations Manager and confirm classification of h event. ECG Section IB Initiating Condition }4 .
Contacts are listed in Attachment 9)
(AU LFoMo notified at lMo hrs on II- 17-W '
name time date
- 2. Complete the NRC Data Sheet with initial data available (pages 4 and 5 of this attachment).
b 3. Notify NRC Operations Center of the event witMin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> .
SNSS Use NRC Data Sheet to record additional information provided to the NRC. (Phone numbers are in Attachment 9) -
O! lhC name notified at IMD hrs on /f-n 9/
time date
- 4. Notify the NRC Resident Inspector. (Contacts are listed in Attachment 9)
NC4 4 rotified at /WO hrs on_U'/7' M name time date 9
SGS Rev. 2 EX IT M P!GE OF,/f PAGE(S)
ECG *
' ATT 12 P9 2 of 5 INITIA1A '
LN55 5. Notify Public Information Manager - Nuclear or Alternate with details of the event. (Contacts are listed in Attachment 9).
bCA M notified at Iho f hrs on_ //-/M/
name time date -
. Notify Telecopy Group E by transmitting the NRC Data Sheet If transmission of this renort to Group E is incomplete, notify the Emergency Preparedness Representative with the description of the event. (Contacts are(EP Rep.)
listed in Attachment 9)
EP Representative:
r.a m e notified at hrs on time date I
'I 7 If a maior loss of communications caoability has occurred (such as ENS, NETS, DID, etc.), then notify both of the followingt ,
EP Representativet (Attachment 9) notified at hrs on name time date Newark Belp Desk operatort Inform the operator that the(201-430-5555 failed systemorisESSXan 5555) -
l "Energency Priority Circuit".
SH33
- 8. Notify External Affaire with details of the event. (Contacts are listed in Attachment 9).
notified at hrs on name time date Y
SGS Rev. 2 EXHIBIT ___sk PAGE / OF /Y PAGE(S)
l ECG ATT 12 -
Pg. 3 of 5 l
. i REPORTING Ilf*
- 1. Ensure that an Incident Report (IR) is prepared.
E'i B l' 2. Forward this attachment, along witn the IR and any 3RH supporting documentation, to the operations Manager.
- 3. Review IR and any other available information for correct classification of event and corrective action taken.
hp 4. Contact the LER Coordinator and request written follovup i M ( required 30 days after event ). Provide this l
attachment and any other supporting documentation ,
. received from the SNSS. i
- 5. Prepara required reports. ECG. Attachment 23 may be UR"d used as a guide for reporting requirements.
l Report or LER Number
- 6. Forward this attachment to the Energency Preparedness DT5 Manager.
- 7. Ensure that offsite (state and local) reporting EPW" requirements have been met.
- 8. Forward this Attachment package to the Central Technical EPM Document Room (CTDR) for microfilming.
SGS Rev. 2 EXHIBIT PAGE // OF/Y_ PAGE(S)
.....,o........................
REMARKS / COMMENTS -------------------------TCMQ324 MhAND INPUT ===> PAGE 4 OF 4 DISPLAY MODE PR HUMBER: 940927126 EVAL TYPE: DTPS EVAL NBR: 01 THE FUNCTIONAL GROUP'TS RESPONSIBLE FOR EVALUATING THE EFFECTS OF AN IDENTIFIED NON CONSERVATISM IN THE POPS SETPOINT.
PRIORITY: 14NG TERM .
BASED UPON THE ATTACMMENT TO PR 940927126, THE PROCEDURE IOP-6 LIMITS THE PLANT OPERATION BELOW 312 DEGREES F TO ONE CHARGING PUMP AND TO ONE RCP WHEN THE RCS TEMPERATURE IS BE14W 200 DEGREES F. IN ADDITION, BASED UPON THE OPERABILITY REVIEW PROVIDED BY THE SYSTEMS ENGINEER, THE OPERABILITY OF THi.
REACTOR COOLANT SYSTLM IS NOT CURRENTLY IMPACTED. THIS DEF IS ASSIGNED A IhfG TERM PRIORITY TO ADDRESS THE POPS SETPOINT NON CONSERVATISMS IDENTIFIED IN WESTINGHOUSE NUCLEAR SAFETY ADVISORY LLTTER PSE-93-204.
