ML20137W288
| ML20137W288 | |
| Person / Time | |
|---|---|
| Issue date: | 02/12/1986 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| ACRS-T-1488, NUDOCS 8602200104 | |
| Download: ML20137W288 (500) | |
Text
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, UNITED STATES
'~# NUCLEAR REGULATORY COMMISSION IN THE MATTER OF: DOCKET NO: ADVISORY COMMITTEE ON REACTOR SAFEGUARDS SUBCOMMITTEE ON REACTOR OPERATIONS LOCATION: WASHINGTON, D. C. PAGES: 1- 162 DATE: WEDNESDAY, FEBRUARY 12, 1986 [ M 8Edf$E N Y ice Jo Xot Removeirom g 22 g 4 860212 AGSOS T-1488 PDR ACE-FEDERAL REPORTERS, INC. Q OfficialReporters 444 North Capitol Street D k Washington, D.C. 20001 (202)347-3700
\ NATIONWIDE COVERACE
CR25727.0 1 REE/sjg I UNITED STATES OF AMERICA 2 NUCLEAR REGULATORY COMMISSION 3 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 4 SUBCOMMITTEE ON REACTOR OPERATIONS S Nuclear Regulatory Commission Room 1046 6 1717 H Street, N.W. Washington, D. C. 7 8 Wednesday, February 12, 1986 9 l The meeting of the ACRS subcommittee convened at 10 1:30 p.m., Mr. Jesse C. Ebersole, chai_rman, presiding. I i 11 j' ACRS MEMBERS PRESENT: 12 > () [ MR. JESSE C. EBERSOLE DR. WILLIAM KERR MR. CARLYLE MICHELSON l 14 DR. DADE W. MOELLER , MR. GLENN A. REED , 15 MR. DAVID A. WARD ! 16 ' DR. IVAN CATTON, Consultant 7 MR. HERMAN ALDERMAN, ACRS Staff Member 18 a , l: 19 1 i I 20 !! 4 il 21 '! l' i, 22 23 l
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() PUBLIC NOTICE BY THE UNITED STATES NUCLEAR REGULATORY COMMISSIONERS' ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WEDNESDAY, FEBRUARY 12, 1986 The contents of this stenographic transcript of the proceedings of the United States Nuclear Regulatory Commission's Advisory Committee on Reactor Safeguards
.(ACRS), as reported herein, is an uncorrected record of the discussions recorded at the meeting held on the above date.
No member of the ACRS Staff and no participant at () this meeting accepts any responsibility for errors or inaccuracies of statement or data contained in this transcript. O
l 25727.0 2 REE I gj 1 PROCEED I NGS 2 MR. EBERSOLE: The meeting will come to order. 3 This is a meeting of the ACRS Subcommittee on 4 Reactor Operations. I am Jesse Ebersole, Chairman of the 5 Subcommittee on Reactor Operations. The other members are 6 Bill Kerr, Carlyle Michelson, Dave Moeller, Mr. Glen Reed, 7 and we have Mr. Ivan Catton as a consultant. 8 The Subcommittee will discuss recent operating 9 events. We will be briefed on the 50-54(f) Improvement 10 ! Program on Fermi and Technical Specification Improvement 11 Program. 12 6 The rules for participation have been announced l 13 Il as part of the notice of the meeting published in the I 14 a Federal Register on January 21, 1986. It is requested that ; f 15 each speaker first identify himself or herself and speak 16 l with sufficient clarity and volume so that he or she can be i 17 readily heard. f 18 , We have received no written comments or requests t 19 l for time to make oral statements from members of the public.
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20 I will now ask if any Subcommittee members or 21 ; consultants have any comments or solicitations to make 22 1 before we get into the meat of this meeting. i 23 j MR. MOELLER: I have one comment. Just to ask, 24 l Carlyle Michelson had brought to my attention this draft 1 25 ; report or report that the AEOD is preparing on the loss of 1 C) i ace-FEDERAL REPORTERS, INC l 202 u7 no Nationmide rmerage m 1 M f 6m
25727.0 3 REE ('N (_) I the ventilation system in the control room at the McGuire 2 plant, I believe it was. We could take it up as a joint l 3 effort. 4 MR. EBERSOLE: Perhaps next time. I l 5 MR. MOELLER: It will depend on when they finish 6 their report. 7l MR. MICHELSON: I wasn't going to get into it 8 lt until the end of the meeting, but there is this question of, i 9l ? IE does a fine job of bringing to us these current events 10 as they come up but we don't hear briefings on AEOD reports 11 k which deal with combinations often of current events and i 12 i final analysis in the longer-range sense. Which l rT 13 l}
! subcommittee is going to listen to the results of those
(_.) J 14 l 1 kind of studies as opposed to current events?
'i 15 h ,
They are also operating history, just from a , 9 16 different viewpoint. We don't normally bring those up here j 17 ; because this is the operating events. l f ! l 18 l MR. EBERSOLE: I suggested that we take this up I l 19 ) as a part of our presentation to the full committee. I 20 ; MR. MICHELSON: That was the question, should 21 ! this subcommittee do it? 22 MR. EBERSOLE: Let's introduce that as a copy i 23 l tomorrow. 24 Well, we have just gotten through a full morning 25 i which I think ellipses most operation events of recent I i ace FEDERAL. RiiPonTI!Rs, INC. 202-347-37(o Nationwide roserage N o 11/v f M6
25727.0 4 REE 1 times. But I see an interesting number here. Let me ask 2 Mr. Dennis Allison to get us going right away. 3 MR. ALLISON: I am Dennis Allison. I am the 4 acting branch chief for the events analysis branch in IE. 5 Next to me is Ron Hernan, principal spokesman for NRR. 6 Before we start the first event, I would like to say, I 7 understand that the team that has been looking at the 8 earthquake at Perry will arrive a little later, about 3:30, 9 So there toward the end of the presentation, we won't be 10 able to move them up until they get here. I think that is 11 , all we had to start off with. The first presentation is on 12 " McGuire, startup with the degraded HPSI system. It is not t 13 ! an air system but a HPSI system. O e MR. GIITTER: My name is Joe Glitter. 14 li I am in 15 the events analysis branch of the Office of Inspection and 2 16 [ Enforcement. At the last Subcommittee meeting, you heard d 17 g about a single failure of a nonsafety-related system, that 1 18 ; resulted in a loss of instrument air and a challenge to the 1 19 safety systems at both McGuire units. This afternoon, I am l i 20 1, going to discuss a set of undesirable conditions that I 21 l existed the following day when unit 1 attempted to start up. 22 The problem was failure to repair volume control 23 tank isolation valve motor operators prior to startup which 24 would have prevented VCT isolation on a safety injection 25 signal. The safety significance of this is that HPSI may Or ace-FEDERAL REPORTERS, INC, 202 m .37m Nationwide Coverage Am3%MM
25727.0 5 REE 1 not have functioned as required on a real demand. 2 As you recall, at 0640 on November 2, a rupture 3 in a flexible pipe at the discharge of an instrument air 4 compressor resulted in a loss of instrument air to both 5l units 1 and 2, a trip of both units 1 and 2 and a safety 6 injection in unit 1. Ordinarily on a safety injection, you 7 would have isolation -- first you would have opening of 8 these two valves here and then isolation of the volume i 9 control tank as suction is swapped from the VCT to the ! f 10 l refueling storage tank. These valves will go closed as l t i 11 required. 4 i 12 l However, it was later discovered that the valve l V i 13 ) motor operators had burned out. The licensee has two l O :: theories that they are looking at as to why these motor l 14 l i' 15 [ operators may have burned out. One is that there was a n ! 16 hammering problem. l 17 In a hammering problem, as a valve closes, the f 18 4 torque builds up which causes a torque switch to open. t 19 When the torque switch opens, power is removed from the 20 valve motor which relaxes torque. When the torque is 21 { relaxed, that torque switch is allowed to renet. If a l 22 close demand signal is still present, once that torque l 23 switch resets, the valve will start to close again even 24 9 though it is closed. It will try to close against a seat. 25 Torque will build up, the torque switch will open, and the Acti FliDI!RAL RiiPORT!!RS, INC. 202 347-37(#1 Nationwide roserage M f L I M (M6
25727.0 6 REE () 1 motor operator will stop -- will stop trying to close the 2 valve, torque will relax, the same thing will happen over 1 1 3{ and over again and the valve will be hammered closed. This 4 can eventually result in overheating of the valve motor. 5l Another theory is that on the McGuire system, I' , 6 there is a miniflow line from the centrifugal charging 7 pumps that is not isolated on a safety injection signal. 8 i So what happened during this event is that the VCT actually 1 9 went solid and the safety relief lifted on the VCT, 10 - pressure setpoint about 75 pounds. This increases the 11 ; static head that these -- that this valve here sees. It 1 12 was believed that there may have been some leakage past i 13 this valve here and when the operators, seeing that the VCT l1 1 14 ) was 4 full, tried to open remote this valve, remote manually, 15 l that they encountered high differential pressures and that I 16 l caused the motor operator to burn out here and subsequently 1 17 .) high differential pressure caused this motor operator to 18 li burn out because this motor operator here wouldn't see the 19 DP that this one would. 20 l It doesn't seem to be fully explained yet. The f 21 licensee is still investigating. But those are two of the 22 theories right now. ! 23 MR. EBERSOLE: This event is still open as to I 24 ' why these valves failed? 25 MR. GIITTER: That is correct. The licensee is O Acti Fl!DERAl. RI:PoRTi Rs, INC 202447-1700 Nationwide rm erage ht W L 11MM6
25727.0 7 REE 1 still -- 2 MR. EBERSOLE: It would have been unlikely that 3 the torque switch would have been, the lockout or the 4 return -- when the torque switch closes or opens to stop 5 the motor, doesn't that result in an interlock which it 6 cannot subsequently attempt to close again until it opens 7 ! again? 8 MR. GIITTER: There are various types of 9 antihammering devices on different types of valves. My
- 10 understanding is that this particular type of valve had an 11 old mechanical latch system that prevented that torque 12 I switch from resetting if the limit switch had sensed that i
i 13 il valve being closed. Why that wouldn't have worked, I don't l l !I l 14 il know. Y ! 15 , MR. EBERSOLE: Was it both -- f 1 16 ij MR. GIITTER: Both of these. 4 17 MR. EBERSOLE: So we have an even validation of f 18 l the single failure theory. . l 19 ! MR. GIITTER: It seems that a single failure did 20 cause the failure of both of these motor operators. 21 MR. EBERSOLE: Well, it is nice to add another 22 one to the list. Like Salem. Carry on. 23 MR. GIITTER: Prior to startup, the operators 24 manually opened the valves, but did not repair the motor 25 operators. So this put them in a situation that if they O ace-FEDERAL RiiPORTI Rs, INC. 202-M7-1700 Nahonwide Cmeraer N st D6 #6 4
25727.0 8 REE () I were to have a safety injection signal, these valves would 2 not have automatically closed. And you would then be in a 3 situation where you were taking suction from the VCT and 4 RWST. Unit startup was commenced at about 6:00 on the 3rd 5 and the unit was in mode 2 for about six hours. The 6 startup was pending repair of the severed instrument line 7 on the secondary side of the steam generator. Because they 8 couldn't get that repaired, they decided to go into cold 9l i shutdown. 10 MR. EBERSOLE: So what would have happened if 11 you had taken common suction on the VCT as well as the RWST. 12 Would you get nitrogen? 13 l MR. GIITTER: I will get to that. (:) 14 There are two concerns. One concern, I think 15 this is a lesser one, that is because of the difference in i 16 3 volume -- this is the volume of the VCT is on the order of 17 several thousand gallons whereas the volume of the RWST is 18 on the order of several hundred thousand gallons. But 19 there is a concern, if you were to draw some suction from 20 the VCT, that the boron injected into the vessel would be 21 lower than that assumed ta the safety analysis. That la a 22 concern more from reactivity perspective than anything else. 23 ; I think the more major concern is could the VCT, 24 the hydrogen and the nitrogen, if this VCT IcVel was low 25 enough, could this hydrogen and nitrogen get entrained into O ace FEDERAL. REPORTERS, INC. 202 347-37t#) N.ition m ide Cm erage N rt 116 (M6
4 25727.0 9 REE 1 the charging pump suction path and ultimately result in gas 2 binding of the charging pumps. I think that was a question 3 you had. 4 MR. EBERSOLE: That is what happened at Palo i 5 Verde. 6 MR. GIITTER: Yes. We asked the licensee about 7 this. And first -- well, I will get to that. 8 The NRC op center was notified at 1218 on 9 January 14, 1986. This event happened on November 3rd. 10 i The only way we found out about this event la that the 1 11 f resident inspector, investigating the previous loss of 12 instrument air event, went back through the shift 13 supervisor's log and saw that they had started up with O 14 l G these valve motor operators being inoperable. That is the
!) I' 15 a only way we found out about it.
f 15 The day after the resident found that out, the - licensee called this in to the op center and notified us. 17 l 18 Apparently what happened is the licensee looked at their 19 tech specs and because they did not consider these valves to be in an ECCS tlow path, didn't think there was any 20 l 21 problem from a tech spec of starting up. And I believe 22 they were planning to repair these valves once in the 23 process of starting up or once they had actually gone to 24 power. 25 MR. EBERSOLE: Why didn't they consider them in O c Acti-FlioriRAI. RiiPOR~i tius, INC. 202 H W m Nat onside rmenge nod 6(M6
25727.0 10 REE () I the ECCS set? One basis would be if they are closed, they 2 are not in the ECCS set. If they are open, they are. 3 MR. GIITTER: Right. 4 MR. EBERSOLE: Do they discriminate like that? 5 That is the mode or the position of the valve in order to 6 determine whether this is within or without the ECCS set? 7 MR. GIITTER: I can't tell you what the 8 licensee's logic was. I don't think that we agree with it. 9 MR. EBERSOLE: Certainly if they are open, then 10 the ECCS function is degraded? 11 MR. GIITTER: That is right. 12 i MR. EBERSOLE: So they are in fact in the ECCS l 13 set if they are shut, are they not? 14 MR. GIITTER: I would have to agree with that. l i 15 MR. EBERSOLE: I am asking in the general 16 ' context, do the licensees do that discriminatory 17 classification to determine whether a valve in a given 18 position is or is not in the ECCS set? 19 l MR. GIITTER: They have some means of making I 20 ! that determination, but I don't know what the logic is. I 21 MR. EllERSOLE : It sounds a little vague to me. 22 MR. GIITTER: That was the official response we 23 got. 24 MR. EllE RSOLE : We got thene things on every 25 plant. Are they all in the ECCS not? (1) ACli FliDiillAl. Rlil'Ol(Tl!RS, INC
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25727.0 11 REE I MR. GIITTER: I am sorry. 2 MR. ALLISON: I don't think we know the answer 3 to that today. We will have to -- 4 MR. EBERSOLE: I am talking about the generic 5 aspect. Carry on. 6 MR. GIITTER: Okay. They did repair the valve 7 motor operator when they went into cold shutdown. They 8! reviewed the design and determined that this -- at first, ! l l l I 9! when it was believed that the hammering problem may have ! I 10 1 been a contributor, they looked at other valves and I i 11 determined that hammering wasn't a problem in any other l 12 ; valve motor operators in the plant. I 13 ; They performed a test -- let me pull this up a l 14 ) little bit -- when they were in operation, they performed a 15 tent. They started the charging pumps and they left those 16 valves open and they sent an operator down thete to see how ! 17 i! long it would take him to go down and close these valves I I l 18 i manually. The first time they conducted tests, it required f 19 19.4 minutes for the operator from the time he left the n 20 j control room or did whatever he had to do to actually 21 closing the valves, it took 19.4 minutes. The licensee 22 concluded at that point in time that the VCT level was down 23 i at around 20 percent. And according to their analysin, 24 ' original analysis, at about the 20 percent IcVel, initially 25 you have a higher pressure hete and the suction is { . O ' Acti 1;iii) lina 1. Riii>on i1.as, INC
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1 l 1 25727.0 12 REE () 1 originally going to be taken from the VCT because you have 2 30 pounds cover gas pressure. But once the VCT level drops 3 to 20 percent, the pressure decays because the gas volume 4 is increasing. And at that point, the primary source of 5! suction is the RWST. So they concluded that at 19.4 i 6 minutes that it would take the operator to get down there, 7 that the VCT level would be at about 20 percent. 8 . Under those conditions, they were assuming the 1 9 VCT initial level was at 100 percent, which it was for this 1 10 particular event because of the failure to -- the design in 11 not isolating the miniflow lines. 12 ; They subsequently did two other tests. Both of i I 13 ! those tests, they sent two different operators down. In l O 14 l one case it took them 15 minutes. The other case took them 15 18 minutes. They concluded that there would be no gas
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) binding of the charging pumps sooner than that 18.5 minutes.
1 17 l Now, there is one question that comes to mind is, i in a real safety injection instance, is it likely, is it 18 l 19 h realistic that an operator could yet down there in that
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20 l amount of time, if they really did have a small break LOCA, l 21 l for example. Is it realistic that they could get down 22 thoto in that amount of time and that they would close 23 those isolation valves. 24 There is an enforcement conference ncheduled for l l 25 February 28. I believe that theno 19nues will be brought t I ( ACli Flii>iinAI. Ittii>on t ains, INC
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25727.0 13 REE 1 up at that point in time. 2 MR. EBERSOLE: The fact that it is two valves 3 there suggests that they needed redundancy and in fact it 4 is almost evident that they had to be considered as safety 5 grade elements of the design, just because there are two of 6 them. Were they in the tech specs, to be periodically 7 examined and tested. The safety grade list? 8 MR. GIITTER: I believe they are. l l 9 MR. EBERSOLE: You told me that the operator ! l l 10 ' didn't even know what classification these valves were in. I 11 MR. GIITTER: They looked at tech specs and they 12 l didn't see any problems. This is my understanding, from ; l l 13 j what the licensee told me, they didn't see any tech speen 14 h that addressed having these valves -- 4 : 15 MR. EllERSOLE : -- function an designed. And l4 f 16 they were not able to dincern from the tech specs or the l 17 physical needs that they were critical to the safety j 18 f function. 19 j MR. GIITTER: They made the determination that i 20 ; they could have thene valves inoperable, prior to ntartup. I 21 MR. EtlERSOLE: They did. 22 MR. GIITTER: They ntasted up with thone valves 23 inoperable. It may have been an overnight on their part. 24 MR. EllEltSOLE : What are you goinq to do about 25 that, an far as I am concerned, a groan minintetptotation 1 O Acti Fiti)l:nAI. Illii>oni tiRs, INC. 3e:w pm Nun. ae < mouc am n6 ua6
25727.0 14 REE 1 of the reliability of the llPC? 2 MR. KERR This was inoperable but closed? They 3 assumed that they were inoperable but closed? 4 MR. MICHELSON: Open. 5 MR. GIITTER: In order to start up, you need to 6 open these valves up. 7 MR. MICHELSON: You need to normally opetate off 8 the volume control tank. 9 MR. GIITTER: You need to ofen these valves up. 10 They manually opened the valves so that if the valves 11 , received a safety injection signal, because the motor i 12 I operators were inoperable, they would have been incapable 13 of isolating the VCT on a safety injection and you would l 14 f get suction from the VCT and the RWST. d 15 1 MR. REED: Whether or not there is a loophole in l 16 , the tech specs in saying these have to be operable. I 17 fthink there is a question of good judgment hete that comes 18 into play. I really think the licensee is ignoting the 19 safety-related aspects of this system and those valves in 20 particular. I would certainly, if I was the staff, putnue 21 how this decision got made. 22 MR. GIITTER: We are currently pursuinq that 23 right now. As I mentioned, there is an enforcement meeting 24 on the 28th of this month to discuss this. 25 MR. EBERSOLE: I hope you will takn that purnuit , Act!.Flini! Rat. RiiPonit:Rs, INC J 202147.17m Nanon* nte rm er age km ) % u m
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25727.0 15 REE 1 into regions more distant from just this one. If they are 2 that lax in interpreting the critical needs of the safety r 3 system, you may find it all over the place. This might not 4 be an isolated case. This sounds like a gross l 5 misinterpretation of a critical functional need. l 6 MR. ALLISON: I don't believe this is a typical i i 7 i kind of thing. j :. I 8 I MR. REED: I have a little concern here, as I !, I 9 , have watched resident inspectors function, I am surprised ! I ! 10 { the resident inspectot didn't catch this the very next )' I 11 ; morning and have his citation book going. I 12 MR. EBERSOLE: That what I was about to say. j 13 ! MR. REED: Normally they review the logs the 1 O 14 I very first thing in the morning, looking at all the I 15 , incidents through the preceding day and ovet the nightfall. I : 16 g This thing just stands out. l 17 MR. GIITTER: The resident was the one that did I 18 ! notice it. l l l 19 MR. REED: But it took him how many days? l l i 20 ' MR. GIITTER: I quess he was investiqating the 21 lons of instrument air. I really don't know what the 22 situation was there. But it ir, ponsible that he could have 23 l been tied up with the loss of insttument air at both units. 24 f MR. EllERSO!E They knew that they wero 25 inoperable? l i ! Aci!.l;Iil)liRAI. l(l!POR II:R% INC.
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l 25727.0 16 REE 1 MR. GIITTER: That is my understanding. 2 MR. MOELLER: Is it part of the resident 3 inspector's duty to go each morning and look at the log? I 4 mean, is there a system of fines for a resident inspector? 5 MR. ALLISON: Looking at the logs every morning 6 is part of their job. 7 MR. MOELLER: Does anyone ask, you know, how he j 8 missed it or she missed it? 1 \ 9 MR. ALLISON: We will ask that and find out. l 10 MR. EBERSOLE: Are the operators commonly asked 11 about the nature of need of these v.alves in an operational 12 ! examination? i l l 13 l MR. GIITTER: I can't annwet that question. I l l 14 can find out for you. ! l! 15 f MR. EBERSOLE: Why don't we do that as a generic l 16 problem. We would like to know whether the operatorn know 17 ! whether or not the nystem -- I don't mean running down to ( 18 the basement to tix it. Whether in the automatic mode 19 these are critical to the llPSI pumpn. I 20 In that all of that? 21 MR. GIITTER: That in it. 22 MR. KERR In there a 100 percent cc:tainty that I 23 ' with those valven open thin nyntem would not work? 24 , MR. EllE RSOf.E I tlather thete in. 25 MI . KERR I haven't heatd anybody nay that. O Acti-FiioliRA1 It1:eoRI1 Hs,1NU l 202 m.rm min
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25727.0 17 REE. - 1 MR. EBERSOLE: You remember the case of -- 2 MR. KERR I am not talking about a case of 3 enything. I am asking a question as to whether there is 4 100 percent certainty. 5 MR. EBERSOLE: You are on the wrong side of the 6 coin. The certainty needs to be that you know that they ; i 7 will, not that you know that they don't. 8 MR. KERR: I want to know, is there a 100 ! 9 percent certainty that they won't work? 10 MR. EBERSOLE: I want to know that they will ! 11 j work. I am not willing to deal with that whether they l l I 12 ! would not. l l 13 MR. KERR I was asking the staff. O 14 , MR. MICHELSON: It in a good question because i
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15 you have to know the relevant pressure from the watet 16 ; storage tank versus the -- 17 $ MR. KERR It isn't clear to me at this point. t 1 18 3 MR. MICilE LSON : On the banin of what in there, 1 19 you cen't tell whethcr the tank -- 20 l MR. ALLISON: The licennee claims that they 21 f would. 22 MR. MICilELSON: Yes. They have done the tent. 23 MR. EHERSOLE: llan he t e n t o<l it? i ! 24 ! MR. ALLISON: Not to our knowledge. l 25 i MR. M I CilE LSON : You got to tent it being altcady ' l ! O i ACli Flii)iinal. l(1;l'on il Rs. INC.
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25727.0 18 REE 1 connected to the refueling storage tank as well. I doubt 2 he did it -- 3 MR. EBERSOLE: You are saying that it is maybe 4 sufficlent? 5 MR. KERR I don't know. The reason I asked the 6 question is because I want to -- 7 MR. EBERSOLE: lia s the operator -- 8 MR. GIITTER: The only explanation they gave me 9 was the one I provided to you where they determined how i 10 l long it would take for an operator to get down and manually 11 close those valves. I 12 i MR. EBERSOLE: The inference you can draw from 13 that is they better get with it because it in going to quit 14 l working. r ; 15 ;i MR. REED: I don't think the tout of how long it 16 takes the operator to do it has any relevance to the innue. 17 The operator could fall in the stairway and kill himself 18 while en route. Ile re is a design. If they want to do a 19 test, they look at the relative headn of the two tanks in 20 their position and location and over prennute of the ganes. 21 But after all, that in pretty much a standard Westinghouse 22 denign for many, many years. I thought evetyone had been 23 trained and knew the importance of thene valven and their 24 functioning and why they are there. 25 MR. MilERSOIE: I think it goen wit.hout saying, O , Acti l'ii.oiinA1. Ri.Pon ifins, INr 202W Um N.shoneM merne e Hr u,h
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l l ! l 25727.0 19 ; REE I 1 I 1 we will put this on the list to be discussed at the full i 2 committee meeting because of its generic inferences. ! l 3 Is that all there was to that? ! l 4 MR. GIITTER: That is all I had. i l 5 , MR. EBERSOLE: Any further comments? i l 6 Let's move to the second one. i ! 7 MS. WEGNER: On January 7, 1986, Carolina Power I 8i & Light Brunswick phoned in the following 5072 report. i I 9 j Test resuits from Wyle Laboratory on 11 SRV, six SRV failed < 10 l to lift at pressures as high as 200 psi above setpoints. j 11 Two SRV were satisfactory and thtee SRVs lifted outside the l i i 12 tolerance band. , I l 13 l These test results wete significantly different O 14 j 1 from other 1iconsees' results reported to NRC recently. y i 15 i The NRC's concern was heightened because of the number of l l 16 valves affected and the extent to which the disc appeared l 17 l to be stuck. ; i i 18 , As a bit of background, the two-stage Tarqet : 1 19 ! Rock safety relief valves were the solution to the ptoblems I 20 l that had been experienced with t.se three-stage valves. I 21 { Two-stage valves exhibited home setpoint drift during 22 testing an became a major concern in July of 1982 when 11 23 I of the flatch 1 SRVs, of which there are 11, failed to open 24 at pressures well above their netpoints following a scram 25 in isolation. Subsequently, three of the valves opened at ! O Atti 171!!)l!RAI IlitPOR'llius. INC
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25727.0 20 REE (( ) 1 1180 psig and the other valves were manually opened later 2 in the transient. 3 The owners group that was formed after this 4 event, the General Electric Company and the Target Rock 5 Company, investigated the causes of the setpoint drift and 6 developed recommendations for solutions to the problem. 7 This is a closeup of the two-stage topworks. 8 (Slide.) 9 MS. WEGNER: This shows the labyrinth scale of 10 . the disc seating area. l 11 The principal causes were determined to be 12 galling in the labyrinth seal area, and corrosion-induced 13 seat-to-disc bonding in this area. 14 MP. MICilE LSON : Which area did you point to? l 15 MS. WEGNER: Steam side of the disc. The seat I 16 ! area here. l MR. EBERSOLE: You said corrosion induced 17 l 18 l bonding? l 19 MS. WEGNER: Yes. l l 20 The recommended solutions were an enhanced 21 l maintenance program and a replacement disc of a material 22 whose oxide film would be less likely to bond to the oxide 23 film of the disc. The seat, sorry. Both the seat and the 24 I disc were of a stellite 6 or 614 25 MR. EBERSOLE: Ilow long a perlod had it been O ACli l?I!!)I!RAI, l(!!POR IliRS, INC 202 347-17m Nahonside Cmcrage km 116 %16
c. 25727.0 21 REE 1 before they were unseated? Ilow long does it take for this 2 corrosion and bonding method to stick the valves? 3 MS. WEGNER: As little as three weeks or it has 4 been observed. 5 MR. EBERSOLE: Normally these things don't get 6 exercised except once every two months. 7 MS. WEGNER: The flatch valves had gone the ! 8 entire cycle without being exercised. ! 9 l MR. EBERSOLE: What sort of bonding does that I 10 i bring about? 11 MS. WEGNER: All 11 valves failed to operate i 12 until three valves lifted at 1180 psig. Since then we have gg 13 { seen some valves, mostly on the test stand, actuate very j %.) l t 14 > much higher than that. Or not actuate at all. i ! 15 , MR. REED: Those valves are valves that are i 16 fbrought in from the field. They haven't been exercised but ; i 17 ! or are they reassembled valves in the test facility? And 18 don't actuate? 19 MS. WEGNER: Okay. The topworks are taken off 20 of the -- off of the main and sent to Wyle Labs. I don't 21 l know whether any other facility can do the test or not. I 22 i That is the only one I am familiar with. The topworks are 1 23 not disassembled until they get to Wyle. The first test -- 1 24 I am getting ahead of myself here, but the first test that 25 they were doing to determine the 'ietpoint was when the O Aci! FliolinAI. RiiPonTiins, INC. 202-347 3700 % tion wide rmerage mWA14tM6
25727.0 22 REE 1 valves came in, they had a body there at Wyle. They took 2 the topworks, mounted it to that body, heated it up until 3 it was in thermal equilibrium and simulated reactor 4 conditions. They did a -- a long time ago they used to do 5 a full-flow steam test on it. 6 MR. REED: They did an as-found-from-the-field 7 test. 8 MS. WEGNER: I am not talking about the t 9 Brunswick test. This is the previous test. There has been 10 some changes which create some problems. I am getting i 11 ahead of myself there. 12 MR. REED: What I want to know is, how did they 13 1 do the Brunswick test where the relieve-it point was much O- 14 i elevated? 15 MS. WEGNER: The Brunswick test and the current 16 t test are done, were done first by backing the stem off of 17 i the disc so the disc would be free-floating, pressurizing 18 under the seat here with nitrogen to determine how much the 19 disc may or may not be stuck. A free-floating disc 20 l shouldn't take any more than 5 pounds of nitrogen to lift 21 it. 22 The facility limit at Wyle was around 200 pai. 23 Six of the Brunswick valves did not lift within the 24 facility limit. They shut down the test. 25 f MR. REED: That is the point I want to key in on. O ACli-171iDiiR AI, Illil'ORTliRS, INC 202. W. Um Nationwide Cmcrage *m Dr> Mm
25727.0 23 REE () 1 These are what I call internal pilot-operated relief valves. 2 I don't have much use for those kind of valves unless the 3 environment in which they work has been studied. This l 4 environment in which these valves work is probably an , 5 environment of hydrogen and oxygen and steam in a 6 condensing recirculating mode. And stripping of the oxygen i 7 and hydrogen is taking place up in those cap areas, I would i 3 8 . expect. Therefore, when you talk about corrosion bonding, You have high l 9 l I think you have a very nice set-up for it. 10 ; oxygen and hydrogen and you are probably going to get all { 11 ) kinds of corrosion and friction factor changes like crazy. 4 9 l 12 j I am not so sure that this is a good application of a valve.
, 13 i MR. EBERSOLE: It is not apparent what the main I l !,
14 1 motivating force is for opening the main bonnet. 15 l MS. WEGNER: This disc, the pressure under this 16 disc, when it exceeds the set pressure, is determined by l 2 i 17 l the spring and bleed off the pressure here. And that i 18 : allows this to compress its spring and open. 1 i 19 l MR. EBERSOLE: So there is an opening factor to i 20 ? the main steam pressure on the big piston. The steam i 21 l pressure applied to the large piston against the spring. I 22 l You simply don't show that in the illustration there. 23 MS. WEGNER: It is applied to the pilot. 24 MR. ALLISON: The question in, what lifts the 25 main disc? j l Acti171!Di!RAl. IlliPOR il:RS, INC N. mon.iae < merage xm w, mu, 20:wxm
25727.0 24 REE ( l MR. EBERSOLE: Yes. 2 MS. WEGNER: Frank, can you help me on this? As 3 I understand it, when lifting the pilot takes the pressure 4 of f here and allows the main -- 5 MR. EBERSOLE: It is just through shaft leakage? i 6 MR. CHURNEY: Frank Churney from NRR. It is not 7 shown on this particular slide. There is a tinge on there I 8 i called the plain piston. It is about the center of the l 9l picture. When you open the pilot, you create a delta P. 10 l MR. EBERSOLE: Where is the supply side of the 11 k delta P? 1 12 ]J MR. CHURNEY: Maybe the picture doesn't show it ' 13 k real well. There is a supply side to this. CE) 1 14 j MR. EBERSOLE: Okay. 15 MR. CilURNEY: You bleed it off and get enough of f i 16 i a delta P so it pops off. 17 )? MS. WEGNER: This maintenance program that I was 18 talking about has been in effect for at least one fuel l l 19 l cycle at most of the plants. Since the issuance of l 20 ! information notice 82-83 in December of 1983, the testing i 21 results that we have seen for a typical plant show a small I I 22 l number of valves within the tech spec limits which are set . I 23 ; pressure plus or minus 1 percent. The majority ate within I
' I 24 a set pressure plus or minus 5 percent. And most plants l l 25 .
would have a single valve stuck. Some maybe more than one, t I I Acti.I;1iolinAI. Riit'ouritRs. INC 202-347 lho %Ininmide rmerage W 116 f 44
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25727.0 25 REE I () 1 MR. REED: So enhanced maintenance is not the 2 answer to the question. 3 ' MS. WEGNER: It is not the entire answer. 4 MR. REED: I will go right back to say the 1 5 problem is the environment in which the valves are being ! 6 j asked to function and the nature of the valve. I 7 MS. WEGNER: It is beginning to look like that 8l Very much. As the data base grew, it was apparent that the { ! 9 ! setpoint drift hadn't gone away. The director of division l l 10 l of licensing, NRC NRR, sent a letter to the owners group in i 11 March of '85 recognizing the benefits of the enhanced g maintenance program but concluding that maintenance alone 12 l 13 was inadequate to solve the problem. (:) 14 The next step taken by the owners group was the 15 selection of a replacement material for the pilot disc 16 lC which would have less tendency to develop this oxide bond l l 17 k with the seat. The material which they eventually selected P f 18 ? was a precipitation-hardenable stainless steel disc. The i 4 19 l replacement disc had recently become available and had been I i i 20 ! installed in 50 percent of the Hatch 1 and the Brunswick l 21 l valves which were tested at Wyle towards the end of 1985. 22 I And the beginning of 1986. 23 i MR. MICHELSON: Wasn't this the same problem i 24 that they had with the three-stage valve and the sticking i 25 problem? f i (:)
.I ACI!.RiDiiRAl. REPORUiRS, INC.
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1 l I l I 25727.0 26 REE l () 1 MS. WEGNER: No, sir. The three-stage valves 2 popped open and stayed open. l l MR. MICHELSON: 3 They had a problem also of 4 gluing shut. 5 MS. WEGNER: I really don't recall that. I know 6l that has been a problem with the electromatics but not 7 these. 8 MR. MICHELSON: To fix them from going from 9 three-stage to two-stage was just to get away from the 10 l sticking open situation? 11 MS. WEGNER: That is right. 12 MR. MICHELSON: Now they got one that sticks g- 13 closed. O 14 MR. REED: I think designers traditionally try l n l 15 $ to look for different materials but if you haven't changed h 16 ltheenvironment, is anyone looking at plugging the 1 17 ! capillary cc.aection to the pilot and putting a loop seal l 18 g in that, have an external line come out on the pilot and i 19 put a loop seal in there to prevent the oxygen / hydrogen i 20 I concentration phenomena? 21 MS. WEGNER: No. I have heard mention of 22 I putting a pressure switch in there so that the valves would 23 b open at a certain pressure, no matter what. But that, too, 1 24 is talk. ! 25 MR. MICHELSON: The problem is that I think when G V ace-FEDERAL. REPORTERS, INC 202 147 379) Nationwide rmerage km ) w ms6
25727.0 27 REE 1 you open this valve, you will blow your loop seal out 2 because it vents at the bottom of the pilot there. 3 MR. REED: It certainly will refill. As long as 4 the valve-functions properly and seal the hydrogen and 5 oxygen concentration, maybe it will work. I am just making 6 a suggestion. Here is another one of these things that we 7 have talked about. We talked about them among PWR and the 8 problem of internal operating va;ves. 9 MR. MICHELSON: Well, if we knew what to swing 3 10 on, we would. 11 MR. EBERSOLE: These valves are typical of 12 boilers. , l i 13 MS. WEGNER: Yes. Quite an enlarged number of 14 the boilers have the Target Rock's. The PWR relief valves, f 15 the majority of them are just spring lift valves that I ( 16 have -- 17 MR. EBERSOLE: They are not -- well, but you 18 have electric -- 19 MS. WEGNER: I am talking about the main steam i 20 safety. ! 21 MR. EBERSOLE: I am talking about second -- 22 MR. REED: She is saying the majority are spring
- 23 action.
l l- 24 MR. CHURNEY: There is one PWR that has 25 pilot-operated safety valves on the primary side. That is , (
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202 347 3700 Nationmide Coverage MD-33MM6 i
25727.0 28 REE [) v 1 Beaver Valley 1. 2 MR. REED: Did you say secondary? 3 MR. CHURNEY: On the primary side. They aren't 4 exactly like this design. My understanding is they are 5 also a two-stage valve. They are more like the valves that 6 are used in the Navy. 7 MR. EBERSOLE: They have to face the problem of 8 corrosion? 9 MR. REED: Well, you got your answer. 10 MR. CHURNEY: They haven't had any real problem 11 I with theirs. 12 MR. CHURNEY: A big U-shaped arrangement of the g ( l 13 l pipe itself. 14 h MR. REED: You can just do it on the pilot. 15 MR. CHURNEY: It could be a little tricky on i 16 these because they are very small valves. They are right 17 '! on very short rises on the steam lines. 18 l MR. MICHELSON: I think they clear the loop seal i 19 ! also when they actuate. f. 20 MR. EBERSOLE: The valves on the PWR that you 21 state are there, those are not the safety valves, are they? 22 MR. CHURNEY: On Beaver Valley 1, there are 23 three safety valves on the pressurizer. Not exactly this 24 design, but they are pilot-operated. 25 MR. EBERSOLE: And they are in a borated (03 ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Coverage Nn1% C6
25727.0 29 REE ,n k) 1 environment with loop seals? 2 MR. CHURNEY: Yes. 3 MR. REED: I believe that is a Westinghouse 4 design? 5 MR. CHURNEY: It is the only one in the country I 6 with those valves. 7 MR. REED: Did you ever ask them why they went 8 to loop seals on pilot-operated relief valves, which is 9 also on another new plant that we just reviewed here about i 10 l six months ago. j 11 g MR. CHURNEY: Well, they have gone -- that is l l i true, Westinghouse has gone to pilot-operated PORV on some 12 ls l I r~ 13 l of the new plant with -- ! k )) m ) { 14 P MR. REED: With loop seals. l 15 k MR. CHURNEY: I haven't checked on that. ; j i 16 q MR. REED: We had another licensee come through l
) I 17 !.~
here four months ago, i i 18 ! MR. CHURNEY: I would imagine it is for the i 19 reason you state, to prevent against the boration problem. 20 MR. EBERSOLE: On the secondary' sides of the PWR, i 21 l they have safety valves and then they have to have 22 electromatics to intercept. What do they use for the pilot 23 or manual remote operation of the electromatics? 24 MS. WEGNER: Did you say BWR? 25 MR. EBERSOLE: PWR. The ones which are manual
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ACE-FEDERAL REPORTERS, INC. 202 347-3700 Nattonwide Coverage WWL1EfM6
l l 25727.0 30 REE l
) 1 remote operated.
1 2 MS. WEGNER: PORVs? ' 3 MR. CHURNEY: They are large PORVs. A lot of 4 them are large air-operated type. 5 MR. EBERSOLE: They are not required to be 6 safety valves. 7 MS, WEGNER: A lot of them use their safeties 8 for that. 9 MR. CHURNEY: They have large relief valves on 10 them. 11 MR. EBERSOLE: But they are just straight 12 air-operated. They are not given the prerogative of being s 13 ! safety valves? s 14 li MR. CHURNEY: Yes. 1 15 ] !: MR. EBERSOLE: Yet they are probably better than 16 l this. E 17 j' MR. CHURNEY: The thing is, it would be very 8 18 l difficult to change these out, for example, and put the i 19 large valves on, that are on the BWR 5 and 6s which are l, 20 more like the big spring safety valves. These are about 21 one-third the size of those. It would be a very difficult 22 thing to change these out. These are mostly on the BWR 4s 23 ! and a couple of the 5s. 24 MR. EBERSOLE: I guess they don't test these l' 25 things every X days because they will stick open. (D \.) ACE-FEDERAL REPORTERS, INC. 202 347-3R0 Nationwide Coverage 810-3 h (M6
25727.0 31 REE (- y ,)
\ 1 MS. WEGNER: After the Hatch incident, there was 2 a requirement for Hatch to open the valves. I believe that 3 was their finding. They did stick open.
4 MR. EBERSOLE: That is a stiff penalty for them 1 5l l to fix the valves. l 6 MR. CHURNEY: As she was saying earlier, some of i ( 1 7 [$ these things start to stick in a couple weeks. So in terms !
! i 8 of finding the frequency that would really be from a !
il i 9l cost-benefit point of view advantageous, it is not really l l 10 l obvious that it really buys you that much. ! 11 I MR. EBERSOLE: You buy some trouble when you 12 , stick them open. j b 13 MR. REED: I could make a suggestion. We did
' l 14 ?' this at Yankee Rowe 25 years ago when we were studying why I
15 !. pilot-operated relief valves didn't work. You could take f 16 j an external thermocouple, put it on that solenoid outside
?
17 the body of the operator, pilot operator, and follow the : il l 18 0 temperature reduction in time of the external casting of ! 19 !* the pilot. That would begin to tell you when you have ! I 20 l saturated the space with hydrogen and oxygen and you have l 21 i stopped the condensate flow and the steam flow up in there, 22 lI and it will tell you how long it takes to get the sticking l 23 I condition. i 24 [ In other words, you have got pure hydrogen and I 25 j oxygen up there. I expect it doesn't take too long. It is ! h (3
%-) I ACE-FEDERAL REPORTERS, INC.
I 202-347-3700 Nationwide Coserage M
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25727.0 32 REE-
- {( ) 1 amazing how low those temperatures will get. I saw the 2 temperature get hot on Yankee Rowe to 280 degrees 3 Fahrenheit. That tells you something is happening.
4 MR. CHURNEY: That is certainly one variable. 5 On the other hand, most of the test data that we see has a 6 whole spectrum of variation. These thingc don't just get 7 to one temperature and all of a sudden stick. There is 8 something else going on there, too. 9 MR. REED: If you have a microleak, you will not 10 get complete strip out. You will continue to keep 11 condensation and keep the temperature up and you won't get 12 solid -- 13 MR. CHURNEY: Those that do leak, and there 14 seems to be fewer and fewer of them as we go along in time, 15 they don't stick as bad. 16 MR. REED: Hey, you got your answer. All the 17 answers are right in what you just talked about here today. j 18 All you got to do is stamp it with a conclusion. . 19 MR. CHURNEY: I think the owner group solution, 20 if it turns out that changing the material to one that 21 won't bind, that sounds . like a pretty good option. It is a 22 fairly cheap option. c 23 MR. EBERSOLE: We could use a buffer gas -- 24 MS. WEGNER: That is one of the gases, the 25 particular element in the environment that is giving them i O ACE-FEDERAL REPORTERS, INC, 202-347-3700 Nationmide Coserage 8m3M4M6 i
25727.0 33 REE \ _.) 1 trouble. 2 This is a detail of the nitrogen lift pressures 3 of each of the valves at the Brunswick plant as per steam 4 line. It is interesting to note that the C steam line had t 5E the best track record. It is even more interesting in view 6 of the fact that the C steam line was isolated for six 7 weeks prior to the outage and was opened approximately a 8 { week before the outage. Also mentioned was a minor water I , i 9 chemistry transient during that period of time. I 10 MR. MICHELSON: I am confused. 11 y MS. WEGNER: It may not be related. 12 MR. MICHELSON: Isn't the safety valve on the
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13 reactor side of the steam line isolation valves? So they x-l 14 i are all pressurized all the time. The fact that you aren't
, i 15 4 drawing steam through the C line might be interesting or l 1'
t 16 j significant. That is all this is saying. That line was { 2 , 17 l J inactive. ! 18 0 MS. WEGNER: The line was isolated -- 3 19 MR. REED: Was there leakage through the valve n 20 $ to the main disc on the C line? t MS. WEGNER: No. None of these valves leaked. 21 l . 22 Or5 the vessel. t 23 I MR. REED: No microleakage or anything? 24 I MS. WEGNER: There was no leakage reported. 25 l ! They have acoustic detectors as well as temperature ACE FEDERAL REPORTERS, INC. 202-347-3700 Nanonwide Cmerage N W A 336 (M6
25727.0 34 REE I fx (,) 1 detectors on the tailpipes to determine leakage. I am not 2 sure of their sensitivity. Temperature rise is what used 3 to be used at Browns Ferry to determine when we were, when 4 the valves were leaking. They were fairly sensitive. 5 MR. MICHELSON: How does the tailpipe 6 temperature detect -- 7 MR. REED: I think I have a theory. If you do 8 not pass the steam through that C line and if you have the 9 : intimately connected safety valve, no steam is going 10 through there, then stochiometric quantities are not going 11 , through there. You are not feeding the cells, so to speak. L 12 1 You have stopped the flow. And that may relate to it. l 13 ! MR. MICHELSON: You can also argue that the
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14 hydrogen and oxygen is rising up and collecting in that 15 j area. It is a noncondensable gas. Therefore it is all i hydrogen and oxygen up there. 16 l i 17 5 MR. EBERSOLE: That is the other side of the 4 18 i coin. I 19 MR. MICHELSON: So it also depends on how much h 20 l that main poppet is leaking. 21 MR. EBERSOLE: Why do you get sticking on the 22 pilot seat without the counterpart and much more serious 23 sticking of the main seat?
