ML20247F487
| ML20247F487 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 03/24/1987 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20247F042 | List:
|
| References | |
| FOIA-89-2, FOIA-89-A-7 NUDOCS 8905300094 | |
| Download: ML20247F487 (78) | |
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%i 1 BEFORE THE l l
2 UNITED STATES 3 NUCLEAR REGULATORY COMMISSION 4 REGION V ,
5 6 In the Matter of: )
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7 INVESTIGATIVE INTERVIEW ) DOCKET NO: NONE
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8 (CLOSED MEETING) -)
9 Sunrise Sheraton Hotel 11211 Point East Drive 10 Rancho Cordova, California 11 Tuesday, March 24, 1987 12 An in v e stig a tive interview was conducted 13 '
with. R. PIERRE OUBRE, commencing at 1:00 p.m.
C. 14 PRESENT:
15 ROBERT G. MARSH, 16 Field Office Director Office of Investigation, Region V 17 Nucient Regulatory Commission ,
18 19 20 21 22 s 23 24
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8905300094 890516 PDR FRIEDMA89-A-7 FOIA PDR ND _
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1 CONTENTS 2 WITNESS PAGE 3 R. Pierre-Oubre 4 Examination by Mr. Marsh ,
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1 10 11 EXHIBITS i
4 12 (None) 13 C 14 15 ,
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o 1 PROCEEDI NGS 2 1:00 p,,,
3 MR. MARSH: On the record. I i
4 Mr. Oubre, would you raise your' rig ht hand 5 and be sworn.
6 Whereupon, 7 R. PIERRE OUBRE 8 was c alled as a witness her ein, and, having fir st been 9 duly. sworn, was examined and testified as follows:
10 MR. MARSH: Let the record show that this la 11 an interview of R. Pierre Oubre, an employee of Sacramento 12 Municipal Utility District, Sacramento, Calif ornia. The 13 time is approximately 1:00 p. m., March 24, 1987. ,
14 Present during the interview is a court 15 ste n og r a ph e r . The interview is being conducted by Robert 16 Marsh, Field Office Director, N u cle a r Regulatory 17 commissio n , Offic e of Investigations, Re gion V. .
18 Mr. Oubre, I have explained the nature of 19 the inv estig ation to you, and I wo uld like to st ate for 20 the record that this is an investigation conducted by the 21 Nuclear Regulatory Commission's Office of Investig ation of 22 matters concerning the release of radio active e ffluents The 23 from Rancho Seco Nuclear Power Generation Sta tion .
24 questions I will be asking you today are r elative to that 25 investigation.
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1 EXAMINATION i 2 BY MR. MARSH: I 3 Q Have you v ol u n t a rily come to answer 4 questions pertaining to this matter? '.
i 5 A Yes, I have.
6 Q Has anyone of your company, or NRC, coerced 7 you, or offered you any reward in any way for submitting 8 to the interview?
9 A No, the y have not.
10 I would like for you to understand that, at Q
11 any time that you feel like you would like stop the 12 interview and not answer further questions, you have that 13 right and may exercise it at any time. ,
14 A I understand.
15 Q Mr. oubre, could you explain for us, please, 16 your current po sitio n with the Sacramento M unicip al 17 Utility District -- and we will use the acronym SMUD from 18 this time forward for that company name.
19 A Pre se n tly, I am the Acting Assistant General 20 Manager of Engineering and Operations.
21 Would you like me to go through the histor y 22 of my time with SMUD at this time?
23 Q. Yes, please.
24 A I first came to work for SMUD in about March 25 1970. I came in the position of Assistant Superin tend ent I ,
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4 1 for Rancho Seco Nuclear Generating Station. I was in that 2 positio n until about the middle of 1978. At that ti m e , I 3 was promoted to the position of Plant A ' o rin tend ent for 4 Rancho Seco Nuclear Generating Station. .
5 It was about July 1983, that I was promoted 6 to the position of Plant Manager -- I take that back. The 7 o f ficial title within the District is Manager, N u cle a r 8 Operations Department. I remained in that position until 9 about Aug ust 19 -- Let me back off. It was 1970 to 1978 10 as Assista n t Superintendent; 1978 to 1983 as Plant 11 Superintendents and the middle of 1983 until August 1985, 12 as the manager Nuclear Operations Department; the n , from 13 1985, Aug ust 19 85 un til D e c e mbe r 1986, I was Program C 14 Manager for Power Opera tio ns Management Program. In 15 January 1987, I became Acting Assistant General Manager of 16 Engineering and Operations.