THE ONLY CONCERN OF THIS DISCREPANCY THAT COULD AFFECT CORE DAMAGE FREQUENCY.
IS IF THE PRESSURIEER IS SOLID, THEREBY INCREASING THE POTENTIAL OF A LOCA.
HOWEVER THE PROBABILITY OF CORE DAMAGE IF A LOCA OCCURS MAY BE SMALL, IF RHE~
IS AVAILABLE.
A SHUTDOWN PRA FOR SALEM STATION HAS NOT YET BEEN COMPLETED, HOWEVER THE FREQUENCY OF AN SI SIGNAL IN MODE 5 SHOULD BE INFREQUENT. IN ADDITION, THE SYSTEMS F.NGINEER MAS INITIATED A PROCEDURE CHANGE TO PRECLUDE SI PUMP INJECTION DUE TO :.N SI SIGNAL IN MODE 5. (PRESS F8 TO CONTINUE)
DISP: RJC CICHON R 10/13/94 RVW2: WMM MC DEVITT W 10/25/94 RVWit AJS SANDERS A 10/24/94 COMP RJC CICHON R 10/27/94 MESSAGE: NO MORE DATA TO SCROLL IN BACKWARD DIRECTION B MY JOB LU #21
~~-------- REMARKS / COMMENTS -------------------------TCMQ324
- OMMAND INPUT ===> PAGE 4 OF 4 DISPLAY MODE PR NUMBER: 940927126 EVAL TYPE: DFPS EVAL NBR: 01 THEREFORE, THIS DISCREPANCY IS ASSIGNED A IANG TERM PRIORITY BASED ON ENGINEERING JUDGEMENT.
9 9
DISP: RJC CICHON R 10/13/94 RVW2: WMM MC DEVITT W 10/25/9e RVW1 AJS SANDERS A 10/24/94 COMP RJC CICHON R 10/27/94 MESSAGE: NO MORE DATA TO SCROLL IN FORWARD DIRECTION B MY JOB LU #21 EX iT PAGE OFM PAGE(S)
.- EVALUATION ---------------------------TCMQ322
-)MMAND INPUT =="> PAGE 3 OF 4 DISPLAY MODE 940927126 LAST UPDATE: 10/27/94 RJC
.ROBLEM RPT NO: Ph 9TATUS-DATE: REVWD 09/27/94 PR TYPE : FR__ INITIATOR : DANAX7BERRICK
==================== EVALUATION DESCRIPTION =======================
EVAL TYPE: DFSE EVAL MBR: 01 EVAL PRI: 01 EVAL STATUS-DT: COMPLT 10/05/94 EVAL DEPT STD ASSIGNED To TECHNICAL ENGINEER LCO-NUM: ~
EVAL INIT DATE: 09/I7/94 EVAL DUE DATE: 10/04/94 K/S NUM:
UNIT / COMP ID : A/S DATE-TIME: ~
SUMMARY
- LTOPS/F6PS NON-CONSERVATIVE SETPOINT ~
DESC EVALUATE OPERABILITY AND VERIFY AS DEF. F6fBP!TKL THCIDENT REPORT.
CONTACT K. O'GARA, X1370 AND M. DANAK, X1872 FOR ADDITIONAL INFO.
REFER TO THE FOL1hWING PAGES FOR EVALUATION BY C. CHEN,_
LER/SR/AR? : ~ TYPE : NUMBER :
REPAIR: UAI ~ CD NUM: EAL NUMBER :
DCR/TMOD7 : -
ECG7 : ~
CAUSE CODE : EK EI
- 10CFR217 : ~
RX PWR FAILURE CODE : KE ~~
10CFR50.597 : ~
14AD : RESOLUTICN CODE: EP ~
100FR50.727 : ~ MRULE SCOPE: ~
MRULE CODE :
10CTR50.737 :
MESSAGE: NA/IGATE TO OTHER SCREENS, OR 'PF4' TO RETURN TO SEARCH OPTIONS SCREEN B MY JOB LU,#21
~
............................ REMARKS / COMMENTS -------------------------TCMQ324 OMMAND INPUT ===> PAGE 4 OF 4 DISPLAY MODE PR NUMBER: ~
940927126 EVAL TYPE: DFSE EVAL NBR: 01 THE EVALUATION OF THE DISCREPANCY IS DIVIDED INTO TWO SECTIONS FOR MODE 4 AND MODE 5. IT IS CONCLUDED THAT THERE IS NO OPERABILITY CONCERN CAUSED BY THIS -
DISCREPANCY AND IT DOES NOT INVOLVE ANY CONDITIONS THAT MAY BE REPORTABLE.