- 24 MS. WEGNER
- The main disc? It doesn't see the l
25 steam. i m 1 l ACE-FEDERAL REPORTERS, INC. l 202-347-3700 Nationwide Coverage Wxt31MM4
25727.0 35 REE (% (_) 1 MR. EBERSOLE: Sure it does. The back side does. 2 Is it that it is a better seal and doesn't leak? 3 MR. MICHELSON: It is that this is a Swiss watch. 4 The pilot of the Swiss watch and the other is a large gear, 5 so to speak. l 6l MR. EBERSOLE: That brings up the old Salem 7 problem of what are the force furctions and what is the 8 margin to do. This is -- do you remember Salem, the margin 9 of force to make the circuit breakers open? Maybe the , 10 , margins of force to execute valve clearance are not enough j l l 11 { for the pilot. I h ! 12 g MS. WEGNER: This is a very thin film that forms ' l g' 13 j here. However, the area in which it forms is very small, ( t 14 too. 15 l MR. EBERSOLE: What is the forcing function that
$ I 16 makes it open? Is it big enough?
17 ! MS. WEGNER: For one thing, this side of the 18 g disc doesn't see the -- a 19 ! MR. EBERSOLE: I am talking about the pilot l 20 [ function over here. When I have an overpressure transient, i 21 ( what are the forces versus the static friction and glding 22 forces that I have to overcome? What is this force balance 23 in this design and what are the margins to overcome this 24 sticking? 25 It is analogous to the old UV trip on the Salem, I ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Coserage m 33MM6
25727.0 36 REE (v) I the ATWSS case, where they had to analyze what force do I 2 need in this case to make those UV trips work versus how 3 much spring tension I had to do that. It is an analysis 4 here of one of the forcing functions to make the valve 5I clear versus the sticking forces I do have to clear. 6 Evidently there has been no analysis of the design in this 7 context. 8l 'l MS. WEGNER: Frank, I don't know about that l 9 i exact question. 10 MR. CHURNEY: I think the way that the main disc 11 k is opened, that what you are talking about there is that
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12 there is much more margin -- I don't think that anyone has i 13 ; done anything real analytical with regard to this. I think 14 ln the operating experience, though, in the way this bonding 15 forms, I think it is fairly intuitively obvious that what 1 ' 16 you are saying is correct. 17 ! MR. EBERSOLE: It is almost, it is a mechanical 18 f version of the Salem case. I 19 Bn MR. MICHELSON: Yes. l 20 !! MS. WEGNER: I know that Salem had a problem. I 21 But I am not familiar with the -- 22 MR. EBERSOLE: I am talking about the breakers. I 23 l But it is the same old thing. I got to move something. 24 With what force do I do it? 25 MR. CHURNEY: When it doesn't corrode, it works b wJ f ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Coverage #n3hfM6
25727.0 37 REE (~) (_j 1 fine. 2 MR. EBERSOLE: But you have to make an allowance 3 for corrosion? 4 MR. CHURNEY: That is what we are trying to get 5 rid of. In order to do that, I think there would be much 6 more substantial design changes. 7 MR. EBERSOLE: I am sure. 8 MR. CHURNEY: That is why I said, changing the a 9i material, assuming that works, is probably the cheapest way. 10 MR. EBERSOLE: Maybe you just need bigger 11 l pistons. i 12 l MR. REED: I tell you what bothers me: We talk ; N ! 13 l about root causes, what really causes valves to malfunction, i 14 l and we try to set up committees to search for root causes.
$ l 15 l And we got root causes all around us. But our recognition ti 16 E is pretty damn poor. I have had recognition of this 0
17 problem for 25 years. And I have been talking about it for f 18 l 25 years. Nobody knocks me down. Any time one of you guys l 19 want to knock me down about the environment of hydrogen and 20 ; oxygen and boron and what it does to internal operator 21 I relief valves, I would like to be knocked down because I 22 will get off this kick. 23 MR. EBERSOLE: They vill wait until they get a 24 sticking of so many of these. It is just like the list we 25 just had on check valves. pJ ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Cmerage ML3%fM6
25727.0 38 REE ( l MR. REED: Are we lacking perception to get to 2 root causes? 3 MR. EBERSOLE: We are not sensitive to anything 4 unless it produces a drastic physical consequence. 5 MR. REED: I think we are sensitive but we lack ; 6 the perception to get the real root causes. I could 7 . comment on Three Mile Island with respect to that. l 8 i MR. MICHELSON: I think one of the difficulties I 9l might be, I don't believe the agency has any activities on 10 ! these types of valves like we do on some of the other ! 11 a valves -- no research activities. I think you are i W l 1 ' 12 depending upon the industry to be doing this and so what we + 4 gS 13 !i need to inquire into is what is the industry really doing \ / I! 14 h about it? And what is the status of the situation? I 15 O don't know if you people have gotten to that stage yet or li 16 l o not. 17 i) r!R. ALLISON: I think what Frank was just a E 18 telling you is what they are doing. They are changing 19 y material. The previous work they have done on the
\ u 20 i labyrinth seals seems to have solved that problem and the
/
21 l n material change will solve the problem and that will be a 22 good solut1on. 23 b MR. CHURNEY: The quickest test program they il 24 could come up with was the test on the reactors. 2 5 ' II MR. ALLISON: It is pretty close to a root cause ACE FEDERAL REPORTERS, INC. l 202 347-37'o Nationwide Coserage 8t N L.13MM6
25727.0 39 REE s/ 1 in that you have the seat and the disc are the same 2 material and so what we think is happening is that the 3 oxide films are the same, they are tight films and so they 4 bond together. 5 MR. EBERSOLE: I didn't hear you say they were 6 the same material. 7 MR. CHURNEY: They are not exactly the same. 8 MR. EBERSOLE: Isn't that a no-no in any design? 9 MR. CHURNEY: I don't think so. l 10 MR. ALLISON: A lot of steam system valves have I 11 stellite discs. 12 MR. EBERSOLE: Okay. 13 ; MR. ALLISON: So if the new material is thought (Jg %- t l 14 4 to have a weak fragile oxide layer, it is different. And < that is the theory anyway. 15 l ; f 9 16 ', MR. REED: To me the root cause is the I 17 1 environment, the oxygen and hydrogen, and you choose to 18 build the dam down stream of a different material. I 19 I MR. ALLISON: Well, if the material works, you 20 j don't have a root cause. 21 MR. REED: You mean, you don't have an event. 22 MR. ALLISOt: Right. 23 MR. REED: But you just said the root cause 24 which is hydrogen and oxygen. You really haven't done 25 anything about that. ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Coverage Mk31MM6
25727.0 40 REE i MR. EBERSOLE: We got to move on. [Q 1 We are going 2 to run out of time. 3 MR. ALLISON: One last thing, these are the set 4 pressures plus the nitrogen pressures. These are the steam 5 pressure plus the nitrogen pressure for the Brunswick 6 valves. And a recent test on the Hatch 1 valves that had 7 , the initial incident can show you basically what BWR water I 8 chemistry control can do towards solving part of the l 1 9 l problem. i 10 MR. MICHELSON: Have any of the valves failed to 11 ; open at all at something approaching what the code would 12 I say they must open up, the safety valves? , l MR. ALLISON: I think the events we are talking (, 13 l 14 I about fit what you are asking. l 15 MS. WEGNER: The original Hatch incident had 16 eight valves that were supposed to open from 1080 to 1100 17 $ and failed to open, period. k 18 i MR. MICHELSON: But as safety valves, they 19 didn't have to open in that range, did they? 20 MR. CHURNEY: According to the code analysis 21 l that was done for Hatch, what she said is correct. They I 22 opened quite a bit later. 23 h MR. MICHELSON: Did they have to open at those 24 l levels? 25 MR. CHURNEY: It just so happened that the l O ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Coserage Mn33MM6
25727.0 41 REE ( l particular transient involved was a very mild transient, 2 very slow. So we have never had an event where the code 3 pressure limit has been exceeded. 4 MR. MICHELSON: That is the question I am asking. 5 MR. SBERSOLE: Do the ASME code committees on 6 valve, are they aware of the evolution of these valves in 71 actual practice? 8 MR. CHURNEY: WE haven't reconsidered that. We 9 have substantially increased the initial qualifications of 10 l the last couple of years for all the safety valves, k 11 g MR. MICHELSON: Does the code address this 12 particular design or is it simply the need for code safety 13 and certain requirements as to what they must meet? 14 MR. CHURNEY: We have not taken any action to e 15 )'; prohibit the use of pilot-operated safety valves. One I l 16 l thing that has come up. You have to realize the way the h 17 code is written. The code has to worry about light water f 18 reactors. One thing that has come up is the concern about l 19 ! how spring actuated safety valves on liquid. They work i 20 l lousy on liquid. If you are going to have any liquid 21 transients, these are much better. So there is a lot of 22 tradeoffs. 23 MR. MICHELSON: So the committees are not yet 24 real worried but they are watching the situation? Is that 25 a correct appraisal? ACE-FEDERAL REPORTERS, INC. 202-347 3700 Nation aide Coverage lu ML 3 EtM6
25727.0 42 REE () 1 MR. CHURNEY: They are watching the situation 2 cautiously. As a matter of fact, there used to be a 3 penalty on the use of pilot valves. You had to put more of 4 them on the system. Recently the penalty was removed but 5 the initial qualification requirements were substantially 6 increased. 7 MR. EBERSOLE: Any further comments on -- what 8 ; is the committee's feeling about taking this matter to the I 9 full committee? Would anyone say no? l, 10 ! MR. REED: I agree with you except that I think 11 l we need to keep all kinds of pressure on to get the 12 f fragmented valve, this misapplied valve situation l
! l 13 straightened out. I think it is one of my burning issues
( i (-) ' and one I wasn't allowed to put on paper because I only 14 l 15 _ could have one. (Laughter.) 16 l 17 h MR. MICHELSON: You didn't choose to put it on 18 ; paper. i 19 ! MR. REED: It wasn't the number 1. l 20 ! MR. EBERSOLE: Let's go to the third item. i 21 MR. KEISSEL: I am Dick Keissel. I am with the 22 the Office of Inspection and Enforcement. I am here to 23 talk about a series of recent check valve failures that 24 occurred at Turkey Point units 3 and 4. These check valves 25 were in the steam supply system to the aux feed pumps and O ACE-FEDERAL REPORTERS, INC. 202 347-37(vi Nation *ide Cmerage WuuW#M6
25727.0 43 REE O) (_ 1 as you know, Turkey Point only has steam-driven aux feed 2 pumps at this time. 3 The safety significance of this problem is the 4l fact that it could have prevented the aux feed system from l 5 functioning properly at Turkey Point, but of a more 6 sweeping nature is that it points out again what happens to 7 check valves, and in this particular case, stop check f I 8 valves, when they are subjected to very low flow conditions. l 5 l 9! By way of history, in late November of 1985, l l l 10 i, Turkey Point became aware of the fact that one of the 12 ! 11 j check valves in question was inoperable. They then ran l 4 i 12 j radiographs of the other 11 and found similar failure to i 4 , 13 3 two more of the check valves. When they opened up the ! t
sl l ?
14 l valves, they found that all of the valves were experiencing H I 15 l signs of mechanical damage from low flow or what they j 16 attributed to be low flow conditions in the lines. i 17 t They replaced the stems -- the disc and the disc 1 l 18 4 guide in 10 of the 12 valves and went back into operation. U I 19 Then in early January, they discovered that, f 4 20 l again, one of the valves had failed. They x-rayed and this i 21 P time they found that a total of four valves out of the 12 22 j had failed. They failed in a similar manner. 23 In addition to making the systems inoperable, 24 'i the loose parts which were generated by this could also 25 ; have done damage to the turbine pumps. ACE.FriotiRAL Rl!PORTliRS, INC.
- fe 3t?.37m Nanon ue reverage p 1% w6
l 25727.0 44 REE () 1 Because of the generic nature of the problem, 2 that being that we have a potential misuse of this type of 3 valve in an information notice, 82 -- I am sorry, 86-09 was 4 issued, this also referred back to a previous information 5 notice, 82-26 which addressed a similar problem with the 6 steam exhaust valves on the HPSI and RPSI systems which 7 were being subjected to low flow conditions when the pumps 8 were run in a test mode. 9 MR. EBERSOLE: I think there is no question. We 10 want to take this to the full committee and present it as an immediate backup to the relevant problem of check valve 11 l l 12 a failures that we spent the whole morning on. This was l 13 again a case of reduced flow and the chattering and damage O 14 , to valves which caused five of them to fail concurrently. 1 15 ] This is just another case of the same sort of event that i 16 { ti happened at SONGS. a 17 E MR. MICHELSON: You just said the licensee i 18 l thought reduced flow was the problem. I haven't heard any 0 19 ! explanation of how reduced flow can cause a failure of this 20 type of valve. . I 21 l MR. EBERSOLE: I read low flow caused vibration 22 and chattering. 23 MR. MICHELSON: It is a little harder to see how
; 24 low flow damages this valve.
25 I MR. KEISSEL: The manufacturer of the valve () Acn-FEDERAL REPORTERS, INC. l . x.mmm % ue cmmu a ne, ra.s
25727.0 45
'REE
) I recommends that they not be used where they see flows of 2 less than 10 percent. These valves are used as isolation 3 valves around the main steam isolation valves to the pumps.
4 And they attribute the low flow condition to leak-by of the 5 isolation valve itself. 6 MR. EBERSOLE: Were you here this morning? 7 MR. KEISSEL: No, sir. 8l MR. EBERSOLE: The failure of the five valves 9! was attributed to a 10 or 15 percent flow reduction. i ! 10 ; MR. MICHELSON: These are stop checks. Look at
! l 11 the drawings.
12 . MR. EBERSOLE: These were deviant from the full (-)g 13 j design flow conditions just as this is. At least this is
%- i 14 q blamed as a causative factor that causes resident action ;
i 15 4, and damage to those valves which in a general context is { 16 F what you are telling me here. W t 17 ll MR. KEISSEL: Yes, sir. , i 18 l MR. EBERSOLE: Does the valve vendor state that i i i 19 ! he will not warrant his valves at flow rates less than X? 20 { MR. KEISSEL: I don't know about the statement l 21 that he will not warrant the valves below a certain amount. 22 j He recommends that they not, that this style valve not be 23 used with less than 10 percent flow. I 24 MR. EBERSOLE: So there are full requirements on l 25 l all valves like this? On those valves? You have full C) ACE-FEDERAL REPORTERS, INC. i 202-347 3700 Nath>n aide Coverag MXL14IM6
, - . . - _ ~
25727.0 46 REE 1 requirement, a recommendation rather than requirement, that 2 they not use these with flows in that range? 3 MR. KEISSEL: Right. The unusual thing here is, 4 of course, we are -- when the system was designed, it was 5 assumed there would be zero flow. And what we were 6 experiencing here was just leak-by of a closed gate valve. 7 MR. EBERSOLE: What is the consequence of 8 failure of these valves. 9 MR. KEISSEL: The system would be declared 10 inoperable -- well, if the valves failed in the open 11 position, there is a possibility that the system would not l 12 respond properly to a line break upstream of them. And you 13 couldn't, because you need to be able to pressurize with 14 , steam from the other plant. 15 Let me show you. I have a schematic of the aux i 16 l feed system. 17 MR. EBERSOLE: It would leave it as an open 18 i system then on aux feed? I 19 I, MR. KEISSEL: Yes, sir. Now you have three aux 20 feed pumps that share their steam source between the two 21 units. If there were a steam -- if there were a line break 22 , upstream, up in this area, failure of these valves then 23 prevent being able to service the pumps with steam from the 24 other unit. 25 MR. EBERSOLE: I see. So I lose the header O ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nation.ide com age 14 n 1m-6 u6
l 25727.0 47 REE
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(_) 1 pressure for steam supply to the aux pumps? 2 MR. KEISSEL: That is correct. 3 MR. EBERSOLE: On a single point basis. Just 4 the failure of one of those valves? 5 MR. KEISSEL: No. Both of them. 6 MR. EBERSOLE: Why is that? 7 MR. KEISSEL: Well, if one of these holds, then 8 a breakdown here would be isolated. 9 MR. EBERSOLE: I thought they were 10 cross-connected. 11 MR. KEISSEL: The cross-connect is through these I 12 headers here. So that your steam flow pattern would be 13 from one unit through the header and out. (~}
's_/ !-
14 MR. REED: I am having some difficulty l a 15 lk recognizing this valve that is called a stop check load 16 valve as a stop check. What it looks to me like is a not 17 I too well designed and is used as a stop valve. How tight l 18 ! is that little thing down there? It looks to me like it is 19 tight. 20 J Now, didn't you just say that the manufacturer 21 says that it shouldn't be used at flows under 10 percent 22 and didn't you say that its real application is to be open 23 or closed? Not to throttle? 24 MR. KEISSEL: That is correct. It is not a 25 throttling valve. ACE-FEDERAL REPORTERS, INC. i 202-347-3Km) Nation ide Coverage m a 6-3 h646 L
25727.0 48 REE ,. m (,) 1 MR. REED: Well, then it is a stop valve. 2 MR. KEISSEL: No, sir. It is a stop check. The 3 valve is in the normally open mode and, therefore, 4 functions as a check valve. 5 g MR. REED: It is normally open and -- well -- 6 MR. KEISSEL: And therefore, it will act as a 7 check valve, blocking flow above the disc but permitting 8 flow from below the disc. l 9 i MR. MICHELSON: The drawing is showing it in the i 10 locked / closed position, 11 . MR. REED: I see there is some way that this I, 12 ! assembly can move up and down on the lower part of the stem h 13 l that isn't obvious in the drawing. g')s \- b 14 l MR. MICHELSON: Well, you know how they work. 15 1) a I t is obvious, but you have got to crank it out. It is in 16 k a locked position. 17 s MR. REED: Do you see what I am talking about? I 18 i MR. KEISSEL: Yes, sir. If this stem is pulled 19 Ik back uD, then the disc is allowed to ride up on it. And it 20 pi is a bit confusing in that the conventional globe valve 21 would have the stem actually filling this piece down here, 22 giving you the color which would permit you to pull it off. I 23 I MR. REED: And a little looseness down there for I 24 a seat. I 25 MR. MICHELSON: This one slides up and down.
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U ACE-FEDERAL REPORTERS, INC. I 202 347 3700 Nation *ide Cos erage Nxk 33M6t6
25727.0 49 REE (') 1 The damage can only occur if the flow is so low that the 2 valve, the weight -- 3 MR. KEISSEL: It just sits here and simmers. 4 MR. MICilELSON: That will wear the seat out. 5 But that is an altogether different problem than having 6 damage to the thing from being in the open position and 7 banking against something. This thing here, as long as it 8 isn't fluttering on the seat -- well, what is happening is, l 9 it is breaking off that guide stem because it is fluttering 10 enough to -- 11 l MR. EBERSOLE: I fail to see why one of these 12 valves failing would deny steam supply to the aux feed _ 13 pumps. Did you not say that was the case? \',) 14 - MR. KEISSEL: No, sir. It provides a path by ,
) I 15 0 which steam can be lost. And it is not just one. They 3
16 to take out two. fwouldhave 17 h MR. EBERSOLE: I didn't see where the blocking 18 lI device was to prevent -- these are not commonly headored 19 I together. I see other check valves up there. Now, it I 20 takes more than one valve failure to deny steam pressure to f 21 the turbines of the aux feed pumps, right? 22 MR. KEISSEL: Yes, sir. 23 MR. EBERSOLE: It would take more than one valve 24 failure? 25 MR. KEISSEL: Yes. (^
\ .
ACE-FrDERAL REPORTERS, INC. 202-347-3700 Nationwide Coverage MH-3 % (M6
25727.0 50 REE e (s s l MR. EBERSOLE: So our problem would only be in 2 front of -- this is a safety problem if we lost more than 3 one? 4 MR. KEISSEL: Yes, sir. But we are talking 5 about a common mode. 6 MR. EBERSOLE: You lost one and you might have, 7 with a little less luck, lost them both? 8 MR. KEISSEL: That is correct. 9 MR. EBERSOLE: If I had lost two, would I have 10 lost my turbine pumps? 11 MR. KEISSEL: You could have lost your flow to 12 them assuming that you also then did not have your -- _ 13 MR. EBERSOLE: I believe you said at Turkey 14 Point, I must depend on these three turbine pumps. I don't 15 have any motor pumps? 16 MR. KEISSEL: That is correct. 17 l MR. EBERSOLE: So we have another mode of loss 18 l of inventory to turbine-driven pumps? i 19 MR. KEISSEL: Yes. 20 MR. MICHELSON: Where are the stop checks on
- 21 your drawing?
22 MR. KEISSEL: Stop checks are this column of 23 valves and this column of valves. (Indicating.) l 24 MR. MICHELSON: You are sure those are stop 25 i check and not just open gates? O ACE FEDERAL REPORTERS, INC. I 202 347 3700 Nationwide Coverage
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25727.0 51 REE
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(~/ (_ 1 MR. KEISSEL: They are defined as stop checks 2 and they were -- 3 MR. MICHELSON: Those are normally just 4 maintenance valves for the purpose of repairing that motor 5 operator valve. I didn't think they were stop checks at 6 all. I don't know why they are. 7 MR. KEISSEL: I don't know if this drawing is 8 the current one. These valves were installed in 1983 or i 9 1984. And they may have been installed as a replacement 10 l for the valves that are there. 11 ) MR. MICHELSON: I would have guessed the stop l 12 g check was the valve to the right without a number on it. , il l 13 1 MR. EBERSOLE: Well -- 14 ] MR. KEISSEL: No, sir. Because the -- I 15 4lj MR. MICHELSON: Go to the topmost line. I i! 16 ) thought that was the stop check. 0 17 E MR. KEISSEL: No, sir. 9 l 18 MR. MICHELSON: Okay. I will stand corrected. 19 ! I am just surprised. l 20 l MR. KEISSEL: These were the valves and they l 21 li have been checked by the number. The reports were by valve 22 numbers. These are the two that failed on unit 4. These 23 are the two that failed on unit 3 in January. 24 MR. MICHELSON: They don't need to be check 25 l valves. Not by the original design. The checking function ( i ACE-FEDERAL REPORTERS, INC. I 202-347-37tu) Nationwide Coverage M(N L 33tr(M6
25727.0 52 REE ("h (_) 1 is provided by the check valves shown on the drawing. 2 MR. EBERSOLE: I am looking at the two check 3 valves that are on the right side of your drawing there. 4 MR. KEISSEL: Just there and there. However you 5 also have a steam cap back this way. It is a very -- one 6 word might be sophisticated, another word, complicated 7 system. 8 MR. MICHELSON: I don't think there is a steam 9 , path back that way though, from the drawing, unless mine is 10 different than yours. That is the same line, it just 11 branches into two branches. If you go further to the right, 12 then there is another check valve. l (' 13 l MR. KEISSEL: I don't see the other check valve. \.s u 14 MR. MICHELSON: We have probably exceeded the 15 ll usefulness of this discussion? l 16 3 MR. EBERSOLE: Any comments, any body? 17 ! Okay. Thank you. p 18 Before you get to this presentation, I suggest 19 li that we include in our full committee presentation with the 20 added aspect of defining the degree of jeopardy to aux 21 turbine feed supply which was not at all clear from the 22 presentation from the E&ID. Okay? I don't know what the 23 degree of jeopardy here is. It was quite clear in the 24 SONGS case what was with us. But it is not here. 25 MR. ALLISON: We can explain that a little more. O ACE FEDERAL REPORTERS, INC. l 202-347 3700 Nationwide Coverage Mk3AfM6
25727.0 53 REE (_). 1 MR. MICHELSON: Would you reconfirm with the 2 licensee which are the stop checks on the drawing and why 3 they are stop checks when they are shown as gates on these 4 drawings. 5 MR. ALLISON: We will do that. I think the 6 discussion of the back leakage has tended to drag out a 7 little bit. I think probably a more realistic threat is 8 loose parts in the turbine. 9 MR. EBERSOLE: That could be, thank you. 10 MR. LICITRA: I am Manny Licitra with NRR. I am 11 I r here to discuss an event at Palo Verde unit 1 on January 9, i 12 l 1986. Palo Verde unit 1 at the time was in the last stages, 1 13 f final stages of its power ascension program. It was *
'~
14 k attempting to do a turbine trip test from 100 percent power. i 15 For that test, there is a certain feature in the plant i 3 1 16 j called a reactor cutback system. For that test -- they, 3 17 l this enabled the reactor cutback system, because it had 18 been successfully tested from a loss of load test at 100 19 percent -- I also believe they tried it at 80 percent and 20 it succeeded at that level -- so for this test, they did 21 j not have the reactor cutback system in operation. What t l 22 they expected was to have a reactor trip due to high 23 pressurized pressure. That is, the turbine would trip. 24 There would be an automatic transfer of nonessential loads 25 to the grid. The pumps, reactor coolant pumps would stay ACE FEDERAL REPORTERS, INC. l 202-347 3700 Nationwide Cmerage 8004tretM 4
__ .- _ ~ - . - - _ _ . _ __ 25727.0 54 REE () 1 on and heat would continue to generate. You raise the 2 pressure and you get a trip due to that event. ' 3 What they got instead is that the house loads 4 did not automatically transfer to the grid because of a 5 frequency mismatch between the grid and the lE buses. 6 MR. EBERSOLE: Was this a delayed -- we just got 7 through talking about the necessity to hold until you had a 8 voltage decay or you had a very rapid transfer. Which kind 9 of transfer is this? 10 MR. LICITRA: It is a synchronization match. 11 There is also a time element in there. I don't fully 12 understand the circuit myself. 13 MR. EBERSOLE: I think if they do it fast enough, 14 they do it fast enough before it gets out of sync. If they 15 wait too late, they have got to wait for a voltage decay. 16 MR. LICITRA: I think time comes into it. 17 MR. EBERSOLE: You don't have any idea of the 18 time spans we are talking about? 19 MR. LICITRA: No. What I do know is that the 20 system performed as designed. It was not a fault in how it 21 was set up. It was set up according to design, but because 22 of the way it was designed, it did not allow this transfer. 23 When you didn't get the transfer, you lose power to the 24 reactor coolant pumps, the main feedwater pumps, the 25 circulating water pumps, the steam dump for the bypass O ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Coverage 800-3 %.us
25727.0 55 < REE 1 control system. They are.all on non-lE buses. 2 MR. EBERSOLE: Aren't the main feed pumps 2 turbine driven? 4 MR. LICITRA: At least one is. I think they
- 5. both are.
6 MR. EBERSOLE: So when you say you lose the main f 7 feed pumps, you mean the booster pumps? [ 8 MR. LICITRA: I believe so, yes. 1 9 Now, when that happened, the pump didn't have 10 power. It is coasting down. Ano that caused a reactor 4
- 11 trip due to a projected, flow projected low DNBR.
12 The steam bypass control system valves did 13 receive a quick open signal but wouldn't close because they 14 lost power. 15 One of the four lowest setting main steam. safety ! 16 valves lifted when the steam generator pressure increased 17 -rapidly. And'the setting for those lowest safety valves is i 18 1250 psig. But only one of the four lifted. It stayed
'~
19 open for about 43 seconds. When that valve receded, five l 20 steam bypass control system valves modulated open, stayed 21 open for about 45 seconds, closed and then they had two l 22 additional open/close cycles for the next 40 seconds. j 23 MR. EBERSOLE: How did they operate when they ( 24 refused to operate just prior to that? You said but reclosed 1 i 25 almost immediately due to loss of power. l l l ACE-FEDERAL REPORTERS, INC. j 202-347-3700 Nationwide Coverage MO-336W44 L
d 25727.0 56 REE () 1 MR. LICITRA: I believe it is the quick-open 2 signal that is on the nonsafety bus. e 3 MR. EBERSOLE: Are you telling me the bypass
- 4 will open in the absence of circulating water? Until you
{ 5 get condenser, low vacuum? 6 MR. LICITRA: That is what happened. I 7 MR. EBERSOLE: That is strange, isn't it? You 8 denied the rapid opening right away due to loss of power. 9 Is that a -- is that the forced function, the loss of power 1 10 that denies bypass? 11 MR. LICITRA: That is what denied it? 12 MR. EBERSOLE: It wasn't loss of vacuum?
-13 MR. LICITRA: Well --
14 MR. EBERSOLE: If it isn't the loss of vacuum, I 15 $ guess I will ask you why, because as long as you have got 16 vacuum, you can bypass. i 17 MR. LICITRA: I guess I do not have an answer to [' 18 that. 19 MR. EBERSOLE: Okay. Go ahead. And then later, 20 you opened -- they modulated open again. 21 By the way, you mentioned that the safety valves, 22 the main steam safety valves lifted. What about the 23 electromatics that are interposed between the safeties and 24 the steam pressure. They are the ones that Palo Verde 25 advertised as being OA and safety grade, et cetera. Why O ( ACE FEDERAL REPORTERS, INC. 202 347-3700 Nationwide Coverage Mn3E(M6
l 25727.0 57 REE ( l didn't they intercept the opening of the main safety valves 2 on the secondary side or are they, are these valves serving 3 no purpose? 4 MR. LICITRA: Only one of the four valves opened. 5 MR. EBERSOLE: Is it a dual purpose electromatic 6 or PORV safety valve or not? You say safety valves. 7 MR. LICITRA: Yes, they are the safety valves. 8 MR. EBERSOLE: Are they the PORVs, or are they 9 electromatic or are they intended to be operational in 10 i front of the safeties? I 11 l MR. LICITRA: As you know, Palo Verde does not 3 12 E have PORVs. 13 j MR. EBERSOLE: I am talking about on the '^ 14 h secondary side. 1 15 )n MR. LICITRA: I am not able to -- 16 MR. EBERSOLE: I have to infer from what you , k i 4 1 17 haven't said is that these are multipurpose valves that 18 perform both a safety fanction and a pressure function l d i 19 0 which they claim is safety grade. Is that true? I 20 q MR. LICITRA: The atmospheric dump valves? 21 l MR. EBERSOLE: That is the one I am talking i 22 h about. 23 MR. LICITRA: These are safety valves. 24 MR. EBERSOLE: The atmospheric dumps, then, did I 25 not work because they are not aatomatically piloted to l
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ACE-FEDERAL REPORTERS, INC. 202-347-37a) Nation *ide coserage 8al-33Mari
l 25727.0 58 l REE I () yx 1 preclude opening of the safeties? Do you follow me? We l 2 put electromatics or dump valves on the secondary side to 3 intercept opening of the safeties because we had reclosing 4 problems with the safeties. All right? 5 What did they do? You tell me the main steam 6 safety valves lifted but you didn't tell me what the dump 7 valves did. 8 MR. LICITRA: Let me see if I can get an answer 9 out of the report. 10 MR. EBERSOLE: I wouldn't look it up. I would 11 rather have you plow on through it and we will look at it 12 later. 3 13 MR. LICITRA: Okay. ,'J f 14 ,f MR. REED: You are right. Normally the 15 [ atmospheric dumps will open for the safeties the same as 3 16 d the safeties, unless the pressure, rate of pressure rises N 17 E too fast. Then the safety will go. 18 MR. EBERSOLE: Right. l 1 19 MR. REED: This is a strange trip here. It is 20 l loss of a lot of vital equipment and it may be that the 21 atmospheric dumps also went open but couldn't take care of 22 the release quantity. So a safety valve opened. 23 MR. EBERSOLE: If so, it is not mentioned. 24 Let's go ahead. We won't have time to pursue the details. 25 Go ahead. A U ACE FEDERAL REPORTERS, INC. 202 347 3700 Nationside Cmerage mit33M6m e
25727.0 59 REE O(_/ 1 MR. LICITRA: Well, after the steam bypass 2 control system valves did modulate open, the operators at 3 that time took manual control of the individual valves and 4 stopped a cooldown. Simultaneously or about the same time, 5 they got a main steam isolation signal due to low steam 6 generator pressure. After the cooldown stabilized, the 7 operators reestablished cooldown via a steam generator, two 8 atmospheric dump valves and one auxiliary feedwater pump. 9 , MR. EBERSOLE: How did they get low steam 10 generator pressure? It was excess opening of the bypasses? l 11 l Was it -- 12 3 MR. LICITRA: They didn't have those valves open. f
~ ;
13 They were losing steam through the valves. (I') ? ! 14 l MR. EBERSOLE: When you say that they were under 3 l 15 ] manual control, they modulated open. They stayed open for l 5 16 l4 45 seconds. And is that, in the course of doing that, did o l 17 [ they lose pressure control? You say -- the bullet below
)
18 . that says manual control was taken? 19 l MR. LICITRA: About that time, they got a main i 20 l steam isolation signal, automatic signal. 21 - MR. EBERSOLE: Due to low pressure? l 22 i MR. LICITRA: Yes. They tried to take over 23 control to stop that from happening but it occurred. 24 MR. EBERSOLE: Did the operators inadvertently 25 let the steam pressure ride through the modulation valves ACE-FEDERAL REPORTERS, INC. 202-347 3700 Nationwide Coserage N n33MAM i
25727.0 60 REE ( ) I too long? I am just curious how they got to low pressure. 2 MR. LICITRA: I do not know the answer. 3 MR. MICHELSON: The inference in your chart is 4 that the steam bypasses were open when they shouldn't be 5 open. That is how you got the low pressure? d 6 MR. LICITRA: Yes. The steam bypass was open. 7 MR. MICHELSON: So it must be the steam bypass 8 i stuck open? I 9 MR. EBERSOLE: That is right. They modulated 10 open and stayed open. 11 ; MR. MICHELSON: But not because of need but l 12 because they -- 13 l MR. EBERSOLE: They just malfunctioned. (m k- 14 : MR. MICHELSON: -- malfunctioned and kept 15 oncoming down instead of closing again. i 16 l MR. LICITRA: That is one of the things they are 17 looking into. Why they did open. 18 j MR. EBERSOLE: Had the circulating water pumps 19 l started? I 20 MR. LICITRA: It is not discussed in their 21 l report. I 22 l MR. EBERSOLE: Okay. Go ahead. k 23 MR. LICITRA: About three minutes after the 24 event, at 13:28, they restored power to the lE buses. And 25 at 14:08, after doing other activities, they got the [Jh ACE-FEDERAL REPORTERS, INC. 202 347-3700 Nationwide Cmcrage N NL3 W%m i
25727.0 61 REE 1 reactor coolant system pumps started again. One at a time. 2 And the event terminated at 14:49, an hour and a half later. 3 MR. EBERSOLE: This is not very well known at 4 l all as to what really transpired. h 5l MR. REED: I think it ought to be brought back I 6 j to the next meeting. i 7l MR. EBERSOLE: I think. I think it is too muddy 8 to discuss it any further because there are no positive l 9 /l statements made about what happened. I didn't learn 10 l anything about whether or not there were these highly l 11 { advertised atmospheric dumps which are supposed to be i 12 safety grade. Palo Verde was proud of the fact that they 13 l said they had safety grade secondary dumps to control the ! ( l 14 i primary pressure up and down. And yet I see no statement i 15 l to the effect that they worked at all or they were even
!i 16 4 there. I think just in summary, we will just go to the l
17 4 next one and anticipate seeing this then. } 18 l MR. HERNAN: I think Manny does have some more l 19 , material as far as what action is going on. I would like i 20 4 to remind you, though, that it is the rules that we bring i these things down. 21 l 22 MR. EBERSOLE: This is why I am saying, I don't 23 1 think we know enough facts here. 24 MR. HERNAN: We will be more than happy to come 25 back. But I kind of get the feeling that the staff is O ACE-FEDERAL REPORTERS, INC. 202-m.noo Nation *ide coverage m33r ru,
25727.0 62 REE r"N (,) 1 being criticized for bringing something down too early. 2 MR. EBERSOLE: I gather that we don't really 3 know enough yet really. 4 MR. LICITRA: I just received the LER for this 9 5 event. 6 , MR. EBERSOLE: I am not criticizing you fot not 7 having all the data. I am just saying, I don't think you 8 have. I don't see the picture very clear as it was handed ( i 9 ! to you, t 10 l MR. HERNAN: We know that we don't have all the l 11 I data. But we didn't want to wait until we do to -- 12 MR. EBERSOLE: There are about four or five 13 blank spots here. [s) '"' b 14 MR. REED: I think a true -- all that is true, l
?
15 !} Jesse. Also, we have s.ome sensitivity with respect to Palo 4 16 l Verde systems. So you can read our letters and perhaps l tailor the presentation. 17 l 18 MR. EBERSOLE: We are most interested in Palo 19 i; Verde and all of its ups and downs. 0 20 MR. LICITRA: The utility has not completed its 21 review of this event. They are still looking at different 22 aspects of it. They don't have a bottom line. 23 MR. EBERSOLE: We will get a rerun of this, say, 24 next month, maybe, or whenever. 25 MR. HERNAN: These meetings are every two months. O ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Cm erage an 33MM6 f
25727.0 63 REE (D s._/ 1 MR. EBERSOLE: Well, who knows. Whatever. Next 2 chance. 3 MR. ALLISON: Next meeting. 4 MR. EBERSOLE: Okay. I am going to put " hold 5 over." Thank you. Let's go to the next one. 6 MR. HERNAN: Did you want him to discuss the 7 action that is going on? 8 MR. EBERSOLE: You have a page 2. I didn't 9 realize that. 10 MR. LICITRA: That just gave the follow-up 11 , actions that they are doing. l 12 i MR. HERNAN: I think that is important to cover.
~'
13 MR. EBERSOLE: Carry on. I didn't know there c_) 14 l was a second page. As a follow-up action, they are I 15 [ MR. LICITRA: E 16 looking into this synchronization check relay circuit to 17 l see what can be done about preventing this problem from 18 ) occurring. They do not have an answer at this time. But 19 } they are looking into it. I I 20 l Because one of the main steam safety valves 21 lifted, they checked the settings on all the valves and 22 they found it okay. So even though the other three didn't 23 lift and one did, apparently it must have been close to 24 that pressure. So they don't see any problem with the 25 settings on the safety valves. I 1 I l I ACE-FEDERAL REPORTERS, INC. 202-347-37(#1 Nationwide Coverage MrAF336MIA
25727.0 64 REE (_) 1 They are also considering providing 2 uninterruptible power to the steam dump bypass control 3 system so they won't run into this problem. They still 4 don't have a final answer on that. 5 After the event, and in evaluation, they 6 determined that the reactor coolant system pump goes down 7 faster than design and faster than considered in one of the 8 safety analysis. That being the loss of flow safety 9 analysis. 10 They also found that because of the fast 11 cooldown, the trip, the anticipatory trip occurred faster j 12 than they previously projected, so they are competing f- 13 effects. Since they didn't know what the overall effect is, t ( :l 14 Il they made an assumption that the faster cooldown does take 15 place but trips at the same time. They previously assumed 16 and imposed some penalties on themselves until they 17 complete that evaluatien. 18 ! And finally, they reran the turbine trip test on l 19 I the 24th, but this time they ran it with the 1, non-1E 1 20 loads powered off at the startup of transformer. So they 21 wouldn't have this problem. 22 MR. EBERSOLE: Okay. And did that one turn out . l 23 all right? ! l 24 MR. LICITRA: That one worked as expected, as l l 25 predicted. , ~h l (V ACE FEDERAL REPORTERS, INC. 202-347 37u) Nationwide roverage FuL11MM6
25727.0 65 REE 1 MR. EBERSOLE: I see. All right. Thank you. 2 Any comments? 3 MR. GIITTER: I am Joe Giitter with the Events 4 Analysis branch of the Office of Inspection and Enforcement. 5 The second event I would like to talk about today is a loss l 6 of off-site power that occurred on January 9 at the 7 Palisades plant with one diesel being out of service and 8 only one diesel in the standby readiness mode. Let me go l 9 ahead right away to the circumstances. I 10 The initial conditions, the unit was in cold { I 11 n shutdown for refueling for about 40 days. About one-third l 12 l 4 of the fuel was new and two-thirds of the fuel was reloaded, i 13 j so there wasn't much decayed heat at the time. The head l
~/ 1 l 14 0 was positioned on the vessel but not tensioned and the !
a l 15 ! 1 vessel was drained about down to the flange level, which is 16 q about 12 feet above the top of active fuel. Diesel ! i 17 ! a generator 1-1 was out of service and would have required i 18 j several hours to make operable. This diesel right here. 1 ! 19 Emergency diesel generator 1-1. 20 The sequence of events, at about 2:41 -- these 21 are p.m. Eastern Standard Time -- about 2:41, smoke was l 22 I observed from the conduit on the 1A bus in this general l l 23 F area here. It was later found out that this smoke was 24 l caused when a cable shorted to ground and heated up some 25 water that had gotten in the conduit, apparently rain water, /~T i (_) I ACE FEDERAL, REPORTERS, INC. 202-347-3no Nationwide cmerage No 33ust6
25727.0 66 REE (~N (_) 1 and it looked as though it was smoke. The operators 2 immediately de-energized bus lA which is a nonvital 4160 3 volt bus, this one right here, with the smoke believed to 4 be coming from here. And then because of fear of potential 5 fire or explosion, they decided to go ahead and de-energize 6 this entire rear bus, this 345 KV bus which feeds these 7 three starter transformers. Before doing that though, they 8 , -- at 2:56, the operators started and loaded the diesel 9 l generator that was in standby readiness. i The diesel 10 } generator function is required and supplied power to this I I 11 l bus, vital bus ID. At that point, at 3:08 they did 12 de-energize the rear 345 KV switchyard bus. That caused a i 13 l loss or caused buses 1C -- well, that was of course not (% 0 n 14 energized -- 1E and 1B to be de-energized. l 15 They did have an RilR pump on this and all of 16 their decay heat removal was accomplished by tying the Q 17 l diesel on this single bus here. I 18 The licensee declared an unusual event at 3:00 19 and the NRC operations center was notified at about 3:19. l l 20 l At 4:50, there was a shift change and the second i 21 shift came on and decided that they were going to establish 22 backfeed to all the main buses and at 4:50 they 23 accomplished this by backfeeding from this front bus 24 through the main transformer through the startup 25 transformer down to the vital buses and to the nonvital ace-FEDERAL. REPORTERS, INC.