17 Q And you are now currently the Assista n t 10 General Manager; you are not acting?
19 A Acting.
20 Q You are still acting?
21 A I am still Acting Assistan t Gener al Manager 22 of Engineering and Operations.
23 Q. As you have e x plain e d , during the time 24 period of the 1984 period, you were the plant manager f or 25 -- or the Manager, Nuclea r Operations Department. Were e
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t' 1 you specific ally responsible for the operations of the l
2 Ranch seco Plant?
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3 A That is correct, within the context of the te c hnic al specifications, t h a t 's right. ,- !
5 Q Can you expand on how you managed that ,
6 o pe r ation ; what kind of org aniz ation did you have I
responsibility for, and the method in whic h you operated 8 as the responsible party?
8 A Let me assume that you are asking me what 10 organizations reported to me at that time ; is that a way 11 of answering that?
12 From an o pe r a tio n al st a n d p oin t , the 13 operation of the power plant, day-to-day operation of the
'C 14 power plant, came up through the position of plant 15 superin te nd en t. The plant superintendent reported to t,he 16 Mang er, Nuclear Operations Department. .
17 Q Can you tell us vho that' plant 18 . superintendent was?
18 A The plant superintendent was George Coward.
20 Another organization that reported to me was 21 calle d the Superintendent of Technical support. And, in and I can't
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22 1984 -- Let me give you the two names, 23 remember when the tr ansition took effect. Half of that 24 time was a gentleman named Dan Whitney; and the other half 25 of the time was a Jim Field. I don't remember when the
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2 Another organization that reported to me was 3 A d minis tr a tiv e Support, and it was the same in divid u al 4 throughout the ti m e . That was Wanda W ell s. She was 5 r e sp o n sible for, as I say, a dminist r a tio n , p a y r oll, 6 lib r a r y, document control, those kind of items.
i 7 And the last org aniza tion was -- no; there 8 is two more, I am sorry. Second from the last 8 organization was Training, and that was Jack Mau; and the 10 last o r g anizatio n was Sc he d ulin g , Plant S c he d ulin g .
11 Ag ain , that was a tr an sition , the first being a g entlem an 0
12 by the name of Tom Tucker, and then Jim Shetler.
13 MR. MARSH: We will be off the record for a 14 second.
15 (Whereupon, a brief recess was taken.)
16 MR. MARSH: Back on the record.
17 BY MR. MARSH: '
18 Q Tom Tucker and -- who?
18 A Tom Tucker and Jim Shetler.
20 Q Who was responsible for the budgeting of the i
21 plant operations, the financial aspects? l 22 A Well, the fin a n cial aspects, from the I
23 standpoint of budgeting, would have been myself through my 24 boss to the Board of Dire c tor s, or through the General J l
25 Manager. All of those positions, that I just quoted you,
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would ' supply their needs into the budget. The budget 2 would be assembled, and then go up through the chain.
3 All right. For the area of c on trolling the Q
e f flue n ts and the releases of e f fluents, who, in your 5 o r:g aniz a tion would have been r e sp o nsible during that 6
period of time?
7 it wo uld have A During that period of time ,
8 been up through the Radiation P r o te c tion ~ Chemistry 8 Divisio n , through the Plant Superintendent. ]I 10 And who would, can you fonow thst c hin l Q
11 down through the org anization?
12 A Through the plant superintendent?
13 Q George Coward, Plant Superintendent, down.
14 A There was a -- I believe he was there the 16 f un time. He has since retired, but I believe he re tired 16 after -- I as sorry; are we talking about 1984 only?
17 1984, rights at the moment, 1984.
Q 18 A That would have been, for that period of 18 ti m e , that wo uld have been Roger Mille r , who was the 20 supervisor of the c he mistry Ra diologic al P r o te c tion 21 Divisio n. He reported to George Coward.'
22 Au right. Following that down from Roger Q
23 Miner, then, do you know who the --
24 A That would have been, then, Fred KeMie s 25 and, then, under Fred K ellie , h's had about four
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I supervisors all r e s p o n si ble for v a rio us parts of 2 I could not tell you who they radiological and chemistry.