A. MODE 4 EVALUATION :
FOLLOWING DOCUMENTS WERE REVIEWED FOR THIS EVALUATION: TECH. SPEC.
3.1.2.4, 3.4.10.3, AND 3.5.3, UFSAR 7.6.3, AND PROCEDURE IOP-6. IN MODE 4, IT IS REQUIRED TO HAVE AT LEAST ONE CENTRIFUGAL CHARGING PUMP IN OFRABLE CONDITION PER TECH. SPEC. 3.1.2.4 AND 3.5.3. BUT FOOTNOTE OF 3.5 e MAKES IT_
REQUIRE THAT ALL SI PUMPS BE TAGGED OUT IF RCS COLD LEG IS < 312 DEG.F.
THIS PRECLUDE ANY POSSIBILITY OF ANY INADVERTENT IHJECTION OF SI PUMPS WHICH ~
MAY CAUSE THE CONCERN ADDRESSED IN THIS DEF. AS SHOWN IN THE ATTACHMENT OF THIS DEF, THE WORST CASE PEAK PRESSURE CAUSED BY AN INADVERTENT INJECTION OF, A CENTRIFUGAL CHARGING PUMP IS BELOW THE TECH. SPEC. LIMITS. THUS THE DISCREPANCY IDENTIFIED IS NOT OF OPERATIONAL SIGNIFICANCE IN MODE 4.
B. MODE 5 EVALUATION:
MORE TECH. SPEC. SECTIONS (3.1.2.3 AND 3.4.1.4) AND PROCEDURES (OP-PT.SJ
-002, OP-SO.SW-002 & 003) WERE REVIEWED FOR MODE 5 EVALUATION. DURING IMPLEMENTATON OF OP-SO.SW-002 OR 003 FOR SERVICE WATER HEADER OUTAGE, THERE__
DISP: RVW2:
RVW1: __ COMP TXC CHEN T 10/05/94 MESSAGE:. REVIEW COMMENTS AND PF4 TO RETURN TO DISPOSITION SCREEN.
B MY JOB LU #21 EXH$lT b PAGED _OFM PAGE(S)
__- - A
......................--..... REMARKS / COMMENTS -------------------------TCMQ324
^7MMAND INPUT ===> PAGE 4 OF 4 DISPLAY MODE PR NUMBERI _940927126 EVAL TYPE: DFSE EVAL NBR: 01 IS A POSSIBILITY THAT ONE SI PUMP CAN BE PLACED IN OPERABLE CONDITION. THERE ~
IS NO RESTRICTION IN THE PROCEDURE TO PREVENT THE OPERATOR FROM MAXING THAT ~
CHOICE. THUS A POTENTIAL OF INADVERDENT SI INJECTION EXITS. NORMALLY THE SI PUMPS ARE TAGGED OUT AND AUTO INJECTION SIGNALS ARE BLOCKED IN MODE 5. A PROCEDURE REVISION REQUEST WILL BE ISSUED TO CHANGE OP-SO.SW-002 AND 003 TO PRECLUDE THE POSSIBILITY OF SI INJECTION IF ANY RCP7S RUNNING AND RCS IS_
NOT VENTED. THIS CAN BE ACHIEVED BY ADDING A NOTE TO PLACE h l RESTRIt.rION ? OF CHOICE OF ANY TWO SAFETY GRADE PUMPS IN THE PREPARATION OF SERVICE WATER HEADER OUTAGES. THE REVISION WILL PREVENT ANY SI PUMP FROM BEING PLACED IN OPERABLE CONDITION BUT STILL CAN BE IN AVAILABLE CONDITION TO SATISFY THE REQUIREMENTS OF TECH. SPEC. 3.1.4.1 IF ANY RCP IS RUNNING.