.02 347 4 700 Nationwide rmerage N o 116fM 6
i 25727.0 67 REE 1 buses. 2 And once that was done, they terminated the 3 unusual event and secured the emergency diesel generator 4 that was tied to the 1D bus. 5 In summary, it was a loss of off-site power 6 caused by a voluntary disconnect and it appears that the ! l 7 licensee did everything right. j i 8 MR. REED: Why do you call it a loss of off-site { i 9 power? It looks to me like a partial loss at the most. ! i 10 MR. GIITTER: From -- as far as the plant, as ! f far as these buses are concerned, it is a loss of off-site 11 l 12 l power, even though they didn't physically lose any of these , 4 i () 13 incoming lines. 14 h MR. REED: Those lines are right in the , l 15 ;! switchyard at the facility, I assume. They are the main i 16 outgoing lines and incoming lines. 17 ! I am worried about these statistics that keep 18 l piling up on a loss of off-site power and total blackout. 19 j I worry that the people that are doing the analytical work 20 are not watching their definitions carefully. To me, I 21 don't see why this is tabbed, called a loss of off-site 22 power. To me it is an internal bus fault in the plant 23 which caused maneuvering and changing of electrical supply. 24 MR. ALLISON: During the time in question, there 25 was no off-site power feeding anything inside the plant. O ACE-FEDERAI, REPORTERS, INC 202 34L37m Nationwide Coverage 8f u 11', 'M6
25727.0 68 REE 1 Is that right? 2 MR. GIITTER: That is correct. 3 MR. ALLISON: No off-site power was reaching 4 into the station. 5 MR. REED: Simply because the diesel was on. 6 MR. GIITTER: The unit was shut down. There is 7 no power being supplied back through the -- 8 MR. ALLISON: There was no electricity coming 9 l from the switchyard into the plant. 10 , MR. REED: Are you telling me that both these i 11 j 345 lines are out of service completely? k 12 j MR. ALLISON: No, they are in the switchyard. I 13 am saying there is no electricity coming from the -- (} 14 f MR. EBERSOLE: They drove themselves into this a 15 [ thing. 16 l MR. GIITTER: They voluntarily disconnected. 17 f MR. EBERSOLEt That is a management decision to l 18 j drive into it. 19 MR. REED: I would like this not to appear as a i 20 loss of off-site power. 21 MR. EBERSOLE: This is not a loss in the context 22 that it was a loss without operator participation. They 23 deliberately drove into it. 24 MR. GIITTER: There may be a terminology problem. 25 MR. EBERSOLE: They decided to risk for a short ace-Fl!DIiRAL RI!PORTliRS, INC. 202 347 37(o Nationwide roserage M n 11MM6 j j
r 25727.0 69 REE 1 while that the diesel would hold. And it did. So I agree 2 with Glen that it is really not a loss in the usual context. 3 It is a -- we need to do something about the statistics 4 where you drive into a configuration like this. Somehow we 5 need the statistics to show that. 6 Any other comments? 7 I don't think we -- we will not carry this to i 8 the full committee. ! I I 9 MR. REED: Agreed. j 10 MR. ALLISON: Mr. Ebersole, we will discuss the i 11 definition for the statistics purposes with the staff l 12 people. () 13 l; MR. EBERSOLE: I think that would be very much 14 I in order. ; 15 Let's take a 10-minute break here. We will be ! b k 16 l back at 25 after. l 17 (Recess.) , 18 j MR. EBERSOLE: We will resume the meeting. I ; I 19 4 believe the next event is the Robinson. Mr. Requa. 20 MR. REQUA: Right. 21 MR. REQUA: My name is Bud Requa. I am the 22 project manager for ll.B. Robinson Unit 2 and this afternoon 23 I will talk to you about another loss of off-site power 24 coincident with a diesel generator out of service. 25 MR. KERR I thought Mr. Reed persuaded the i l Acti-FliolinAI. III!!'ORTiins, INC I 202 30-17(n Nationw ale Cmerage mo1%(M6
25727.0 70 REE 1 staff that this would be a loss of off-site power? 2 MR. REQUA: In quotes. 3 In January about the time this happened, on 4 January 28, the plant was in a closedown for refueling for 5 carly February. With a diesel generator out of service, a 6 fault occurred on the emergency bus E2 which caused a spike 7 on the instrumentation bus number 4. A false rod dropped e 8 signal was received, causing a turbine runback. The 9 reactor tripped on high pressurizer pressure from 80 { 10 percent power. At about 9:18, a loss of off-site power 11 occurred coincident with a fast transfer to the startup , i 12 transformer. We originally worded it that way because at ! I () 13 the time we didn't know why we lost off-site power but f 14 flaterwe found that the C phase balance relay opened due to 15 ' a phase mismatch. 16 MR. KERR What is the significance of the term i 17 '
" loss coincided with fast transfer to startup transformer."
i 18 l Does that mean that it actually didn't transfer to the 1 19 startup transformer? 20 MR. REQUAt Yes. It happened 80 fast it 21 couldn't tell what, which happened. But it turned out that 22 the -- at the time they didn't know what happened. 23 MR. KERR: Was the connection made to the 24 startup transformer? 25 MR. REOUA: The connection was made but retuned O Acti-lifiDIiR Al Illil'ORIliRS, INC
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25727.0 71 REE
- I because the relay had opened, refusing it.
2 MR. KERR: Okay. 3 MR. REQUA: At 9:35 -- I am sorry. The A diesel 4 generator automatically started and loaded on bus El. The 5 reactor coolant pump tripped and natural circulation was 6 established. 7 At this same time frame we received a safety 8 injection signal and the MSIVs were found closed. 9 MR. KERR: What does it mean if there is no 10 off-site power, to have the RRCPs tripped? 11 MR. REQUA: We didn't have them. 12 MR. EBERSOLE: When you say " fault on emergency j {} 13 bus E2." The bus fault is a rare thing? 14 l MR. REQUA: I understand this is what they were i W 15 1 speculating and they still don't know. There are a lot of i 16 unknowns on this one still. I f 17 5 MR. EBERSOLE: Suppose it had been a bus fault i
! l 18 on El, what would the emergency diesel have done then? Can 19 ! it get to E2 or is it locked out by circuitry? What I am 20 trying to say is, if it had a true bus fault, that really 21 kills that bus. If that happened to be the unfortunate 22 diesel generator which was working, we would be in trouble, 23 wouldn't we?
24 MR. SWENSON: I am Warren Swenson in NRR. It 25 does not appear to have been a true bus fault per se. I O Acti Fliolinal. RIil>okilius, INC. 202 147 170s) Nationwide rm erage No1%(M4
f i 25727.0 72 REE O 1 may be playing with the terminology here, but one of the 2 pieces of equipment for something like this that was 3 connected to the bus fault and they had not -- as the day 4 progressed, they had difficulty determining which specific 5 piece of equipment had produced the fault. The whole bus 6 itself apparently did not go dead, but it was sufficient to 7 lock the bus out temporarily. 8 MR. EBERSOLE: But it would have precluded 9 loading the diesel onto it. 10 MR. SWENSON: Apparently, yes. 11 MR. EBERSOLE: If that had been the case and if 12 that had been the diesel still operating, where would we 13 have been then? ( 14 MR. SWENSON: A station blackout. l 15 i MR. EBERSOLE: Okay. I just wanted to know how l 16 close to trouble we might be. Carry on. 17 l1 MR. REQUA: At this same time frame, about 9:18, 18 f we received a safety injection signal, found the MStVs l 19 l closed. This is another big unknown. We don't know what l 20 ' happened first as yet. At 9:35 an unusual event was 21 declared and open telephone lines established between the 22 NRC operating center and the Robinson site. 23 At 9:46, the B diesel generator was restored. 24 At this time the reactor coolant temperatute was 535 25 degrees F, pressure 2000 psig and the plant was stable. O Acti Fl!DI!RAL Rfil'ORII!Rs, INC. 202 m nm Naion*ide emerav om "'> "* ,
l 25727.0 73 REE O 1 MR. KERR: What good did it dn to have the B 2 diesel generator restored if there was a bus fault on bus 3 E2? 4 MR. EBERSOLE: Right. ! i 5 MR. REQUA: I -- 6 MR. SWENSON: This is Warren Swenson. As this l b 7 event progressed, like I said, it apparently was not a true i 8! ! bus fault but the, they shed the loads on that patticular ! l 9 bus. And as the event progressed, they reloaded the j 10 equipment back on the bus one piece at a time, taking their l f 11 i time and making sure that there were not additional 12 !I problems that they didn't drop out the bus again. It was () 13 . not the bus itself, a separate piece of equipment that 14 caused the failure. They didn't want to repeat the process 15 , and lose power. j i 16 l MR. KERR: Okay. Thank you. I i i 17 MR. EBERSOLE: I guess sooner or later you will 18 i tell me where the fault was. l 19 MR. REQUA: I still can't tell you where it was. 20 ! They still haven't found out. I 21 MR. FBERSOLE: That is another one of these 1 22 hanging cases. I 23 l MR. MICIIELSON: Spurious.
}
24 l MR. REQUA: At 10:52 the pressurizer heaters I 25 l were restored. At 12:28, not shown on hete, the C steam O Acti-Flii)iinAI. Riii oRilins, INC, m2 m rm moonmide nnnue nm m 'a
l l 25727.0 74 REE O 1 generator PORV was manually opened to balance loop 2 temperature for natural circulation. The valve stuck open 3 due to freezing of the instrumentation lines and they got 4 another safety injection signal on high steam line delta P. 5 The valve was closed a few minutes later by opening the air 6 supply valve and the partial blowdown was secured. At 12:55, , i 7 the El bus was connected to the off-site power. At 16:03, 8 the E2 bus was connected to off-site power. 9 Realigning the electrical system was delayed, as 10 Mr. Swenson said, because they wanted to take extreme care l l l 11 I to not load equipment on that may be still full. 12 ! The plant was taken to hot shutdown, and a day : () 13 or two later Carolina Power & Light decided to remain shut 14 l down for refueling, which is where they are now. Their k 15 )4 follow-up is that they are in a 45-day refueling outage and 4 16 the licensee is investigating the cause of the loss of l s 17 l off-site power with Region 2 reviewing their findings prior 18 l to startup. I 19 Also, the licensee's and INPO team are 20 investigating the high frequency of reactor scrams at g 21 Robinson, of which they had four in January. Region 2 was 22 also evaluating this along with Carolina Powet & Light, and l the results will be evaluated prior to startup. 23 l 24 i MR. REED: I would like to just clarify. You I i 25 say up there, stuck open steam generator PORV. If you were ! O Act? FrioriaAI. REpoirilias, INC. 2e m.rm Nanon.nle nn erare smtw<w,
25727.0 75 REE O- 1 asked to use the words -- instead of PORV -- what does PORV 2 mean to you? 3 MR. REQUA: That is a pressure-operated relief 4 valve? j 5 MR. REED: I believe the terminology is -- PORV 6 is power-operated relief valve. I have seen in various f 7 descriptions around here and a lot of -- and some news i 8 releases, that PORV means " pilot." I want to make sure we ' 9 have our nomenclature straight. I would have called this 10 an ADV. That is what it is, isn't it, an atmospheric dump i valve? Isn't it? 11 l 12 I MR. REQUA: Yes. Yes. , I i 13 MR. REED: We sort of, I think the terminology 14 [ goes like this, that PORV or power-operated relief valves i i 15 i are on the primary side of a PPWR. ADVs are on the } i I i 16 i secondary side. Just nitpicking, thank you. l l 17 j MR. REQUA: I appreciate that. { 18 i MR. EBERSOLE: Let me ask why a signal is 4 19 produced caused by closure of main steam isolation valves 20 to produce a safety injection signal, since the first 21 consequence of that is to lose heat transfer and to cause a 22 pressure rise in the primary loop which has nothing to do 23 with inventory. 24 MR. REQUA: That is another one of these 25 . mysteries that have to be solved. We don't know; they do Acti-Fl!DIiRAI, Rlil'OR~ll!RS, INC. I 202-347 17(s) N itionwide Cmcrase mn 3 m f6M
25727.0 76 REE 1 not still at this time know why they got that safety 2 injection signal. This was one of the reasons why they 3 didn't want to -- 4 MR. EBERSOLE: So that is spurious? 5 MR. REQUA: Right. 6 MR. EBERSOLE: The second signal caused by stuck 7 open, that is due to cctitraction of the -- or you don't say, 8! it hasn't yet -- there is nothing automatically producing a 9l safety injection signal because that occurred, is there? I 10 l MR. REQUA: Are you talking about the second -- 11 MR. EBERSOLE: The stuck open S/G PORV. That 12 doesn't automatically produce a safety injection signal. 13 Only the secondary effects of it, which is shrinkage? O 14 l k MR. REQUA: Right. j 15 4] MR. EBERSOLE: Which you say was caused by that. 2 16 l As though it were a directly produced signal by this stuck 17 open PORV. Do you mean it is by the secondary effects of 1 18 E it? 19 3 MR. REQUA: Yes. When these are condensed down, f ! 20 then it shrinks. l 21 l MR. EBERSOLE: So it is the secondary effects of 22 it. So the first signal was -- I don't know why that was 23 caused. The second issue was legitimate? 24 , MR. REQUA: Right. 25 MR. KERR: Did you say why that fast transfer i O i Acti-Flil)liRAl. Riii>oRTliRs, INC 202.m 17tni N.ition.nte cmcrage msr 116 ru6
1 25727.0 l 77 REE I failed to be successful? 2 MR. REQUA: Yes. Because there was a phase 3 mismatch in that the C phase balance relay opened. It says 4 a phase mismatch. I don't know what percentage it is. If 5 it is a certain percent, then the relay opens up and won't 6 accept the transfer, as I understand it. Is that correct? 7 MR. SWENSON: That is my understanding. j 8 MR. EBERSOLE: In summary, this is a real, this 9 is a close approach to total blackodt. Not like the 10 earlier blackout. This was not invoked by the operators. 11 ; It happened? 12 MR. REQUA: That is correct. j () i 13 MR. EBERSOLE: So it is a legitimate off-site 14 power failure down to a degraded emergency power supply. 3 15 j And it happened -- it was lucky enough it was on the right l 16 crisscrossed pattern of buses, a t 17 f One thing that bugs me, I see about 20 minutes ! 18 there gone while they figured out whether it was a bus j 19 l fault or not a bus fault. What were they doing, opening i ! 20 ' the breakers to clear the bus and verify the bus was cicar 21 and then put the loads on one at a time? 22 MR. REQUA: Yes. 23 MR. EBERSOLE: Did they over find where the bus 24 fault was? 25 MR. EBERSOLE: I take it it was a load fault O Acii Flini!RAI. RiiPonTiins, INC,
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25727.0 78 REE 1 that didn't clear? 2 MR. REQUA: As of this moment, we don't know.
- 3 MR. EBERSOLE: We don't know yet. Well, it 4 wasn't enough to clear by independent opening of the 5 circuit breakers? They didn't clear it? You say there was 6 a ground fault. Wait a minute. You say, fault on 7 emergency bus E2.
8 MR. REQUA: No. A spike -- on E2, yes. 9 MR. EBERSOLE: Ilow did you determine that? What 10 was the signal that says there is a fault on that bus? 11 MR. REQUA: As we said before, the bus itself i 12 I probably was not faulted. It was some equipment on there. l T 13 , MR. EBERSOLE: What is the signal that told you
) i 14 I it was a bus fault? Ilow did you conclude it was a fault or i
15 even erroneously? 16 MR. SWENSON: I don't think we have all that i 17 ! information. It was either a degraded voltage or loss of i 18 [ pcwor to the bus. And that is a question that -- 1 19 l MR. EI3ERSOLE: You don't know whether it was 20 1 incoming or outgoing problems? 21 MR. REQUA: No. We don't know. 22 MR. EBERSOLE: Okay. 23 l Any questions? 24 MR. REED: No questions. It doesn't look like 25 this should come before committee? O Acti-Fl!Dl!RAI, Illil'OR iliRS. INC. 2f>2 347 37m Nanonwide Cmerage M81 M6 WA
k 25727.0 79 REE O \/ 1 MR. EBERSOLE: No. This is just a close 2 approach to loss of power. 3 MR. REED: Did you happan to see that Piper Cub 4 that flew into the high lines last night? 5 MR. EBERSOLE: Maybe it was one of those. A bus 6 fault. 7 Any further -- no questions. Thank you. 8 MR. REQUA: Thank you. 9 MR. EBERSOLE: Our next one is design deficiency. I 10 l This is a B&W plant?
- i 11 MR. VISSING: Right.
12 MR. EBERSOLE: Peedwater in a B&W plant. That (~} 13 l is already interesting, say no more. \/ 14 ; MR. ALLISON: Mr. Ebersole, I don't think I j I 15 s wrote down an answer about Palisades, but I take it you ! I 16 don't want Palisades -- 17 . MR. EBERSOLE: No, you are correct. l 18 i MR. VISSING: I am Guy Vissing with the -- 19 project manager for Arkansas Nuclear One, Unit Number 1. I 20 am with the Office of Nuclear Reactor Regulation. 21 I am going to discuss with you today a design 22 deficiency in the emergency feedwater system of ANO 1 which 23 was discovered January 14. The significance of this event 24 would be a potential loss of all emergency feedwater and 25 blowdown of both steam generators during a steam line break O Acti Fl!DIiRAI. Rt!PORTiins, INC. 20214717m Natkmaide Cmcrage xm 1M (M6
7 l 3_ 25727.0 79 REE
^
()
\~) 1 MR. EBERSOLE: No. This is just a close 2 approach to loss of power.
3 MR. REED: Did you happen to see that Piper Cub n 4 that flew into the high lines last night? 5 MR. EBERSOLE: Maybe it was one of those. A bus 6 fault. 7 Any further -- no questions. Thank you. l [ 8 MR. REQUA: Thank you. ' 9 MR. EBERSOLE: Our next one is design deficiency.
,. 10 This is a B&W plant?
11 MR. VISSING: Right. 12 MR. EBERSOLE: Feedwater in a B&W plant. That 13 is already interesting, say no more. (']N u \ , 14 l' MR. ALLISON: Mr. Ebersole, I don't think I j I 15 i wrote down an answer about Palisades, but I take it you 16 don't want Palisades -- 17 MR. EBERSOLE: No, you are correct. 18 MR. VISSING: I am Guy Vissing with the -- 19 project manager for Arkansas Nuclear One, Unit Number 1. I 20 am with the Office of Nuclear Reactor Regulation. 21 I am going to discuss with you today a design 22 deficiency in the emergency feedwater system of ANO 1 which 23 was discovered January 14. The significance of this event 24 would be a potential loss of all emergency feedwater and 25 blowdown of both steam generators during a steam line break Acti I;IioliRAI. Riti>oRTiins, INC. 202 E U m Nanon*ide rmnage Mm H6 ua4 x
25727.0 80 REE O' I with a single failure of one'AC bus. 2 I would like to go through the narrative here 3 and then I will show you some schematics that more -- in 4 more detail explain the event. This deficiency was 5 discovered by an inspection team from the Office of 6 Inspection and Enforcement in which they postulated a steam 7 line break concurrent with a loss of the red AC power bus. 8 The turbine-driven and motor-driven emergency feedwater l s 9l i pumps would be lost and there would be a possible blowdown 10 ! of both steam generators. I
- . 11 j MR. EBERSOLE: How old is this plant? It is
,' 12 many years old?
i i () 13 MR. VISSING: This plant was licensed in April 14 3 1974. 9 15 MR. EBERSOLE: So we have got 11 or 12 years? 16 t' MR. VISSING: Yes. However, I would like to add 17 that this emergency feedwater system has recently been 18 } upgraded to a safety grade system and it was partially a
\
19 ; result of that upgrade, i 20 ; MR. EBERSOLEt~ The upgrade in fact ran it into 21 this failure mode?
- 22 MR. VISSING: Well, yes. I would like to get to 23 , that.
24 The design would not meet the single failure 25 criteria. The plant was at 89 percent power at the time, Acli-17EDl!RAl. REl'ORTI!RS, INC. I 202 347 170) Nationwide Cmcrase Ni n t U6 (M6
25727.0 81 REE O' # 1 and concurrent with this they had a failure of the diesel 2 generator and they were shut down. 3 The licensee confirmed the existence -- 4 MR. KERR: I am sorry. Concurrent with what did 5 the diesel genera *.or fail? 6 MR. VISSING: The diesel generator tripped on a 7 high pressure in the crank case and they had a controlled 8 shutdown to repair the diesel generator. It was the diesel, 9 really. 10 The licensee, to correct this, they did indeed 11 install check valves in the steam lines from both steam i 12 l generators before startup. Our staff at NRR did review l () t 13 this issue and concurred with the inspection team's 14 I findings. I would like to show you the emergency feedwater i 15 l system of Arkansas. You know that Arkansas Unit 1 is a B&W i 16 l reactor. The emergency feedwater system is a two-train 17 system. It has the adversity in that one pump is 18 motor-driven and one pump is steam turbine-driven. i 19 l I wanted you to note that the motor-driven pump i
- 20 ! is powered from the red bus, and if we take a take a look 21 at the steam system of the turbine --
22 MR. REED: Where is this newly installed check 23 valve? Is that the one right next to the steam generator? 24 MR. VISSING: I haven't shown that yet. Yes. 25 This would be in the steam part of the -- the seem steam i i Acti FriotinAL RiiPonTiins, INC. 202-3 3 3700 N.oion*ide Cmcrue m W, um
25727.0 82 REE 1 system of the emergency feedwater steam turbine. 2 MR. HERNAN: It may be helpful, if you want to 3 use both projectors, you can turn the other one on there. 4 MR. VISSING: Okay. Let's see here. That is 5 the hydraulic part of the system. 6 MR. EBERSOLE: Before you go further, isn't it l 7 true that if you have only a two-pump aux feed system, it i I 8 you lose steam supply to the turbine-driven pump and then 9 you invoke a theoretical AC power supply to the single l i 10 remaining pump and it doesn't have a transfer capability, . i 11 f you are dead? l l 12 MR. VISSING: That is right. , 13 j MR. EBERSOLE: That is just built into the 14 fdesignconfigurationtobeginwith? ; 1 15 MR. VISSING: That is right. 16 O MR. EBERSOLE: So was that not accepted as a h a 17 basis? 18 l MR. ALLISON: The difference here is the steam 19 line break loses the supply to the pump and that is the 20 accident that the system is there to mitigate. It is true 21 you can go down and fail at -- you can fail the 22 steam-driven pump, but that is a failure. However, when 23 you have a steam line break, that is not really a failure. 24 MR. EBERSOLE: Why not? 25 MR. ALLISON: You then need to have a redundant O V ACE-FEDERAL REPORTERS, INC. ( 202 347 3700 Nationside Coserage M O-3 3MM6
I 25727.0 83 REE
; I aux feed system to deal with the steam line break.
1 2 MR. EBERSOLE: Well, then that is not done. t 3 MR. ALLISON: Yes, they missed it here. 4 MR. EBERSOLE: Are there not quite a few plants 5 that only have two pumps, aux feed pumps, one of which is 6 turbine-driven? 7 MR. VISSING: That is right. Z 8 MR. EBERSOLE: Doesn't every one of them face ; 9 the possibility of loss of steam flow to one pump? 10 MR. ALLISON: If you wait a little, he gats to 11 the check valve question. 12 MR. EBERSOLE: Maybe it is something -- 13 MR. VISSING: I would like to show you the {} 14 l scenario that would happen with this configuration if they l { 15 a had a steam line break in the 36-inch line here, and you
- t l 16 will note that this valve is a normally opened valve and it 17 is on the red bus. And the motor over here is on the red r j
18 bus. Now, having the break here,' you would then have the 19 valve here normally opened and the valve here opened and 1 20 you would have a blowdown of the steam generator and no 1 21 steam to the auxiliary -- the turbine-driven pump. t 22 MR. EBERSOLE: Right. 23 MR. VISSING: Therefore -- and also you wouldn't 24 have power to the red -- to the motor-driven pump. So, , 25 therefore, you would lose, first you would lose your main I
! ACE-FEDERAL REPORTERS, INC.
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25727.0 84 REE (~h ! \/ 1 feedwater because it would trip off because of the steam 2 line break. 3 MR. EBERSOLE: Sure. J 4 MR. VISSING: And then you would lose your 5 emergency feedwater pumps. 6 Now, what is normally done here is what they did 7 do to correct this situation: they, Arkansas, recognized 8 what they did and they installed check valves here and here 9 (indicating) and they switched this, these two pumps, these I 10 !I two valves around so that this would be a closed valve, 11 { locked, closed valve and this would be the operating valve i 12 l that would operate off the red bus. - I 1 () 13 Now, therefore, if you have a postulated break 14 L in the 36 inch line, you would then lose the main feedwater il 15 y pumps, and you would lose -- if you had a single failure in s il 16 S the red bus, then you would lose this valve here, but it B 17 f would have a check valve and you wouldn't have the blowdown 18 j of the steam, of this part over here and you would still 19 have steam to your turbine of the turbine-driven motor. 20 What happened here, in their upgrade of their 21 system, they did initially have in their design such check 22 valves. However, subsequently, because of problems that 23 they had experienced on Unit 2, with check valves 24 chattering, they thought that they were going to have a 25 much more reliable system if they didn't have those check ACE FEDERAL REPORTERS, INC. 202-347-3700 Nation ide rmerage No3% W6
25727.0 85 REE em N_) 1 valves, and not recognizing that they didn't adequately 2 look at the single failure criteria. 3 And so they did indeed install check valves, the 4 swing-type check valves. And they are going to upgrade 5 those check valves with -- when they can -- with the 6 designs that are similar to what they used in Unit 2 to 7 correct the problem. 8 MR. REED: Is there any traceability on -- 9 apparently a design decision was made way back when to 10 remove some valves in a very critical system. I 11 MR. VISSING: Yes. 12 MR. REED: Is there any traceability as to who I ( 13 ! was responsible, who verified, approved such a design 14 change? j 15 i n MR. VISSING: Well, that is an issue that isn't I I 16 altogether clear to us. They explained their -- the 17 ' scenario during the inspection process and it wasn't 18 altogether clear as to the thoroughness of their review for 19 that change. 20 MR. REED: Do you think it was a designer 21 failure or a utility / owner failure? Who was -- was it the 22 utility that caused the deletion of the valves or the 23 designer? 24 MR. VISSING: The design was originally there. 25 The valves were originally in the design. Subsequently, [
\
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25727.0 86 REE 1 they were indeed removed and there was a conscious decision 2 to remove them from the design.
- 3' Now, who indeed were the contributing people 1
4 that made that decision, I don't know. 4-5 MR. EBERSOLE: In a turbine-driven aux pump, 6 isn't it just like breathing? When I look at the steam , 7 line, I will always see two checks? 8 MR. VISSING: Yes. , t 9 MR. EBERSOLE: But it wasn't done? 1 10 MR. VISSING: That is right. r 11 MR. EBERSOLE: So they weren't breathing very 12 well. That is astonishing, isn't it? (} 13 MR. VISSING: Yes. 14 MR. EBERSOLE: How did they escape all this 15 ferocious review in their engineering section? I want to 16 know how it got through? 17 MR. VISSING: I can't answer that. I don't know. 18 MR. EBERSOLE: How did it escape NRC?
.19 MR. VISSING: It did not escape NRC. Their 20 original design was submitted to NRC --
21 MR. EBERSOLE: I mean the modification? 22 MR. VISSING: The modification, they did it 23 under a 5059. 24 MR. EBERSOLE: What happens to that? 25 MR. VISSING: We did not review that.
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25727.0 87 REE 1 MR. REED: I don't see how this could go into 2 5059. It certainly degrades, changes the safety analysis 3 by a lot. 4 MR. MICHELSON: They just didn't recognize it. 5 MR. EBERSOLE: But it is up to them to recognize 6 that. So here is a basic flaw in the whole rationale. t 7 Unless thcv can see it, you will never know. hnd this l 8 f would apply to anything. l 9l MR. MICHELSON: You see it at the end of the 10 year when you get the yearly report from the utility. 11 MR. REED: This was done some time ago, l 12 ! apparently. This is an old plant. I l {} 13 14 MR. VISSING: The design was originally submitted to us in the latter part of 1981. 15 l MR. REED: This modification to take out the l l 16 checks? l 17 MR. VISSING: No, the initial design of the 18 upgrade of the system was submitted to the NRC in December 19 of 1981. 20 MR. EBERSOLE: And it escaped that -- 21 MR. VISSING: It had the check valves in that 22 design. 23 MR. EBERSOLE: It had the check valves in 1981? 24 MR. VISSING: Yes. 25 MR. EBERSOLE: And then they took them out. A U ACE FEDERAL REPORTERS, INC. I 202-347-37(X) Nationwide Coverage 8(xk3164M6
25727.0 88 REE /~N 1 MR. VISSING: The upgrade was installed in the 2 last refueling outage which was in December of -- it was 3 December of 1984. 4 MR. REED: I think I am becoming confused. I 5 think you are talking about the checks being put back in 6 and I am focusing on when, how, and what time were they 7 deleted from the original design. That must have been back 8 in 1974 or -5? 9 MR. VISSING: No. They had submitted an upgrade l 10 of the emergency feedwater system in December of 1981. At ( 11 l that time that design had had check valves in the system. i 12 ! Tne system was not installed until December of 1984. k (} 13 p MR. EBERSOLE: You mean it took three years -- 14 f MR. VISSING: Yes. b 15 ij MR. EBERSOLE: -- for that to materialize? 16 g MR. VISSING: That is right. And between the 17 f time that the design was submitted to the NRC and the time 18 it was installed, there was a change made in the design. I 19 MR. EBERSOLE: How did that get supervised? i l l 20 l MR. VISSING: We did not see that change. 21 MR. EBERSOLE: That doesn't come to you? 22 MR. VISSING: No. 23 MR. REED: I guess what you are telling me is 24 that this 12-year old plant had a Three Mile Island backfit 25 modification going and that didn't hit the fan until 1981? () V ACE-FEDERAL REPORTERS, INC. l 202-347-37M) Nationwide Coverage h3 W6M6
25727.0 89 REE m 1 Is that right? 2 MR. VISSING: Right. 3 MR. REED: I wonder what they had prior to Three 4 Mile Island. 5 MR. VISSING: I tried to find that out. I don't 6 know. 7 MR. REED: They probably had the right system. 8 That's just a joke. I 9 On this one, I see this only as a statement to l l 10 , point out that auxiliary boiler feed systems are very l 1 11 vulnerable to all kinds of problems and design aspects. I l 12 don't know that I support auxiliary feed system as the only
/~}
NJ 13 way to take care of decay heat removal on PWRs. l 14 l! MR. EBERSOLE: I was wondering if we should ; 15 y bring this to the full committee of how important things 16 can slip through a supposedly rigorous review system and 17 and then be picked up as late as this. 18 MR. REED: In the gut issue, the root issue is 19 decay heat removal. 20 MR. EBERSOLE: But here is an explicit -- i 21 MR. KERR: I don't think we have to convince 22 anybody that decay heat removal is important. What is it 23 that we are trying to persuade the full committee to do? 24 MR. REED: To adopt primary blowdown for certain 25 PWRs? l ACE FEDERAL REPORTERS, INC. 202-347-3700 Nationside Coserage 8rn336 6M6
25727.0 90 REE C 1 MR. EBERSOLE: But is there an administrative 2 control or move here that is indicated to cl.ose these holes 3 in the review process so that we don' t let these things lay 4 around like this? What is the NRC action on this in the 5 generic context. 6 MR. KERR: My impression was that this was not 7 done deliberately, it was just that somebody overlooked 8 something. Are we suggesting this they go back and review 9l every plant? 10 MR. EBERSOLE: Aren't there some general rules 11 i like bre athing, that you always look for two check valves? 12 k MR. KERR: Well, there are some general rules I 13 like one never makes a mistake. { 14 6 MR. ALLISON: This type of inspection is a good 15 thing to do where you go and look at the design of the 16 $ g system in some detail and find cases like this and take
!i 17 [ enforcement action and publish information notices. That 18 is really our -- what we are doing.
I 19 ! MR. EBERSOLE: What is contemplated here is an 20 ! enforcement action other than -- 21 MR. ALLISON: I don't know specifically. I 22 guess we can find that out. 23 MR. EBERSOLE: What about the quality of the 24 tility review? 25 MR. ALLISON: I think that will get a closer ACE FEDERAL REPORTERS, INC. l 202-347-3700 Nationwide Cmerage 8m 3W6t>M
i i 1 l 25727.0 91 REE (~)
\/ 1 look by the Region now.
2 MR. EBERSOLE: Is that enough? 3 This has been sitting there for how long? Three 4 or four years? , 5 MR. ALLISON: Right. About that long. 6 MR. REED: Jesse, I would like to repeat my i 7 position, it seems to me that it is worthwhile for the full j 8 committee to know that here is another example of the 9 vulnerability of aux feed as the only path of decay heat 10 [ removal for some reactors. i 11 MR. EBERSOLE: Let's put down aboat five minutes i 12 for this one, just to mention it to the full committee. l
~'
13 MR. KERR: I will agree to that will if you will 14 ?I assure the full committee that this is being put to them i l 15 [ because of Glen Reed. i H 16 2 MR. EBERSOLE: Okay. I will mention this in a 1 : R 17 very quick way. }' 18 l MR. HERNAN: Do you want the presentation or is i this just something -- 19 l , j 20 MR. EBERSOLE: No. We will make a statement to 1 1 21 this effect. You can make it or I will make it. Whatever. 22 Let's go to the next one. l 23 MR. WEISS: The next item is the River Bend l 24 startup review. It is not an event per se, but it is an l 25 action that was taken because we saw significant number of n\_/ l ACE FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Coserage N O-33MM6
, 25727.0 92 REE (3 V 1 '5072 reports for a new plant that was just starting up and 2 we decided to go out and have a look and see if we could 3 divine exactly what was going on. 4 MR. KERR: Will you forgive me if I ask you what 5 a 5072 report is? 6 MR. WEISS: A 5072 report is a report to the f 7 operations center made over the emergency notification 8, system. A red phone call. If you go in the control room i 9 ) of every nuclear power plant, there is a red phone. You I 10 pick it up and call within either one or four hours of an 11 l event occurring. The criteria for 5072 are very similar to 12 those required for LERs. So generally they are the same i ()
%)
13 l thing. l 14 l MR. KERR: Thank you. 15 j MR. WEISS: To get to the bottom line first, we, 16 f after our visit we came to the conclusion that we will 17 expect the number of events that this plant is producing to 18 decrease; generally we thought their management appeared 19 sound. However, as a check, we asked the licensee to come 20 to the Region 4 for a progress review, and I believe the 21 date is March 4. There were a number of us that went. I 22 was the only representative from IE. Steven Stern, the 23 project manager who is here, and three other individuals 24 from NRR went. The Region 4 section chief, John Jordan and 25 his project manager showed up. John is here. O ACE-FEDERAL REPORTERS, INC. l Il 202-347 3700 Nationside Coverage Nn336W46
25727.0 93 REE k- 1 When we visited the site on January 28 through 2 30, we went over with the staff their recent events. 3 Recent scrams and other significant events, examined root 4 causes. We witnessed a shift turnover in the control room. 5 We had a tour of the plant, including the diesel generators 1 6 and the feedwater system. The feedwater system was I 7 particularly important since this plant is having a , 1 8 significant number of problems with their feedwater system. l 9 We went to the Fancy Point substation, which was I 10 significant because they had a loss of off-site power event
- L 11 early in January that was precipitated at this substation.
12 We witnessed a surveillance procedure on the reactor water l l I 13 3 coolant system. The largest single contributor was ("} (- l 14 y inadvertent isolation of reactor water cleanup. It was
$ i 15 also significant from the point of view of the l
4 16 l s demonstrating their ability to deal with jumper control , 17 l problems. 18 { We also got a walkdown of the control room
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19 I panels, and that was revealing in a number of ways, not l 20 I only to familiarize ourselves with the plant, but to also 21 with some of the problems they had. I noticed that they 22 had a 3 by 5 card pasted on the panel with instructions 23 warning the operators to be careful not to have the suction 9' to shut down cooling and LPSI open at the same time. They,
.; l 3.<e most plants, have had one of those draindown events.
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25727.0 94 REE (% \ l In the exit interview, the project manager, 2 Steve Stern, urged the licensee to continue open 3 communication with the NRC. We thought this was important I l 4 because the licensee was indeed making a significant number ! 5 of reports that were voluntary in nature and weren't really 6 required by the regulations and we wanted that to continue. 7 There were such things as half scrams and half 8 isolations that told us something about the plant or 9 generic component problems and so forth. 10 We urged the licensee to resolve the outstanding 11 equipment and human factors problems. He is undergoing a 12 significant number of problems, as would any plant during (') 13 L
]
startup. And we did criticize them and urge them to 14 h upgrade their communication with other plants, because we 3 15 thought there was a significant number of problems that if
]
16 l they were not preventable, at least they could have found 2 17 8 out earlier by having more open, more frank discussions 18 l with their counterparts at other plants, other similar 19 boilers with a year or two more experience. We urged them 20 to conduct these particularly at the lower levels rather 21 than just at the highest levels of plant management. 22 I could discuss a large number of plant-specific 23 problems, but I just want to mention a few to give you a 24 flavor for some of the things that we found. They had a 25 temporary alterations program that I guess was best O ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Coverage 8(WL1W6M6
l 25727.0 95 REE k/ 1 characterized like a quick 5059 that was maybe appropriate 2 during the latter stages of construction but we thought 3 should come to a halt soon. It was appropriate for an 4 operating plant. 5 They had, like most new boilers, they have more 1 6l i instrumentation than the old boilers and a small number of 7 annunciator windows. One of those small windows had 8 something like a diesel generator problem. There were, I 9 . think, over 15 different conditions that could have caused I 10 li that window to flash, some of which were more in the nature 11 l of nuisance alarms and others were mote serious. However, 12 l they wouldn't get a reflash when a more serious problem l {} 13 came in, and the operators that had asked that this be 14 ! fixed and it is being fixed. 15 $ The feedwater problems we thought were the most
)
16 ) serious of those that we had seen out there. They had 17 l vibration, for example, in the short-cycle loop that this j 18 I had actually torn an anchor out of the wall. They had a 19 packing leak in that room when we visited it. 20 MR. KERR: What is a short cycle? l 21 l MR. WEISS: That is a loop that takes you back l 22 i to the condenser for the purpose of cleaning up the l 23 feedwater system. There is both a short cycle and a long 24 cycle, depending on how much of the feedwater system you 25 want to clean up. They had erosion in the long-cycle loop O ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Coserage 8m 336 (M6
I l 25727.0 96 ; REE I l T [/ \- 1 and were installing heavier schedule piping, more 2 erosion-resistant piping. They had sticking of the 3 feedwater regular valves, actual mechanical binding. The 4 operators noticed that they did not actually have real 5 valve position indication in the control room. They had a 6 l demand signal. That was being changed. 7 They had steam tunnel ventilation problems. And 8 I understand that they are being rectified with additional a 9 coolers. But one of the most interesting aspects of the 10 visit was this was a particularly productive visit in terms ll 11 ! of identifying generic problems for us. We are lucky if we I i 12 get one good problem out of a visit that we can tell other 13 plants about. Here we came up with three that seemed to be 14 v interesting. 15 $ You may recall from reading regional daily 16 reports earlier this year they had a feedwater size 17 isolation valve with its operator fallen off on the floor. 18 There were about three contributing factors to that event 19 -l that ultimately I think will result in an information 20 notice. There was inadequate thread engagement. The bolts 21 were too short. There is a difference of opinion as to 22 what was the proper torque value between the various 23 vendors involved. And there was also a question of whether 24 the valves should operate or should have been torqued with 25 a valve on its, seat because that appeared to have ACE-FEDERAL REPORTERS, INC. sn.97 3700 Nationwide coscrage *xt3 % ua6 m --_.. ,
25727.0 97 REE (-) 1 transferred some torque into the valve stem. Then when the 2 valve was cycled, the torque was relaxed on those bolts. 3 Another generic problem occurred with a 4 particular variety of temperature switch they were using in 5 their reactor water cleanup isolation. The licensee 6 informed us of the need to retrofit in a significant number 7 of capacitors and to add a resister for the purpose of 8 noise suppression. And it is our understanding at this 9 point that General Electric will be issuing a service 10 information letter on that subject. 11 Now, the fiber -- 12 MR. KERR: Was that a new design? /~ 13 $ MR. WEISS: No. That was one of the things that \-) 14 h we thought that they could have learned about if they had i
! i 15 been speaking to other late model boilers, would have had i 16 similar problems.
17 MR. KERR: GE didn't know about it because GE is [ 18 just now issuing something from what you said. Or did I 19 misunderstand? 20 MR. WEISS: That is correct. I understand that 21 GE will be issuing something on the subject. Exactly how 22 much GE knew and when they knew it, I am not in a position 23 to say. But we have seen -- 24 MR. KERR: Had others previously had the problem 25 and done repairs? (q/ ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nation ide Coverage N3WMA
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sl 1 MR. WEISS: Yes. 2 MR. KERR: But they just haven't told anybody 3 about it. 4 MR. WEISS: Sometimes when you replace a 5 component, you think, well, maybe it is a random component 6 failure. But after one sees a pattern at several plants, 7 one becomes disillusioned with that point of view and 8 begins to think that there is a generic problem. 9 MR. KERR: I taought it was a matter of 10 redesigning the circuit. 11 MR. WEISS: There was a noire suppression 12 resistor that River Bend thinks is advisable so that when
/~N 13 one moves a switch to read to function on the -- this O j 14 l particular temperature switch, one doesn't get a noise 15 spike that causes an inadvertent isolation.
16 l I think, I don't know whether other plants have 17 adopted that change. But it is certainly worth telling 18 other people about. 19 There is an interesting problem in the 20 switchyard. The plant added a diverse method of 21 controlling breakers in the switchyard which consisted of a 22 fiber optics multiplex system. They suffered a loss of 1 23 off-site power event around January 1 of this year. There 24 were no valid targets set that -- in other words, there 25 were no real faults identified, electrical faults (')') ACE-FEDERAL REPORTER $, INC. 202-347-37m Nanonside Coverage m k ) in-t<.46 i
25727.0 99 REE x gh t, 1 identified to cause this event. Testing subsequent to the 2 event has shown that any one of several different kinds of 3 hand-held walkie-talkies will cause tripping of the 4 tone-relaying equipment that is associated with the fiber 5 f optics system. I l 6 ' The utility has adopted shielding; they 7 installed some plywood panels and put copper shielding i 8l around the room that contains the equipment. They have 9 installed an events recorder so that, should it happen 10 again, they will have much more detailed information about 11 ! it. They have put up signs in the building and on the 12 ', fence around the substation and installed nonduplicating t i I i () r 13 locks at the entrance. They have trained their people, 14 l1 cautioned their people and supplied additional procedures 15 ] in the substation for resetting breakers, should the need 16 arise. j 17 ! They are also doing some filtering and l 18 separation of power supplies. 1 19 , In general, -- l 20 MR. KERR: I am a little puzzled. Fiber optics, l 1 21 I thought, transmitted rather high frequency 22 ! electromagnetic radiation. 23 MR. WEISS: That was the impression I was under. 24 And every time I bring the subject up with an electrical 25 engineer, he says it can't happen. That is why we will O ACE FEDERAL REPORTERS, INC. I 202. m -U m Nahonude rmeage W 3 % u46
I 25727.0 100 REE (~)s (- 1 probably write an information notice on the subject. What 2 people ferget is that at the end of the fiber optics line, 3 at the end of this glass or fiber tube, you will have a 4 demodulator receiver of some kind which will operate in the 5 radio frequency regime. 6 MR. KERR: My point was, it seems to me the 7 difficulty is not with the fiber optics but with the 8 termination? 9 MR. WEISS: With the end, which is exactly tight. 10 You may recall that there was two loss-of-off-site-power 11 events, I believe, at Palo Verde that were caused by fiber 12 optics systems. 13 I am reasonably sure that the licensee there did 14 not evaluate whether a walkie-talkie could have l 15 y precipitated that problem. But that licensee decided to 16 remove breaker control by fiber optics. Now their breaker i 17 f control is strictly hard wire. 18 l In any case, there were several things that h 19 f pleased us when we went out there. One I wanted to mention 20 ! is their condition report system. Apparently anyone in 21 plant can write a condition report on any condition and 22 this goes to, directly to the upper level management, which 23 then makes a decision on its ultimate disposition. One of 24 the strengths of their system is their ability to focus 25 resources quickly. ACE-FEDERAL. REPORTERS, INC. an-347 37oo Nationwide coverage sco i m.ius
25727.0 101 REE 1 One of the things that I was most concerned 2 about when I arrived was their jumper control. They have 3 instated a strong program there of locking cabinets and 4 then when one wants to perform a maintenance procedure or 5 surveillance of some kind, then you go to the jumper l 6 I control officer who then gives you a serialized tag for i 7 each jumper that is installed and there is an open log s entry for each jumper which then must be cleared. That 9 is in addition to the double verification that would be j 10 fperformedonorderSTPs. i > 11 N Another interesting aspect of their jumper l 12 control problem was they had a significant number of l j i i 13 events that were caused by alligator clips falling off 14 the terminal strips. They tried several different 15 things. They settled on a solution that appears to be j 16 f quite desirable. They used banana plug, banana jack-type il 17 assemblies. They install a banana jack beneath the f 18 l terminal lugs that are to be used in a surveillance b 19 ji procedure. Then when the banana plug is used for a test i I 20 instrument, it will not fall out. 21 It also has the added feature, I noticed, of 22 when you open the cabinet, you may see several hundred 23 possible connections, but you see only four brightly 24 l colored banana jacks, which greatly cuts down on the 25 possibility for putting it in the wrong place. O ACE FEDERAL REPORTERS, INC. 202-m-37m Nationwide coverage 8m-33uu6
25727.0 102 REE
' 1 I have already mentioned the work they did with 2 fiber optics. Another thing I notice that they did in the 3 control room was a large number of their meters were 4 colored in such a way that the normal operating range was 5 something like green and abnormal range was in some other 6 color. And I asked the assistant operations supervisor 7 whose idea that was. He said it was mine. So I think that 8 indicates that the operators are having a substantial 9 feedback into their plant design and management is 10 listening to them.
11 MR. EBERSOLE: That was the plant that provided 12 the number 3 diesel cooling water from the number 2 diesel
/~T 13 i supplies? Did they ever fix that?
V 14 MR. WEISS: Yes. s d 15 j MR. EBERSOLE: Okay. I 16 h MR. WEISS: I noticed a number of interesting c 17 f design features out there. I haven't gone to that many 18 j plants, but it was interesting. I 19 ; MR. KERR: That location is fairly remote. You 20 f may be getting all those telephone calls just because they i 21 l get lonesome and want to call somebody up. 22 MR. REED: Let me try another theory. It seems 23 to me that what I an hearing, because they have lots of 24 events, reportable events, is that there is an aggressive 25 utility, informed, perhaps capable organization that see , ACE FEDERAL REPORTERS, INC. 202 347 3700 Nationwide Cmcrage 8m 3 4 6M6 l
I i I 25727.0 103 REE D I weaknesses and vulnerabilities in design and they are not 2 just accepting that and laying down with it. Is that what 3 I am hearing? 4 MR. WEISS: That is an entirely possible thing. 5 We have the project manager here who has a series of charts 6 that show that on a month-to-month basis, they have a fair 7 number of reports. But if you look at it from a 8 milestone-to-milestone basis, they are doing pretty well. 9 MR. REED: They have a convergent seam. 10 MR. WEISS: One of the benefits of going out 11 (t there is that you get some individuals willing to speak 12 very frankly. I had my eyes opened up to some of the l n (~) 13 ) techniques that plants use to eliminate reportable events. l
\/ j !
14 j For example, if you have got a system that you know is , 15 cantankerous and is likely to cause a red phone call, one 4 i 16 5 thing you can do is take it out of service, declare it l 3 a 17 ; inoperable and perform your surveillances on it while it is i 18 : inoperable. Then it doesn't become reportable. I hadn't 19 heard of that twist before but now I am a little bit wiser. l 20 l The plant, by the way, has not been doing that. 21 i, MR. KERR: I bet Mr. Reed has heard of it. 22 (Laughter). 23 MR. REED: I hadn't really. 24 MR. KERR: I didn't mean that you'had done it. 25 , I just meant, I bet you had heard of it. ACE FliDERAL RIiPORTiins. INC. I 202J47-3RN) Nationa nle Cm erage Nuk 116 fM6
25727.0 104 REE '-- 1 MR. REED: I am getting the feeling that you 2 have a couple people probably selected by aptitude testing, 3 well trained, who are causing quite a stir with this new 4 plant and finding a lot of problems. 5 MR. EBERSOLE: You are an optimist. 6 MR. KERR: We have got to have another report on 7 aptitude tests. 8 MR. EBERSOLE: Is that it? 9 MR. WEISS: That's it. Thank you very much. 10 MR. EBERSOLE: I believe we have had a 11 substantial increase in interest. We would like to i 12 l introduce Bob Benaro, director of the PWR licensing l (} 13 j division. 14 MR. BENARO: I would like to get into the Perry I 15 t earthquake event. This event occurred on January 31st and 16 simultaneously I approached people on the committee to get I' 17 on the agenda for the committee's reaction or review of 18 what we were doing in safety evaluation of the event. And 19 simultaneously two members of the Congress wrote to you and 20 I think you must have seen those letters by now. I will be 21 talking to Mr. Fraley and others in order to do what I can 22 to expedite the ACRS attenr. ion to this natter. 23 The first formal report of the event has just 24 become available today. We had an all-day meeting at the 25 site yesterday on the event. And the owner of the plant in O ace-FEDERAL REPORTERS, INC. l 20' 34747m B'ationwide Cos erage m n l it,(M6
25727.0 105 REE \> 1 here with his consultants and I have a good number of staff 2 members here and I would like to go right in by turning it 3 over to the owner who will give you a substantive briefing 4 on what happened and what their evaluation of the event 5 shows. 6 So Mr. Murray Edelman, who is the senior vice 7 president, if I recall his title correctly, of Cleveland ! l 8 l Electric Illuminating Company, I will turn the floor over 9 to him. , { 10 h MR. EDELMAN: Good afternoon. I am Murray l
! I 11 l Edelman, vice president of CEI in charge of the Perry 12 project in our nuclear program. I am glad to be here this
() 13 ! 14 afternoon on what actions we have taken with respect to the
; earthquake that occurred in the vicinity of our plant on i 15 , January 31.
16 Yesterday afternoon, we had the opportunity to f i 17 P brief the NRC staff and their consultants at the site as to l 18 l what actions we have taken since the earthquake occurred on 19 January 31st. This is the overhead that we used yesterday. 20 j What I plan to do is to cover the introduction and the 21 overview of what actions we have taken the plant status and 22 our response that was given by our operations general 23 supervisor yesterday. And a little bit about the 24 earthquake analysis and seismicity that was performed for 25 us by Weston Geophysical. Mr. liolt is here. I am not a O ace-FEDERAL REPORTERS, INC l 202 347 37m Natum ide Cmcrage Mt1%'M6
25727.0 106 REE 1 geologist but I will try and cover that portion of the 2 program. Then we would like to have Dr. Chen, who was 3 responsible for the design and seismic analysis of plant 4 for Gilbert Associates, our architect engineer, review our 5 design criteria and actual experience that we had during 6 the earthquake on January 31st. 7 As an overview, the event occurred on the 31st, I 8 about 11:45 in the morning. Our plant people, even though 9 we do not have an operating license and are about 10 case 10 days away from, in our opinion from receiving a license, 11 l even though we're not in an operating mode, chose to go into 12 our emergency procedures for several reasons. One, it is {} 13 the best way we have learned, through training over the 14 ! last several years, of marshaling all the resources of CEI, 15 hn the county, the NRC and the state of Ohio. It gives us the 16 opportunity to ensure accountability of anybody who was in I 17 the plant at the time. Since our security system is in d' 18 l full effect, we used that to get a site accountability of 19 everybody on site and we went into our emergency plan and 20 ' activated our facilities as well as the facilities in 21 Chicago and Washington. 22 We then also sent our people out in the plant 23 after that to assess the status of the plant. We ran 24 numerous plant inspections with our operating people to 25 insure that the plant was in a safe condition. We then i. . ACE FEDERAL REPORTERS, INC. l 202 347 37(o Nationwide roverage Hu t 1 W rMI,
25727.0 107 REE C\ kl 1 downgraded our emergency event from a site emergency down 2 to an alert and after conferences with the state, the NRC l 3 and ourselves, we terminated the emergency about 2:30 in 4 the afternoon. At which point in time we instituted our 5 recovery organization, the same as we would have in a 6 normal emergency if one had existed. We are still in that 7 mode where I am the recovery manager to bring the site and 8 response to the technical issues raised from the earthquake. 9 We did conduct numerous walkdowns by our people. 10 About 65 of our people spent the entire evening and next i i 11 j day while the NRC sent people in from both Chicago and 1 12 I Washington in response to a mutually agreed upon l {} 13 14 confirmatory action letter from Region 3. We froze all equipment and inspected everything and Washington came in i i 15 j to inspect the equipment. After inspections by our people 16 i and the NRC, the confirmatory action letter was modified 1 17 !i which allowed us to go back to finish our testing and 18 operation of the plant. l 19 In addition, immediately after the event 20 l occurred, we got a hold of all our special consultants who 21 have worked on the Perry project. This included Gilbert
- 22 Associates, Weston Geophysical, Mr. Engdahl who 23 manufactures the seismic information; the people from 24 Kinemetrics; Dr. Hall, Mr. Stevenson and a number of other 25 outside consultants to review the plant conditions.
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25727.0 108 REE O 1 Close examinations were also prepared in nearby 2 ateas to record any aftershocks that may have occurred 3 after the event took place. 4 We also started to compare the plant data that 5_ _ we had recorded during the earthquake to see if it had any 6 engineering significance in our overall design. We did 7 record some high frequency acceJorations which we didn't 8 consider to be a problem, which.is a conclusion of our 9 l report which you have in front of you, because they were
~10 _
low energy, very short duration and low velocity. The 11 subject of the high frequency accelerations is a generic l 12 subject which we know the staff is looking at at a number l t
, ( 13 of sites around this country. We have reviewed and are 14 l still reviewing the plant'structbres and equipment to make i
15 ! sure that there is enough conservatism in the design of the I 16 l plant and considering our loading combinations and i 17 r allowable stresses. 18 I would like to go through what our operations 19 f people looked at in terms of what actions we have taken. 4 20 I As the event started, we had ongoing testing and 21 calibration activity going on. We were preparing to run 22 our division 2 diesel generator testing. We were ready to 23 move the startup sources from the upper pool toward the 24 reactor and we had a number of systems in standby and 25 operating mode throughout the plant. What occurred during, O ACE FEDERAL REPORTERS, INC. 202 347-17tv) Nas sonwide Cmerage 8m 1%#AA
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1 after and while the operation of the event took place is 2 all systems remained operable. All safety systems that 3 were designed to be operating'through an earthquake j 4 operated through the earthquake -- all relays, all the 5 safety systems we had in the' plant. 6 Immediately following that -- 7 MR. MAOL: What was the status of the plant at 8 the time the earthquake occurred? ) 9 MR. EDELMAN: We were in our final surveillance
- 10 testing. We had moved fuel from the lower pool to the i
11 upper pool. We were calibrating our source range monitors 12 in the upper pool and running our final surveillance test
- 13 on some of our major systems such as the final test that
)
14 you have to run on diesel generators. ; We sent our operators out after the event.took 15 [ 16 place and examined the plant to see if there was any i,
- 17. structural damage. We did walkdown by maintenance people.
'18 A systematic identification of everything in the plant. We f 19 asked our people to identify any abnormalities they might
! 20 see, no matter what they were, whether it was a burned-out 21 lightbulb or a cracked or peeled paint. Each one.was j 22 evaluated to see if it had any impact on the earthquake ' 23 itself. ? 24 We ran site surveys and settlement surveys which l 25 we normally run on a monthly basis. Part of our recovery J i ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Coverage 800-3364M4
25727.0 110 REE (ql 1 plan. We asked those surveys to be taken the next day to 2 see if we had seen any settlement throughout th7 plant. We 3 looked at our cooling tower and walked down the structures. 4 We have certain areas in the plant where we have identified 5 what we call seismic violations. One pipe may be too close 6 to another. There were 26 of those that we had not yet 7 fixed so we went and looked at those specifically to see if 8 anything had interacted. Nothing was identified. 9 We studied all the relays that were energized to 10 make sure that they stayed energized through the event, and 11 ; they were. 12 And we in addition, and consistent with our 13 .' confirmatory action letter, we put a new procedure in as we (~)3
~
14 were running our surveillances. If any surveillance test k 15 ' didn't pass, we would go back and analyze that to make sure 16 3 that was not a cause of the earthquake. 17 Since that time, we are running these 18 l surveillances, we have not identified any that have worked 19 without our management review or NRC review because of the j 20 earthquake itself. 21 Today we are back finishing our surveillances 22 and our preopen test right now. We met with the staff and 23 reviewed our operational readiness. We anticipate to be 24 complete, ready for fuel loading in about 10 days. 25 Now I would like to cover some general USGS data G O ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Coverage Mb3 W6646
25727.0 111 REE fs ( ) N/ 1 on the status of the earthquake itself. 2 MR. KERR: Excuse me. One of the things that we 3 have encountered in discussion of seismic event and fix, at 4 least those that might be hypothesized, is the possibility ! 5 that relay chatter will occur and relays will put things in 6 modes of operation that are unexpected or unusual. 7 Did your investigation include any effort to see j 8 whether that sort of thing had oc:urred? i 9 MR. EDELMAN: To the best of my knowledge, yes, l 10 I and to the best of my knowledge we found nothing where that l 11 l occurred. The relay that did trip in the field was one 12 that was on our generator breaker which was supposed to (~') 13 !; trip and the volt damage to it was off at the time and
\m/ 4 14 [ would not go back. But of all o our safety systems, we ,
3 . 15 ] had no relay chatter or anythin' go into a different mode i 16 hl l because of the earthquake. l 17 MR. KERR: Thank you. l 18 j MR. MICHELSON: Were all of the safety systems 19 fully energized and armed at the time of the earthquake so 20 that you could see if relay chatter was causing anytnir.g l 21 unusual to happen? 22 MR. EDELMAN: I don't think all of our systems 23 were energized but I can give you a list of about 40 that 24 were. Our control rod drives -- 25 MR. BENARO: For convenience, it is on page 3.6 V'O ACE FEDERAL REPORTERS, INC. l 202-347-3700 Nationwide cmerage NnDuM6
25727.0 112 REE q i
'- 1 of that big report. Section 3, page 3.6.
2 MR. EDELMAN: I think we listed in there all the 3 safety systems that were energized. 4 MR. MICHELSON: Of those that were energized, 5 was there any potential for relay chatter to even occur? 6 MR. BENARO: Ask him, not me. 7 MR. MICHELSON: Have you actually analyzed your 8 system to see if there was a potential to observe the 9 effects of relay chatter to know if there was even a 10 question about it? 11 MR. EDELMAN: Our manager of engineering is here. 12 MR. STEAD: We have looked at that. In fact, at 13 the time the event occurred, we were in preparation for a f~'/) x_ q 14 response time testing and it was fully lined up and in a l t 15 j standby mode ready to do the test. This is where you 0 16 9 actually initiate a signal to the bus and put all the loads 17 ! '.t after the diesel starts. All the systems were lined 18 up, ready to go into operation, waiting for a LOCA signal 19 to initiate. So all those relays were in position whereby 20 any movement to have started a LPSI pump or a high pressure 21 injection pump or caused a valve to open or close. So 22 those things are all in that position. So any of those 23 systems that are related to emergency core cooling, which I 24 think is probably the best test, in response to your 25 question, were in a mode that that could have happened and (O
%)
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25727.0 113 REE b '~' 1 wouldn't be immediately recognized by the control room 2 operators by indication in the control room. We saw no 3 indication of that during the event or after the event. 4 MR. MICHELSON: Not knowing the details of your 5 particular test, it is a little hard to see how the systems-6 were fully armed, though, because in order to be so, there l 7 must be certain temperature pressure permissives, et cetera, 8 in the system. You went in and jumpered out a great deal 9 of ECCS initiating pressure or temperature. l 10 MR. EDELMAN: If a plant is not running, you 11 can't have some of those. 12 MR. MICHELSON: But within the spectrum of what E (>) 13 ]y you had, you did not see any relay chatter effects and you 14 [ were already armed and ready to go on that? ; 1 i 15 { MR. STEAD: That is correct. ! n il 16 y MR. MICHELSON: Thank you. l 17 c MR. EDELMAN: The data I am showing up here, all l 18 j this information and slides that we have are all in the i 19 report that we have. What we have handed out to the staff 20 l and yourself today is the USGS data which locates the type I 21 and latitude and longitude of the earthquakes, the depth of 22 the earthquake and based on 64 stations worldwide, the 23 magnitude on the Richter scale was about 4.96. It was 24 located in Geauga County about 11 miles from the plant site. 25 MR. MICHELSON: I am assuming that you observed
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25727.0 114 REE /~N 1 no nonsafety systems' spurious operation? 2 MR. STEAD: No, we did not. We had a strip of 3 our instrument air compressor on high vibration. 4 MR. MICHELSON: I am not sure what you mean. 5 Was it induced by the earthquake? 6 MR. STEAD: Yes. It was a shaft vibration line. 7 MR. MICHELSON: That is the sort of thing you 8 are looking for, just like relay chatter is induced by the 9 earthquake. I 10 l MR. STEAD: This was in the normal mode of its b 11 f operation. It is a centrifugal compressor. l 12 [ MR. MICHELSON: What I am asking you, was there 4 13 Oj anything unusual that happened in the plant that might have ("/) 8 14 ',' been coming from the effects of the earthquake? ll 15 j MR. EDELMAN: The only two things that tripped il 16 l that we describe in our report are the instrument air 17 ] compressor and a building heating boiler which tripped. I 18 All other equipment remained in operation. 19 MR. STEAD: I think on the next page is a list 20 ; of all the nonsafety systems that were in operation at the 21 i t.me and gave no indication of any spurious operations. 22 MR. MICHELSON: Your air system was nonessential? 23 MR. STEAD: That is right. 24 MR. KERR: I know you don't like this, but it 25 seems to me the data that you are collecting is rather O
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25727.0 115 REE O 1 important in terms of sort of a general indication of what 2 happened. 3 MR. MICHELSON: huat-you have-to compare it with 4 is what is the design basis seismic situation, which I 5 assume you are going to tell us about in a minute. 6 MR. SIESS: Have you got any idea of what the 7 range is all about? 8 MR. EDELMAN: I didn't hear the question. 9 MR. SIESS: Do you have any idea of what the 10 range of natural frequencies are for relays? 11 MR. EDELMAN: For relays? I do not know. 12 MR. KERR: That is an easy question to answer. (} 13 The answer is no. 14 MR. EDELHAN: The map of the site area which is 15 in your book shows the plant site on the top, a 5- and a 16 10-mile radius ring. The epicenter was here on January 17 I 31st. The recorded aftershocks were about three miles west 18 of the epicenter. The highest one I believe is about 2.4. 19 We had no additional readings from these aftershocks on any 20 of our plant instrumentation. 21 This slide, the site at Perry is located -- and 22 this is kind of a skewed picture -- we are showing the 23 focus of the epicenter about 11 miles offset from the site 24 and about 5- to 6000 feet below or 30,000 feet below the 25 surface of the earth in the Precambrian rock. The site is O ACE. FEDERAL. REPORTERS, INC. l 202 347 3700 Nationwide Coserage 80b33MM4
25727.0 116 REE b) N' 1 actually located on top of undifferentiated Paleozoic 2 sedimentary rocks on shale, as I call it. 3 During the -- we did extensive review of the 4 seismology during the PSAR and FSAR stage of the plant. We 5 did find geological anomalies which were glacially formed 6 folds in the sedimentary rock well above the Precambrian 7 rocks and the earthquake occurred down here. Our people 8 have looked at it and the conclusion that our geologists 9 and seismologists have reached from their analysis is that 10 the tectonic approach that we took in our PSAR and FSAR 11 f stage is still valid. That there is no capable fault, no i 12 h tectonic structure. That our safe shutdown earthquake I {} 13 14 intensive scale is 7. That our site-specific spectrum was 5.3. 15 We did see, based on extensive instrumentation
]
16 i;i that we had in place and operatir.g at the time, an 17 exceedance and short duration of a high frequency. That is 18 ; about 20 Hertz of a spike on the upper levels of the 1 19 i containment building. Dr. Chen will go through that in 20 l detail. But that was, again, high frequency and was above, t 21 ! exceeded the 84 percentile spectrum that was put in the 22 k FSAR. 23 Dr. Chen now, I have gone through this pretty 24 quickly to allow Dr. Chen enough time to review the details 25 of the specifics of the site and the design of the plant. (_-) f ace-FEDERAL REPORTERS, INC l 202-347 37(o Nationwide Coverage 8 n 3 M(M6
25727.0 117 REE O 1 I would like to turn it over now to Dr. Chen, who is our 2 structure recall designer for our plant. 3 MR. CHEN: I am the manager of the civil 4 structural department of Gilbert. Before we go into a 5 detailed discussion of the comparison of this 1986 Ohio 6 earthquake event versus our design, I would like to 7 summarize the nature of the earthquake which will give all 8 of us a proper perspective of the size of the earthquake. 9 i The recorded 1986 Ohio earthquake is of the i 10 l nature of high frequencies, low velocity, low displacement j 11 -I and low energy, short duration earthquake. In comparison i 12 i with the kind of design earthquake we used for Perry, it is h (} 13 of the broad frequency band: long duration, high velocity, 14 I idgh displacement and high energy earthquake. i
, i 15 To show you how we reached this conclusion, we 16 h will show you the comparison of the recorded time history d
17 :; versus the time history we used in design. This time ti 18 l history at the top is our design time history in the I 19 north / south direction on the top of the foundation mat. l 1 I 20 The one at the bottom is the recorded time 21 history in the same direction, at the same location. As 22 you can see, at the top the design time history has a 23 duration of 22 seconds. The recorded time history with the 24 str ng motion, part of the record is less than one second. 25 And the one at the bottcm has much higher frequency content, t l ACE FEDERAL REPORTERS, INC. 202-347-37(x) Nationwide rm erage W 14 tMI>
25727.0 118 REE ( i 1 The high frequency content which translates into low 2 velocity and low displacement. 3 This is the north / south component. We will show 4 you the rest of the components. 5 This one is east / west. As you can see, the 6 expiration value for east / west is also lower than the 7 design time history. Not only the duration is shorter. 8l i The next one would be the vertical component. Also of 9 !, similar nature. Also the scores were low. 10 MR. REED: In the lower graph, does it actually 11 i stop and then is it a straight line on the right or? l 12 j MR. CHEN: Totally stopped. Whatever is left is 13 : the background noise. (_~/') 3 14 ( u MR. REED: Okay. 1 15 3 MR. CHEN: Next we show you the comparison of i' 16 the upper elevation which is at the steel containment. 17 This one is also north / south component. l 18 , similar conclusion, that the curation of the 19 j recorder is much shorter and that frequency is much higher. 20 There is some exceeders in the expiration which is a 20 i ! 21 Hertz region. We will come to discuss that comparison 22 ! later. l 23 The next one is in east / west direction. As you 24 , can see, the acceleration is also less than what we used in i 25 ! t design.
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25727.0 119 REE 1 This one is in the vertical direction, a similar 2 conclusion. 3 Then in comparison, we can see our seismic 4 design basis is of the nature of broad band frequency 5 design with smoothed 84.1 percentile spectrum. And it has 6 much higher energy and long duration. 7 The spectrum we used I think everybody is 8 familiar with. 9 Next we will show you the location of those 10 instruments. And the type of the instrument. At the top 11 of the Kinemetrics SMA-3, time history recorder. They are 12 located at the top of the reactor building map and also at { 13 14 the elevation 686 of containment. The three at the bottom are the Engdahl recorders. One out of the three during the 15 earthquake was in maintenance. That is why there was no 16 record of the other one. 17 There are four more instruments on this table. 18 These four instruments are Engdahl response recorders. 19 They are located at reactor building foundation mat, 20 reactor building 630 platform and the other two are at the 21 basement of aux building. 22 This one shows you the location in the plant l 23 view of those recorders. 1 and 2, for example here, are j 24 the triaxial time history recorder at the top of the 25 foundation mat and at the steel containment and the rest 4 l ! ACE-FEDERAL REPORTERS, INC. f 202 347-3?00 Nationwide Coserage 8M3%fM6 i
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25727.0 120 REE tm, I
' '} l are Engdahl instruments.
2 This one shows you the elevation view of the 3 location of the instruments. 1 and 2 are Kinemetrics 4 triaxial time history recorders. And the rest are Engdahl 5 instruments. 6 MR. MOELLER: The 4, 7 and 8 are different, are 7 they? 8 MR. CHEN: 4, 7, and 8. 3 and 4 -- let me see. 9 3 and 4 are the peak spectrum recorders. 5, 6, 7, 8 are l 10 the triaxial response spectrum recorders. One records the 11 peak. One records the response at 12 discrete frequencies. 12 i All those instruments with regard were used in l ('-)S n 13 i the comparison with our design data. The first design 14 barrier we compare with is the so-called ZPA variable which 15 means zero period acceleration. 16 The conclusion of this comparison is that the 17 recorded ZPA value varied from way below OBE value to 74 18 percent of the SSE values. P 19 I Except at containment vessel elevation 686, as 20 we showed earlier in the time history comparison. But at 21 this point, the relative displacement, in other words, the 22 stresses is very low. 23 The reason for that is because the nature of the 24 earthquake is of high frequency and low displacement 25 ; earthquake. l l ACE-FEDERAL REPORTERS, INC. ! 202 mJ100 Nationwide Coverage MWL3 % MM L
25727.0 121 REE 1 Now we will show you the table, how we reached 2 this conclusion. 3 There are the five columns and the five 4 different locations of the recorders with the ZPA values. 5 'ne i first column is at the foundation mat of the aux I 6 building. The second column is at the toundation level of 7 the reactor building. The third column is at the top of 8 the recirculation pump. The fourth column is at the 9 elevation 630 of the reactor building. And the fifth i 10 column is elevation 686 of the steel containment. And the i 11 l first row here is in the north / south direction. And the i 12 second row is in the east / west direction. And the third (} 13 row is in the vertical direction. Since during the seismic 14 I . design we considered three components simultaneously, the 3 4 I 15 q proper comparison is the square root of sum of squares 16 which is at the fourth row here. i 17 i If we look at the SRSS comparison here, at a 18 foundation mat of aux building, the recorded value is 19 i comparable to OBE value. At the foundation mat of the 20 reactor building, the recorded value is between OBE and SSE. 21 At the top of recirculation pump, the recorded value is way 22 l below OBE value. And at elevation 630 of reactor building, 23 the record in the vertical direction cannot be determined. l 24 So the only thing we can compare is in the north / south and l. 25 east / west direction. In these two directions, the recorded ace FIIDI!RAL lttil>oRTI!Rs, INC. 202 M7 37(W) Nationwide rtaserage N n 33(r(M6
25727.0 122 REE V 1 values are below OBE value. 2 The only thing of some exceedance is in the last 3 column in the reactor building elevation 686. The SRSS 4 value of the recorded acceleration is comparable to SSE 5 with slight, less than 4 percent exceedance. 6 Now we were looking to dislocation about its 7 relative displacement. At elevation 686, with the recorded 8 value exceeding the design value by less than 4 percent, we 9 are looking at a displacement. The first column is the 10 displacement at the foundation mat. The second column is , 1 11 the displacement at the elevation 686. And the third 12 column is the difference which is the so-called relative l 13 l displacement. Relative displacement is a measure of f')S s ; 14 in the containment. fstresses 15 If we look at a comparison in the last row, t 16 s which is the SRSS value, the relative displacement is about l 17 ; one-half of OBE value. So that is why we said that the l 18 l nature of this earthquake is of very small displacement, 19 even though the acceleration is high, the induced stress is I 20 low. 21 In addition to the comparison of ZPA values, we 22 also compared response spectra. This is the conclusion of 23 our response spectra comparison. 24 First, Perry design spectra are far above the 25 recorded spectra in the frequency range below 14 liertz. O 1 ace-FEDERAL. REl'ORTERS, INC.
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25727.0 123 REE $ O
\~) 1 Second, certain recorded response spectra 2 exceeded design spectra around the frequency region of 20 3 Hertz. But, this exceedance at 20 Hertz corresponding to 4 very small displacement, for example, at the foundation mat, 5 which is less than 3/100ths of an inch, because of this 6 small displacement at 20 Hertz, it doesn't really have any 7 engineering significance. Also, the recorded velocity 8l spectra, which is a measure of energy input, is also much i
9 lower than the design velocity spectra. The north / south 10 components of the foundation mat of the reactor mat -- by i 11 !, the way, this is a paper used in 160. I presume everybody 12 is f amiliar with this type of paper. (~% 13 l At the frequency region below 14 Hertz, this is (-) ) e 14 l the design spectra. This is the recorded spectra. As we 15 n can see here, the design spectra is far above the recorded 16 spectra, except with this spike here, which corresponds to 17 $ 20 Hertz. But if we look at the displacement value at this 18 l spike, which is somewhere here, it is 2 times 10 to the 19 ; minus 2 inches, that is .02 inch, 2/100th of an inch 20 ! spectra displacement. It doesn't really have any l 21 l engineering significance because of the low displacement. 22 l This north / south component -- we will look at I 23 j the east / west components in the same location. The 24 exceedance is even less in the high frequency region here. 25 Same thing in the low frequency region. Design is well O Acti-I;liDI! rat RiiPORTliRS, INC. 202-347-37m %tionwide roverage Nt o-33MM6
25727.0 124 REE O 1 exceeded by the design spectra. Again, the peak 2 displacement here is about 1.5 times 10 to the minus 2. 3 That means .015 inch, which is very small. 4 Next we will look at vertical component at the 5 top of the foundation mat. Same thing. Below 14 Hertz, 6 the recorded spectra is way under the design spectra. 7 Again, with a 20 Hertz spike here which corresponded to 8 about 1.5 times 10 to the minus 2. In other words, .015 9 inch. Which is very small. I 10 I will now show the rest of the comparisons 11 which are all in the book. 12 l After this comparison, we want to show you why 13 this earthquake is insignificant. First of all, all the (~)h u , 14 I energized equipment went through the earthquake as designed.
)
15 ! Second, as to the other industrial design criteria. During 16 (1 the blasting or pile-driving operation, the criterion used 17 for nondamage, nondamage threshold of non-engineered l 18 ) building is 1 inch per second. The maximum recorded 19 velocity at the site on the foundation mat is only .87 inch 20 l per second. That means that the maximum recorded velocity I 21 on site is less than what people would experience during a 22 pile-driving operation. 23 Another thing is, in comparison with IEE 344 1 24 j which is the seismic qualification of class IE components l 25 in a nuclear power plant, section 77.5 of this IEE 344 O ace-FEDERAL REPORTEns, INC. 20:447.n m Nanon*ide nnerage sm 3 w rom
25727.0 125 REE O 1 prohibits the use of shock tests to qualify the equipment 2 for the reason of high frequency and short duration input. 3 MR. MICHELSON: I guess one can say from that 4 that I guess we can't derive too much comfort from the fact i 5 that there were no electrical disturbances in terms of 6 relay chatter or whatever. So 1 guess it still leaves us 7 without any good data on responses of electronic equipment 8 of that type. 9 MR. EBERSOLE: My experience in listening to 10 j this is that you thought maybe in the beginning you had i 11 l something here and then when you analyzed it, you found you i 12 - didn't. l (} 13 ; MR. CHEN From the plant safety point of view, 14 you are right. It is nothing. l 15 l1 MR. EBERSOLE: But you didn't know that. ! i 16 f MR. CHEN: We didn't know until we looked into i 17 / it. I 18 ' MR. EBERSOLE: I guess we have to take that sort I
, I 19 , of news to the full committee. There is a special session 20 l set up for it.
I 21 MR. CHEN: Yes, sir. 22 ; MR. MICHELSON: So it is inconc ' 'is i ve from the 23 ! viewpoint of providing us some indirect evidence of 24 survivability of these plants from the electtical telay 25 chatter viewpoint at least. O Acit 1;iiDIiRAI. Illii>oR'11!Rs, INC. MM.Foo Naion* Me Omane xm mto,
25727.0 126 ; REE
/~T kl 1 MR. CHEN: You are right, sir.
2 MR. EBERSOLE: And that is because of the 3 absence of any duration. 4 MR. CHEN: Short duration and high frequency 5 input. 6 MR. SIESS: I still think the high frequency 7 might have been a better test of relay chatter, but 8 something else. 9 MR. KERR: You only need one half cycles. 10 MR. SIESS: You had several cycles of the high 11 frequency. But somebody has got to tell me more .. bout -- 12 MR. MICHELSON: You could go back and do soine -- {} 13 14 MR. SIESS: MR. MICHELSON: With respect to relays -- You might be able to do nome 15 laboratory work to correlate -- MR. SIESS: If it turns out they are sensitive 16 l 17 to high frequency, we could probably do some pile-driving i 18 j and not wait for an earthquake. I 19 j MR. ETHERINGTON: Or just a couple sticks of 20 dynamite at the appropriate distance. 21 MR. EBERSOLE: But you did find one point where 22 you had high acceleration on top of containment? 23 MR. CilEN Yes, but that high acceleration i 24 corresponded to 20 liertz with very low velocity and 25 displacement. ( i Aci!-Fliolinal. REI'ORTI!Rs, INC 202447-UU) Nationaide Ometage Nd M IMM6 __ . . _ . . _ _ . _ . . ~ , . _ . _ . , . _ _ . _ _ _ _ . - - _ _ . - _ _ _ _ . _ . - . _ . - _ _ - . _ _ - - , _ - - - - , - . . _ . - . - -- . . . -
25727.0 127 REE O 1 MR. SIESS: What was the frequency of those 2 angle brackets that were held -- 3 MR. CilEN: After the earthquake event, the 4 vendor, which is Kinemetrics, came in and took some tests 5 of the brackets. The conclusion at that time was around 6 100 liertz. 7 Well, there are other slides. I don't think 8 there is any need to show you here because I think you have 9 seen them 100 times. 10 MR. EBERSOLE: In view of the fact there in 11 going to be a full committee presentation by some sort of 12 l edict that I don't really understand, I guess we can cover () 13 that tomorrow. 14 f MR. CilENs Okay. 15 i Now the conclusion here we have is, again, this 16 l 1986 Ohio earthquake is of such a short duration, high 17 frequency, low velocity, low displacement and low energy 18 event it really doesn't have any engineering significance. 19 l Thank you. l 20 ' MR. EBERSOLE: Thank you. 21 MR. ETilE RINGTON : fising the same basis of 22 evaluation, how much does the addition of this carthquake 23 to the previous history of earthquakes add to the SSE 24 acceleration? 25 MR. CllENs flow does thin compare with SSE? l ACli Flini RAI. Rii1>oR'rlins, INC. I 202 W Um Nnin a Mc Cmcrue N o W, (M. L
1 1 25727.0 128 REE O \-) 1 MR. ETHERINGTON: Is it a significant addition 2 to previous history of seismicity in the region? 3 MR. EDELMAN: Would you like to respond to that, 4 Dick? Dick Holt from Western Geophysics. 5 MR. HOLT: It certainly adds to our data base of 6 eastern United States earthquakes of which we have very few. 7 We have some records from the New Brunswick series of 8 earthquakes and some from New Hampshire. This is the other 9 addition. 10 MR. EDELMAN: The inclusion of this event would i 11 ! not change our design. That is what I would like to go 1 12 through now briefly. This proof test, which may not have 13 proved everything that we might have looked for as far as equipment chatter since it was such a short duration, took 14 l 15 ! place at the plant, the structures and the systems were l 16 ! unaffected by the earthquake. It does not change any l 17 l conclusion as to the geology and seismology of our site. I 18 l The design earthquake bounds of the 1986 event and that the 19 inclusion of the event does not change the design basis; 20 that the plant seismic capability is there to accommodate 21 the January earthquake, specifically this earthquake of 22 short duration which we have covered; and that we have a 23 number of follow on confirmatory programs that we at CEI 24 are participating with the rest of the industry to gather 25 as much information as we can. O Aci!.FliorinAi. Riii>onTiins. INC. 202-347 1700 Niionwide rmerage mat tis ian
25727.0 129 REE O 1 We have worked with EPRI to gather as much data 2 as they can and worked with us and provide our data to the 3 staff and to EPRI and to USGS for additional what we think 4 might be useful to the industry. Even though it proved 5 there was no damage to our plant. We are part of the 6 seismic owners group, CEI, and we are dealing with them in 7 providing the data. They have been on site. The EPRI 8 people looking at our data, analyzing, and we are providing 9 ; all that information to them. We worked closely with all 10 these people to provide that data. We think it will 11 provide additional basis of information for the industry. 12 I am not very happy that we were the plant to do a proof ! () 13 l test 10 days prior to fuel load, but it did prove that the 14 I plant was designed and capable of doing that. j 15 i MR. EBERSOLE: What happened to the people in 16 . the control room? Did the swivel chairs roll at all and 0 17 r the emergency manuals fall out of the bookcases. 18 l MR. EDELMAN: Nothing. No, sir. None of that i 19 ! happened. 20 ) MR. EBERSOLE: You couldn't even feel it. 1 I 21 MR. EDELMAN: We looked at our control room, no 22 spurious annunciators. When I was recovery manager on site 23 that evening, I talked to the shift that actually went on, l 24 i I talked to the shift on after that. They said they 25 noticed no spurious alarms in the control room. They know O An 1 liol!RA1. Illii>on Ilins. INC.
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l 25727.0 130 ; REE l I O k- 1 the buildings shook. 2 MR. EBERSOLE: No lights flickered? I 3 MR. EDELMAN: No. We did have the alarm that l 4 went off which allowed us .o go into our emergency 5 procedures. But they reacted very well and knew what to do 6 and followed the procedures. There were no spurious 7 actions in that respect. 8 MR. KERR If you are asking was it felt, it was 9 felt. 10 MR. EDELMAN: It was felt. 11 MR. EDERSOLE: In the control room they even 12 felt it? t 13 MR. EDELMAN: Yes. It was felt.
}
14 i MR. SIESS: Could I follow up on i 15 - Mr. Etherington's questions. You have a map in here, i 16 l figure 4.3, and there is nothing in the area of the site l 17 { the size of 6 historically. 18 l MR. EDELMAN: I think I have a copy right there. 19 MR. SIESS: This shows 5 as the highest. 20 MR. EDELMAN: The site -- there were two 21 earthquakes used to bound our site consideration. One in 22 the Attica earthquake in New York and the one in the Anna 23 earthquake. 24 MR. SIESS: But in your immediate vicinity, this 25 is the highest? O Aci!.Flini!RAI. Riii>oRTiins, INC x 4:7.17m Nan.m ue nnera,e em i u, u,u,
25727.0 131 REE 1 MR. EDELMAN: Yes. 2 MR. SIESS: There would be no instrumental 3 records for any of those other earthquakes? 4 MR. EDELMAN: Not to my knowledge. 5 MR, SIESS: This just bears out my theory that 6 nuclear plants attract earthquakes. 7 (Laughter.) 8 MR. EDELMAN: I guess I will not comment on that. 9 That concludes our remarks and with some 10 direction from you or from your staff as to how much detail I 11 we covered tonight that you would like covered, it was 12 about a half hour presentation to the full committee. () 13 MR. EBERSOLE: That is about the best I can 14 define it. - l 15 l MR. EDELMAN: I don't know if the staff wishes 16 l to make any comments now. Mr. Stefano is our project i 17 l manager. 18 { MR. STEFANO: Good afternoon. I am John Stefano. ! I 19 I am the Perry project manaqer. llave been the project 20 manager since February 1982. I had the pleasure of 21 presenting the initial SSER to the SACRS Sabcommittee which 22 was headed at the time by Dr. Ray and Dr. Ebersole was on 23 that committee. 24 All I intend to do right here is to try to give l 25 l you some sort of a feeling for where the NRC staff is
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25727.0 132 REE O 1 coming from; what they are going to be doing since we have 2 received this information. The books that you have before 3 you are hot off the press. We received them the same time l 4 you did. There are additional copies here which can be 5 disseminated tomorrow if need be or anybody else you need 6 to disseminate them with today. We have additional copics 7 back at my office. 8 I would like to say that immediately upon being 9 notified of the earthquake, an inspection team was l l l 10 l immediately assembled by Region 3. They were dispatched f 11 immediately, the next day, to the plant site to begin their 1 1 12 investigation of plant damage. NRR joined them the same 13 day. I was a member of the NRR team that went out there. 3 I 14 We spent the first and second day going through the plant 15 from a preliminary standpoint. Our initial f indings in l 16 walking through the plant at that time, which was around l i 17 1 the 2nd, 1st and 2nd of February, was to confirm pretty l 18 l much what the utility found by way of minimal damage. We l 19 did see some hairline cracks in concrete, most of which I l 20 understand the utility has records of as having occurred 21 prior to the earthquake. We did confirm the leak in the l 22 flange of the hot water space heating tank which is l 23 nonsoismic supported. And I understand that certain of un 24 went out to check the circuits that tripped which were 25 designed to trip. O i Acti Fl!!)i:RAI. III:i'on II Rs, INC,
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i. 4 1 Where we are right now is in the very i
) 2 preliminary early stages of our review. We really do not i
i 3 have a position on what the utility has analyzed thus far. i 4 We will certainly use what the utility has given us. We i 5 will also use the resources we have available within 6 research, within NRR, our consultants, to do our own 7 independent assessment of what we found and to hopefully 8 present -- come to a conclusion that is similar to the 9 utility's. We recognize very well that they are in a i j 10 j near-term licensing condition and we do intend to try to i i l j 11 l support that the best we can. We want to assure you, we j . i
- j 12 are not going to rush this because they ate 10 days away l 13 from licensing.
l 14 i Essentially, the actions planned are as noted on i ; j 15 , this sheet. We are in the initial process of 1 ! 16 characterizing what that earthquake means to basically [ 4 l 17 reaffirm the seismological and geological asnumptionn that i 1 i 18 l we used in initial plant design basin, as in described in j i j 19 i the FSAR and as we have evaluated in our SER, issued in May i l 20 I have 1982. >
- 21 In terms of thin, we will also include a i
l ) 22 structural denign review where we will be comparing the l' 23 measured versus predicted responsen, the effect and impact l 24 of thin no-called short duration high f requency exceedance i 2S I to the design basis of the structures themselves. ) lO ! i 1, i i ACli l?til)liRAl l(lil'OR !!!HS, INC. l l 202 WMe Ne pin
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25727.0 134 REE 1 In addition to that, we will be looking at 2 piping and equipment which may be sensitive in this high 3 frequency region to see if there is any impacts there. 4 We have amassed a team of our research people, 5 people within NR7 -- to name a few, Dr. Leon Rider; we have 6 Mr. Bob Herman, who is heading up our mechanical equipment 7 qualification design group; we have Phyllis Sobel and we 8 have Arnold Lee, who are also, by the way, here to answer 9 any specific questions you may have. 10 All this we hope will culminate in obtaining 11 some preliminary internal findings by the 21st of this I 12 j month which will be a week from this Friday. I plan to 13 amass all of the people within NRR, NRC only, to try to 14 I come up with a picture of where we are. I 15 , I want to emphasize that the follow-up dates on l 16 s SSER issue and Perry licensing are strictly vety tough 17 estimates and targets. Those dates really can't be firmed i 18 l up until we meet on the 21st to decide where we really are 19 l on this thing. 20 As I understand it from our seismologists and 21 geologists, this is the first plant outside of Califotnia 22 where an earthquake occurred and there were instruments 23 inside that were used to record it as opposed to tryinq to 24 set charges outside someplace to nee what the response l 25 would be in the building. So we do have -- we really have ' f O Acti Flii)iil<at. Riii>onTI Rs. INc 202,147 )?m Nat h m* kk rmerage *n H6 (M6
25727.0 135 REE 1 a unique set of data here. I am not sure at this point 2 whether we really understand what it means. So I think 3 when you look at SSER data and Perry licensing date, that 4 is really at this point sort of red. 5 flowever, we do wish to indicate to the ACRS that 1 6 we will try to get your oversight of this draft SER as we 7 put it out. Ilopefully before the 7th. And we would 8 request your assistance in that regard, if you could 9 provide that to us. I 10 Let us know where we are going, is there 11 anything we are not covering, what is good and what is not 12 good before we finalize that. It is my intent to try to 13 , beat that 7/7 date with a draft SER if we can do that. 14 I One last point. Just coming here, in fact after 15 coming here we found out that there are some preliminary l 16 l unconfirmed report received from the USGS that there are 17 ! some indications - at least they think that the earthquake i I 18 may have been precipitated by an underground injection 19 plant which apparently is located somewhere between the 20 plant site and the site of the earthquake. We don't know 21 any specifics about that. As I say, this is just unconfirmed. 22 We are looking into that. As I understand it, the 23 injection is of agricultural waste which is injected quite 24 a largo depth into the ground, I understand something like 25 6000 feet below ground. So I don't know if that has any O Aci!.Fiti)iinA1. RI!I>oniI!Rs,1NC. l 202 147.17(o Nalionwide rm erag *H lI6 tMr.
l l 25727.0 t 136 REE
' 1 input at all as to what we are finding here or any measure 2 whatsoever. I just thought I would let you know that I
( 3 because we just found that out. I 4 ; Are there any questions? i i l 5 l MR. SIESS: I have got two questions. I don't l l 6 , know who can answer that. The first one is fairly simple. t 7l Was there any instrumentation in the free field? l ' 8J MR. STEAD: To the best of my knowledge, no. ! 9 ' MR. RIDER: There was a dam some seven 10 h) kilometers to the south. i Thete is a hospital in Erie, I 11 i Pennsylvania which may have recordings of that. There are ) l 12 . two records in distance which were not in the free field. ,
~
( '; 13 MR. SIESS: The seismic analysis and design, - l 14 what structures, components or systems would be affccted by ; 15 the portion of the reg guide spectrum in the range above 15 1 16 liertz? ! 17 MR. CllEN: We just found reactor building as l 18 ) second higher. I 19 : f1R . SIESS: llow much would that affect the ! j i 20 ; design? l 1 ! 21 1 MR. CilEN: As we just showed, in compatison with ! 22 the steel c'intainment, even though acceleration was 23 ! exceeded, relative displacement is below OBE value. 24 q MR. SIESS: The design of the building would be
! l 25 ! forces from mode 22, I think that is the one, the 18 Il e r t z ;
o . ( ) v . 6 l l
! i ,
Acti-I lioliRAI. IttiPouliins, INc l .msmo Nanona nle Gncrw m) W. udh
25727.0 137 REE O k 1 mode -- 2 MR. CHEN: We took the -- in the original design 3 we took the modes out to 33 Hertz. 4 MR. SIESS: How much contribution did you get 5 from the modes above -- 6 MR. CHEN: The contribution, the participation 7 factor at the mode of 18.4 Hertz is somewhat, about 8 one-third to one-half of the first one. So as to 9 contribution to which building, we haven't looked into them 10 and determined that. So the only one we have looked into 11 in detail is the containment which we showed the 12 displacement. /~h 13 d" MR. SIESS: And the piping wouldn' t be falling (_/ 14 in that range? i 15 . MR. CHEN: Piping, usually the frequency range 16 is lower. Especially for this kind of displacement, 4 17 [ usually the piping has a gap of about 1-1/16 of an inch. 18 MR. SIESS: Thank you. 19 MR. EBERSOLE: Any further questions? l 20 l MR. BENARO: I would just like to add one thing. 21 This is indeed a very valuable data point for eastern 22 earthquakes and we are trying as we go into this thing to 23 break our attention into two categories. One is Perry 24 plant-specific questioning, whether the seismic design 25 basis and structural design are appropriate for Perry. But ace FEDERAL. REPORTERS, INC 202 347 37W Nationwide roserage Nt at 31MM6
25727.0 138 REE O \~l 1 secondly, there are a lot of things that we could learn 2 from this about earthquake instrumentation or it doesn't 3 make sense to put so many frequency range meters of 4 earthquake response in a control room and have the shift 5 supervisor interpret what they mean. There are many 6 generic lessons to be learned. We are trying to not lose 7 those and I am sure the committee will be interested in 8 that for its follow-on value. 9 , MR. EBERSOLE: Would that be something that I 10 might be mentioned tomorrow? 11 MR. BENARO: I was just going to make the same l 12 point tomorrow. (} 13 MR. EBERSOLE: Well, all I can point out is that 14 tomorrow -- well, we are only 30 minutes tomorrow. From 15 what . '..e a rd , the operator deserves at least 20 of that and 16 the staf f 5 and perhaps you can cover the other 5. How is 1 17 that? 18 MR. BENARO: I would say 25 and 5. 19 MR. STEFANO: Exactly. 20 MR. EBERSOLE: That is the kind of ratio we are 21 heading for. 22 MR. HERMAN: We will take a look at the 23 equipment. 24 MR. SIESS: That was the containment building. 25 MR. CHEN: Yes, sir. l' o) ACE-FEDERAL REPORTERS, INC. 202 347-3700 Nationaide Cosetage 8m-33MM6
25727.0 139 REE O 1 MR. SIESS: To what extent do the seismically 2 calculated stresses govern the design of that construction? 3 MR. CHEN: Very small. The loss of coolant 4 pressure design, thickness of the shell -- it is a small 5 portion of that. 6 MR. EBERSOLE: Thank you. 7 There may be people who would want to go out. 8 We are going into other topics. I will just pause for a 9 minute if you want to clear. 10 (Pause.) 11 MR. GREENMAN: Good evening. I am Ed Greenman 12 from the division of reactor projects in Region 3. I have 13 with me Jeffrey Wright. The topical area on the agenda is ( 14 the Fermi 50.54 improvement program, which while again not 15 an event, was the culmination of a number of events and a 16 number of' problems at the Fermi station. I think it might 17 be helpful in the way of background to go over a very, very 18 brief history of what has transpired at the Fermi station 19 since March of last year culminating in the issuance of the ' 20 50.54 letter in late December this year. 21 The lower power license for Fermi was issued on l 22 March 20, 1985. It was a Commission briefing for the full 23 power license on July 10 of last year. And the license was 24 issued on July the 15th. 25 On July 1, the evening of July 1, into July 2, a f ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Coverage 8(xF336-(M6 __ .~. . ~ - . _ - - - - ._ . - - _ . _ - _ - . _ - - - .