THE CONCERN IDENTIFIED BY THE "WOULD BE POSSIBLE" SCENARIO OF THIS DEF ~
TS THEREFORE DEEMED TO BE NO OPERATIONAL SIGNIFICANCE AND DOES NOT INVOLVE ANY CONDITION THAT MAY BE REPORTABLE. _
BY G. CHEN X 2754 DISP RVW2:
RVWit COMP: TXC CHEN f 10/05/94 l MESSAGE: NO MORE DATA TO SCROLL IN FORWARD DIRECTION B MY JOB LU #21 e
EXH81T PAGE N OF /YPAGE(S)
4 1
EXHIBIT 33 -
r Case No. 1 95 013 Exhibit 30
O PSEG Pubhc Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 .
~
Salem 0enerating Station December 14, 1994
- d. S. Nuclear Regulatory Commission ,
Document Control Desk Washington, DC '20555 Attn.: Document Control Desk '
~
SALEM GENERATING F'ATION" LICENSE NO: DPR-7v DOCKET NO: 50-272 UNIT NO: 1 LICENSEE EVENT kEPORT NO. 94-017-00 t
This Licensee Event Report is being submitted pursuant to the requirements of Code of Federal Regulati'on 10CFR50.73 (a) (2) (1) (B) . Issuance of this report is required within thirty (30) days of event discovery.
Sincerely, -
, . . Hagan Gene @ 4anager -
s Salem Operations HJPJ:vs SORC 94-094 C- Distribution
- f LER File CASE NO.
1-95-013 EMBR N 11epcuerisinymtImls . .. . _
PAGE
/ OFhPAGE(S) nwnm Ag IMue ~
aa
l N~C f ORM 366 U.S. NUCLEAR REGULATORY COMMISSION l pas, APPROVED BY OMB NO. 315%104 EXPIRES 5/31/95 (sTIMATED BURDEN MR RESPONSE TO CXA>Ly vnm yygg LICENSEE EVENT REPORT (LER) "o E E "s"R E EsT0 E io 7E w o,, ,n'aN c u * * * "
AND MO3@s MANAGEMDdT BRANCH (MMBS ffggjjg ggjegggg MOULATORY CUMMISSKN. WASHINGTON. DC pos44008. AND TO (See teverse for required rymber of digits / characters for each block) THE PAPEMWORK EDUCTION PRCUECT (3150cl% CFTQ OF MANAGEMENT AND BUDGET WASHINGTON, DC 20503 pacaufv naut gij ooca.tT MuuaER (2)
Salem Generating Station - Unit 1 05000 272 pAaC m 10F S.
TTnau Anacequate Margin tor Pressurizer Overpressure Protection During Low Temperature Conditions (Applicable to Unit 1)
EVENT DATE (5) LEel NUMBER (6 REPORT NUMBER (7) OTHER FACluTl"S INVOLVED (8)
MONTH Day vtAR vtAR I OU "D"E MogM NVMRER DAY YEAR a HUMPEn 05000 11 17 94 94 - 017 00 12 14 94 ' " * "**"
05000 OPERATING } THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 1: (Check one or ,nore) (11) -
MODE (9) 20 402(b) 20 405(c) 5013(a)(2)0v) 73.71(b)
POWER * * "U N "*MU # (*M*"#3 100% # # "*3 LEVEL (10) 20 405(a)(1)0i) 50.36(c)(2) 50.73(a)(2)(vir) X OTHER Sp"eyin Abstruct
% ,n U ., os W
20.405(a)(1)0li) 50t7N<(2)0) 50 73(a)(2)(viii)(A) i Of+
A ir x 4
S 20.405(a)(1)0v) X $073la)(2)0i) 5013(a)(2)(viii)(B) N i
20 405(a)(1)(v) LOJ3(a)(2)fii) 5013(a)(2)(x) 10CFR21 LICENSEE CONTACT FOR THI3 LER (12)
Naut TELEPHONE NUMBD PnWOe Atta Cedel Michael J. Pastva. Jr. LER Coordinator 609 339-5165 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) causE sYstEu coupoNENT MANUFACTUREP . CAUSE SYSTEM CouPONENT MANUF ACTURE R z $@s
' hf+
mob ga;p g P l
SUPPLEMEIITAL REPORT EXPECTED (ta) EXPECTED MONW DAY YEAR vss X SUBMISSION p vei. me=ie EXPECTED suBuiSslON DATE)
DATE (15)
' bSTRACT (Umit to 1400 spaces, t e.. approximately 15 single. spaced typewritten lines) (16)
On 11/17/94, it was determined that the following realistic assumptions could place Unit 1 outside the design / licensing basis for Pressurizer Overpressure Protection System (POPS) analysis should a safety injection (SI) signal occur:-Reactor-Coolant System (RCS) temperature < 200*F, 1 Reactor Coolant Pump (RCP) in operation, Positive Displacement Charging Pump (PDP) in service, and power available to a maximum of 1 Centrifuga' Charging Pump (CCP). Undet: these conditions, an SI signal could rer. ult in a peak RCS pressure of 474 psig from combined flow from the PDP and the CCP,.which exceeds the current design basis pressure limit of 450 psig.