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i f 25727.' 140 REE 1 premature criticality event occurred. The NRC was not 2 aware of that event at the time of the Commission briefing. j 3 Nor was it mentioned at the Commission briefing. That , ! 4 event was not significant in and of itself technically, i' 5 MR. KERR: I have heard reporta of this 6 statement before. It seems to me, in f a s t ness to the 1 7 licensee, you should point cut that although you perhaps 8 were not aware of it being a criticality event, I must say 9 I am still skeptical that the thing ever went critical. l 10 You certainly did know that an event had occurred. 11 MR. GREENMAN: That is a correct statement. 12 Nonetheless the Commission was not briefed on that event by 1 (} 13 either the NRC or the licensee at the time. 14 MR. KERR: The resident inspector was. He knew , 15 the details of the event, he knew about the rad withdrawal 16 and that there had been instrumentation observed. 17 MR. GREENMAN: As I know the Subcommittec is 18 aware, that subject is still under discussion and is part 19 of an OIA investigation internal to the staff. 20 At the time the full power license was granted, 21 the NRC's view of licensee performance was favorable. 22 There had been minimal operational-type events. There had 23 been minimal difficulties in making a transition from an 24 engineering organization into an operational phase. Since 25 the time the license was issued and subsequent to l I ACE FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Coverage 800-33MM6
4 f 25727.0 141
- REE 1 notification of the premature criticality, a confirmatory l
2 action letter was issued by Region 3 that restricted power
! 3 operation of that utility to 5 percent power. The utility 7
4 still has not operated above 5 percent power. There was no 4 J 5 operation up until a scheduled shutdown which occurred in 6 October of last year which was scheduled to do work on a 7 remote shutdown panel. That outage was originally i 8 anticipated to have been concluded late November, early 9 December of 1985 and the utility still remains in a 10 shutdown condition today. ! 11 Projected availability from the standpoint of
; 12 resolution of technical issues is now anticipated to be 13 sometime in late February or early March, dependent upon
}
14 resolution of some engineering issues which I will discuss , 15 a little bit later. 16 MR. MOELLER: When you say the utility or the 17 plant remains shut down, that is in accord with NRC 18 regulations? 19 MR. GREENMAN: They are shut down with the 20 constraints of the confirmatory action letter which 21 restricts them to 5 percent power. They are also shut down 22 within the constraints of the 50.54 approach. However, 23 there are a number of technical issues. I guess my view is 24 from what we know today, the diesel generators which we 25 briefed the Subcommittee on a month or so ago still remains ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Cmerage 800-336-6M6
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' ( 1 as a critical path issue for the utility, dependent upon 2 the resolution of the engineering issues. I 3 As I mentioned, we briefed you on the diesel 4 generators a month or so ago. Nothing really new has j 5 happened with respect to the diesel generators. They have t 4 6 embarked upon a demonstration program for the NRC involving
- 7 two of the four Fairbanks Morris diesels. They recently 8 encountered a situation where they identified a portion of 9 a scraper ring on the top of one of the crank lines, 10 reevaluated and reinspected all of the pistons and did not f
11 find any damage. We have also recently identified a case J j 12 of a viscosity change in the oil, which was a Shell product, () 13 an additive type product, where foaming was encountered in i 14 the oil. And they are involved in their test runs. 15 The utility has submitted an action plan to the 16 NRC. We have just received that within the last few days J 17 and we have not made a determination as to the 18 acceptability of that plant. ! 19 It involves a series of starts and extended runs 20 to demonstrate the reliability of the diesel generators. l [ 21 It will also involve our assessment of what types of i 22 inspections they do after the fact to assure the 23 operational capability of those diesel generators. 24 Since the plant started up, we have also
- 25 encountered cracking in the turbine bypass lines. That I
f ACE FEDERAL REPORTERS, INC. 202-347 3700 Nationwide Coverage 800-336-6M6 i
25727.0 143 , REE ' 1 piping has been replaced with nominal one-inch wall 2 thickness. NRC inspection of that area is still in 3 progress. 4 There was a major failure of one of the two 5 feedwater pumps at that utility. The south feedwater pipe, 6 it is a TDI product. It experienced excessive vibration. 7 The root cause, the licensee and we have attributed to 8 failure on the part of the engineering department to 9 translate warmup instructions into the operations procedure. 10 Also an inadequate engineering assessment of the meaning of 11 the vibration alarms and the fact that the computer that 12 monitored what the vibration alarms were telling them was 13 scaled off by about 2.5. , 14 This resulted in catastrophic bearing failures,
- 15 problems with the pedestal and a number of other issues.
16 That particular feed pump has been restored and is ready to 17 be returned to service and I think it will perform 18 acceptably. 19 There is an outstanding issue with NRR that the l 20 licensee has requested a waiver for inability to meet i 21 general design criteria 56 on the traversing in-core probe. i 22 NRR is reviewing that request. One possible solution to
- 23 that would be the installation of two ball check valves i 24 outside of containment and a request for relief to meet GEC 25 56 until the first refueling outage.
1 l ACE-FEDERAL REPORTERS, INC. 202-347-3XX) Nationwide Coverage 14X)-336 646 [
25727.0 144 REE ,T \J l One personnel error has occurred that involved a 2 rupture of the condensate storage tank up near the top of 3 that particular tank and the loss of about 100,000 gallons 4 of water. 5 The weld failure has not been repaired. The lie 6 licensee is planning on using that as is and we are still 7 in the process of evaluating the acceptability of that. 8 MR. REED: On one bullet up there you have the 9 l south reactor feed pump turbine. Is that a main feed pump i 10 turbine? l 11 MR. GREENMAN: Yes. 12 MR. REED: Then you have the condensate storage 13 l [J
't tank. Are these safety-related pieces of equipment?
14 j MR. GREENMAN: They are balance of plant
'l 15 equipment. Feedwater pumps and turbines are not 16 l i safety-related. The same thing is true of the condensate i
17 ; storage tank. 18 l MR. REED: And they could run at 60 percent 19 power wich that south reactor feed pump out of service? 20 i MR. GREENMAN: They could run at reduced power 21 on one pump. However, I would emphasize that the 22 engineering repairs and the reviews of those repairs have 23 been completed and that is simply awaiting testing now to 24 return to service. 25 The same problem was not encountered on the (O/ ACE FEDERAL REPORTERS, INC. 202-347-3700 Nationwide Coverage 800-3 % I146
25727.0 145 REE n NE north feed pump. 1 When we looked at all of the engineering 2 calculations, all of the alignments, they were all aligned 3 properly. 4 MR. REED: I am having a little problem why you 5 looked at all the engineering calculations on the pump. It 6 is not in the NRC jurisdiction, is it? 7 MR. GREENMAN: We have encountered a number of 8 problems with calculations at the Fermi station. That was 9 also the subject -- with respect to the south feedwater and 10 I pump, that was also the subject of a specific allegation 1 11 j that the NRR received from an outside source. 12 MR. REED: If it was an allegation, would it be I'l 13 in your horizon? NJ i. 14 l MR. GREENMAN: As far as a regulatory ; 15 I requirement, no, I don't have any regulatory requirement. 16 l MR. REED: I know there is a study going on or 17 't will be a study on regulatory involvement in balance of 18 plant. It seems to me that we are already involved. So 19 I why the study? 20 MR. GREENMAN: I think we are already involved. 21 I think it is also of value to note that even though that 22 is nonsafety-related, every time you have a problem with 23 components of that type, you also put yourself at risk on 24 challenges to safety systems in the plant. 25 MR. REED: If you ran at 60 percent power, one (~j',
~
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25727.0 146 REE
- s 1 boiler feed pump.
2 MR. EBERSOLE: Loss of one source of coolant 3 water, you are not supposed to run that. 4 MR. REED: I am worried about empire building 5 here. 6 MR. EBERSOLE: If you only have one feed pump 7 continuously, wouldn't you worry about that? 8 MR. REED: I can look at these as indicators 9 that they have equipment trouble on the secondary side and 10 they have equipment trouble on the safety-related side. 11 But I worry that what we are doing is creating a stairway 12 to total involvement in, unto the toilet. 13 l MR. EBERSOLE: On the other hand, the feedwater i 14 l system is an entity. I 15 g MR. REED: The reason I worry about it is l 16 - because I really think that the Nuclear Regulatory is over, 17 in over their heads a great deal right now. That if we 18 l were to settle down to the real safety-related stuff and 19 stay with it and do it right, we wouldn't have any time to 20 l be fiddling around in these other areas. 21 MR. EBERSOLE: If you did that, what you would 22 have is great safety stuff that would run every time you 23 needed it, every day. 24 MR. REED: Right. And that takes care of the 25 core and we wouldn't have any problems with the core. ACE-FEDERAL REPORTERS, INC. 202-347-37ft) Nationwide Coverage Mn33MM6
,). i 4 2- 25727.0 147 REE 1 MR. EBERSOLE: But every day is too often. } 2 MR. KERR: I am going to urge that this i 3 discussion continue. 4 MR. EBERSOLE: Thank you, sir. a 5 MR. GREENMAN: Other problems that have been 6 encountered by the utility involve either an absence of 7 sufficient documentation at the time to support the reviews 8 of environmental qualifications and absence of 9 documentation to support the adequacy of seismic reviews, 10 absence of reviews and potential problems with the 11 calculation for concrete embedments and the hangers that 12 they are hooked to, which have a potential impact on safety (} 13 systems, problems with one of the RHR pumps, which appears 14 to be in one case a manufacturing defect, which has been 15 l resolved with replacement of the motor. There is a 4 ] 16 potentially generic issue there which the NRC is reviewing 17 on whether or not the laminated material was the correct 18 application. 19 MR. EBERSOLE: You are looking at this plant ' j 20 pretty much the way Vogtle was being looked at for startup l i I 21 readiness review. j
, 22 MR. GREENMAN: Yes, sir. And in this case I i
l 23 think it has been a combination of initiation on the part l 24 of the NRC, various questions and problems identified and
,i j 25 then an approach by the licensee to resolve them.
i l ACE-FEDERAL REPORTERS, INC. { 202 347 3700 Nationwide Coverage Mxk33MM6
I 25727.0 148 REE ym 1 MR. EBERSOLE: What is wrong with the remote 2 shutdown panel? 3 MR. GREENMAN: The problem with that, that 4 installation, that is what the October shutdown was 5 scheduled to do. The startup or the completion of the 6 outage was delayed when during the design reviews it was 7 identified that there was some inadequacies in a couple of 8 , areas with separation and fire protection issues. And that 9 has been resolved and the last couple of modifications on 10 that particular issue are just undergoing a final details 11 design review 'ight r now. d 12 The most recent development is a finding that (} 13 9 occurred in late January where 7hanges, modifications, 14 l stress reports and calculations to drawings were not 3 15 !) updated since September of 1984. The impact of this is not 16 fully known at this point in time. The licensee is doing 17 an extensive e"aluation and will be meeting with the NRC 18 lI this Friday to discuss t*1e results of their evaluation. 19 It did result in an immediate impact from the 20 standpoint of the Detroit Edison Company in declaring 21 certain systems inoperable. This involved fire protection 22 systems and ECCS systems. 23 MR. MICHELSON: What is the problem on the RHR 24 pump? 25 MR. GREENMAN: The RHR pump motor problem, there ACE-FEDERAL REPORTERS, INC. 202-347 3700 Nationwide Coserage 8m3%(M6
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- I I was a seizure that broke off a couple of components in the i
1 2 pump. It was shipped off-site and a spare motor from i ( 3 Brown's Ferry was brought in. 4 MR. MICHELSON: This related to a bearing 5 clearance problem and so forth? 6 MR. GREENMAN: The initial installation, the j 7 pressure plate when it was installed in the pressure , 8 fingers were unbalanced on the Fermi pump and they broke 9 and that gave a short circuit in the motor. It is these. I
- 10 MR. MICHELSON
- The thrust bearing failed? Is 11 that what you are saying?
4 12 MR. GREENMAN: No. Not on the Fermi facility. 13 It was a problem in windings. 14 MR. MICHELSON: I am sorry. Okay. It is , , , 15- entirely a motor problem. F 16 MR. GREENMAN: Yes, sir. } 17 MR. MICHELSON: That is not generic to any other 18 plant then? 19 MR. GREENMAN: Well, the use of an epoxy might i i 20 be generic to some other facilities. And the Office of r 21 Inspection and Enforcement is looking at that right now. 22 MR. MICHELSON: I see. Thank you.
'23 MR. EBERSOLE: Is that it?
24 MR. GREENMAN: On the technical issues 25 themselves. O%) i ACE FEDERAL REPORTERS, INC. 202 347 3700 Nationside Coverage 800 3364 616
25727.0 150 REE 1 Performance issues at the utility can be 2 characterized into several broad areas. Detroit Edison has j- 3 experienced a number of personnel errors; they have had ! 4 problems with implementation and direction of their I 5 security program. While that particular topic was very 6 briefly addressed in their response to the 50.54 efforts, a 7 separate improvement program and a meeting will be held t 8 with the NRC to discuss that. So their resolution of that 9 area is not known at this point in time. 10 An inability to disposition known problems, this 11 is all part and parcel of the engineering issues that we 12 have encountered. It is taking them a long time to resolve {} 13 14 the feedwater pump problem. It is taking them a long time to resolve the diesel generator problem. It is taking them 15 h a long time to work their way out of engineering 16 calculations and design reviews to satisfy themselves if 17 all of their initial reviews were adequate. On the up side 18 of that, to date none of the engineering reviews, reviews 19 that have been done, have resulted in any situation that 20 has required any hardware change at that plant. Phrasing 21 that another way is that there was no hardware impact as 22 the result of not doing the appropriate level of review or 23 not documenting the appropriate level of review to date. 24 Ineffective communications, I think the example 25 of the feedwater pump, I think problems with communicaticns O ACE-FEDERAL REPORTERS, INC. 202-347-3700 Nation *ide Coverage 8akituM6
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i 25727.0 151 ) REE I' 1 that resulted in transfer of information with respect to 2 the premature criticality all fall into that arena.
- 3 To give you an idea of the personnel errors, we 1 4 did a review from March 20 to December 25 of the 78 5 licensee event reports. 41 of those involved personnel ,
l 6 errors. Taking two subsets of that, from July 10 through t t 7 September 10, nine of the 25 involved personnel errors; 8 from September to November 25, almost half, eight of the 17, 1 9 involved personnel errors. 1 10 Personnel errors in and of themselves on a new 1 11 plant, people don't always perform perfectly. It has 12 manifested itself at Fermi in a situation where there have : (} 13 14 been multiple violations of technical specification requirements, problems with meeting limiting conditions for 15 operations in one case it involved having an ECCS system 16 out of service. 17 MR. MICHELSON: Have you looked at other plants l 18 from the viewpoint of seeing whether this number of 1 19 personnel errors in relation to total LERs is out of normal? i 20 MR. GREENMAN: The number of personnel errors is 21 out of normal. I wouldn't be in a position to say that i
- 22 total number of LERs for that time frame is out of normal.
} 23 MR. MICHELSON: You say that personnel error 24 were out of order. I was kind of under the impression that l 25 about 40 percent of the LERs involve personnel error and i j ACE. FEDERAL REPORTERS, INC. 202 347-3700 Nationwide Coverage 8M336 6M6
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25727.0 152 4 REE ] 1 looking at your chart, about 40 percent of them involve 2 personnel error. 3 MR. GREENMAN: In some cases we have exceeded 50
- 4 percent.
i 5 MR. MICHELSON: Give or take 10 on something { 6 like that is not unusual. : } 7 MR. GREENMAN: I think that figure is a figure 8 that you have to be careful for. I say that for this 9 reason: There is a disparity in the way licensees classify 1 10 personnel errors. And -- 11 MR. MICHELSON: These are the new LERs we are j 12 dealing with here. Whose classification of personnel error (} 13 are you using? The one on the data base that NRC maintains l 14 or are you using the one off the LER? j 15 MR. GREENMAN: The one off the LER. 16 MR. MICHELSON: If you use it off the. sequence 17 coded data base, that is where the 40 percent numbers come , 4 18 from? 19 MR. GREENMAN: That is correct. 20 MR. MICHELSON: That is not unusual to see that i 21 high number of other plants. That is why I wondered why l l 22 Fermi is dif ferent. l 23 MR. GREENMAN: It is unusual when you look at 24 what the personnel errors caused. 25 MR. MICHELSON: This is just a nose count here. l i ACE FEDERAL REPORTERS, INC. 202 347 3700 Nationwide Coserage 8m 3% u46
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25727.0 ) 153 REE O 1 MR. GREENMAN: I am not trying to compare 2 utility A to utility B. 3 MR. MICHELSON: I was just trying to figure out 4 why Fermi was pointed out for this particular problem. 5 MR. GREENMAN: It is the types of personnel 6 errors. 7 Given all of this, in December of this year, the 8 NRC collective views of Region 3, NRR and IE, we issued a i 9 50.54 letter to the utility requesting additional
- 10- information on how they would address certain issues. The
! 11 basic items that that particular letter addressed, ]. 12 requested the utility-to provide information on were the a 13 adequacy of their management structures and systems to ! 14 operate the plant, the actions that they had taken and l 15 ' planned to take to insure their readiness to restart and i f 16 support power ascension to operate the plant, and what 17 short-term and long-term actions they plan to take to 18 improve not only their regulatory performance but their 19 operational performance. 20 The licensee responded formally by letter, 21 January 29 of this year, and acknowledged that there was a 22 need for improvement. 23 The basic bullets that were contained in their 24 response were as follows: The chief executive officer, 25 chairman of the board, in conversations with Region 3 and O ACE. FEDERAL REPORTERS, INC. 202- m .M 00 Nationdie cmerage ExkMu4
25727.0 154 REE O 1 in his letter, assumed the direct management and control to 2 support the restart effort. As a matter of fact, he is 3 planning as chairman of the board to spend three to four 4 mornings a week at that plant. He also directed the 1 5 president of the company to oversee the quality assurance I 6 program as another senior manager.
- 7 He tasked the organization to form an i 8 independent overview committee consisting of outside 9 consultants, including a member of the MAC organization 10 that had been on site before, to independently make 11 recommendations to him and to the board of directors. That 12 oversight committee held their first meeting and briefing
) 13 on February 7 a week ago last Friday. We are expecting to 14 have a copy of their report shortly. We don't have it at i
^ 15 this date so I don't have any information as to what the i~ 16 overview committee told the board of directors and the
- 17 chairman.
i 18 He has also established a reactors operations 19 improvement plan which contains some 60-odd different areas 20 which the utility has addressed to improve all aspects of 21 their operation. A number of those have been completed and 22 are under review by the Region 3 office. That particular 23 program is in place. 24 MR. EBERSOLE: Is there anything like a safety 25 review staff like we hear about with TVA that is O ACE-FEDERAL REPORTERS, INC. 202 347-37ft) Nationside Coserage Mk3WIM6
25727.0 155 REE \/- 1 contemplated? 2 MR. GREENMAN: Not what I would characterize as 3 a safety -- the independent oversight committee, but not a 4 safety review staff that goes beyond the normal 5 organization. 6 There have been some managerial changes that 7 have been made to strengthen their performance and there 8 are more contemplated. As an example, the utility is going 9 outside to seek a senior vice presidential level type 10 individual to move into that organization. That would !, 11 result in a pyriamid of almost two-on-one senior vice i 12 j presidents on-site reporting to the corporate office. Both (') N.s 13 f j in the engineering support as well as operations. j i 14 ! The utility is also developing a nuclear ; I 15 operations improvement plan with full implementation of 16 l that anticipated in May of this year. 17 One thing that is significant, in order to go into 18 a power escalation phase, the combination of 19 i recommendations from the on-site management organizational 20 I structures, the combinations of recommendations and changes 21 that are put in place and accepted from the overview 22 committee will all be given to the chief executive officer 23 and he will personally authorize any change in power. 24 MR. REED: I am sitting here having some deep 25 thoughts about what I see up here. The CEO of this company ACE-FEDERAL REPORTERS, INC. 202-347-37m Nationwide Cmerage am-336-6646
25727.0 156 REE ,m
~ 1 is a long time involved, old-time nuclear person. We won't 2 mention any names. That company established years ago a 3 safety review board right at the top, very top level of the 4 company and had a nationally known person who was a member 5 of the board with the CEO as president. It also had an 6 in-between plateau of safety experts and reviewers who met 7 on a regular basis. What I see up here is more of the same.
8 I would be looking for a different solution. It seems to 9 me over the years they have had all this stuff with great 10 coupling and nationally known people and so on. Maybe 11 l there is something different rather than more plateaus and J 12 l reshuffling of review boards and things like that. [ i 13 ,I I have to say that I may have a slight conflict [s'. / y 14 !t of interest here. I did serve for a short term on one of 15 !1 their boards. Then I unfortunately made the mistake of 16 joining the ACRS. So do you think this is going to d 17 r determine something or improve something? To me, I am 18 quizzical. I 19 l MR. GREENMAN: I think the fact that the Detroit i 20 f Edison Company perceives that strengthening of management 21 controls are needed is positive. I think the fact that 22 they have established an independent overview group is 23 positive. I am not of the opinion, based on the results 24 j that we have seen to date -- and part of this program is i 25 very, very new -- that the problems are solved at Fermi.
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25727.0 157 REE g> k/ 1 They have identified a number of goals. 2 MR. REED: They are the company that in addition 3 -- which I thought was way up front in management -- they 4 are the company that put all their offices on the job site 5 in a special building and only a mile away from the plant. 6 What I would think from my recollection is that this thing 7 was just overloaded with management review and attention 8 and so on. Is there something fundamental here that maybe 9 is related to their large number of LERs or human errors or 10 i something? i
! l 11 MR. KERR: Wait a minute. We don't necessarily 12 l have a large number of LERs.
13 MR. REED: The proportions are the same. [T s/ ; i 14 l MR. GREENMAN: I think I can only respond that , i 15 : given all of that, their performance has been disappointing. i i 16 MR. EBERSOLE: Can you come in on the 17 f accountability problem. Who is in charge of what? Can you 18 check back to the accountability locus in this 19 organizational structure? 20 MR. GREENMAN: I think accountability, I think 21 accountability, it is clear that accountability and where 22 accountability resta in response to the 50.54 letter, I 23 think it was clear that accountability was established 24 between engineering organizations and operational 25 organizations at the time the facility was licensed. It is O ACE-FEDERAL. REPORTERS, INC. I 202-347-37(m) Nationside Coverage k(M b 3 ?MM6 L- __
25727.0 158 REE 7-1 not clear that communications were adequate. It is not 2 clear that the organizations talked to each other properly. 3 MR. EBERSOLE: That is the perennial problem: 4 Who is in charge of that? 5 MR. REED: I just can't figure this out. These 6 people were the first people to put their total engineering 7 and offices on the job site right at the plant. All kind 1 8 of company were supposed to be there, if I was an 9 inspection person, I would be looking for some key here. 10 If they in fact have gotten more real problems than they 11 should have at this point in time. l 12 We talked about River Bend. It is a boiling f f' ; 13 lj water reactor and somebody said, they have got a lot of v 14 f problems. We tried to explain that away. But here is 15 l another boiling water reactor, perhaps a little earlier 16 vintage. I am not saying it is reactor-related. You look f 17 at the diesels, hey, we know what that is related to: 18 lousy manuf acturing and design. And that is throughout the 19 industry. Maybe if you threw away some of these 20 discardable aspects, you could get down to whether or not 21 they need to do a complete reorganization of what looked to 22 be outstanding at one time. 23 I could throw in the other thing, do they use 24 aptitude testing for their rank and file people and their 25 management? I l l l l ACE-FEDERAL REPORTERS, INC. l 202-347 370) Nationwide Coserage MUMM 6
25727.0 159 REE 1 MR. EBERSOLE: Is that it? 2 Any further comments? 3 Let me make a proposal to shorten this meeting 4 by asking Ron Hernan a question. I believe we are going to 5 be obligated to take the tech spec improvement program to 6 the full committee in any case. And that your observations 7 on it will be just as comprehensive then as they would be 8 now. 9 MR. HERNAN: The way it is, there is agreement, 10 the leader of the tech spec improvement group will be here 11 Friday with his people to make the presentation. 12 MR. MICHELSON: How much time is he going to 13 have? ( ! 14 MR. HERNAN: We would like to have a half hour i l 15 for him. j 16 MR. EBERSOLE: Since we have an hour and a half 17 anyway. { l 18 MR. MICHELSON: Are we going to get the fire i ! 19 protection questions out of the way at that time then? 20 They have questions on fire protection in tech spec changes? 4 21 MR. HERNAN: I personally don't think we are i 22 going to get into that much detail. I 23 MR. MICHELSON: Well, when are we going to hit l 24 it? i 25 MR. HERNAN: Our purpose of being here today and i ACE FEDERAL REPORTERS, INC. 202 347 37(r) Nationwide Coverage 8M.136.fM6
I 25727.0 160 REE f k# 1 Friday is to follow up on Harold Denton's comments to the 2 full committee last month. 3 MR. MICHELSON: I thought this was following up 4 on my request. 5 MR. HERNAN: Mr. Denton said that we are kind of 6 at a milestone point in the tech spec program and we will 7 come down and tell the committee. 8 MR. MICHELSON: I had some concerns on taking 9 the fire protection aspects out of the tech specs and I 10 guess putting them back in to the PSAR itself somehow, and 11 how this was going to work in view of the kind of details 12 associated with fire protection tech specs. [} 13 MR. HERNAN: We may be able to address that I 14 question fairly simply on Friday. There is some discussion 15 l in the Commission paper which we -- 16 ! MR. MICHELSON: That wasn't too hopeful. I 17 don't think they have much details understood of what a i 18 l fire protection tech spec was. 19 MR. EBERSOLE: Let me poll the full committee 20 for a second. We have an internal conference going on. 21 . MR. MICHELSON: Let's ignore them. I 22 MR. EBERSOLE: I am granting a 30-minute 23 presentation for the tech specs Friday, okay? 24 MR. HERNAN: Okay. 25 MR. EBERSOLE: We will put out a block of time G V ACE-FEDERAL REPORTLIRS, INC. 202-347 3700 Nationwide rm erage WM MW46
) t I
. 25727.0 161 l- REE
) 1 for a half an hour or so for that. There are three other l 2 topics then that I have marked which I am going to suggest 4 3 we carry to tech specs, that is McGuire, Turkey Point, ANO i'
4 1. And they will get about 10 to 15 minutes apiece. That 5 will leave ample time to cover the tech specs. l j 6 MR. REED: I thought we had McGuire, I 7 Brunswick -- ' ]' 8 MR. EBERSOLE: We have McGuire, Brunswick, and ) l, 9 ANO 1. i j 10 MR. REED: Yes. ! l 11 MR. EBERSOLE: And that is it. We are going to I 12 automatically get Perry, the earthquake, the prior day. l l ql[) 13 That ought to do it. Any further comments? No further 4 14 comments? ' j 15 MR. HERNAN: On Brunswick and ANO 1, you do not ! 16 need a staff presentation for that? 1 j 17 MR. EBERSOLE: No. I i ', 18 MR. HERNAN: So you need staff on McGuire and b - l 19 Turkey Point. i i 20 MR. EBERSOLE: Yes.
- . 21 MR. ALLISON
- No staff on ANO I?
22 MR. EBERSOLE: Let's see. That is a design 23 deficiency. I didn't say any staff support on that. That ; i j 24 is that single failure design problem. I think we can just ! t 25 get staff support on that. t i (:) , i < ACE FEDERAL REPORTERS, INC. 202-347-3700 Nanon*ide Cmcrage 800 3wua6
25727.0 162 REE f' w. 1 MR. HERNAN: We -- you decided earlier in the 2 meeting that we didn't need it. 3 MR. EBERSOLE: That is the single failure -- 4 MR. REED: What we said was, have a full 5 committee quick statement of the complexity and 6 vulnerability of auxiliary feed and whether or not maybe we 7 ought to be thinking beyond auxiliary feed. 8 MR. KERR: Do you want all this recorded? 9 MR. EBERSOLE: Off the record. 10 (Whereupon, at 6:00 p.m., the meeting was 11 concluded.) 12 13 [ 14 15 Iq 16 17 18 19 20 21 22 23 24 25 O ACE-FEDERAL. REPORTERS, INC.
! 202-347-37(x) Na:ionwide Coverage Mn 3M M-M
CERTIFICATE OF OFFICIAL REPORTER
/
C This is to certify that the attached proceedings before the UNITED STATES NUCLEAR REGULATORY COMMISSION in the matter of: NAME OF PROCEEDING: ADVISORY COMMITTEE ON REACTOR SAFEGUARDS SUBCOMMITTEE ON REACTOR OPERATIONS DOCKET NO.: PLACE: WASHINGTON, D. C. ( DATE: WEDNESDAY, FEBRUARY 12, 1986 were held as herein appears, and that this is the original transcript thereof for the file of the United States Nuclear Regulatory Commission. (sigt) /fW C'f f' s (TYPED) REBECCA E. EYSTER Official Reporter ACE-FEDERAL REPORTERS, INC. Reporter's Affiliation O
..a+>. .w..-
O . i NRR STAFF ACRS PRESENTATION T
~~ '
i i PERRY EARTHQUAKE
SUBJECT:
t January 31.1986 DATE: m O JOHN J. STEFANO PRESENTER: PROJECT MANAGER /PD#4/ DBL /NR PRESENT5R'S TITLE / BRANCH /DIV: i i 492-9473 PRESENTER'S NRC TEL. NO.: OPERATIONS l SUBCOMMITTEE: O .. E
'I
h RESULTS OF INDEPENDENT NRC WALKDOWN OF PLANT - CONFlRMS () CEl FINDINGS OF NO SIGNIFICANT PLANT DAMAGE o NRC ASSESSMENT OF EVENT IN EARLY STAGE OF REVIEW - ACJ,10NSPLANNEDARE: t
- CHARACTERIZING EARTHQUAKE TO REAFFIRM SEISMOLOGIC/
GEOLOGIC ASSUMPTIONS USED FOR PLANT DESIGN BASIS AS DESCRIBED IN FSAR/SER
- STRUCTURAL DESIGN REVIEW:
- COMPARISON OF MEASURED / PREDICTED RESPONSES
- EFFECT/ IMPACT OF SHORT DURATION /HI FREQUENCY EXCEEDANCE OF DESIGN BASIS SPECTRA ON STRUCTURES
- PIPING / EQUIPMENT DESIGN REVIEW:
- COMPARISON OF MEASURED / DESIGN BASIS SEISMIC LOADS
- EFFECT AND QUANTITATIVE ASSESSMENT OF IMPACT OF SHORT DURATION /HI FREQUENCY EXCEEDANCE ON PLANT PIPING / EQUIPMENT i
l SCHEDULED ACTIONS l.
- NRC STAFF / CONSULTANTS PRELIM. FINDINGS / DATA NEEDS }
2/21/86 j
- PERRY SPECIFIC DESIGN BASIS t
- OTHER RECOMMENDED GENERIC ACTIVITIES
- SSER ISSUED 3/7/86
- PERRY 1 LICENSING TARGET 3/14/86 Jn - . - _ _ -
t 1 PROPOSED REVISION OF 10 CFR PART 20 I
" STANDARDS F,0R PROTECTION AGAINST RADIATION" ROBERT E. ALEXANDER CHIEF, HEALTH EFFECTS AND OCCUPATIONAL RADIATION PROTECTION OFFICE OF NUCLEAR REGULATORY RESEARCH JANUARY 16,1906
-~ _ _ _ . . - - . _ _ - . ._._ _ . _ . _
0 Q REASONS FOR REVISiliG_PART 20 l
- o UPDATE PRESENT PART 20 PROMULGATED IN 1957 o
l IMPLEMENTAPPROPRIATECURRENTItECOMMENDATIONSOFICRP j o liiPLEMEiiT EPA PROPOSED FEDERAL RADIAT10ll PROTECTION GUIDANCE FOR OC EXPOSURES , o lil GENERAL -- ' t 10 ESTABLISil A SCIENTIFICALLY SOUND AND EXPLICIT llEALTil PROTECTION BASIS FOR PART 20 STANDAllDS AND OTilER NRC REGULATORY ACTIONS I i i i
- i i 2 -
Q Q . 1 i liEEDED IMPROVErIENTS IN lilE PRESENT PART 20 o UNIFORM RISK BASED SIANDAl(D F0lt DOSE a OTilER LIMITS
- o UPDATENUCLIDElilTAKELIMITS,(NPC'S) o DELETE 5(ti-18) RULE PERMITTlHG 12 REMS /YR o
INTEGRATED LIMITS FOR IfiTERNAL Ailu EXTERiiAL DOSES ESPECIALLY InPORTAllT FOR NMSS LICEllSEES
- o PROVIDE EXPLICIT DOSE LIMITS Full NEMBERS OF THE PUBLIC o
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- 6. IN ACTUAL CASES MEE [11LTIPLE ORGANS ARE IRRADIATED, FRm INTERNAL APO EXTERNAL RADIATION, TIE TOTAL EFECTIVE DOSE AND TlE RISK WILL BE LOWER UNDER ILRP-26 SYSTEM OF DOSE LIMITATION.
, f
RISK COEFFICIENTS COPPARISON - ICRP 26 RISK PER REM CANCER PDDEL USED: LINEAR ABSOLUTE PROJECTION 12.5 x 10-5 GENETIC EFFECTS (FIRST TWO GENERAT10tS) 4 x 10-5 J i PEIR Ill CANCER MODELS PESFMED: LINEAR AP60LLITE PROJECTIOff 16.7 x 10-5 LIEAR RELATIVE PROJECTION 50.1 x 10-5 LINEAR CUADRATIC ABSOLilTE PROJECTION 7.7 x 10-5 LIEAR CUADRATIC RELATIVE PROJECTION 22.6 x 10-5 . CBETIC EFFECTS (FIRST TWO GENERATI0tS) 3 x 10-5 i NOTE: ALL ITDELS ASSifE "N0 Tl!RES10LD" i .! l, i 5
Q I dRIEF_IllSIGRY OF RUIFMAKUlG i o 1977 PUBLICATION OF ICRP-26 i i o 1979 NMSS MEMO TO RES SUGGESTED REVIS10k 0F PART 20 o INTEROFFICE WORKING GROUP, SCOPED EFFORT, IDLNTIFIED ISSUES i l 0 MARCil 1980 ANPl! i i o ESTABLISill1EllT OF DRAFTlHG GROUP t i o EPA DRAFT GUIDAllCL Dil DCCUPAll0NAL EXPOSURES -- SECY-81-232 & COMMISS10d RLSP o MEETli4GS WITil INTERESIED PARTIES 4 o NCRP/ACRS RECOMEND WAITING FOR NCRP RECOMMENDAT10tlS, NEG. EDO RESPONSE o INTERiiAL NRC REVIEW, ComENT, CONCURRENCE
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i 1 4 IGP-26 SYSTEM 0F DOSE I IMITAIl0li o JUSTIFICATION OF ANY RADIATION EXPOSURE i o OPTIMlZATION OF RADIAT10f1 PROTECTlull ! o LIMITATION OF DOSES i
" EFFECTIVE DOSE E00lVALENT" CONCEPT COMBil4ES IllTERNAL AND EXTERNAL DOSES INCLUDES ALL ORGAliS, WEIGilTING FACTORS
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,. EXTERNAL OCCUPATIONAL DOSE LittlTS, M10LE BODY 1.25 REMS /0TR 3 REMS /0TR AND 5 REMS /YR OR Af;D j GWlGE._ID
- 3 REMS /0TR AND
" PLANNED SPECIAL EXPOSURES"
! 5 REMS /YR AVERAGE 5 ADDITIONAL REMS /YR i ! 5(N - 18) FORMULA IllTil SPECIAL JUSTlFICATION 25 RENS LIFETIME LIMIT FOR SUCil EXPOSURES i l' AN'lUAL LIMIT INCLUDES INTERNAL DOSE 1 0 SLIGitTLY MODIFIED ICRP-26 C0tlCEPT 1
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o o O DIllER EXTOh1AL OCCUPATI0t{AL DOSE LIf11TS EXTREMITIES: 18-3/4 RENS/0TR GIBlGE_IQ 50 REMS /YR ' SKIN: 7-1/2 RElIS/0TR OlN1GE TO 50 REMS /YR LENS: 1.25 REMS /0TR, OR ' Q!aEE_IQ 15 REf1S/YR 3 REMS /0TR AND 5 REMS /YR AVERAGE 0 EilANGES FULLY CONSISTENT lllTl! ICRP-26 TREATED AS PART OF l!!!0LE BODY l
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i 10
"9 Q Q
i INIERilAL OCCUPAU ONAL DOSE LIMIIS l ! OUARTERLY INTAKE LIMITS F0l! EACll NUCLIDE CllAN(iE_IQ AllNUAL INTAKE LIMITS FOR EACll NUCLIDE j (COMPLIANCE PREVENTS MOST llIGilLY EXPOSED (COMPLIANCE PREVENTS RISK 3 ORGAtl FROM EXCEEDING ITS OLD ICRP DOSE LIMIT) TO ALL AFFECTED ORGANS FROM EXCEEDING TilAT FROM 1
- 5 REMS /YR TO Ifil0LE BODY)
I LIMITS MUST INCLUDE l EXTERNAL DOSE O CHANGES FULLY CONSISTENT llITil ICRP-26 l f, I
~' o o o JMIERNAL OCCllPAll0Xal DOSE COMIROL 0 CHANGES WOULD PRIMARILY AFFECT NMSS LICENSEES 0 PISK FROM AIRBORNE IluCLlDES AT NPP'S IS SMALL (98% ORGAN BURDEt!S <2% OF PERMISSIBLE) 0 INTAKE LIMITS FOR SOME NUCLIDES WOULD IIE LOWERED, PRIMARILY ALP!1A EMITTERS ENC 0UNTERED IN NMSS-LICENSED FACILITIES, E.G. URANIUM - FACTOR OF 6, Til0Riufl - FACTOR OF 60, AMERICIUM - FACTOR OF 50 0 INTAKE LIMITS FOR SOME iluCLIDES WOULD DE INCREASED, PRIMARILY BETA EHITTERS 0 SPECIAL PROVISIONS IN REVISED PART 20 FOR ALPilA EMITTERS DIFFICULT TO HEASURE A HEW LittlTS (ANNUAL DOSE CONTROL RATilER TilAN 50-YEAR INTEGRATED DOSE) 0 ALL CilANGES CONSISTENT HIT!I ICRP-26 I h I . g 12
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0 OCCllPAU_0Hal ALARA CONCEPT 920.1(c) SAYS EXPOSURES O Wl6E_ID REQUIREMENT TO DEVELOP, SHOULD BE MAINTAINED ALARA DOCUf1ENT AND IMPLEMENT A RADIATION PROGRAM THAT IIICLUDES ALARA INVE.SUf2AUDN_ LEVELS CONCEPT NOT USED IN PRESENT DIANGE TQ INVESTIGATION LEVELS BELO!! PART 20 REGULATORY LIMITS REQUIRED -- SET BY LICENSEE
~ DE_I!InllilS EXPOSUPE._ LEVELS PRESENT PART 20 SETS LEVELS PROPOSED PART 20 WOULD SET A ' BELOW HHICH CERTAIN REQUIREMENTS CUT OFF LEVEL 0F 1 MREM FOR DO NOT APPLY: THESE LEVELS REIAllLSit11LAR COLLECTIVE DOSE CALCULATIONS ESTABLISH CONTROL POINTS, NOT LEVELS Bill ADIk TRIVIALITY NO DE MINIMIS LEVEL IS PROPOSED FOR INDIVIDUAL HORKER EXPOSURES UK NATIONAL RADIOLOGICAL PROTECTION BOARD Ill JANUARY 1985 ADVISED USE OF Tile FOLLOWING DE MINIMIS LEVELS: 5 MREM /YR INDIVIDUAL MEMBERS OF THE PUBLIC (1% OF ANNUAL DOSE LIMIT FOR PUBLIC) 0.511REH/YR UHERE THE INDIVIDUAL HAY BE EXPOSED TO SEVERAL "DEMINIMIS"SOURCts
o o O - PROTECTION OF THE PUBLIC RalllATION LEVELS IN UNRESTRICIflLAREAS RADIATI0fl LEVELS AND RADI0 ACTIVITY CONCEtlIRATIONS Ill UNRESTRICTED AREAS PRESENT PART 20 (TO AN INDIVIDUAL): 2 HREMS IN ANY HOUR 100 MREMS IN ANY 7 DAYS 500 MREMS IN A CALEllDAR YR ' - 500 MREMS/YR, EXPLICIT, EFFECTIVE (IMPLIED) DOSE FR0fl ALL EXTERilAL/ INTERNAL SOURCES 100 HREM/YR REFERENCE LEVEL QIANGE TO BAD 10 ACTIVITY IN EFFLUENTS PRESENT PART 20 (ANY TYPE FACILITY): PRESCRIBET C0.'lCENTRATION VALUES, AVERAGED OVER 1 YR ' f PRESCRIBED C0t:CENTRAT10N VALUES ADDITIONAL LIMITATI0f1S IF FOOD PAIllWAY INCf. EASES DAILY litTAKE COMPLIANCE HITil 110 CFR PART 190 l REQUIRED OF FUEL CYCLE /LHR LICENSEES l
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' 15
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+ O O 1 1 1 i i CONTROVERSIAL ISSUES i l IN THE PROPOSED i i 10 CFR PART 20 i PROTECTION OF EMBRYO / FETUS - DE MINIMIS LEVELS f l ANNUAL EXPOSURE REPORTS - OPTIONAL IMPLEMENTATION PERIOD l COMMITTED VS. ANNUAL DOSE i f, i 1 ,, .,
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o o O EFFECTS OF MATERNAL FACTORS ON PREGNANCY OUTCOME NUMBER OCCURRING EXCESS OCCURRENCES DUE FACTOR EFFECT FROM NATURAL CAUSEs TO MATERNAL FACTOR RADIATION RISK CHILDHOOD CANCER RADIATION DOSE OF CANCER DEATH 1200 PER MILLION 600 PER MILLION 1000 MILLIREMS RECEIVED BEFORE BIRTH ABNORHALITIEs A. RADIATION DOSE OF 1000 MILLIRADS RECEIVED DURING SPECIFIC PERIODS AFTER CONCEPTION. ' 4-7 WEEKS SMALL HEAD SIZE 40 PER THOUSAND 5 PER THOUSAND 8-11 WEEKS 40 PER THOUSAND 9 PER THOUSAND B. RADIATION DOSE OF 1000 MILLIRADS RECEIVED DURING THE FOLLOWING PERIOD AFTER CONCEPTION
- 8-15 WEEKS MENTAL RETARDATION 4 PER THOUSAND 4 PER THOUSAND 18 .
-O O O EFFECTS OF MATERNAL FACTORS ON PREGNANCY OUTCOME (CONT.)
IlUMBER OCCURRING EXCESS OCCURRENCES DUE FACTOR EFFECT FROM NATURAL CAUSES TO MATERNAL FACTOR NONRADIATION RISKS OCCUPATION WORK IN HIGH RISK STILLBIRTH OR 200 PER THOUSAND 90 PER THOUSAND OCCUPATIONS SPONTANEOUS ABORTION
- ALCOHOL CONSUMPTION 2-4 DRINKS PER DAY FETAL ALCOHOL l-2 PER THOUSAND 100 PER THOUSAND SYNDROME (TOTAL BIRTHS) '
4-10 DRINKS PER DAY FETAL ALCOHOL 1-2 PER THOUSAND 200 PER THOUSAND SYNDROME (TOTAL BIRTHS) MORE THAN 10 DRINKS PER DAY FETAL ALCOHOL l-2 PER THOUSAND 350 PER THOUSAND (CHRONIC ALCOHOLIC) SYNDROME (TOTAL BIRTHS) MORE THAN 10 DRINKS PER DAY PERINATAL 23 PER THOUSAND 170 PER THOUSAND (CHRONIC ALCOHOLIC) 19
o o O 1 EFFECTS OF MATERNAL FACTORS ON PRFGNANCY OUTCOME (couT.) NUMBER OCCURRING EXCESS OCCURRENCES DUE FACTOR ErrECT l FRoa NATURAL [AUSEs TO MATERNAL FACTOR - SMOKING I - 3ABIES AT BIRTH TEND - TO WEIGH'LESS THAN i BABIES OF NON SMOKERS t 4 LESS THAN 1 PACK PER DAY PERINATAL INFANT 23 PER THOUSAND l 5 PER THOUSAND BIRTHS j DEATH j ONE PACK OR MORE PER DAY PERINATAL INFANT 23 PER THOUSAND 10 PER THOUSAND BIRTHS I
, DEATH l
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O _ O O . PROTECTION OF EMBRY0/ FETUS PRINCIPAL ALTERNATIVES INFORMED CONSENT (PRESENT I;RC POSITION) ONUS FOR PROTECTION ON WORKER LOWER DOSE LIMIT FOR ALL WORKERS HIGHER COLLECTIVE DOSES AND COSTS LOWER DOSE LIMIT FOR ALL WOMEN JOB DISCRIMINATION, INAPPROPRIATE LIMIT FOR SOME WOMEN LOWER DOSE LIMIT FOR FERTILE WOMEN JOB DISCRIMINATION, INVASION OF PRIVACY LOWER DOSE LIMIi FOR WOMEN KNOWN TO BE PREGNANT JOB DISCRIMINATION, INVASION OF PRIVACY, MODERATE OVEREXPOSURE POTENTIAL ONUS FOR PROTECTION ON EMPLOYER LOWER DOSE LIMIT FOR WOMEN DECLARED TO BE PREGNANT JOB DISCRIMINATION, HIGHER OVEREXPOSURE POTENTIAL ONUS FOR DECLARATION ON WORKER EPA IS PROPOSING THE FINAL ALTERNATIVE IN NEW GUIDANCE TO FEDERAL AGENCIES s 21
~~
~o O O PROPOSED EMBRYO / FETUS LIMIT PROVIDES DOSE LIMIT OF 0.5 REM TO EMBRYO / FETUS FOR ENTIRE GESTATION PERIOD.
DECLARATION OF PREGNANCY WOULD BE VOLUNTARY. ONCE DECLARED, THE RESPONSIBILI'TY OF MEETING LIMIT WOULD SHIFT TO LICENSEE. AN ADDITIONAL 0.05 REM TO THE EMBRYO / FETUS PERMITTED IF EMBRYO / FETUS LIMIT HAS BEEN EXCEEDED BEFORE DECLARATION OF PREGNANCY. WOMEN WHO CHOSE NOT TO DECLARE THEIR PREGNANCIES WILL BE SUBJECT TO THE SAME OCCUPATIONAL LIMITS AS ANY OTHER NUCLEAR WORKER. CONSISTENT WITH EPA PROPOSED FEDERAL GUIDELINES, NCHP, ICRP, PREVIOUS COMMISSION GUIDANCE AND OTHER FEDERAL AGENCIES. I, 22 ,
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~~ O . DE MINIMIS LEVELS IN UNITED KINGDOM 5 MREMS/YR, INDIVIDUAL DOSE FROM ALL NON-MEDICAL, CONTROLLED SOURCES (RISK OF /10-6 ya, naa74t:7y) 0.5 MREM /YR, INDIVIDUAL DOSE FROM ANY ONE NON-MEDICAL, CONTROLLED SOURCE INDIVIDUAL DOSES LESS THAN 0.5 MREM /YR NOT INCLUDED IN COLLECTIVE DOSE ESTIMATES APPROVAL PURPOSES IF COLLECTIVE DOSE IS LESS THAN 100 PERSONREMS THESE LEVELS WILL AFFECT: GOVERNMENT APPROVALS - IHE DEGREE OF REASON EMPLOYED IN RADIATION PROTECTION DESIGN AND PLANNING - UNNECESSARY EXPENDITURE OF RESOURCES OPERATIONS - DECISIONS OTHERWISE AFFECTED UNDULY BY IRIVIAL CONCERNS WASTE MANAGEMENT - EXAGGERATED PROTECTIVE MEASURES THAT EXACERBATE OTHER HAZARDS DECOMMISSIONING - PusLIC UNDERSTANDING THAT RADIATION RISK IS DOSE-DEPENDENT PROPOSED RULE IN PART 20 WOULD ALLOW A 1 MREM /YR CUT OFF IN COLLECTIVE DOSE CA DE MINIMUS LEVEL) 25
" ~~ O O O QUASI DE MINIMIS RULE IN PROPOSED 10 CFR PART 20 1 MREM /YR CUT OFF IN COLLECTIVE DOSE CALCULATIONS MADE TO OBTAIN NRC APPROV (COMPARABLE UK LEVEL IS 0.5 MREM /YR) WOULD ELIMINATE CONSIDERATION OF VERY LOW DOSES DELIVERED AT VERY LOW DOSE RATES OVER EXTREMELY LONG PERIODS OF TIME TO EXTREMELY LARGE NUMBERS OF PEOPLE COLLECTIVE DOSES CALCULATED FROM INDIVIDUAL DOSES LESS THAN 1 MREM /YR ARE A HIGH PERCENTAGE OF THE TOTAL RISKS ASSOCIATED WITH 1 MREM /YR OR LESS SHOULD BE OF NO CONCERN TO THE EXPOSED INDIVIDUAL (ONE-FIFTH OF THE COMPARABLE LEVEL ADOPTED IN UK), ARE TRIVIAL WITH RESPECT TO REGULATORY CONCERN, AND SHOULD NOT AFFECT DECISION-MAKING PROCESSES IF A LONG-LIVED RADIONUCLIDE WERE RELEASED SUCH THAT EVERY US CITIZEN CONTINUO - RECEIVED 1 MREM /YR (WORST CASE), THE THEORETICAL NUMBER OF PREMATURE DEATHS PER YEAR WOULD BE ABOUT 21 IN A POPULATION OF 220,000,000 PEOPLE. THE US DEATH RATE FROM CANCER IS ABOUT 400,000 PER YEAR AND FROM ALL CAUSES IS { ABOUT 2,000,000 PER YEAR. f I 26 .
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~O o O APPLICATION OF DE MINIMIS CONCEPT TO INDIVIDUALS WCULD ESTABLISH ANNUAL DOSE LEVEL SO SMALL THAT INDIVIDUALS USUALLY WOULD NOT CONSID THE RISK IN ARRIVING AT DECISIONS REGARDING TilEIR ACTIONS.
APPLICATI0flS WOULD BE MUCil BRA 0 DER TilAN CALCULATIONAL " CUT 0FF". NO REGULATORY ACTIONS WOULD BE IMPOSED FOR DOSES BELOW THIS LEVEL. COST REDUCTIONS WOULD BE APPRECIABLE (E.G., RADWASTE DISPOSAL, EFFLUENT PROCESSING). WAS CONSIDERED DURING DEVELOPMENT OF PROPOSED RULE. COULD RESULT IN: UNAPPROVED INCORPORATION OF RADI0 ACTIVE MATERIALS INTO CONSUfER PRODUCTS. RELEASE OF VERY LOW-LEVEL WASTE STREAMS WITHOUT EVALUATION. RAISING LEVELS, AS IN NPP TECilNICAL SPECIFICATIONS, AT WHICH CERTAIN CONTROLS OR EQUIPfENT MUST BE INSTALLED OR OPERATED.
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SUPPLEMENTAL INFORMATION OF PROPOSED RULE DISCUSSES AND SPECIFICALLY REQUESTS COMMENT r 3 9 .
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- Sir.CIFICAI.LY IFulEST 1111 tlc CGIGT ON lillS ISSlE Af0 (E A lf MINIMIS lEVFL RU lilllVilMJALS.
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O O O - EXPOSURE REPORTS Tile NATIONAL INTEREST THE FEDERAL G0vERNMENT IS EXPECTED TO BE INFORMED, AND TO DISSEMINATE INFORMATION, REGARDING HEALTH AND SAFETY IN THE WORKPLACE DOL BUREAU OF LABOR STATISTICS (RADIATION EXPOSURE NOT INCLUDED) EXPOSURE DATA ARE VITAL IN, PLANNING FOR THE ADMINISTRATION OF JUSTICE (WORKMANS COMPENSATION, IORT LAW) NRC INTERESTS NRC IS PRIMARY SOURCE OF INFORMATION FOR CONGRESS, ADMINISTRATION, OTHER FEDERAL AGENCIES, STATE 6OVERNMENTS, LABOR DNIONS, SCIENTIFIC AND INDUSTRIAL ORGANIZATIONS, SPECIAL INTEREST GROUPS, NEWS MEDIA, UNITED NATIONS, FOREIGN GOVERNMENTS, AMONG OTHERS DATA BASE FOR DECISIONS REGARDING: BUDGET NEEDS; RESEARCH AND STANDARDS DEVELOPMENT PRIORITIES; EVALUATING LICENSEE PERFORMANCE; AREAS TO EMPHASIZE IN LICENSINGJ INSPECTION AND ENFORCEMENT PRIORITIES DATA BASE FOR FACTORING WORKER RISKS INTo DECISIONS ON PLANT SAFETY REQUIREMENTS DATA BASE FOR CONTROLLING POTENTIAL TRANSIENT WORKER PROBLEMS TRIGGERING IIMELY CORRECTIVE ACTION DATA BASE FOR EVALUATION OF RADIOLOGICAL RISKS IN NRC-LICENSED ACTIVITIES AND LICENSEE PERFORMANCE 31
o O O EXPOSURE REPORTS... CONTINUED INDUSTRY INTEREST UNBIASED GOVERNMENT DATA ARE USED BY INDUSTRY TO VERIFY THAT SAFE WORKING CONDITIONS ARE MAINTAINED DATA ARE USED BY INDUSTRY TO IDENTIFY ITS STRONG AND WEAK PERFORMERS AND TO EFFECT IMPROVEMENTS WORKER I!1TEREST ASSURANCE THAT llEALTH PROTECTION IS ADEQUATE, AS SUPERVISED BY GOVERNMENT CONFIDENCE THAT ADEQUATE LEGAL RECORDS ARE AVAILABLE IF NEEDED ! GOVERNMENT IERMINATION OF UNSAFE CONDITIONS i;
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OPTIONAL IMPIFMENTATION PERIOD i ISSUE: SHOULD THERE BE A TRANSITION PERIOD, FOLLOWING PUBLICATION OF THE NEW PART 20 AS A FINAL RULE, DURING WHICH COMPLIANCE WITH EITHER THE PRESENT OR NEW PART 20 WOULD BE ACCEPTABLE? -
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COMMITTED VS ANNUAL DOSE , ISSUE: SHOULD CONTROL OF LONG-LIVED RADIDACTIVE MATERIAL INTAKES BE BASED ON THE DOSE ACTUALLY RECEIVED EACH YEAR OR ON THE DOSE INTEGRATED OVER A PERIOD OF 50 YEARS? l l _i
o 9 O - COMMITTED VS ANNUAL DOSE FOR CONTROL 0F LONG-LIVED AIRBORNE RADIONUCLIDES BY FEDERAL AGENCIES CONTROVERSY HIGHLIGHTTED BY EPA / DOE /NRC REACTION TO ICRP-26 ICRP HAS ALWAYS USED COMMITTED, DOSE TO INTERNAL ORGANS (50-YR INTEGRATED D INCLUDING ICRP-26 AEC-REG./NRC HAVE ALWAYS USED COMMITTED DOSE l AEC/ERDA/D0E HAVE USED COMBINATION OF COMMITTED AND ANNUAL DOSE EPA, IN NEW GUIDANCE TO FEDERAL AGENCIES, IS PROPOSING COMBINATION OF COMMITTED j AND ANNUAL DOSE 10 CFR PART 20 REVISION WOULD USE COMBINATION OF COMMITTED AND ANN FEDERAL AGENCIES ARE IN AGREEMENT ON THIS ISSUE THE DEGREE OF PROTECTION PROVIDED FOR NRC-LICENSEE AND DOE-CONTR WOULD CONTINUE TO BE VIRTUALLY THE SAME. e t I I 41
o o o COMMITTED VS ANNUAL DOSE CONTROVERSY IN HEALTH PHYSICS COMMUNITY THIS ISSUE INVOLVES CAREER INTE'RFERENCE, INTERNAL DOSE RECORDS AND REPORTS TO EMPLOYEES IN THE MANAGEMENT OF LARGE DEPOSITIONSJ NOT COVERED IN FEDERAL GUIDANCE. 1 HEALTH PHYSICS COMMUNITY DIVIDED ON THIS ISSUE. I ISSUE WILL BE VOTED ON BY THE HEALTH PHYSICS SOCIETY -- THE FIRST TIME VOTING OF THIS TYPE HAS BEEN HELD. THE RESULTS OF THIS VOTE WILL BE INFLUENTIAL ON THE STAFF AS PUBLIC COMMENTS ON 10 CFR PART 20 ARE ANALYZED. i i
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CAREER INTERFERENCE: RECORDING: REPORTING A WORKER ACCIDENTALLY RECEIVING A SUFFICIENTLY LARGE INTAKE WOULD RECEIVE AN EFFECTIVE DOSE EQUIVALENT LARGER THAN THE 5 REMS PER YEAR LIMIT THE REST OF HIS/HER LIFE. SINCE THE RISK FROM THIS DEPOSITION WOULD NOT BE ALTERED BY SUBSEQUENT INTAKE OR EXTERNAL EXPOSURE LIMITATIONp, WHILE CAREER INTERFERENCE COULD IMPOSE SEVERE ECONOMIC (AND OTHER) PENALTIES, PRESENT NRC REGULATIONS (AND THE NEw PART 20) WOULD, AT THE BEGINNING OF THE NEXT QUARTER (YEAR) DISREGARD THE PRESENCE OF THIS DEPOSITION. THE LICENSEE WOULD BE REQUIRED TO RECORD THE INTAKE AND AN ESTIMATE OF THE COMMITTED DOSE EQUIVALENT (EFFECTIVE COMMITTED DOSE EQUIVALENT AFTER REVISED PART 20 BECOMES EFFECTIVE). THE LCIENSEE WOULD BE REQUIRED TO REPORT THIS INFORMATION TO THE NRC AND TO THE WORKER. FOR CERTAIN NUCLIDES DIFFICULT TO MEASURE IN QUANTITIES ASSOCIATED WITH THE DOSE COMMITMENT, RECORDING AND REPORTING OF THE ANNUAL EFFECTIVE DOSE WOULD BE ALLOWED BY THE NEW PART 20, AND LARGER INTAKES WOULD BE ALLOWED UNDER PRESCRIBED CONDITIONS. INDIVIDUAL DETERMINATIONS WILL BE CONTINUED FOR DOE WORKERS SO EXPOSED, INCLUDING COMPENSATORY DOSE LIMITATIONS IF CONSIDERED NECESSARY BY DOE MEDICAL AUTHORITIES. DOE WILL CONTINUE TO RECORD AND REPORT ANNUAL DOSES. I rc
o '. 9 e S S C E I S E I S V N E E R C I 0 L 2 D T E R Y A E P V R R U F S C T 0 A l i 1 T T S T Gj O S T l C O C U N L C A R A L T E I A N o M F l P i I r T I N I U N M O C D E R F O O S 0 0 N 0 ,0 O S
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PERSPECTIVE ON ESTIMATED COSTS o Mlitil 0F ECONOMIC IMPACT TilAT IS ASSIGNED TO PROPOSED REVISION OF PART 20 IIAS BEEN, IS BEING, OR WILL BE COMMITTED WHETilER OR NOT REVISION IS PROMULGATED. i o MANY LICENSEES VOLUllTARILY IMPLEMENTING RECOMMENDATIONS OF ICRP 26 AND 30, BECAUSE RECOGNIZED AS " GOOD PRACTICE" AND LIKELY TO BE IlELPFUL IN MITIGA LIABILITY CLAIMS o COST OF SPECIAL EQUIPMENT, SUCil AS LUNG C00flTERS, PROCESS CilANGES, ROBOTICS, 4 AND OTilER MAJOR MODERNIZATION ACTIVITIES, llAVE AND WILL BE "CilARGED" TO Tile PART 20 REVIS10ll, ALTil00Gil IllEY ARE NOT REQUIRED AllD WOULD BE If1 CURRED AllYWAY FOR OTilER REASONS i o ESTinATED COSIS TAKE NO CREDIT FOR SAVINGS FROM CllANGES IN TECll SPECS, LIC. CONDITIONS, ETC. TilAT COULD RESULT FROM PROMULGATION OF PROPOSED REVISION l
's 45
j BENEFITS OF PROPOSED REVJM0!i t REGui.ATION WILL REFLECT ICRP C0llERENT RISK-BASED SYSTEM AND USE WIDELY-I ACCEPTED CONTEMPORARY SCIENTIFIC KNOWLEDGE o AtlllUAL AND LIFETIME DOSES TO WORKERS RECEIVING llIGilEST EXPOSURES & URAll. MILLS & FUEL FABR. WILL BE l(EDUCED 3 o illLL PROVIDE METil0D FOR SUMMIilG EXTERNAL AND INTERNAL EXPOSURES -- ESPECIALLY IMPORTANT FOR SOME HMSS LICENSED ACTIVITIES l 0 PUBLIC DOSE LIMITS ARE CLEARLY IDEllIIFIED o WORKERS AND PUBLIC Sil00LD BETTER UNDERSTAND llEALTil RISK BASE AND PR PROVIDED o CUT 0FF ON COLLECTIVE DOSE EVALUATIONS WOULD ELIMINATE CONSIDERATIO INSIGillFICANT llEALTil RISKS o IMPROVES REQUIREliENIS UN RADIATION SAFETY; E.G., TO PREVEf!T ACCESS TO VERY l11611 RADIATION AREAS, POSTING OF AREAS USED FOR MEDICAL RADIATI0il TREATMENTS, ! REQUIRED APPLICAT10tl 0F ALARA o WOULD INTRODUCE SI (METRIC) RADIATION UNITS INTO NRC REGS i
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'o o o SIAFF REC 0lHENDAUDNS o CONCURRED IN BY ALL OFFICES IhVOLVED o
PROPOSEEXTENDEDIMPLEMENTATI0hPEltl0DOF5YEARSFROMPUBLICATIONOF FINAL RULE o PUBLISH PROPOSED REVISION OF PART 20 FOR PUBLIC COMMENT o SPECIFICALLY REQUEST COMMENTS ON CONTROVERSIAL ISSUES o ALLOW EXTENDLD PERIOD (120 DAYS) FOR COMMENT I i
Agenda for ACRS Subcommittee w Meeting on February 12, 1986 1:30 p.m. Room 1046, H Street RECENT SIGNIFICANT EVENTS Presenter / Office Date Plant Event telephone h 11/3/85 McGuire 1 Start-Up with Degraded HPSI J. Giitter, IE 2. System 492-9001 4 1/7/86 Brunswick 2 Target Rock Two-Stage SRV M. Wegner, IE Y Setpoint Drift 492-4511 1/7/86 Turkey Point Stop Check Valve Failures R. Kiessel, IE I 492-8119 1/9/86 Palo Verde 1 Reactor Trip E. Licitra, NRR 8 492-8599 1/9/86 Palisades Loss of Offsite Power J. Giitter, IE /O 492-9001 1/28/86 Robinson Loss of Offsite Power G. Requa, NRR 492-9798
/3 1/14/86 ANO 1 Design Deficiency in Emergency G. Vissing, NRR /Y Feedwater System 492-8796 1/29/86 Riverbend Review Of Start-Up Test Program E. Weiss, IE /8
- 492-9005 1/31/86 Perry Earthquake J. Stefano, NRR .2 9 492-9473 Fermi 50.54(f) Improvement Program E. Greenman, Reg 3 (
312-790-5518 puf Technical Specification R. Hernan, NRR 2/ Improvement Program 492-95]9 l i J 4 i L
McGUIRE UNIT 1 - START-UP WITH DEGRADED HPSI SYSTEM . NOVEMBER 3, 1985 (J. GilTTER, IE) 1
- PROBLEM: FAILURE TO REPAIR VCT ISOLATION VALVE MOTOR OPERATORS PRIOR TO START-UP WOULD HAVE PREVENTED VCT ISOLATION ON Sl SIGNAL.
SAFETY SIGNIFICANCE: HPSI MAY NOT HAVE FUNCTIONED AS REQUIRED ON A REAL DEMAND. DISCUSSION:
- AT 0640 ON NOVEMBER 2, A LOSS OF INST. AIR (SHARED BY BOTH '
UNITS) CAUSED BOTH UNITS TO TRIP AND SAFETY INJECTION IN UNIT-1. VCT ISOL. VALVES CLOSED AS REQUIRED; HOWEVER, THE VALVE MOTOR OPERATORS WERE LATER FOUND BURNED OUT.
- PRIOR TO START-UP, OPERATORS MANUALLY OPENED VALVES, BUT DID NOT REPAIR THE MOTOR OPERATORS.
UNIT START-UP COMMENCED AT ABOUT 0600 ON 11/3. THE UNIT WAS IN MODE 2 FOR ABOUT 6 HOURS. ("N - TWO MAJOR CONCERNS 1) IS BORON CONC. LESS THAN THAT ASSUMED
\
IN SAFETY ANALYSIS? AND 2) COULD VCT COVER (Hz AND Nz) BECOME ENTRAINED IN THE CHARGING PUMP SUCTION PATH RESULTING IN GAS BINDING OF PUMPS? NRC (OP CENTER)'WAS NOTIFIED AT 1218 ON 1/14/86. FOLLOW-UP: VALVE OPERATOR REPAIRED. REVIEWED DESIGN OF OTHER MOTOR OPERATED VALVES IN PLANT. PERFORMED TEST TO DETERMINE TIME REQUIRED FOR OPERATOR TO CLOSE ISOLATION VALVES. TIME REQUIRED WAS 19.4 MINUTES. AT 20 MINUTES VCT LEVEL WAS AT 20%. A U 1
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OS ' BRUNSWICK 2 TARGET ROCK TWO-STAGE SRV SETPOINT DRIFT, 1/7/86, MARY S. WEGNER 1/7/86 CPal REPORTED 6/11 BSEP 1 VALVES LIFTED AT 200 PSIG OVER SETPOINT AT WYLE LAB HISTORY THREE STAGE VALVE SPURIOUS OPENINGS AND FAILURE TO RESEAT TWO STAGE REPLACEMENT BEGUN 1978 - HATCH 1 HAD 11 0F 11 FAILED TO OPEN AT SETPOINT OWNERS GROUP FORMED TO INVESTIGATE PROBLEM, POSE SOLUTION (]) . CAUSE ATTRIBUTED TO LABYRINTH SEAL GALLING AND SEAT-TO-DISC BONDING ENHANCED MAINTENANCE SUGGESTED RECURRING PROBLEMS NRR LETTER TO OWNERS GROUP SEAT REPLACEFENT PROGRAM l
- THE BRUNSWICK INCIDENT WYLE TESTING OBSERVATIONS AT WYLE HATCH 1 RESULTS l
MILLSTONE 1 RESULTS GENERIC ISSUE B55
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I TURKEY POINT UNITS 3 AND 4 - STOP CHECK VALVE FAILURES NOV. 1985 IHRU JAN. 1986 - (R. KIESSEL, IE) O AFW STOP CHECK VALVE GUIDE PIN FAILURES PROBLEM: SAFETY SIGNIFICANCE: AFW VALVE / SYSTEM CONCERNS AND GENERIC IMPLICATIONS OF LOW FLOW THROUGH CHECK VALVES CIRCUMSTANCES: BETWEEN NOV. 1985 AND JAN. 1986, NUMEROUS FAILURES OF STOP CHECK VALVES IN STEAM SUPPLY SYSTEM TO AFW PUMPS STOP CHECK VALVES LOCATED UPSTREAM AND DOWNSTREAM 0F MOTOR OPERATED VALVES THAT OPENS WHEN REQUIRED TO INITIATE FLOW STOP CHECK NORMALLY OPEN - PREVENTS BACKFLOW IN EVENT OF STEAM LINE BREAK MODE OF FAILURE - DEGRADATION OF DISC AND DISC NUT FAILURE DUE TO LOW STEAM FLOW CONDITIONS CAUSED BY SLIGHT LEAKAGE OF NORMALLY CLOSED MOV O - t0W Ft0W CAUSED VIBRATION AND CHAT 1eaiNG BReAxiNG DISC guide FROM DISC , LOOSE DISC GUIDE PREVENTED FULL CLOSURE AND FULL OPENING OF VALVE. ALSO FREE TO TRAVEL
- FOLLOWUP:
AFW SYSTEM WAS INSPECTED ALL MISSING GUIDE PINS LOCATED AND REMOVED l
- l ALL VALVES REPAIRED WITH HIGHER STRENGTH MATERIAL USED IN DISC GUIDE l
FAILURE ANALYSIS AND METALLURGICAL EXAMINATIONS PERFORMED LICENSEE COMMITTED TO REGULAR RADIOGRAPHIC EXAMINATION OF VALVES I LICENSEE CONSIDERS THIS AS INTERIM REPAIR PENDING COMPLETION ! 0F THE STUDY BY ITS AFW ENHANCEMENT TASK FORCE INFORMATION NOTICE 86-09, " FAILURE OF CHECK AND STOP CHECK O vAtveS SUB;eCreD T0 t0w Ft0w CONDITIONS ~ wAS issued FEBRUARY 3, 1986 - l 5
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O , PALO VERDE UNIT 1 - COMPLICATIONS TUEEUWTNU TURETNE TRTF TEKT--- FRUR TUUY FUWER JANUARY 9, TVE6 - (E. ETflTRA, NRR) l 't CIRCUMSTANCES:
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TURBINE TRIP TEST STARTED FROM 100% (2250 PSI'AND 565*F COLD LEG) AT APPP.0XIMATELY 13:25 NON 1E HOUSE LOADS FAILED TO TRANSFER TO 0FFSITE POWER DUE TO FREQUENCY MISMATCH BETWEEN GRID AND NON 1E BUSES POWER LOSS AFFECTED RCS PUMPS, MFW PUMPS, CIRCULATING WATER PUMPS AND STEAM DUMP / BYPASS CONTROL SYSTEM 3 REACTOR TRIP DUE TO FLOW PROJECTED LOW DNBR ! STEAM BYPASS CONTROL SYSTEM (SBCS) VALVES RECEIVED QUICK l OPEN SIGNAL BUT RECLOSED ALMOST IMMEDIATELY (DUE TO LOSS l OF POWER)
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- ONE OF THE FOUR LOWEST SETTING (1250 PSIG) MAIN STEAM SAFETY VALVES LIFTED WHEN SG PRESSURE INCREASED RAPIDLY - STAYED OPEN FOR 43 SECONDS WHEN SAFETY VALVE RESEATED, FIVE SBCS' VALVES MODULATED OPEN-STAYEDOPENFOR45SECONDSWITdTWOADDITIONAL OPEN/CLOSE CYCLES DURJNG NEXT 40 SECOND3 MANUAL CONTROL TAKEN OF INDIVIDUAL SBCS VALVES TO STOP COOL-DOWN - ALSO MSIS OCCURRED DUE TO LOW SG PRESSURE AFTER STABILIZATION, OPERATOR RESTABLISHED C0CLDOWN VIA ONE SG, TWO ADVs AND ONE AFW PUMP l
l POWER RESTORED TO NON 1E BUSES AT 13:28 (3 MINUTES AFTER START 0F EVENT) l RCS PUMP FLOW RESTORED AT 14:08 EVENT TERMINATED AT 14:49 Er l L _ ._
FOLLOW UP: CONSIDERING DESIGN CHANGE TO SYNCHRONIZATION CHECK RELAY ASSOCIATED WITH AUTO TRANSFER MAIN STEAM SAFETY VALVE SETTINGS CHECKED AND FOUND OK CONSIDERING UNINTERUPTABLE POWER TO STEAM DUMP / BYPASS CONTROL SYSTEM RCS PUMP COASTDOWN FOUND FASTER THAN DESIGN - ONLY AFFECTS LOSS-0F-FLOW SAFETY ANALYSIS COLSS AND CPC PENALTIES IMPOSED WHILE CE EVALUATES FAST C0ASTDOWN EFFECT i
- TURBINE TRIP TEST SUCESSFULLY RERUN ON JAN 24, 1986 (NON 1E LOADS POWERED OFF OF STARTUP TRANSFORMER) 4 l
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PALISADES - LOSS OF OFF-SITE POWER JANUARY 9, 1986 (J. G. GilTTER) PROBLEM LOSS OF OFF-SITE POWER WITH 1 DIESEL-GENERATOR OUT OF SERVICE. VOLUNTARY DISCONNECT FROM GRID DUE TO STEAM OBSERVED FROM 4160V SYSTEM. SAFETY SIGNIFICANCE: LOSS OF OFF-SITE POWER WITH 1 OF 2 D/Gs OUT OF SERVICE (TORN DOWN); VESSEL DRAINED DOWN TO APPROXIMATELY FLANGE LEVEL; HEAD NOT TENSIONED. CIRCUMSTANCES: INITIAL CONDITIONS: UNIT HAD BEEN IN COLD SHUTDOWN FOR REFUELING SINCE NOVEMBER 30, 1985 (40 DAYS). HEAD WAS POSITIONED ON VESSEL BUT NOT TENSIONED. VESSEL ( WAS DRAINED TO APPROXIMATELY FLANGE LEVEL (%12 FT ABOVE FUEL). D/G l-1 WAS OUT OF SERVICE (WOULD HAVE REQUIRED SEVERAL HOURS TO MAKE OPERABLE). SEQUENCE: (TIMES ARE PM, EST) 2:41 " SMOKE" (STEAM) OBSERVED FROM CONDUlT ON TEE 1A BUS. OPERATORS IMMEDIATELY DE-ENERGl2ED NON-VITAL 4160V BUS 1A. IT WAS DECIDED TO DE-ENERGlZE THE R 345 KV SWITCH YARD (S/Y) BUS. (DE-ENERGlZING ALL 3 S/U XFMRS.) 2: 56 - OPERATORS STARTED AND LOADED D/G l-2 ON THE ID BUS. 3: 08 - R 345 KV S/Y BUS DE-ENERGlZED. 1B, (IE NON-VITAL BUSES) LOST- 1C VITAL BUS LOST. 1D POWERED FROM l-2 D/G, UNUSUAL EVENT (GE) DECL'D AT 3:00 AND NRC OP CENTER N NOTIFIED AT 3: 19.
/0
2-O - 4: 50 - BACKFEED ESTABLISHED TO ALL BUSES THRU MAIN TRANSFORMER AND STATION POWER TRANSFORMER. UE TERMINATED AT 5: 00. FOLLOW UP: CABLE BETWEEN 4160 V BUS 1A AND STARTUP TRANSFORMER 1-2 WAS REPLACED PRIOR TO ELECTRICAL REAL!GNMENT. 2 O O
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l i O SIMPLIFIED ELECTRICAL DISTRIBUTION DIAGRAM FOR PALISADES I 345 KV BUS "f" 345 KY BUS "R"
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MAIN XfMR l wW WW WW ww Mm S/U m m S/U M m S/U m m XfMR 1-1 XEMR 1-2 XEMR 1-3 l i l' pwR _ l _ l XFMR 4160 v Bus 1 A 4160 v Bus 1B 1 1 a , , , M AIN GEN ? l El ' E WI - E MmW l 2400 v Bus 1C 2400 v Bus ID 2400 v Bus IE DEDG 1-1 D' EDG 1-2 O
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H. B ROBINSON UNIT 2 - REACTOR TRIP WITH LOSS OF 0FFSITE POWER JANUARY 28, 1986 (G. REQUA, NRPJ PROBLEM : LOSS OF 0FFSITE POWER (LOOP) WITH "B" EDG OUT OF SERVICE SAFETY SIGNIFICANCE PRECURSOR TO STATION BLACK 0UT DISCUSSION 09:17 WITH "B" EDG OUT OF SERVICE, FAULT ON EMERGENCY BUS E-2 CAUSED SPIKED ON INSTRUMENT BUS #4. FALSE RCD DROP SIGNAL CAUSED TURBINE RUNBACK. REACTOR TRIPPED ON HIGH PRESSURIZER PRESSURE. 09:18 LOSS OF 0FFSITE POWER COINCIDED WITH FAST TRANSFER TO START UP TRANSFORMER. "A" EDG STARTS 8 LOADS ON BUS E-1.
. RCP's TRIPPED. NATURAL CIRCULATION ESTABLISHED.
09:35 UNUSUAL EVENT DECLARED - OPEN TELEPHONE LINE ESTABLISHED
' O 09:46 BETWEEN NRC OPERATIONS CENTER AND ROBINSON SITE "B" DIESEL GENERATOR RESTORED 10:52 PRESSURIZER HEATERS RESTORED
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12:55 E-1 BUS CONNECTED TO 0FFSITE POWER 16:03 E-2 BUS CONNECTED TO 0FFSITE POWER. REALIGNMENT OF 0FFSITE POWER DELAYED BY
- 1. TWO SI SIGNALS - IST SIGNAL CAUSED BY CLOSURE OF MSIVs; SECOND SIGNAL CAUSED BY STUCK OPEN S/G PORV
- 2. ADDITIONAL BREAKER PROBLEM BETWEEN #3 BUS 8 E-2
- 3. OPERATOR DECISION TO PROCEED WITH CAUTION IN RESTORING ELECTRICAL CONFIGURATION PLANT WAS TAKEN TO HOT SHUTDOWN. CP8L DECIDED TO SHUTDOWN FOR REFUELING ONE WEEK EARLY FOLLOWUP PLANT IN 45 DAY REFUELING OUTAGE; LICENSEE INVESTIGATING CAUSE OF LOOP; REGION II WILL REVIEW PRIOR TO RESTART LICENSEE 8 INP0 TEAM FORMED TO INVESTIGATE HIGH FRE0llENCY OF Q^ REACTOR SCRAMS AT ROBINSON; REGION II TO EVALUATE RESULTS PRIOR j
l TO RESTART
/3
ARKANSAS NUCLEAR ONE UNIT 1 - DESIGN DEFICIENCY IN EMERGENCY FEEDWATER SYSTEM (EFW) O JANUARY 14, 1986 - (G. VISSING, NRR) PROBLEM: DESIGN DEFICIENCY IN THE EMERGENCY FEEDWATER SYSTEM (EFW) SIGNIFICANCE: POTENTIAL FOR LOSS OF ALL EFW AND BLOWDOWN OF BOTH STEAM GENERATORS, DURING STEAM LINE BREAK WITH SINGLE FAILURE OF
, ONE A/C BUS CIRCUMSTANCES:
DEFICIENCY DISCOVERED BY IE INSPECTION TEAM DURING POSTULATED STEAM LINE BREAK CONCURRENT WITH LOSS OF RED A/C POWER BUS, TURBINE DRIVEN AND MOTOR DRIVEN EFW PUMPS WOULD BE LOST POSSIBLE BLOWDOWN OF BOTH S/Gs O
- DESIGN DOES NOT MEET THE SINGLE FAILURE CRITERION
- PLANT, AT 89% POWER AT THE TIME OF THE DISCOVERY, WENT INTO CONTROLLED SHUTDOWN DUE TO FAILURE OF EDG FOLLOWUP: LICENSEE CONFIRMED EXISTENCE OF DESIGN DEFICIENCY LICENSEE TO INSTALL CHECK VALVES IN STEAM LINES FROM EACH OTSG BEFORE STARTUP. NRR (PEICSB) REVIEWED ISSUE AND CONCURRED WITH INSPECTION TEAM'S FINDINGS O
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O RIVERBEND START-UP REVIEW (Eric Weiss, IE) CONCERN: NUMBER OF REPORTABLE EVENTS SINCE OPERATING LICENSE CONCLUSION: NUMBER OF EVENTS SHOULD DECREASE RIVERBEND MANAGEMENT APPEARS SOUND SHOULD MEET IN REGION 4 TO REVIEW PROGRESS IN ONE MONTH NRC SITE VISIT JANUARY 28 - 30, 1986 REPORTABLE EVENTS REVIEW WITH PLANT STAFF SHIFT TURNOVER DIESEL GENERATORS FEEDWATER SYSTEM FANCY POINT SUBSTATION SURVEILLANCE PROCEDURE ON RWCUS CONTROL ROOM PANEL WALKDOWN URGED LICENSEE TO: CONTINUE OPEN COMMUNICATION WITH THE NRC RESOLVE OUTSTANDING EQUIPMENT AND HUMAN FACTORS PROBLEMS 1 l UPGRADE COMMUNICATION WITH OTHER PLANTS PLANT SPECIFIC PROBLEMS TEMPORARY ALTERATION PROGRAM SHOULD BE STOPPED ANNUNCIATOR REFLASH FEEDWATER - VlBRATION, EROSION, STICKING FW REG VALVES STEAM TUNNEL VENTILATION l O i
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.5 POTENTIALLY GENERIC PROBLEMS O VALVE OPERATOR PROBLEM TEMP SWITCH PROBLEMS
- FIBER OPTICS SYSTEM INTERFERENCE EXAMPLES OF PROBLEM RESOLUTION 4
CONDITION REPORT SYSTEM JUMPER CONTROL - TAGS, LOGS, BANANA PLUGS FIBER OPTICS - SHIELDING, GROUND LOOP ELIMINATION, EVENTS RECORDING, SIGNS AND NON-DUPLICATING LOCKS COLOR ON METERS I ( O
/7
R" PERRY 1 AND 2 - EARTHOUAKE
- /* JANUARY 31, 1986 (J. STEFANO, NRR)
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PROBLEM EARTHOUAKE NEASURING 5.0 ON RICHTER SCALE WITH EPICENTER APPR0XIMATELY 10 MILES FROM SITE. SIGNIFICANCE EVENT MAY BE OUTSIDE DESIGN ENVELOPE CIRCUMSTANCES PLANT CONSTRUCTION COMPLETE-WAS UNDERG0ING SYSTEM SURVEILLANCE / OPERATIONAL READINESS TESTS FOR LICENSING. (NUMEROUS SAFETY SYSTEMS WERE IN OPERATION). - SEISMIC INSTRUMENTS WERE BEING CALIBRATED AT TIME OF EVENT. TWO NON-SAFETY EQUIP. TRIPPED ON PROTECTIVE SIGNAL AS DESIGNED; SOME HAIRLINE CONCRETE CRACKS OBSERVED IN AUX., INTERMED. AND RADWASTE TREATMENT BLDGS. PIPE FLANGE ON WATER TANK IN 'h
- RADWASTE BLDG. LEAKING.