This concern is not applicable to Unit 2. This event occurred because PDP operation was not considered in any design basis analysis, based on a POPS lift setting of s/= 375 psig. Current operating procedures limit RCP operation in Mode 5 to 1 pt.mp, and require power be removed from the SI pumps in Mode 4 (350*F >Tave, > 200*F) . Residual Heat Removal relief valve RH3 is required to be available when RCS temperature is </= 200'F and when the PDP is ,perating. Code case relief will be requested to increase tpe pressure / temperature limits operating margin for the POPS during Low '
Temperature Overpressure conditions. This LER is also intended to satisfy reporting requirements of 10CFR21.
I ll =:. . J I
~
EXHIBIT &
FAGE d 0F $IPAGE(B) .
4 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page Unit 1 50-272- 94-017 2 of 5 Plant and- System Identification:
Westinghouse - Pressurized Water Reactor Energy Industry Identification System (EIIS) codes are shown in the text as {xx) .
Identification of occurrence:
Inadequate Margin For Pressurizer Overpressure Protection During Low Temperature Conditions (Applicable To Unit 1)
Event Date: November 17, 1994 Report Date: December 14, 1994 This report was initiated by Incident Report 94-419 This report is intended to also satisfy reporting requirements of .
Initial Conditions: -
Mode 1 , Reactor Power 100% Unit Load 1150 MWe Description of Occurrence:
The current bases for Technical Specifications (TSs) 3/4.4.9.3 states that one Pressurize'r Overpressure Protection System (POPS) relief valve, at a lift setting af </w 375 psig, provides adequate relieving capacity in the event of an overpressure transient that includes inadvertent start of a safety injection (SIl pump (mass addition transient) into a water solid Reactor Coolant System (RCS). Subsequently, it has been determined that the following realistic mass addition transient assumptions could place Unit 1 outside the design and licensing basis POPS analysis should an SI s3gnal occur:
- RCS temperature </= 200*F "
- One Reactor Coolant Pump (RCP) in operation
- Positive Displacement Charging Pump (PDP) in service
-Power supply available to a maximum of One Centrifugal Charging Pump -(CCP)
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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Docket Number LER Number Page Unit 1 50 272 94-017 3 of 5 Description of Occurrence: (cont' d) ,
At 1746 hours0.0202 days <br />0.485 hours <br />0.00289 weeks <br />6.64353e-4 months <br /> on November 17, 1994, the Nuclear Regulatory Commission was notified of this event, pursuant to the requirements of 10CFR50.72 (b) (1) (ii) .
Under the above conditions, an SI signal could result in a combined flow from tisc .PDP and the CCP with a peak RCS pressure of 474 psig. This exceeds the current design basis pressure limit of 450 psig for Salem Unit 1.
Analysis of Occurrence:
Background
POPS protects the RCS from excceding the TS pressure / temperature (P/T) limits for plant heatup (reference TS Figure 3.4-2) and cooldown (reference TS Figure 3.4-3) by opening the Power.