PRELIM. DATA SEEMS TO INDICATE SHORT DURATION (41 SEC) SEISMIC MOTION MAINLY FALLS WITHIN OBE/SSE DESIGN SPECTRUM EXCEPT POSSIBLY AT HIGH FREQUENCY RANGE (>15Hz) FOLLOWUP NRR/ REG.III TEAM VISIT TO SITE TO LOOK FOR DAMAGE 8 OBTAIN SEISMIC DATA ON 2/1/86 - 2/2/86 NRR/ REG.III FOLLOWUP VISIT TO SITE ON 2/5/86 - 2/7/86 CEI ANALYSIS UNDERWAY TO DETERMINE TO WHAT EXTENT PLANT STRUCTURE / COMPONENT DESIGN WAS EXCEEDED. PLANT LICENSING DECISION PENDING NRC STAFF EVALUATION AllD DETERMINATIONS RE EVENT. (O
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([) NRR TECHNICAL SPECIFICATION IMPROVEMENT PROGRAN (TSIP) (E. Butcher, NRR, DHFT) NEED FOR TSIP
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T00 MANY SPECS - LESS IMPORTANT ONES THUS DETRACTING FROM IMPORTANT ONES SIZE AND COMPLEXITY OF TS INDUSTRY AVAILABILITY RECORD NOT OPERATOR ORIENTED FINDINGS OF NUREG-1024 " TECHNICAL ([) SPECIFICATIONS-ENHANCING THE SAFETY IMPACT" 1 PROBLEM IDENTIFICATION IDENTIFIED PROBLEMS BY INTERVIEWS, DOCUMENT REVIEWS, AND CONTRACTOR ASSISTANCE 1 IDENTIFIED THREE PROBLEM AREAS LACK OF WELL-DEFINED CRITERIA FOR TS HUMAN FACTORS AND TECHNICAL WEAKNESSES RELUCTANCE OF THE NRC STAFF TO USE TOOLS l OTHER THAN TS O
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O TECHNICAL SPECIFICATION IMPROVEMENT PROJECT RECOMMENDATIONS (1) A COMMISSION POLICY STATEMENT SHOULD BE ISSUED WHICH DEFINES THE SCOPE AND PURPOSE OF TECHNICAL SPECIFICATIONS AND ENCOURAGES LICENSEES T0 IMPLEMENT A PROGRAM TO UPGRADE THEIR TECHNICAL SPECIFICATIONS."
~
(2) THE NRC SHOULD GIVE INCREASED ATTENTION TO CHANGES O MADE BY ticeNSEeS uSiNG THE 10 CeR 50.59 PROCESS. (3) THE NRC SHOULD REVIEW AND REVISE THE STANDARD TECHNICAL SPECIFICATIONS TO CORRECT HUMAN FACTORS AND OTHER TECHNICAL WEAKNESSES THROUGH A PROGRAM OF TECHNICAL ASSISTANCE AND DEDICATED IN-HOUSE TECHNICAL ! RESOURCES. ! (4) THE NRC SHOULD ENC 0URAGE THE CONTINUED DEVELOPMENT AND APPLICATION OF PROBABILISTIC RISK ASSESSMENT METHODS TO ADDRESS TECHNICAL SPECIFICATIONS REQUIREMENTS." i O xt m
t 1 CRITEPTA FOR TS CONTENT AN INSTALLED SYSTEM THAT IS USED TO DETECT, BY MONITORS IN THE CONTROL ROOM, A SIGNIFICANT ABNORMAL DEGRADATION OF THE REACTOR COOLANT PRESSURE BOUNDARY, A PROCESS VARIABLE THAT IS AN INITIAL CONDITION OF A DBA ANALYSIS,
! ($)
- A STRUCTURE, SYSTEM, OR COMPONENT THAT IS PART j OF THE PRIMARY SUCCESS PATH OF A SAFETY SEQUENCE ANALYSIS AND FUNCTIONS OR ACTUATES TO MITIGATE A DESIGN BASIS ACCIDENT, i
l I j 33
ONGOING ACTIVITIES 4 O - TRIAL USE OF TSIP CRITEPIA MEETINGS WITH INDUSTRY OWNERS GROUPS AND AIF SHORT TERM IMPROVEMENTS TO EXISTING STS FIRE PROTECTION TECHNICAL SPECIFICATION ACTION STATEMENTS FOR MISSED SURVEILLANCE TESTS BWR RPS SURVEILLANCE INTERVALS AND A0Ts l (NEDC-30851P) O BWR ECCS INSTRUMENTATION SURVEILLANCE INTERVALS AND A0Ts (NEDC-30936P) EVALUATION OF COMMENTS ON TSIP REPORT
~~ .. g FUTURE ACTIVITIES l, DETAILED IMPLEMENTATION PROGRAM PLAN - 03/86
, PROPOSED COMMISSION POLICY STATEMENT - 06/86 ULTIMATE LONG TERM OBJECTIVE I lO l
"A COMPLETE REWRITE / STREAMLINING" OF THE EXISTING STS BASED ON THE RECOMMENDATIONS OF THE TSIP REPORT N
I O l BR Er \G FOR ACRS - l Fermi 2 O Wec nescay r e aruary ' 2, ' 98'S 4 7 secion I a i
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r , . O i FERMI HISTORY Significant Events Date i Low Power License March 20,1985 Premature Criticality July 1,1985 Event Commission Briefing July 10,1985 ]O Full Power License July 15,1985 !, Unit Shutdown for October 11, 1985 Planned Outage Prbjected Availability March,1986** l for Startup l i
** This may be delayed, dependent upon resolution of j Stress Report and Hanger Design Calculations. .
1O (PNO-lll-86-011) n s l ,,
D TECFNICAL ISSUES o Diese Genera': ors o Turaine Byaass Lines o Sou<: 1 Reac':or reec Dump Turaine o TP Purge Une o Concensate Storage Tan < l l i b o Remote Slutcown '3 _) Pane l o Environmen':a Quaifications/ Review o Seismic Review o Concre':e Emaecments o RF R Puma o Ou':co':ec S:ress Reacrts and -anger l Ca cu ations l l Ox l
O PERFORMANCE ISSUES i l o Excessive Personne Errors o Security Program ima emerr:a: ion O i o Faiure to Disposi': ion known
. Proaems l .
o Ine"ec':ive Communica': ions l s
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O Fermi 2 PERSONNEL ERRORS l l l A Review of 78 LERs No. No. of LERs of Caused by Covered During LERs Personnel Errors 03/2/85-11/25/85 78 il 07/10/85-09/10/85 25 9 O 09/10/85.11/25/85 17 8
~*~---e~< - - - -
O 10 CFR 50.54'f} < LETTER e Acecuacy o" Management anc Managemen: S:ructures anc Sys: ems
;O e Actions to Ensure Reaciness to Res<: art anc 3ower Ascension e
Ac: ions to Imarove Recu a:ory and Oaera iona ?erformance 1 I P
!! O l LICENSEE RESPONSE o CEO Assumes Direct Managemen': Contro 1
o ncependent Overview Commi': tee o Reac<:or Operations Imarovement Plan i O o Nucleor 0aera<: ions improvemen': Plan o CEO tc Autacrize Power Esca ation i; o Sing e A/E On Si':e o ndeaencent Overview by Ccracrate Officers o P on to -ire A Senior V? l (O
. Attachment 1 IN 82-41 October 22, 1982
. Page 2 of 3 DIAPHRAGM TYPE PNEUMATIC ACTUATOR
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, IN 82-41 PN ' '
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PERRY POWER P/.A%7 JANUARY 31,1986 EARTHQUAKE SEISMIC EVENT EVALUATION O ' l
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4 1 i SEISMIC EVENT EVALUATION 1 i 1 S PEPORT t 1 1 PERRY NUCLEAR POWER PLANT s 1 DOCKET NOS. 50-440; 50-441 3 ? I 4 1 I ! f 4 j THE CLEVELAND ELECTRIC ILLUNINATING COMPANY i ! FEBRUARY 1986 t i 4 l l l t I i i l I 1 1 8 I I i ! f i
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( 'u TABLE OF CONTENTS
1.0 INTRODUCTION
2.0 OVERVIEW OF EVENT 3.0 PLANT STATUS AND IMPACT ASSESSMENT 4.0 EARTHQUAKE ANALYSIS AND SITE SEISMICITY 5.0 SEISMIC INSTRUMENTATION DATA EVALUATION 6.0 PLANT SEISMIC DESIGN EVALUATION 7.0 CONFIRMATORY PROGRAMS 8.0
SUMMARY
& CONCLUSIONS APPENDIX A: Strong-Motion Data from the Perry Nuclear Power Plant Seismic Instrumentation (Kinemetrics)
APPENDIX B: Report on the Peak Shock Recorders and Peak Acceleration Recorders Installed at the Perry Nuclear Power Plant during the Seismic Event on January 31, 1986 (Engdahl Enterprises) APPENDIX C: A Preliminary Evaluation of the Significance of the Seismic Event on January 31, 1986 at the Perry Nuclear Power Plant (Stevenson) APPENDIX D: Response Spectra Plots APPENDIX E: Results of Specific Inspections Evaluation of Walkdown Items Plant Settlement Readings Seismic Clearance Walkdown Cooling Tower Walkdown Review of Energized Circuits O
O
/
1.0 INTRODUCTION
The purpose and scope of this report is to provide the results of The Cleveland Electric Illuminating Company seismic event evaluation for the Perry Nuclear Power Plant. The discussions contained herein provide the basis for CEI's conclusions that the January 31, 1986 earthquake in the vicinity of the Perry site:
- 1) did not adversely effect the plant structures, systems or components,
- 2) was within the design capability of the Perry Nuclear Power Plant, and
- 3) does not change the licensing basis or conclusions regarding the site geology, seismology or design basis earthquake.
This evaluation report addresses the key issues related to the January 31 earthquake including the immediate response to the event, and the plant status and impact assessments following the earthquake. Detailed evaluations of the geological and seismological implications of this event and an analysis of the plant seismic design basis capabilities are presented. In addition, a description is picovided of the confirmatory programs to monitor post seismic event activity, to continue the evaluations to identify any earthquake related effects, and to participate in generic industry studies. O
2.0 SEISMIC EVENT OVERVIEW Event At approximately 11:48 a.m. on January 31, 1986, an earthquake occurred, which was located about 10 miles south of the Perry site and had a Richter magnitude of approximately 5.0. CEI implemented the Perry emergency plan in response to the seismic event as described in the attached chronology. A site area emergency was declared as a precautionary measure for site personnel accountability and for informational notification to local officials. Timely notifications were made and plant staff responded professionally and successfully implemented the plant procedures for this type of an event. Plant Response and Assessments Immediately following the earthquake, plant operations personnel were dispatched into the plant to survey for any major damage. The initial reports indicated no damage. Subsequently, a team of approximately 65 engineers and technicians was organized to perform a detailed walkdown of all plant areas. These inspections found no damage to any systems, structures or components. The hairline cracks in concrete walls that were observed have been reviewed and found to be typical of reinforced concrete structures whb h have not experienced seismic events. Numerous safaty-related systems in operation or standby readiness continued to operate without incident. Earthquake Analysis Based on United States Geological Survey (USGS) recorded data, the earthquake of January 31, 1986 was centered about 10 miles south of the Perry Site and had a Richter magnitude of 4.96. This is a lesser magnitude than the earthquakes for which the Perry Plant has been analyzed and had substantially lower total energy content than the O
O) Perry design response spectra. The January 31 earthquake is consistent with the previously established geology and historical seismicity of the region, as described in the Final Safety Analysis Report. The earthquake does not change the conclusions of the FSAR on the geology and seismicity of the site area. 2 Seismic Design Evaluation Acceleration data taken from the in plant seismic recorders showed recorded floor response spectra in certain locations outside the design spectra at high frequencies. The design spectra are based on a statistical envelope of historical earthquakes (84th percentile) and, therefore, some instances of recorded responses exceeding predicted floor responses are expected. The possibility of high frequencies outside the spectra has been evaluated at other nuclear plant sites and concluded to have insignificant effect on plant structure and components. () CEI analysis shows the high frequency accelerations involved are of a very short duration and the velocities are well below those which could cause damage even to non-engineered structures. The total energy associated with these high f requency accelerations is small, and therefore has no adverse impact on plant structures and equipment. Thus, the high frequency accelerations have no engineering significance and the effects of the earthquake experienced at Perry are well within the seismic capability of the plant. O V l l l l l
January 31, 1986 Earthquake
<- Chronological Summary of Events
(/ w Time of Occurrence Event 11.46:42.3 (USGS data) Seismic event occurs 1148 Control room reports nosie & vibration to to Systems Operation Center 1150 Main generator breaker reported open, isolating main and auxiliary transformer, automatically shif ting to startup transformer. Auxiliary boiler trips noted Seismic alarms received in P680 1155 Trip of instrument air compressor noted 1200 Visual inspection of lower areas of Turbine Building, Auxiliary Building, Intermediate Building and transformer yard satisfactory 1201 Shif t Supervisor sounds Plant Emergency Alarm 1204 Visual inspection of Turbine Building, Turbine Power Complex, Intermediate Building, Auxiliary Building and Control Complex Ox satisfactory. 1206 Shift Supervisor delcares precautionary Site Area Emergency, makes Evacuation Announcement. 1211 Auxiliary boiler restarted 1216 Notifications to CEI emergency personnel pursuant to Emergency Plan began 1218 Initiated retrieval of seismic plates and magnetic tapes from seismic instrumentation 1219 Visual inspection of service water and emergency service water pump house satisfactory 1225-1240 Initial notifications of Site Area Emergency provided to Lake, Geauga and Ashtabula counties, the State of Ohio, Coast Guard, NRC 1230 Visual inspection of cooling towers and basins satisfactory 1232 Operational Support Center (OSC) activated 1 l
Time cf Occurrence Evsnt 1235 Technical Support Center (TSC) activated 1251 Initial inspections of all Unit 1 and Common areas completed with satisfaeory results, only minor problems noted. 1254 Visual inspection of suppression pool satisfactory 1257 - 1301 TSC completes precautionary Site Area Emergency Follow-up Notifications to Counties, State, Coast Guard 1300 Walkdown of all Unit I areas has finds no major equipment damage and noted minor flange leaks. 1302 Site Area Emergency downgraded to Alert 1303 - 1315 Initial Notification of downgrading to Alert made to Counties, State, Coast Guard and NRC. 1305 Three teams dispatched for additional system walkdowns; six maintenance teams dispatched to investigate equipment O 1340 - 1401 Follow-up notification of Alert status provided to Counties, State, Coast Guard and NRC 1341 INPO contacted 1420 NRC - Bethesda and Region III concur on
- termination of emergency 1425 Termination of Emergency Event 1431 - 1442 Termination of Emergency Event reported to Counties, State, Coast Guard, & NRC.
1440 INPO notified of termination of Emergency Event 1531 Deactivated TSC. 1552 St.ismic alarm P969 reset 1630 Recovery organization met to review seismic event, emergency response and confirmatory actions.
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3.0 PLANT STATUS AND U1 PACT ASSE3SMSNTS v 3.1 PLANT STATUS Prior to the earthquake that occurred on January 31, 1986, numerous' testing, calibration, and work completion activites were being conducted in preparation for fuel load. One major activity was preparation for the Division II Diesel Generator response time testing. As part of this work, all of the safety related components powered from the Division II Diesel were energized and in standby readiness. All of this equipment behaved normally through the event; that is, there were no spurious starts or alarms. Preparations were also underway to move the startup soer es. This work had not yet begun when the seismic event occurred. The sources were never actually moved, and remained stored in the upper pools. , /7 In support of the ongoing test and surveillance activities, a G) significant number of systems were in operation. In addition, numerous other systems were energized and in the standby mode. Lists of the specific safety and non-safety systems energized or operating prior to and during the earthquake are included as Tables 3.1 and 3.2. All of the operating safety-related systems continued to operate through the event. None of the safety-related systems in the standby mode experienced any spurious initiations. As noted in Table 3.2, a large number of non-safety systems were operating or in the standby mode, and maintained their status throughout the event. Two non-safety items tripped on protective signals as intended by the design. These were the Unit 1 instrument air compressor, which tripped on high vibration, and the auxiliary steam boiler, which tripped due to actuation of one of its protective 3.1
circuits. The instrument air compressor is a centrifugal machine
that operates at greater than 40,000 rpo and as part of its protective devices has a very sensitive vibration switch. The auxiliary steam boiler has several protective circuits of which one tripped during the earthquake. The boiler was successfully restarted after the event.
The only other non-safety items of equipment that tripped during the earthquake were the Unit 1 main and auxiliary transformers, which tripped due to the closing of the generator protection relays. These relays although open at the time of the seismic event, did not have voltage applied as a result of an ongoing outage. Laboratory testing of these relays since the event has confirmed that the presence of voltage oa the relays significantly increases the force required to close these relays. Had the voltage been supplied to these relays, they would not have closed during the event. This is substantiated by the fact that other similar open relays with voltage applied did not close during the e' vent. Investigation is ongoing to determine the cause of an indicated 1 1/2 inch increase in suppression pool level. No basis for a physical change in the water level has been identified. The water level transmitters were found to be out of calibration, though not enough to account for the entire indicated level increase. The same transmitters in other applications did not show any anomalous behavior. In addition to the emergency plan actions previously discussed, immediately following the event the plant operators performed initial surveys o'f the plant. Areas visually inspected included the Transformer Yard, lower elevations of the Turbine, Auxiliary, Intermediate and Radwaste Buildings, as well as the Control Complex, Turbine Power Complex, Heater Bay and Water Treatment Building. The reports back to the Control Room indicated that the areas were found g in satisf actory condition with no major damage. In addition, the 7/
\-- General Supervisor of Operations and the Senior Operations
~
3.2 a.
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[T/ Coordinator made a specific survey of below grade areas. They found
\/ no unusual or abnormal conditions. purther steps taken to assess and evaluate the status of the plant included additional walkdowns by teams of plant maintenance personnel dispatched from the Operations Support Center.
3.3 PLANT IMPACT ASSESSMENT As part of CEI's response to the earthquake, a team of approximately 65 engineers and technicians was organized on the evening of January 31 to perform systematic and thorough walkdowns of all plant areas. These walkdowns were performed using drawings of each area and checklists of components to inspect for any abnormal conditions. These included such items as piping, hangers, snubbers, valves, pumps, instrumentation and other components. The results of these walkdowns were recorded and compiled into a list of approximately 480 observations, many of which were later determined tb be preexisting g conditions. None of the observations involved structural damage to ,f \~ I the plant or equipment. The 480 observations are typified by minor hairline cracks in concrete, burned out light bulbs and leaking valve or piping flanges, all of which are normal and expected conditions that would be identified in any comprehensive walkdown. In the inspections that were conducted following the earthquake, plant personnel were instructed to document all unusual or abnormal conditions. Those conducting the inspections did not attempt to determine whether the conditions were the result of the earthquake; instead, discrepant conditions regardless of potential cause were documented to insure that the status of the plant following the earthquake would be fully documented for subsequent evaluation by engineering. Each of the observed discrepant conditions was subsequently evaluated by engineeering to determine whether the ; condition was caused by the earthquake and whether rework or repair was required. The engineering evaluation of the items concluded that 77% were preexisting conditions, and only two minor items, were ! directly attributable to the earthquake. The remainder, 3.3 l
C) approximately 100 items, have been classified as inleterminate, i.e. , it could not be definitively established that the condition existed prior to the earthquake. About 25% of the approximately 480 items will need rework or repair. (See Appendix E). These will be processed in accordance with a special procedure instituted in response to the earthquake. A number of other inspections were also performed to determine the effect, if any, on specific plant structures and conditions. A site survey was performed to assess any impact of the earthquake on the site environs, and in particular on the shoreline bluff. No evidence of any earthquake impacts could be found. A survey of settlement monitoring points was ordered to determine if the earthquake had any effect on building settlement. Monitoring points at various locations around the perimeter of the plant buildings are surveyed on a monthly basis to monitor building settlement. The results of the surveys were that the recorded movements were consistent with those measured in the past, including the amount of change from prior surveys and the absolute elevations. For example, a comparison of the Reactor Building reading with that of February 1985, found that the two readings were identical. Th a . it is concluded that the earthquake had no impact on building settlement. (See Appendix E). A walkdown of Unit 1 Cooling Tower was performed to determine whether any damage had resulted from the earthquake. The areas inspected included the basin walls, tower columns and footers, internal support columns, baffle system, discharge pipe, and veil. While all inspections were done from ground level, any significant cracks in O 3.4
g- the veil would have been readily apparent since they would have baan ( ,/ saturated by the previous day's rain. No structural damage was found in any area of the cooling tower. Water was observed seeping through the north and south vertical joints where the basin plume wall and pump house fiume wall meet. Seepage at this joint has been noted in the past and stopped by the application of mastic material. (See Appendix E). As part of the design program for the plant, seismic clearance criteria were estab'ished to assure that a seismic event would not cause any impact on a safety system either by causing swaying or by impact from a non-safety item. Instances of these criteria not being met are termed Seismic Clearance Violations (SCV's). SCV's are forwarded to engineering for evalution to determine whether repair is required. At the time of the earthquake, there were 29 SCV's that had been dispositioned for repair, where the repair had not yet been completed. Following the earthquake, inspectors wefe directed to ' reinspect these SCV's to determine whether the seismic event affected () the SCV condition. These inspections found neither damage nor , dimensional change. (See Appendix E). As previously noted, the plant systems, both safety related and - non-safety related, operated properly during and following the seismic event. Recognizing the sensitivity of electrical components to high frequency response, a detailed engineering study was undertaken to identify the number and types of electrical equipment that was energized during the earthquake. The components included motors, transformers, relays, switchgear breakers, switches, batteries, contacts, valve operators, chargers / inverters, meters, recorders, and transmitters. A wide variety of suppliers was represented. More than 70 separate systems were involved. The study showed that over 47,000 electrical components were energized and experienced no adverse effects in terms of spurious systec actuation (See Appendix E). V(~h 3.5
TABLE 3.1 , ,j
'I SAFETY RELATED SYSTEMS .
ENERGIZED DURING THE SEISMIC EVENT
. 0F JANJARY 31, 1986 SYSTEM DESCRIPTION ~
2 -a.. .
- .s .- ..-
C11 Control Rod Drive .. C41 Standby Liquid Control _. .. C71 Reactor Protection System __ . , . . . ;g ,j.g; P1 ant R aHeat d i a tRe=oyal i o n M o n7-y_y'("g'g-i t o r s :y; t :'f.;-ffC'. i :.O.q[:.. .. {.{.- ' ' , .' : D 1
- t E127 -- ? Residual' {t."$;il'!ly E21 Low Pressure Core Spray . . . r ,... . c. . . c . -
E22 High Pressure Core Sray....;J...,....... - Fuel P o ol C o o li n g an d Cle anu p 1"0 7.'" . :,;. . N" G41
'M15 ' - - ' ~ Annulus Exhaust Gas Treattie'ntji.hh. N-(-
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M24 Battery Roo= Exhaust - 9.
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M25 Control Room HVAC ~ M26 Control Room E=ergency Recirculation M32 ESW Pu=phouse Ventilation ,- - M40 Fuel Handling Building Ventilation -
. M43 Diesel Building Ventilation _ c' -
P11 Condensate Transfer and-Stora'ge'Z u5 - i .. -.
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.. P22 Mixed Bed De=inerali::er
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P41 Service Water 1 ; . 9'3Et@y;; - r. ; . P42 Emergency Closed Coolin6 ~ ~ d I.T. I . $ ' ^ P43 Nuclear Closed Cooling " , ' ' ' - J ci-196' - - P45 Emergency Service Water .
r -
7 P;7 Control Complex Chill Water \ _,/ P;9 ESW Screen Wasr P52 Instrument Air P5e Fire Protectien' C95 E=ergency Response Information Syste: P51 Service Air R14 110 VAC Vital Inverters R22 Metalclad Switchgear __ R23 420 V Load Centers , R24 Motor Control Centers - R25 Distribution Panels - 120, 208 & 480 volts R '. 2 D. C. Syste R43 Standby Diesel Generator (SDG) R'5 SDG Fuel Oil R46 SDG Jacket Water Coolant Es7 SDG Lube Oil R61 Main Control Roc = Annunciator t_ 3.6
TABb 3.2 - i NON-SAFETY RELATED SYSTEMS ENERGIZED DURING THE SEISMIC EVENT OF JANUARY 31, 1986 s, _/ SYSTEM DESCR'IPTION i F42 Fuel Transfer Equipment G33 Reactor Water Cleanup
- M11 Containment Vessel Cooling.' -
',' M13 Drywe11 Coo 1 ins ,
. _.f.M,y.iji}..y;.'[., '. ,_. ,' . . [. ...
- Controlled Access HVAC t.27.G ycn y r.g;jg-y't:dy G fi. ,
~ -
M21 - . M27 Computer Room HVAC '- -- MT Turbine Building Cooling'&_Ve'nt'ilation 1
~ "
~
M35 M36 O f f -G a s 'B u i l d i n g E. x. h a u s t '. '. . .'.<#~ ~~I". . .. .. . . . '. .
". ~.
M41 H e a t e r B ay V e n t il a t io n 77 C O.e'.Q.5p ' -S-M45 Circulating Water Pump House' Ventilation -7' ?.:.;= .?@-- "
' ~
N21 Condensate - .' V.;h. .
/ - _
~ ;';~ '
N23 Condensate Filtration ~ N24 Condensate De:ineralizers N32 Turbine Control (EHC) - N71 Circulating Water . P20 Water Treatment
- P21
~
P44
. Turbine Tw o B eBuilding d D e =inClosed e e a1iz Cooli e'. .$
r g;W#.(.i-%'dh -
.y.).5.(_ 9 . e.. , . .
. P55 Building Heating . _ fr ' - i .
'~
. P61 Auxiliary Stea: . , Ti - ,
P62 Auxiliary Boiler Fuel Oil. ' - 1 f :~, -- P72 Plant Underdrain .. ._ . . fs C91 Process Cc puter . - -v ( ) C94 Health Pr.ysic Cc puter P5c, Security R11 Station Trar.sfer:ers R15 Technical Support Center UPS R36 Heat Tracing & Anti Freeze Protection R44 SDG Starting Air . R51 Intra Plant Communications R52 Maintenance & Calibration
-_ Exclusicn Area Paging Syste:
R53 - R57 Radio & In-Plant Antenna Systes R71 Lighting S11 Power Transfer =ers ~ S41 Step Up Station - 1 FN i d 3.7 ) l
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v 4.0 EARTdQUAKS ANALYSIS AND SITE SSISMICITY An earthquake of magnitude 4.96 M blg ecurred on January 31, 1986 at 11 hours, 46 minutes, and 42.3 seconds approximately 11 miles (17.7 kilometers) south of the plant. The depth of the earthquake is presently calculated to be 6 miles (10 kilometers) deep and is located at 41.640 W and 81.098 N by the National Earthquake Information Center of the United States Geological Survey (USGS). This location is near the intersection of Highways 86 and 166 in Thompson Township, Geauga County. The location of this earthquake is shown on Figure 4.1 of this report. Earthquakes which have occurred within 200 miles in historical times, and an update for those occurring within 50 miles of the plant site are shown in Figures 4.2. and 4.3.
4.1 BACKGROUND
GEOLOGICAL & SEISMOLOGICAL STUDIES RELATED TO THE PERRY NUCLEAR POWER PLANT , .
,n V As required by the regulations governing the siting of nuclear power plants, a thorough study of the geological and seismological characteristics of the Perry Nuclear Power Plant site and its regional surroundings was made as part of both the Preliminary Safety Analysis Report (PSAR) and Final Safety Analysis Report (FSAR). The purpose of these investigations was to assure that the site was geologically suitable for the construction of a nuclear power plant and to provide e basis for the determination of a Safe Shutdown Earthquake (SSE) and the site ground motion resulting from the occurrence of such an earthquake. The information contained herein is cummarized from the detailed discussions contained in Chapter 2 of the PSAR and FSAR, as reviewed and accepted by the NRC in the Safety Evaluation Reports and Supplements.
These studies were extensive, consisting of a compilation and analyses of published and unpublished literature; field geological check'ing and mapping including wide scale and local geophysical l studies to characterize geological conditions at depth; borings; J laboratory analyses; and detailed engineering analysis of the site foundation materials. 4.1 i
,/) Based on these studie< nnd following Appendix A of 10 CFR Part 100, a k-s! correlation of earthqaakes to a particul.ar f ault or series of f aults which would be designated as " capable" could not be made. In addition, no "large scale dislocation or distortion" of the earth's crust designated as a tectonic structure could be identified to which earthquakes could be correlated. Consequently, earthquakes were identified with a " tectonic province", representative of a region within which there is a relative consistency of geologic structural features.
To select the SSE, a Modified Mercalli Intensity of VII was chosen as the maximum intensity earthquake at the Perry site. This intensity corresponds to an acceleration value of 0.15g, based upon a number of developed relationships which relate peak accelaration to earthquake
- intensity values; the principal relationship was developed by Trifunae and Bradv. (Trifonac, M.D. and 3rady, A.G., 1975, on the Correlation of Seismic Intensity Scales with the Peaks of Recorded Strong Ground Motion
- Bulletin of the Saismological Society of
-) \~ / America, v. 65, No. 1, pp. 139-162). The response spectra representing the SSE were then developed by adopting a NRC Regulatory Guide 1.60 response spectral shape. The design response spectra are shown on Figures 4.4 and 4.5.
During the review of the FSAR, the NRC staff requested that site-specific spectra be constructed for the Perry site. In response to this request, site-specific response spectra were constructed using a set of ground motion accelerograms from actual earthquakes of magnitude range 5.3 f; .5 recorded on rock (to simulate the foundation conditions at Perry) at epicentral distances of 0 to 25 kilometers; t'11s represents the earthquake "at the site" as required by Appendix A and is shown on Figure 4.6. Eleven (11) earthquakes representing 22 components of motion were chosen. A subset of recards accepted by the staff as representative of an Anna, Ohio type earthquake had an average magnitude of 5.53,f;
.3 at an average distance of 8.5 miles (13.66 f; 4.5 kilometers).
A smoothed 84 percentile of this data set fell below the desid n 4.2
i e response 3pectra rape.:sented by a RJdulatory Guide 1.60 spect ra se: at an acceleration of 0.15 3 These spectra ara representative of free field data recorded at locations away from the influence of buildings and structures, and are shown on Figure 4.7. 4.2 REGIONAL GEOLOGY AND TECTONICS The Perry site is located in the central part of Eastern Stable Platform Tectonic Province, characterized by an upper Precambrian crystalline basement and overlain unconformably by a sequence of Paleozoic sedimentary rocks. Basement rocks of this tectonic province comprise a complex sequence of high grade metamorphics and include: schists, gneisses, marbles, and granu11tes consolidated during the Grenville Orogeny (950 mya) onto the North American craton. , The basement rocks are overlain by a 5000' thick sequence of , /' sedimentary rocks, Cambrian to Carboniferous in age, which dips less
# than 5 to the south. (Fig. 4.8). Sedimentary rocks within this sequence of Paleozoic sediments includes shales, salt, sandstone, dolomites, and limestones. In the epicentral region the sedimentary sequence is approximately 2 kilometers thick with the main shock focus well within the crystalline basement.
A thin veneer, generally less than 100' of variable thick Pleistocene deposits, lies unconformably on the sedimentary sequence. These deposits include a lower till, dense and compact (approximately 30' thick) overlain by less compact till, lacustrine deposits and beach deposits. Post consolidation tectonic deformation in the province includes the following structual elements. Paleozoic structures include broad upwarps: Cincinnati arch, Findlay arch, Kankakee arch, Ozark uplift, Nashville dome, and intervening Michigan and Illinois basins. Uplift
-~ and subsidence produced localized faulting and folding. The north
. I
~-/ northeast-trending Waverly arch of west central Ohio is the nearest upwarp structure.
4.3
l u . l l
) Faults in the site restoa 11clude: \j o Chatham sag faults o Peck fault, Howell-Northville anticline faults o Bowling Green fault o Anna Ohio faults o Cincinnati arch faults o Eastern Ohio faults o Western New York faults o Appalachian Plateau and Northern Valley and Ridge faults Within the region only the Clarendon-Linden f ault system in western New York is considered active. -
4.3 SITE GEOLOGY In conjunction with the PSAR and FSAR preparation an,d reviews, , intensive geological and geotechnical investigations were conducted f-
\ - .s at the Perry site including:
o test borings (maximum depth 730') o 42" drilled exploratory shafts o in-site testing, plate load tests o permeability determinations o piezometer installations o seismic analyses o seismic refraction and seismic shear wave determinations o geologic mapping of excavations, tunnels and trenches Two bedrock structural styles were observed by Gilbert Commonwealth, NRC staff, USGS, and the Corps of Engineers. Gentle northeast-I trending folds with two to three foot wavelength and 6" amplitude were attributed to depositional processes. Two larger folds and several related faults were also examined. The folds terminated below foundation grade. Faults with characteristic north over south (, directed motion become bedding plane detachments at depth. One to 4.4
/ ) three Inch thick goose accurs in the fault 7.o ne s . Absence of foreign F ' /
mat erials, no reerystallizstion of country rock or cryst allizstion within fault zone or adjacent f racture zones is interpreted to result , from localized low temperature, relatively low stress deformation. In summary, an approximately 45 foot thick layer, between excavation grade of the deepest onshore foundation excavations and the base of a i boulder layer defining the bottom of structureless basal till, experienced deformation (folds, faults) including bedding detachment rotation and buckling, and slight upward thrusting. These features occur in glaciated terraine and are attributed to glacial loading, unloading and/or ice push mechanisms. Similar faulting was studied in the Warner Creek area with the same conclusions. 4.4 DEFORMATION - INTAKE AND DISCRARGE TUNNELS Three minor low-angle north-northeast striking thrust faults occur in (N the intake and discharge tunnels to the north beneath Lake Erie. Displacements range between 0.5 and 2.5 feet, upward to the northeast. Studies undertaken to define tunnel fault geometry included: o detailed mapping of tunnel walls o reconnaissance of lake bottom o lake shore reconnaissance o exploratory borings . o borehole logging, of f shore and onshore magnetic surveys o review of existing geophysical data o isotopic analyses of Lake Erie and fault seepage water Studies to date fault included: o x-ray diffraction s o clay mineralogic analysis o microcrack o consolidation of gouge 4.5 d
u , Misedi.neous studies included: (/p) s. o borehole stress o structure contour maps o interviews with knowledgeable Ohio geologists Investigations of the vertical and lateral extent of faulting indicated that the faulting did not extend upward to the lake floor. Borings at the projected western shoreline intersection showed no faulting. Conclusions reached from detailed mapping of the tunnel faults, geophysical surveys, borings, and analysis of fault gouge and seepage included: o faults are genetically related; same fault or an echelon o faults confined in Chagrin shale; limited lateral and vertical extent o date of last motion is Pleistocene or older o motion sense indicates faults originated in northwest directed stress field, approximately 90ofrom present stress field i s ,) o possible mechanisms of nontectonic glacial origin include ice sheet traction, differential downwarp, differential rebound, surficial stress relief (" pop up") o geologic processes responsible for initiation and latest motion are nonteetonic and no longer operative; therefore faults are not capable according to Appendix A to 10 CFR 100 a O 4.6
/O V
4.5 CURRENT SEISMOLOGICAL AND GEOLOGICAL STUDIES , Immediately after the occurrence of the earthquake, CEI undertook a number of geological and seismological investigations to provide a thorough understanding of the earthquake and assess any impact on previous studies performed for the siting and licensing of the Perry Nuclear Plant. In addition to the investigations undertaken by CEI, USGS, as well as various universities and private groups, have deployed instruments to study earthquake aftershocks. 4 1 Portable Seismographic Network At the request of CEl, Weston Geophysical Corporatio,n installed sta , fx portable analog seismographs (Sprengnether Instrument Co. MEG-800) in k.- the epicenter area of the January 31, 1986 earthquake during the period f rom approximately 10 hours to 30 hours af ter the event. These seismograph stations are located at the Perry Nuclear Plant and in the communities of Chardon, Chesterland, Middlefield, Hartsgrove, and Thompson. A seventh station was installed on February 4, 1986 in the town of Concord. This spatial distribution of the stations is designed to form a symmetrical array around the preliminary epicental area of the main shock, which was located in the basis of more distant stations. All instruments are operated continuously and all seismograms are recovered and analyzed daily. The purpose of this network is to obtain accurate locations of any recorded af tershocks, to refine the original location of the main shock, and to determine whether or not their occurrence reveals anything about the causative geologic structure. Five other portable instruments integrated into this network are ( operated by Woodward-Clyde Consultants and deployed in a similar configurat8on to provide additional locationing capabilities. 4,7 1
Five small microear:.hquakes have been detacted. The parameters of V these earthquakes ace located on the Table 4-1. Preliminary analyses indicate that the focal depths for these microcarthquakes range from 2.3 to 8.9 kilometers. The largest of these microearthquakes, a magnitude 2.4 event on February 6,1986, was the only event to be felt. These microearthquake locations are slightly to the west of the preliminary location of main shock provided by the National Earthquake Information Center. Felt Intensity Investigation A questionnaire survey is being conducted to evaluate the distribution of ef fects, including a general description of how people experienced the event and accounts of any damage that have been incurred. The questionnaires are being distributed using several parallel approaches to obtain broad coverage of the affected areas. Analysis and compilation of questionnaire resu,lts will be , p used to produce an isoseismal map" or plot of intensity levels U measured on the Modified Mercalli Scale. The purpose of such a map is to enable a comparison of effects of the present event with a well-known epicenter to the effects of soae historical. events located in the site area that have no well-determined instrumental epicenter. Weston Geophysical personnel have been conducting personal interviews on perception and other effects of the enrthquake in the epicenter region. Questionnaires have been distributed at establishments such as fire departments, grocery stores, schools, etc. with instructions to distribute these to persons near the earthquake epicenter. These reports will be used to recover information on the range of ef fects. A preliminary evaluation of returned questionnaires indicates that most of the reports in the epicentral area are evaluated as representative of an Intensity VI on the Modified Mercalli Scale. Maximum observed or reported effects include a few instances of damaged chimneys above the roof line, cracks in concrete and cinder block walls, cracked or fallen plaster, and few broken windows. Some disturbances including silting of well-water have also been reported. 4,8
- . . . . - . . - . _ - . .- .-_ = -- - - -
4 O Geologic Studies Weston Geophysical geologists have conducted preliminary reconnaissance of bedrock exposures in the epicentral area to determine whether or not any surface expression resulted 2 rom the - earthquake. No significant expression of surface disturbance has i been observed. Although several occurrences of minor rock slides and soil slumps have been documented and photographed, these are not
. considered unusual, since they occur in unstable, undercut steam banks where they could have been caused by ordinary weathering 2 processes or induced vibratory ground motion from the earthquake.
4 Previously mapped f ault locations on Paine Creek have been exa:nined. No evidence of recent fault movement was observed. Also, no slumping ] or sliding of the steep slope was apparent. No evidence suggestive of a " capable fault" has been observed. t l On-going work includes examination of other geological features, as well as an investigation of sites of unusual felt reports such as foundation damage and water-well disturbance. A field observstion
)
and evaluation of soil and rock conditions at such sites is being made to determine whether or not there is a correlation between the higher intensity values and geological conditions. 4 4 e 4 e 1 I 3 l 4.9 1 4
. . _ , _ _ . , _ , _ _ _ . . . _ . . . . _ _ , . , _ .. . , , ____,,_,r. , , _ .. , _ , _ . , ,
J
4.6 CONCLUSION
S REGARDING OF THE JANUARY 31, 1986 ORIO EARTHQUAKE The earthquake, both as regards to magnitude and intensity, is below the maximum earthquake selected to represent the Safe Shutdown Earthquake. The intensity of the Safe Shutdown Earthquake was selected as intensity VII. It is estimated that the present earthquake is best represented by an intensity VI. The magnitude 4.96 M of the January 1986 earth'uake is below the magnitude of b1g 5.3 + 5, used in establishing the site specific response spectra.
- Based on the initial data evaluation, it appears that the free field design response spectra constructed to represent the SSE may have been exceeded. An accelerogram at the foundation level of Uni I showed a peak acceleration of 0.18 g at approximately 20 Hz on the north-south component. The duration of the motion on foundation ,
above the smoothed ground response spectra (SSE) is less than 0.1 (/) s', second. Since both the Regulatory Guide 1.60 ground motica and the site-specific spectra represent a smoothed spectra at the 84;h percentile for a number of strong motion accelerograms, exceedances above the smoothed spectra are not unexpected. At the high frequency end of the spectra, where the 20 Hz exceedance exists, it is important to look at the other parameters of ground motion. The particle velocity associated with the 0.18 g. is 0.55 inch per second and the displacement is 0.004 inch. This velocity value would be f ar less than the 1 inch per second generally accepted by the US Bureau of Mines as the threshold of damage at the 20 Hz frequency: cracking of plaster walls, etc. to ordinary structures. (Siskind, D.E. et al., 1980, Structure Response and Damage Produced by Ground Vibrations from Surface Mine Blasting, Bureau of Mines RI 8507). Structural damage therefore is not a problem. 4.10
f' The area and region in ahich the January 31, 1986 earthquake >ccurred is one of loa seismicity. Prior to 1986, the tardest earthquake to occur within 50 miles of the site occurred in 1943. The 1906 Ohio earthquake is slightly largar in magnitude (4.9 vs. 4.7) and intensity (VI vs. V) than the March 9,1943 earthquake which occurrer! approximately 12 miles west-southwest of the 1986 earthquake. Although somewhat larger than historical earthquakes within 50 miles of the plant site, it is smaller than those within 200 miles of the site, as well as those on which the plant design is based. This earthquake is consistent with the seismicity of the area and the area and region are still of low seisinicity. Geological investigations to date have not uncovered any evidence suggestive of a " capable fault" as defined in 10 CFR Part 100, nor has the investigation revealed a cause for any geological ' concern. The 1986 earthquake does not change the conclusions in the FSAR on the geology and seismology of the Perry site. . . 7--) v . 4 4 l l
~
4.11
. - . _ _ _ . ._. ~. . -- ~- . ..
I f) Table 4.1 RECENT EARTHQUAKES IN THE SITE VICINITY i ORIGIN ( } PRELIMINARY TIME LATITUDE LONGITUDE DEPTH (KM) MAGNITUDE DATE 2.7M () 22-JANUARY-1983 07:46:57.9 41 51.24' 81 11.46' 5 31-JANUARY-1986 16:46:42.3 41 38.84' 81 05.30' 10 4.9 Mb ( } 01-FEBRUARY-1986 18:54:49.7 41 38.39' 81 09.99' 3.1 -- t 02-FEBRUARY-1986 03:22:49.1 41 38.37' 81 09.81' 2.3 -- 1 03-FEBRUARY-1986 19:47:19.6 41 39.19' 81*10.27' 9 -- O 4 05-FEBRUARY-1986 06:34:02.4 41 39.93' 81 09.11' 6 -- l 2.4 06-FEBRUARY-1986 18:36:22.6 41 38.66' 81 09.80' 5 i J 1 I (1) UNIVERSAL Time Unless Noted As Local Time l (2) SOURCE: University of Michigan i (3) SOURCE: National Earthquake Information Center (NEIC) l i O
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Figure 4.4 ANpp PERRY HUCLEAR POWER PL ANT THE CLEVELAND ELECTRIC Y ILLUMINATING COMPANY Safe Shutdown Earthquake Design Response Spectra - Os Vertical Motion 1 Figure 3.7-2 l
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. PERRY HUCLEAR POWER PLANT THE CLEVELAND ELECTRIC CN5583 ILLUMlNATING COMPANY Safe Shutdown Earthquake Design Response Spectra -
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i 5.0 SEISMIC INSTRUMENTATION DATA EVALUATION f Three different types of seismic monitoring instrumentation were used to record the 1986 Ohio. Earthquake. Table 5.1 and Figure A through H and J delineate the specific instrument number, type and location. One type of instrument used is the Kinemetrics Model SMA-3 strong motion triaxial time-history accelerograph. This system detects and records three mutually perpendicular components of acceleration over the entire duration of the earthquake onto cassette magnetic tape. I Power to the unit is-supplied by internal rechargeable batteries which are kept in a charged state by 120 VAC line power. Two instruments of this type were used and were located on the Reactor Building Foundation Mat at an elevation of approximately 575 feet. Their latest calibration was December 1,1985. See Appendix A for further instrumentation details and data tabulation. The second type of instrumentation used was the Engdahl PSR 1200-H/V response spectrum recorder. This totally mechanical system also records three mutually perpendicular components of acceleration. The instrument used twelve reeds fabricated of varying lengths and weights of spring steel, one for each frequency (ranging from approximately 2 Hz to 25 Hz). A diamond-tipped stylus is attached to the free end of each reed to inscribe a permanent record of its deflection on one of twelve record plates. The record plates are { made of aluminum and plated with successive layers of nickel, tin and lead-tin. This system is totally self-contained and requires no outside power source. Four instruments of this type were used - two on the Auxiliary Building Foundation Mat and an elevation of approximately 568 feet, one at the Reactor Building Foundation Mat at an elevation l approximately 575 feet, and one at the Reactor Building Inside ! Drywell Platform at an elevation of approximately 630 feet. Except for the one instrument located on the Reactor Building platform which was calibrated on January 30, 1986, all instruments of this type were calibrated during January 1985. See Appendix B for further i instrumentation details and data tabulation. 5.1
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The third type of instrument was the Engdahl PAR 400 peak accelerograph. This totally mechanical system records three mutually perpendicular components of peak local ecceleration (i.e., the zero period acceleration). A diamond tipped scriber at the end of an amplifier arm records a permanent mark on a record plate made of aluminum and successive layers of nickel, gold and burnt gold. Again, this system is totally self-contained and requires no outside power source. Two instruments of this type were used and were located on the Auxiliary Building Foundation Mat at an elevation of approximately 568 feet and on the Reactor Recirculation Pump at the elevation of approximately 605 feet. The latest calibration date for the Auxiliary Building instrument was January 30, 1986, while the calibration date for the Recirculation Pump instrument was December 4, 1985. A third instrument of this type was out of service at the time of the earthquakt because it was being recalibrated. See Appendix B for further festrumentation details and data tabulation. All recorded data f rom the in plant seismic instruments have been i ( used in the evaluation. O 5.2
Page1 PERRY NUCLEAR POWER PLANT UNIT NO.1 SEISMIC MONITORING INSTRUMENTATION T A B L E 5.1
," Type Manufacturer / Model Number Location References Reactor Building D51-N 101 (1) Kinemetrics / SMA-3 un at e ,(ation 9 y _ g. F gures A and 8 Azimuth 175' Reactor Building nt "
D51-N111 (1) Kinemetrics/5MA-3 f,e ,at 5 Figures A and C 686* 0. ' Azimuth 174* Reactor Recirculation Pump D51-R 120 (2) Engdahl / PAR-400 I '"$'d r d' Ele a 0" Figures A and D 5 A pro ate ) Azimuth 145' i
! D51-R130 (2) Engdahl / PAR-400 OUT OF5ERV1CE Auxiliary Building D51-R140 (2) Engdahl / PAR-400 Foundation Mat l
(HPCS Pump Room) Figures A and E Elevation 568*-4" I
- 1. Triaxial Time-History Accelerograph j 2. Triaxial Peak Accelerograph
- 3. Triaxial Response Spectrum Recorder
- - ~ __-___
O Page ? o D PERRY NUCLEAR POWER PLANT UNIT NO.1 SEISMIC MONITORING INSTRUMENTATION T A B L E 5.1 In r ent Type Manufacturer / Model Number Location References Reactor Building U"" d D51-R 160 (3) Engdahl / P5R- 1200-H / V-12A Figures A and F o 7 ' 10 Azimuth 225* Reactor Beilding 630' Platform D51-R170 (3) Engdahl / P5R- 1200-H / V I "$'d r I ., E a 0 Figures A and G Azimuth 238' Auxiliary Building D51-R180 (3) Engdahl / P5R-1200-H / V F undati n Mat ( HPCS Pump Room) Figures A and H Elevation 568'-4" Auailiary Building D51-R190 (3) Engdahl / P5R-1200-H / V Foundation Mat p; , ( RCIC Pump Room) Elevation 568*-4* l l 1. Triaxial Time-History Accelerograph l 2. Trianial Peak Accelerograph 3 Triaxiai Response Spectrum Recorder I
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1 I 6.0 PLANT SEISMIC DESIGN EVALUATION The seismic design basis for the Perry Nuclear Power Plant is established by requirements in 10 CFR Part 100, Appendix A and NRC Regulatory Guide 1.60. These regulations require nuclear plant structures and safety class systems and components to be designed to withstand loads induced by a " Safe Shutdown Earthquake" (SSE) for the particular site. The SSE is the strongest earthquake in terms of magnitude of vibratory ground motion that is ever expected to occur at a particular site. The SSE is the design basis earthquake considered for plant licensing. A second seismic event also considered in designing nuclear plants is the " Operating Basis Earthquake" (OBE). The OBE is the strongest earthquake considered likely to occur at a particular site and is at least one-half of the SSE. Operations may resume following an earthquake which exceeds the OBE after demonstrating that no functional damage has occurred to safety-related plant features. (10 CFR Part 100, Aphndix A, III(c), V(a)). The SSE can be described by means of a " response spectrum;" which depicts the maximum acceleration, velocity or displacement response to an input excitation (here the SSE) at a specified damping value for single degree-of-freedom oscillators of varying natural
^
frequencies. The high frequency end of a response spectrum indicares the "zero period acceleration" (ZPA) associated with the event. The ZPA is equal to the maximum ground acceleration of the ,SSE itself. In the design of any plant, it is difficult to predict the exact shape of postulated earthquake acceleration time-histories and associated ground response spectra. Appendix A of 10 CFR Part 100 therefore requires an expected SSE to be developed by statistically combining the response spectra from multiple his'torical earthquakes. Following this guideline, the NRC has provided in Reg. Guide 1.60 standardized response spectra that can be used in lieu of spectra developed for each site (see Fig. 6.1). These standardized spectra were derived by normalizing and combining spectra calculated from e
v, . numerous sets of historically recorded acceleration time-histories. From these sets of spectra, smoothed response curves (acceleration, velocity and displacement) were generated at a level equal to one standard deviation greater than the mean of the responses. This method provides an 84% level of statistical confidence that responses at any particular frequency will not be exceeded by any future event. Thus, in lieu of having to develop site-specific SSE ground response - spectra, the standardized response spectra of Reg. Guide 1.60 can be used. The standardized spectra need only be scaled up or down to reflect the effective maximum ground accelerations (i.e., ZPA's) expected for the SSE at that site. The SSE design response spectra i are used to dynamically analyze a lumped-mass model of the power plant structures. l l
, 6.1 DESIGN OF THE PERRY PLANT l . .
The Perry design response spectra were derived by using the standard response spectra of Reg. Guide 1.60 scaled to a ZPA of 0.15 g determined for the Perry site. These spectra served as the design response spectra at the foundation elevations for use in designing the plant buildings. From these spectra, a simulated SSE time-history of ground accelerations was developed for each directional component (N-S, E-W, and vertical). The conservatism of these simulated time-histories
- was checked and confirmed by assuring that the response spectra generated from the simulated time-histories envelop the Reg. Guide 1.60 design response spectra (see Fig. 6.2).
Seismic Category I structures were analyzed by applying the simulated time-histories to a lumped-mass model of the entire structure, as shown in Figure 6.3. From this analysis, time-history accelerations at each floor elevation were also derived. These time-histories were then used to derive response spectra for each floor of each main building. The floor response spectra were used in designing the safety class equipment, components, and systems.
- - - - - - - - - - - - - - , + < ,- -s---e, .-,n -,,-.--g,e , nm ,_, ,----,_,w- ,,,r, ,-.,n,, , - - - --n,- -- -- .--, p - , - - .
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! In addition to the conservatism included in the derivation of V response spectra, there were numerous other conservatisms included in the overall, structural design of the Perry structures, systems and components. Examples of some of the more significant conservatisms are as follows:
- 1. Broadening the Envelope of Floor Response Spectra Frequency bands of floor responses spectra were artifically broadened (typically by 15%) to account for possible frequency variations. Responses used for design were thus overestimated for systems having more than one dominant frequency falling into the broadened frequency bands of the floor response spectra.
- 2. Equipment Qualification by Test Equipment qualified by shake table testing used* time-histories simulated from the floor response spectra. The simulated time-histories were generated in such a way that their calculated response spectra envelop the broadened-floor response spectra, which in turn already envelop the original design response spectra. The conservatism of the time-histories f.s increased by this " envelope on top of an envelope" process. Moreover, this process results in simulated time-histories with maximum accelerations much higher than the ZPA's of the floor response spectra.
- 3. Strain Hardening Not Accounted For and Static A11owables Used for Dynamic Load In equipment design, material is assumed to behave linearly up to the yield point, then to deform continuously to collapse when the O
( ,f external ioad is maintained. All material used in equipment design exhibits characteristics of strain hardening. This means that resistance'to deformation increases after the deformation exceeds the yield point. Furthermore, even if no strain hardening is assumed, the material can resist dynamic loads having peak values higher than the yield strength through the absorption of energy in the plastic region.
- 4. Loading Combinations The plant was designed to withstand loading combinations with a very low probability of simultaneous occurrence. For example, some load combinations included seismic loads, hydrodynamic loads, and loads due to a hypothetical loss of coolant accident.
This results in design capability well above the loads associated with seismic alone. () 5. Allowable Stresses Computed seismic stresses used in design were considered to be primary, non-self-limiting stresses instead of secondary stresses with a self-limiting nature. The actual behavior of seismic stresses is somewhere between a primary and secondary nature. Consideration of seismic stresses as primary stresses results in overestimated values used for design.
- 6. Damping Values Conservative damping values were employed at Perry pursuant to NRC Regulatory Guide 1.61. The recent ASME code case N-411 allows increased damping values to be used in the design of nuclear power plant piping systems.
O
l v, . i l l O One example of just how significant these types of design conservatisms are is the response of the El Centro Steam Plant (in California) to the 1979 Imperial Valley earthquake. The El Centro Steam Plant was designed to withstand a 0.2 g static lateral load. The recorded peak horizontal load at the site was 0.5 g. The station tripped when station power was lost. One unit was restored to service in 15 minutes and another one in 2 hours. According to calcuations performed by Lawrence Livermore Laboratories, the actual loads experienced by the plant were 2 to 9 times higher than the design values. The plant, however, suffered essentially no damage. The El Centro case shows that an engineered structure can indeed resist seismic loads many times higher than their design values. 6.2 EVALUATION OF THE JANUARY 31 EARTHQUAKE The USGS determined the magnitude of the January 31, f986 earthquake to be M = 4.9 with an epicenter at about 11 miles (17.6 Km.) south b of the Perry Power Plant site. This is of much less magnitude than the earthquake for which the plant was designed (the SSE) and contained substantially lower total energy than the Perry SSE. Evidence of the low energy content of the January 31 earthquake is shown by a comparison of the acceleration time-histories it induced at various elevations with the corresponding design acceleration time-histories. (See Figs. 6.4 through 6.9). The time-histories used for design are 22 seconds long and of sustained high magnitude (strong motion). By contrast, the January 31 time-histories are about 5 seconds long and contain strong motion in only less than a one-second interval (total) of the event. A comparison of Figures 6.1 and 6.10 gives a further indication of the low energy content of the January 31 earthquake. These figures show that the Reg. Guide 1.60 spectra used for design have much broader frequency contents than those of the recordad earthquake, which contain strong motion only at high ' frequencies. The design earthquake therefore contains much greater total energy.
The maximum relative displacements f rom the recorded time-histories of the recurded earthquake are shot:n in Table 6.1. A comparison of the total square-root-of-the-sum-of-the-squares (SRSS) recorded relative displacements with the SSE and OBE t'alues shows that the recorded displacements were all far below those values. For example, the overall relative disp'lacement shown in the Table is 0.36 cm for the SSE and 0.10 cm for the actual event. Since stresses in the structures are proportional to relative displacements, and the recorded relative displacements were far less than the SSE design values, the stresses induced by the 1986 earthquake were all well within design capabilities. Table 6.2 compares the structural response ZPA's of the recorded data with those of the SSE and OBE. The SRSS comparison indicates that the recorded values of the 1986 earthquake vary from significantly below OBE values to 74% of SSE values, except at elevation 686 feet of the , keactor Building Containment Vessel. At that location, the N-S and Vertical acceleration components exceed SSE values, while the E-W acceleration component is less than the SSE value. However, the recorded relative displacements are far less 'han t their design values, as shown in Table 6-1. In addition, recorded response spectra accelerations show that the design response spectra accelerations in certain instances were exceeded at the high frequency end of the spectra. At lower frequencies (at or below approximately 14 Hz) the recorded accelerations are all well under the design values (see response spectra comparisons in Appendix D). The measucement of accelerations outside the predicted responses at the high frequency ends of certain response spectra has no engineering significance. This is explained by the interrelationships among the frequencies, accelerations, velocities, and displacements associated with a seismic event. In general, high frequency acceleration responses have correspondingly low velocity and displacement responses. The 1986 earthquake accelerations occurred at very high frequencies. Therefore, despite some recorded maximum i i 1
y acceleration responses which exceeded SSE values at higher frequencies, corresponding velocities and displacements (and resulting stresses) were nevertheless acceptably low. As discussed, the significant indicators of structural stresses are the relative displacements, and Table 6.1 indicates that relative displacements (and thus stresses) caused by the 1986 earthquake were very small. This is consistent with the high frequency nature of the disturbance. The high frequencies combined with the short duration resulted in an earthquake that contained very low total energy compared to the SSE. The maximum recorded velocity at the top of the Reactor Building foundation sat during the 1986 earthquake was 0.87 inches /sec (2.21 cm/sec). This can be compared with the Bureau of Mines (B0H) velocity threshold for no damage to non-engineered buildings, which is 1 inch /sec (2.54 cm/sec). This shows that the BOM considers it
' acceptable for blasting work to induce velocity waves in nearby residential housing foundations that are greater than the maximum velocities induced by the 1986 earthquake st the Perry Plant. This example helps provide perspective on just how low the velocities and energy content of the 1986 event were.
As discussed earlier in this report, extensive plant inspections have indicated that no structural damage resulted form the 1986 earthquake. This is as expected based upon the low energy, short duration, and low velocity and displaceant of the event. Although some hairline cracks in the structural concrete were documented during plant walkdowns, this does not constitute damage. Reinforced concrete structures are expected to shor aairline cracks. Regardless of their cause, such cracks have no effect on the strength and integrity of the structures. Moreover, such cracking is judged not to be attributable to the 1986 earthquake because of the low magnitude of the event. O
O Section 7.5 of IEEE 344 " Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations," was employed at Perry. This standard recognizes that short duration /high frequency / low energy input motions will not cause significant structural stresses. Instead, it requires qualification
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by long duration / broad band frequency /high energy testing to provide conservatism. As discussed earlier in this report, all energized plant equipment functioned during this event as designed. To confirm the lack of
- impact of the high frequency accelerations on plant equipment, CEI is comparing the qualification data for equipment listed in Table 6.3 against recorded response spectra. Although still ongoing, the evaluation to date shows that the original conservatism in the equipment qualification was more than adequate to accomodate the ,
recorded event. , 6.3 EVALUATION OF SPECIFIC DATA In light of the above discussion, recorded responses at particular locations can be evaluated. At all four instrument locations recording response spectra, SSE design spectra are all well above the recorded spectra in the frequency range of 1 Hz to 14 Hz (see Figs. D1 through D12). These figures compare recorded data with the appropriate design spectra at adjacent elevations. These figures also compare the data from different types of seismic instrumentation at the same elevation. At higher frequencies, the design spectra are exceeded by recorded values in certain cases. However, the corresponding displacements based on recorded data are all extremely small (on the order of several one-hundredths of an. inch) at 20 Hz. These extemely low displacements conform to the above analysis demonstrating that the stresses at higher frequencies are insignificant despite acceleration exceedences. O
In evaluating all the spectra data recorded at the various locations, it was noted that the acceleration responses at the Reactor Building Platform outside the Biological Shield Wall varied from the general pattern of responses recorded at the other three locations. The recorded N-S and E-W acceleration components for this location are all well-enveloped by the entire range of the SSE spectra while the recorded vertical acceleration component exceeds the SSE spectra at the high frequency end (see Figure D-9). This response may be due to the fact that this particular Engdahl PSR-1200 instrument is located near multiple supports and piping system snubbers and components. Actuation of snubbers or local loads induced by nearby components may thus have influenced the recorded vertical response. Such impacts would be of a local, secondary nature. Regardless, the low energy, short duration, high frequency nature of the event indicates that these accelerations had no structural significance. Indeed, the recorded displacement spectrum value is only 0.023 inches (0.06 cm) at 25 Hz at this location. O In general, the high frequency acceleartion content of ground motion will be filtered out by buildings and thus wil1 not appear at higher elevations. This is due in part to the low participation factor generally associated with modes at the higher frequencies. This phenomenon is exhibited by the responses recorded at the Reactor Building sat and elevation 686 feet of the Reactor Building Containment Vessel. A very high frequency p-wave was recorded at the Reactor Building foundation mat. The time-histories shown in Figures 6.4 through 6.9 indicate that this p-wave (appearing during the first second or so of the time-histories) was filtered out by the building and did not appear at elevation 686 feet. There was a response in the range of 20 Hz that was transmitted to the higher elevations. The explanation for this involves the structural characteristics of the buildings on the Reactor Building foundation mat. The Reactor Building consists of multiple structures sitting on O
i a common foundation mat--a concrete shield building, steel contain:nent vessel, concrete drywell wall, and biological shield wall. The structural response of each building influences the responses of the others. The frequencies, mode shapes and participation factors of the two most dominant vibration modes are at roughly 4 Hz and 18.4 Hz, as shown in Figures 6.11 through 6.13. These two dominant frequencies correspond to the peaks at 4 Hz and 20 Hz on the recorded spectra for the Reactor Building at the mat and elevation 686 feet. The input motion at 20 Hz (corresponding to the s-wave) was amplified by this latter mode with some rigid body motion. The 20 Hz input was thus not filtered out but did appear at the higher elevation. As discussed, the acceleration peaks at 20 Hz at this location correspond to very small relative displacements and thus are not significant in an engineering sense.
6.4 CONCLUSION
The 1986 Ohio earthquake was a low energy, high frequency, short V duration, low velocity, and small displacement event. As a result of these characteristics and the above discussions, the 1986 earthquake had no adverse effects on the Perry structures, systems, or components, and no changes to the Perry seismic design basis are . required. O ~
O TABLE O 6.1 O i s
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Comparison of Design Displacements' VS Recorded Displacements' ( Expressed in centimeters /one inch = 2.54 cm)
; COLUMN 1 COLUMN 2 C O L U M N 2 minus C O L U M N 1 Reactor Building Reactor Building Foundation Mat Containment Vessel Relative Displacements Elevation 574*-10" Elevation 686* for the SMA-3 ( Kinemetrics) SMA-3 ( kinemetrics) Containment Vessel D51-N101 D51-N111 Recorded 0.09 0.17 0 08 NS SSE 0.044 0.28 0 24 OBE 0.023 0.17 0.15 Recorded 0.16 0.21 0.05 EW SSE 0.044 0.28 0.24 OBE 0.023 0.17 0.15 Recorded 0.05 0.07 0 02 VERT. SSE 0.02 0.37 0.017 OBE 0013 0.022 0.009 Recorded - -
0.1 8 SRSS SSE - - 0.34 OBE - - 0.21 1 Displacements based on same time-step to determine relative displacements
- 2. Square-root-of-the-sum of the squares
i O O O TABLE 6.2 Comparison of Design ZPA's' VS Recorded ZPA's ( Expressed in g values) Auxiliary Building Reactor Building Reactor Building Reactor Building Reactor Building Foundation Mat Foundation Mat Recirculation Pump Platform Containment Vessel i Elevation 568' Elevation 574*-10" Elevation 605' Elevation 630' Elevation 686* l PAR 400(Engdahl) SMA-3 (Kinemetrics) PAR 400(Engdahl) Inside Drywell SMA-3 (Kinemetrics) j D51-R140 D51-N101 D51-R120 PSR 1200 (Engdahl) D51-N111 D51-R170 Recorded .17 .18 .32 .09 .55 NS SSE .17 .18 1.06 .48 .40 OBE .10 .10 .86 .40 24 Recorded .06 .10 .11 .16 .18 EW SSE .20 .18 1.06 .48 .40 OBE .10 .10 .86 .40 .24 Recorded .03 .11 .05 Note 2 .30 VERT. SSE .20 .18 .47 .28 .24 OBE .10 .10 .38 .16 .15 Recorded .18 .23 .34 Note 2 .65 SRSS' SSE' .33 .31 1.57 .73 .62 OBE .17 .17 1.27 .59 .37 l
- 1. Zero period acceleration of structural response l 2 ZPA indeterminable from available data
- 3. Square-root-of-the-sum of the squares
- 4. Licensing basis is SSE
i TABLE /u3 ' EQUIPMENT LIST AT AUXILIARY BUILDING ELEVATION 568' 1H22P0001 LPCS Instrument Rack 1H22P0017 RCIC Instrument Rack 1H22P0018 RHR Instrument Rack A 1H22P0021 RHR Instrument Rack B 1H22P0055 RHR Instrument Rack C 1C61N0001 Differential Press Transmitter 1E12N0007A,B Differential Press Transmitter 1E12N0015A,B,C Differential Press Transmitter
, 1E12N0026A,B Pressure Transmitter 1 Pressure Transmitter 1E12N0028 1E12N0050A,B Pressure Transmitter 1E12N0051A,B Pressure Transmitter IE12N0052A,B,C Differential Press Transmitter 1E12N0055A,B,C Pressure Transmitter 1E12N0056A,B,C Pressure Transmitter 1E12N0058 C Pressure Transmitter 1E21N0003 Pressure Transmitter 1E21N0050 Pressure Transmitter 1E21N0051 Flow Transmitter 1E21N0052 Pressure Transmitter
" O 1E21N0053 1E21N0054 Pressure Transmitter Pressure Transmitter Pressure Transmitter 1E31N0075A 1E31N0077A Pressure Transmitter 1E31N0083A,B Pressure Transmitter 1E51N0003 Differential Press Transmitter IE51N0050 Pressure Transmitter
; IE51N0051 Differential Press Transmitter 1E51N0053 Pressure Transmitter IE51N0055A,B,E,F Pressure Transmitter 1E51N0056A, E Pressure Transmitter 1E12C002A RHR Pump & Motor IE1200028 RHR Pump & Motor 1E12C002C RHR Pump & Motor
( 1E21C001 LPCS Pump & Motor l 1E220001 HPCS Pump & Motor l O i
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7.0 CONFIRMATORY PROGRAMS Within hours of the earthquake, CEI's geophysical corsultant had set up seismographs in the area of the epicenter to monitor any aftershocks. These remain in place at this time and the monitoring will continue until it is determined that no further af tershocks are anticipated. In addition, CEI is cooperating with the U.S. Geological Survey and others who are studying the earthquake. CEI has instituted a specific procedure (OM19A: GTI-003) to ensure proper documentation, review, and reporting of all potentially earthquake related conditions in the plant. Under the procedure, all
, of the items identified within 24 hours following the seismic event have been documented as Earthquake Inspection Team Items ("EITI's").
Engineering has evaluated each EITI to determine whether the item was a direct result of the earthquake. The results of the evaluation are shown in Appendix E. The two EITI's determined to have been caused by the earthquake, and those with an " indeterminate" cause (i.e.,
/O where it cannot be definitively establisned that' the condition U existed prior to the earthquake), were identified and documented as discussed above. None of these items 1s associated with any plant structural damage. It is anticipated that minor rework or repair will be done on some of the items in accordance with CEI's normal program to correct nonconforming conditions. CEI's procedure provides that all potentially earthquake related EITI's will be maintained in the "as found" condition until reviewed by CEI and released by the NRC. .
4 New Work Requests (WR's) (for conditions other than those already covered by EITI's), are also being reviewed in accordance with CEI's new procedure for earthquake related items. e 7.I
Engineering evaluation results for these items are being docum'ented and tra'cked. As with the EITI's, any potentially earthquake related [ conditions associated with new WR's are being maintained in the as-found condition until reviewed by CEI and released by the NRC. CEI has not identified any plant structural damage associated with potentially earthquake related items identified on new WR's. On a longer term basis, CEI is participating in several industry efforts to study the effects of seismic events on nuclear plants. The organizations performing these studies include the Seismic Owners Group (SOG), the Seismic Qualification Utilities. Group (SQUG), and Electric Power Research Institute (EPRI). These industry groups are examining various gneneric seismic issues which have been under consideration by the NRC. For example, SOG has been focusing on eastern seismicity hazard analysis, with EPRI managing the program effort. SOC will review the Perry earthquake as part of this work. SQUG has focused its effort on the seismic qualification of electrical equipment. SQUG intends to review the . m . Perry data presented in this report, and will integrate this informatton into their studies. EPRI has been supporting SQUG by sponsoring projects to resolve issues associated with equipment qualification, focusing on test data, adequacy of equipment anchorages, and post earthquake investigation programs. These industry groups all visited the site shortly after the seismic event. A SOG/EPRI team installed in-plant and field instruments within a day of the seismic event to collect aftershock data. An SQUG team conducted a plant walkdown. The team informed CEI that the seismic event at Perry was much smaller than others they have evaluated (Coalingo, Chile, Mexico City, Morgan Hill), and that the SQUG data base generated from these previous earthquakes would predict no damage from the Janaury 31, 1986 earthquake. This prediction was confirmed by the group's plant walkdown. The EPRI equipment qualification program manager concluded that Perry's response to the seismic event was properly handled. The Perry experience will be used in EPRI's development of generic post-earthquake investigation methods. ( l 7.2 i
i l
")/ 8.0
SUMMARY
AND CONCLUSIONS
%/
The seismic event which occurred on January 31, 1986 has been thoroughly studied and its effects on the Perry Nuclear Power Plant
. analyzed in dttail. The earthquake itself was of smaller magnitude and intensity than the postulated earthquake which was used as the basis for the plant seismic design. The occurrence of the 1986 earthquake does not change any of the conclusions previously reached as to the geology and seismology of the site. Considecation of this event does not result in any change in the Safe Shutdown Earthquake licensing basis for the Perry plant.
The earthquake confirmed the adequacy of the plant's seismic design. The plant structures and equipment were essentially unaffected by the earthquake. The large number of safety and non-safety related systems which were operating or energized at the time of the earthquake responded in accordance with their design. Extensive .
,<~3 ,
plant walkdowns and inspections revealed no structural or equipment h damage. The seismic characteristics of the earthquake have been reviewed and compared the plant's seismic design. The high frequencies which typified the 1986 earthquake are of no significance with rego to the adequacy of the plant's design. In contrast to the scismic - design basis, the earthquake was of short duration, with low energy, low velocities and small displacements. Although certain of the
' recorded response spectra exceeded the design response spectra in the high frequency range, such exceedances are consistent with the analytical methods of Regulatory Guide 1.60 and are of no engineering significance. In the frequency range of significance for plant structural design (below 14 Hz), recorded spectra are f ar below the design response spectra for Perry.
The January 31, 1986 earthquake, in effect, constituted a proof test of Perry's seismic design. By any standard the Perry Nuclear Power Plant passed that test. The earthquake presents no new information ( which would change the previously accepted licensing basis for the plant. 8.1
O APPENDIX A STRONG-MOTION DATA FROM THE PERRY NUCLEAR POWER PLANT SEISMIC INSTRUMENTATION KINEMETRICS O l 0
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$ 5.0 EARTHQUAKE of 1147 EST, JANUARY 31,1986 PERRY, OHIO i
RECORDED ON THE PERRY NUCLEAR POSTER PLANT STRONG MOTION ACCELEROGRAPHS l for i Cleveland Electric Illuminating Company Requisition No. NED-E-860006 8 R Kinemetrics/ Systems 222 Vista Ave.
. Pasadena, CA 91107 I Sales Order C-K6028
- [ February 4, 1986
A y M TABLE OF CONTENTS . a L Page l
1.0 INTRODUCTION
........................................ 1 2.0 INS TRU MEN TAT ION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.1 Model SMA-3 Accelerograph............................I
'B 5
2.2 Calibration Deta.....................................I 3.0 D ATA PRO C ES S I NG . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3.1 Digitization........................................ 2 3.2 VOL1 Processing..................................... 2 3.3 VOL 2 Pr o c e s s i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
~, 3.4 VOL3 Processing..................................... 3 DATA PLOTS i
Uncorrected Acceleration Corrected Acceleration, and Integrated Velocity and Displacement Velocity Response Spectrum with Fourier Spectra
~
Tripartite Presentation of PSV,PSA and SD for triaxial response at each of: Reactor Building Foundation, El 575', l Containment Vessel Annulus, El 682' APPENDICES -
" Conditioning and Correction of Strong Motion Data on on Analog Magnetic Tapes" SMA-3 Data Sheet l
1 -
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1.0 INTRODUCTION
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On January 31,1986, a (Mt 5.0) local earthquake was recorded by the strong-motion instrumentation at Perry Nuclear Power Plant, Perry, Ohio. The FM analog magnetic tape cassette records from two Kinemetrics Model SMA-3 accelerographs were retrieved from the instruments and provided to Kinemetrics for analysis.
; This report describes the processing of these strong-motion records and presents the results. Included are the uncorrected 3 accelerograms, corrected acceleration, velocity and displacemt.nt L time series, and response spectra. 1 I
2.0 INSTRUMENTATION 2.1 Model SMA-3 Accelerograph The SMA-3 is a multi-channel, centralized recording, FM analog I magnetic tape accelerograph system designed to detect and record strong local earthquakes and record the three orth ogonal acceleration signals on cassette tape. The SMA-3 remains in a standby mode until its vertical trigger detects an earthquake. The trigger then actuates recording in less than .10 seconds. l] The force balance accelerometers in the SMA-3 have a nominal W natural frequency of 50 Hz and damping of 65% critical, providing flat (-3dB) response from DC to 50 Hz. The nominal 3 sensitivity of each of the three channels is 2.5 volts /g with a E full scale response of 1.09 The dynamic range of the accelerograph is nominall7 40 dB, giving it a resolution of approximately .01g. The trigger in the SMA-3 has a flat (-3dB) response from 1 to 10 Hz and a nominal trigger level of 0.01g. Power is supplied to the SMA-3 by internal rechargeable batteries. These batteries are kept in a charged state by 120 VAC line power. o ,
2.2 Calibration Data The three Model SM-3 accelerographs which recorded' the event
._ were factory calibrated in January, 1985, and the sensors were recalibrated for sensitivity by the Perry NPP personnel in December of 1985. These most current calibration data are given in Table 1 below.
1 i l I Ser. No. 165-1 Channel long Sens., v/g 2.48 Nat. Freq., Hz 52.3 Damping t critical 65 tran 2.49 53.7 65 vert 2.47 50.6 64 165-2 long 2.48 52.6 67 tran 2.48 52.2 72 vert 2.65 50.5 66 l TABLE 1: Calibration Data 3.0 DATA PROCESSING 1 Data from the Model SMA-3 accelerographs were played back using a Kinemetrics Model SMP-1 Playback System through a Data Compenstor, digitized using a Kinemetrica Model DDS-1105 Digital Data System and processed as described in Kinemetrics' Aplication Note No. 7 " Conditioning and Correction of Strong Motion Data on Analog Magnetic Tapes", appended to this report. 3.1 Digitization The magnetic tapes were digitized using the DDS-1105. The 1024 Hertz FM time reference recorded on channel 4 of the cassette is output from the SMP-1 and divided down by four (256 Ez i deviation) and used as the timing signal for the digital
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, units using the Table 1 calibration data. The mean was then subtracted from each acceleration time history. The new time histories file.
were then written in a Kinemetrics' VOL1-format disk The three uncorrected acceleration time histories from each SMA-3 record were then plotted; these plots are included in the data section of this report. 3.3 VOL2 Processing The recorded accelerograms were then instrument and baseline corrected using Kinemetrics' VOL2 program. This program is based upon the VOL2 program developed at Caltech (Trifunac and I Lee, 1973). No major modifications to the original VOL2 algorithms have been made. The data were bandpass filtered using Ormsby filters. The low-pass filter had a cut-off frequency of 35 Hz and a termination frequenc of 40 Hz. The high-pass filter had a cutoff frequency of .625 Hz and a termination frequency of 0.4 Output of this program consists of a plot of corrected 'W acceleration, velocity and displacement for each conponent of
,. recorded data. These plots are presented in the data section of this report.
.- 3.4 VOL3 Processing Linear response spectra were calculated from the corrected acceleration time histories using the algorithms developed by Trifunac and Lee. Response spectra were calculated for damping l ratios of 0, 1, 2, 4, and 7 percent. The period range of these spectra was 1.68 to 0.0283 seconds (0.59 to 35.4 Bz) with oscillator response calculated at 1/24 th octave intervals. Two types of plots were produced and are included in the data
, section of this report. The first type is the traditional tripartite log-log plot of pseudo-velocity vs. period. Th e second is a linear plot of velocity response and Fourier spectrum vs. frequency.
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G g l 0 R N 1 1 1 P M L A I)I lI D , g I
, 1
. 0 I. I '
_ 0 8 A 1 1 - _ - - - - - - O 0 o, 0 0
- o. '
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! i
__ 2 y rw 'ans una nei - m' -> > .
.. ..e C) # 1.O, RELAT1VE VELOC1TY RESPONSE SPECTRUM ML 5.0 EARTHQUAKE JANUART 31. 1986
! 11AB001 PERRY NUCLEAR POWER PLANT COMP WEST SMA3S/N 165-If DAMP 1NG VALUES ARE 0, 1. 2. 4. 7 PERCENT OF CRITICAL 10 0 i l l i l i : l SV O j w 3.0 - -------- FS _ N i 2". i , - I. > L.0 - ! d 5
< 4.0 -
i u
\
9, i 20 - I I 1 p\ ~
\
W ,'
\
l J.1 ji -
'.. / ~.. _
.. 1 I I l l t
* *0 10 20 30. 40 50.
FREQUENCY - H2
/// /
NNs s N KiNEMEiRICS f
g m '" ' * *
- ggese _ m '" ~ ~
amve ' e %, , ,, , ,_,
~l RELATIVE VELOC1TY RESPCNSE SPECTRUM ML 5.0 EARTHOUAKE JANUARY 31, 1986 11A8001 PERRY NUCLEAR POWER PLANT COMP UP SMA3S/N 165-IV DAMP 1NG VALUES ARE 0. 1. 2. 4. 7 PERCENT OF CRITICAL , 20 0 i l l l i l s s I l SV o
w ig,o _ -------- FS __
.S.
D.= u12 0 - 3 N c - 3 s.0 - E i -
?
a.e - t - k I
, I N
- _i '
v ' N l 0.0 .' 4) 0 f i 10.
' 'l20. !
__,_-N
' ' f !
J0. 40. 50. FREQUENCY - H2
/// / '
NN s \ s KiNEMEiRICS
! ML 5.0 EARTHOUAKE JANUARY'31, 1986 i- 1 liA8001 *ERRY NUCLEAR POWER PLANT COMP SOUTH SMA35/N 165-1L l DAMPING VALUES ARE 0. 1. 2. 4. 7 PERCENT OF CRITICAL f'm FREQUENCY - HZ _3 V e 10 siis e i i 1
- e i i 10 3 i i e i e i i sis i i-10
_ d _ PSA - G SD - IN
~
lL'* 0 1 rv - g _ sjg , 10
, SD - CM _-
l
. 1
. 10 -
0'
~
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3 7 1 [ ~
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0 ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' lO' 1 10 PERIOD - SEC / /# [
--~ - - - - ..--_. pt 1 N f ft f f 9 i C I m
~
ML 5.0 EARTHOUAKE JANU RY 3i, 1986 11A8001 PERRY NUCLEAR POWER PLANT COMP WEST SMA35/N 165-1T
~
DAMPING VALUES ARE 0. 1. 2. 4. 7 PERCENT OF CRITICAL
CREDUENCY - HZ _i a 10 1 10 3 i..i . . . g.. . . . . g... .-10 1 -
PSA - G SD - IN I'* i - 0 1 l 2
. si g -
io
~
SD - CM 10 _ g
~
9 m . ' Io , .
' ~ )
g'. % . g s,- I
~
[ i . o-O P
, 3
_1 _ 10 1 10 PERIOD - SEC p 1 re E ti t T R 1 C E
ML 5.0 EARTHOUAKE JANUARY 31. 1986 IIA 8001 PERRY NUCLEAR POWER PLANT COMP UP SMA35/N 165-1V 2
, DAMPING VALUES ARE 0. 1. 2 4. 7 PERCENT OF CRITICAL
[ FREQUENCY - HZ _; A 10 1 , 10 3 i i . ii - i iiii i i . i i i siiii-10
- 1 PSA - G SD - IN ~
I ? 0
~
1
~
~ 2 I IO
%1[
- l. -
SD - Cn : I 1
. 10 -
6 i
?
1 0
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l t t t t t t t l e t t f f f f f f f I t f t t t t I
~
' 10' 1 10 PERIOD - SEC'
-- -----__ _. ____ _ - . pt 1 ft E r1 E I ft 1 C E
ML 5.0 EARTHOUAKE JANUARY 31. 1986 IIA 8001 PERRY NUCLEAR POWER PLANT COMP 5007.9 SMA35/N 165-1L f DA!1 PING VALUES ARE 2 PERCENT OF CRITICAL
'l FREQUENCY - HZ ~'
d 10 1 , 10 3 i i i.. ii i . . . .. . . . . . 4si i i-10 s . - PSA - G SD - IN 10 - W j0 18 h $ '.',
'. sig -
10 2
,l -
s0 - Cn l - 1
,'l 10 .-
0 -
.O' 1
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- 1 -
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.g 10 1 10
~
L
ML 5.0 EARTHOUAKE JANUARY 31. 1986 IIA 8001 PERRY NUCLEAR POWER PLANT COMP WEST SMA3S/N 165-1T DAMPING VALUES ARE 2 PERCENT OF CRITICAL FREDUENCY - HZ _. 10 1 - 10 3 i... . . . . ,i.... . . . . ii . -10 1 , PSA - G SD - IN 2 - L' 0 1 10 50 - Cri 1 I i
, 10 0 -
N , , 3 ry ' 1 0
~
i.t._, E
. u, s,
10 m . . O - _' 2 . E
'1 0' ' -
0' f ~ 1 [2 . o g
~
3 _1 _
-1 ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' '
s0 -1 10 1 10 PERIDO - SEC
. _ _ . . ._ . _ . . . _ _ _ _ _ . 5 I N E N E I ft 1 C E
; ML 5.0 EARTHOUAKE JANUARY 31, 1986 11A8001 PERRY NUCLEAR POWER PLANT COM? UP SMA35/N 165-1V ~
DANPING VALUES ARE 2 f'ERCENT OF CRITICAL FREQUENCY - HZ _3 ] 10 10
..-103 l 1 .
iiie i a e i i gi.. . i . . i ei id
~
PSA - G SD - IN ~ 2 g 10 - iR - 0 1 up e 0'
. ,lg , ~
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10 _- , ,
- _3 i 40 i -
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1.1
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l 10* - S _3 _1 _ l f f f I f f l t t f I t t f f l e t t i t t 8 t I I
.. 10 1 10 PERIOD - SEC
, ML 5.0 EARTHOUAKE JANUARY 31, 1986
- 11A8001 PERRY NUCLEAR POWER . PLANT COMP SOUTH SMA35/N 165-1L
'". DAMP 1NC VALUES ARE 4 PERCENT OF CRITICAL l
FREQUENCY - HZ V) 10 1 . 10 _3 3
..i . 6.. . . . . i iii. -10 e p PSA - G SD - IN e . -
L10 -
~
^ *
,0 1d
- ' 2
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- _i 5 s
1 0 o NO p { 10 2-
@ 0 C
l(
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D ' g # U '1 0' ' - U 1
- -2 g-s g _
i
/
6 _1 , SW Sus t t I 1 if f l f f f I f f I f ! f f f I f f ! _3 10 1 10 PERIOD - SEC /
. _ _ . _ E178Ef1ETRICE
r
; :1L 5. 0 E ART HOU A KE JANUARY 31, 1986 11A8001 PERRY NUCLEAR POWER PLANT COMP WEST EMA35/N 165-1T
~
DAMPING VALUES ARE 4 PERCENT OF CRI.iCAL
*REQUENCY - HZ -3 1, 10 1 10 3 4 . .
iiii e i i sia i i e i i sia ii-10 PSA - G SD - IN (10 , 0 1
~
l
. 2
~
'sj g 7 10 Il -
. s0 - Cn :
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1
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10 -3 10 i 10
, PERIOD - SEC m M [ fi t T R 1 C E
ML 5.0 EARTHOUAKE JANUARY 31. 1986 _ 11A8001 PERRY NUCLEAR POWER PLANT COMP UP SMA35/N 165-1V
'~
DAMPING VALUES ARE 4 PERCENT OF CRITICAL ( cREQUENCY - HZ 3 10 1 10 3 iii. . . . . . . . . . . . . .....-10 e
.PSA - G SD - IN ,
2 . 10 - 0 1 2 sjf 10
. SD - CM :
B ', .
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rv W - 1 > 10 5 {-- -
, e 10 E f
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\\
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+
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, PERIOD - SEC
- . _ - . _ _ _ . - . . _._ rincncir es
J .7= -
., ._=..-.-.- _, .
N g . v I
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?
Conte.inment Vessel Annulus, Elevation 682 Ft. SMA-3 Serial Number 165-2 g Tag Number D51-Nill Longitudinal channel - South Orientation
, Transverse Channel - West Orientation L Vertical Channal - Up Orientation 8
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NNsN N , K1NEMETRICS ML 5.0 EARTHOUAKE JANUARY 31. 1986 11AB002 PERRY NUCLEAR POWER PLANT COMP UP SMA3S/N 165-2V AttELEROGRAM 1S BAND-PASS FILTERED BETWEEN 0 400- 0.625 AND 35 00- 40.00 HERT2 cf PEAK VALUES ; %CCEla *297.21 CN/SEC/SEC VEL = 43.09 CM/SEC DISPL-0 07 cn
; +500 -
r.
~
e i : g3 a . m_ _ p.,___,__ gu - 300 ' ' J 43.) - n 1 _ b* O O p: -~ .[ jp:-:mev.-=:=- . -- .
- + 0 .1 -
N ui - l l8 0 j~^a j,~.ww x-v ,^n b . 1
'o o
_0.i 0 5 10 15 20 71ME - SECONDS l
I m m p pr=t e men e tw 5 m . . . _ _ , ., , t , w' RELATIVE VELOC1TY RESPONSE SPECTRUM ML 5.0 EARTHOUAKE JANUART 31. 1986 11AB002 PERRY NUCLEAR POWER PLANT COMP SOUTH SMA3S/N 165-2L DAMP 1NG VALUES ARE 0. 1. 2. 4. 7 PERCENT OF CRITICAL so.o , , , , , , , , SV ho,o -. -------- FS __ 2 E o. g30.0 lh - e li' I i _ W ,
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io. 20 ao. 4o. so. FREQUENCY - H2
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MINEMEiR1CS s
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J [ RELATIVE VELOCITY RESPONSE SPECTRUM
, ML 5 0 EARTHOUAKE JANUART 31. 1986 11AB002 PERRT NUCLEAR POWER PLANT COMP UP SMA3S/N 165-2V DAMP 1NG VALUES ARE 0. 1. 2. 4. 7 PERCENT OF CRITICAL 20 0 i i l l i i l l SV u
w ig,o _. -------- FS ._ S i t l G20 O l - 1 I w D I
; s.0 - ,
ci - t c h \ Q l 1 4.0 - d
/' 1 -
g f ? D j I
.l i .- 's s' ~__s IV g 'j, J ' ' N 0.0 O. 10. 20. J0 40 50 FREQUENCY - H7 1
/// /
< < ( < <
\\s s \
KINEMETRICS ,
ML 5.0 EARTHOUAKE JANUARY 31, 1986 IIA 8002 PERRY NUCLEAR POWER PLANT COMP SOUTH SMA3S/N 165-2L DAMPlNG VALUES ARE 0 1. 2. 4. 7 PERCENT OF CRITICAL Os FREQUENCY - HZ s) . . . 10 1
. i i i e 10
_3 iiii-10 3
- i. i PSA - G SD - IN 2 ~
.0 1 2
., e
. e0 rl d . So - Cn :
~
- ( -
1 !g 10 - [ x y 3 -g
-i i -
. M
- /
t l , f 'p _ 1, 0, .N . l.. ? _ fhh j
.e.
s *g _3 - _ 1
~
f f f I f f f f l t f f I f f f f ! f f 1 ! f f f f ! I 10 1 10 PERIOD - SEC //
- -- -- - p 1 es r M r i e 1 C C
ML 5.0 EARTHOUAKE JANUARY 3'. 1986 11A8002 P.ERRY NUCLEAR POWER PLANT COMP WEST SMA35/N 165-2T
.~ DAMPING VALUES ARE 0. 1. 2. 4. 7 PERCENT OF CRIT 1 CAL CREQUENCY - NZ _3 10 1 -
10 3
. . . 34 4 ,a a a e a siis i i i a i 6.i -10
-s
- d _
PSA - G SD - IN 0 1
. 2 s3g -
10 ih - SD - CN : i - 3
'10 0 -
~
Q, . 10 {Ro _ r I (
~ ~
_i -
'10 1
N -2 -3 1 . 40
- ~
'1 0'
~
, 3 -
_s _ e ., 10 'N ' '
_i 10 1
/ x':
10 PERIOD - SEC ____. . _ . . ._}}