Operated Relief Valves (PORV) during low temperature overpressure (LTOP) transients (RCS cold leg temperature below 312*F) . Per existing design bases, either of the two PORVs has adequate relieving capacity to protect the RCS from overpressurization wnen the transient is limited to either (1) the start of an idle RCP with the secondary water temperature less than or equal to 50*F above the RCS cold leg temperature (heat addition), or (2) the start of an SI pump and resultant injection into a water ,
solid RCS (mass addition). The pressure limit at the low temperature end of the P/T curves is presently 450 psig for Unit 1, as read from the current heatup and cooldown curves. -
The Nuclear Steam Supply System (NSSS) vendor identified in a letter, dated March 15, 1992, a non-conservatism in the calculation for peak pressure for the heat input and mass addition transients that affects both Salem Units 1 and 2. The concern was that the difference between the wide range pressure transmitters (PT403 and PT405) elevations, which sense hot leg pressure and the reactor vessel midplane (where the TS heatup and cooldown P/T limits are defined) with the RCPs operating was not considered in the original Salem POPS analysis. Calculations were performed to address the identified concern and necessary ,'
limits were met.
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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION -
Salem Generating Station Docket Number LER Number Page Unit 1 50-272 94-017 4 of 5 Analysis of Occurrence: (cont' d)
Present Situation Since the satisfactory completion of this evaluation, it has been determined that the PDP, if already in operation, would continue ~
to operate upon initiation of a'SI signal if offsite power remained available. During this, postulated event, letdown would ,
automatically isolate as part of the SI actuation. The additional flow from the PDP is a concern for the limited period of time when the RCS is </= 200*F (Mode 5), the PDP is in operation, and one (1) CCP has its associated power supply available. The combined flow of 665 gpm from the PDP (105 gpm) and the CCP (560 gpm) is now considered the most limiting mass addition transient.
PSE&G has re-analyzed this mass addition event using the GOTHIC computer code assuming a bounding maximum pump flow rate of 675 gpm. The resulting peak pressure is 474 psig, which exceeds the current limit of 450 psig for Salem Unit 1.
For the heat addition transient (i.e., the start of an idle RCP with the secondary water temperature </= 50*F above the RCS cold leg temperature), the peak pressure is 449 psig, below the POPS limit of 450 psig for Salem Unit 1.
Additinnal margin on the TS P/T curves can be obtained when operating with POPS (RCS cold legs </.= ' 312*F) by taking credit for ASME Code Case N-514. The code case allows exceeding the P/T limits calculated in accordance with 10CFR50, Appendix G, by 10%.
As compensatory action, administrative controls ensure that Residual Heat Removal (RHR) relief valve RH3 is available and the associated RHR isolation valves are in the open position. PSE&G has deterrined that RH3 has similar relieving capacity to that of a PORV. This action is only necessary when RCS temperature is
</= 200*F, the PDP is in operation and a CCP has power available.
Apparent cause of occurrence:
This event-is attributed to " Design", as classified in Appendix $3 of NUREG-1022. This occurred because the NSSS vendor had not considered the PDP operation as part of any design basis analysis.
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- LICENSEE EVENT REPORT (LER) TEXT CONTINUATION -
Salem Generating Station Docket Number LER Number Page Unit 1 50-272 94-017 5 of 5 Prior Similar Occurrency No other prior similar occurrences have been identified related to this design deficiency.
Safety Significance:
This event is reportable in accordance with the requirements of 10CFR50.73 (a) (2) (ii) (B) , due to the POPS not being able to meet its current design basis. This event had minitral safety significancy, based upon the additional relieving capacity of RH3 and/or wun the 10% allowance, permitted by use of code case N-514.
Corrective Action:
3 As interim corrective action to ensure compliance with the POPS design basis, the following administrative controls are in place:
Salem Unit 1 operating procedures limit RCP operation in Mode 5 to one pump, and require that power be removed from the SI pumps in Mode 4 (350*r >Tave, > 200*r) . RH3 is required to be available to ensure the current P/T limits are met when RCS temperature is
</= 200*F, when the PDP is in operation. Similar administrative controls for Salem Unit 2 are in place, although RH3 is not needed to meet the corresponding Unit 2 P/T limit.
A submittal will be made to the NRC requesting application of Code Case N-514. This additional margin increases the P/T limits operating margin for the POPS during LTOP conditions, i
J. J. n General Manager Salem Operations 1
MJPJ REF: SORC Mtg.94-094 -
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