3 were at this time, not since 1984.
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, Q How frequent would your contact.be with the 5 Roger Minor or Fred Kellie level within the plant?
6 A contact, if you meant just see'ing the m7 -
i 7 No; I am t alking about actu any passing on Q .
8 in st r u c tio n s , or me e tin g to work out your plans of 9 o pera tion s .
10 A I wo uld , thinking back to 1984 time frame, I 11 wo uld thin k no more than -- and I as p rio b ably being 12 optimistic on that -- maybe once a wee k.
13 Q So, it would be a g e n e r ally a c c ur a te G 14 statement 'to say that you administered your program, in 15 that regard, through George Coward, down through Roger 16 Mille r ; and, if you wanted so m e thin g done, you w o uld 17 generany ten George Coward what you wanted done?
18 A That is correct.
18 In September 1984, a report, No. 84-07, was Q
20 submitted to the NRC concerning the release of effluents.
21 tare you f amiliar with that report?
22 A I would have to see the report that you are ,
23 speaking of. There are a lot of reports.
24 Q I win show this report to you.
25 (The document was proffered to the witness.)
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MR. MARSH: Let the record r e fle c t that I 2 have provided a copy of the September 27 Report 84-07 for 3 review.
- saw THE WITNESS: I suspect that[J it, f 5 because I am f amiliar with some of the numbers that are in 6 the first page that you are talking about there. That was I
a general subject that was brought up at almost all sta f f 8 m e e tin g s. And there was a lot of, I want to call it 8 research, if you would, that was occurring at that tim e .
10 I knew it, if yo u wo uld, as an informational thing. The 11 researen, the work that was being performed was being 12 performed through the Nuclear Engineering Department.
13 And who was r e spo nsible for the Nuclear Q ,
14 Engineering Department?
15 A The Nuclear Engineering Department was Lee 16 Keilm an , and he reported to the Assistant , General Manager 1
17 of Nuclear. That would have been -- at the time, he was 18 c alled the Ex ec utive Director of Nu cle ar ; and some time 18 later, he was made an assistant general manager.
20 Q Who was that?
21 A Ron Rodriguez.
22 Now, when you say it was a general topic of Q f 1
23 disc ussion at staf f meetings, who would be present at I 24 staff meetings; what staff meetings is that? j 25 A This would have been one of Mr. Rodrig ue z's I ,
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11 I staff meetings, which he held almost every Friday morning.
2 And that would have been nis direct reports, which would 3 have been myself, as the o p e r a tio n s Department; Lee j
- Keilman, as the Engineering Department: Andy Shhweiger, as 5 the Qu ality Assurance Department; Bob -- my mind went 6 blan k for names -- Dieteric h, Bob Die te ric h , Lic en sin g ;
7 Yes; that was all.
and I believe that is all.
8 Q Do you r e c a ll sp e cific all y what your 8 involvement was in putting together this submittal to NRC?
10 A My involvement?
11 Right.
Q 12 A None. I had no in v olv e ment in put tin g 13 together this submittal. This is outside of my expertise.
C 14 Q Was it coordinated with you in any way?
15 A It would not have been coordinated with me 16 dire c tly; it may have coordinate d with Ron columbo. In 17 fact, that is who you direct at the very end. The split 18 within the lic e n sin g o r g a ni z a tio n was, that, 18 communications with the Region cone out of Ron Columbo at 20 the plan t. Ron reported to me as a licensing individual.
21 And, coordination with Washington-type NRC, it would have 22 out of Bob Dieterich.
23 So, in this case ,- it probably was written.
24 I don't know that for a fact. It was probably written by
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25 someone within the engin e e ring department, through Ron
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Ro drig u e z 's sig n a tu r e , be ca use Columbo, for of the 2
asso cia tio n with the Re gio n .
3 Now, sin c e you d e s c rib e d that you are Q
somewhat f amiliar with the memo at this time, ,this me mo is 5 basically describing the operations at Rancho Seco aro und 6 and c o mmits the SMUD the release of e f flu e n ts, 7 certain thing s over a short and org aniz ation to doing 8 long-term process. Is that usual, that you would not have 8 any coordination with you on such a matter?
10 No, no. It would not The que stion you A --
11 asked -- did I have any coordin ation in prod u cing this 12 document? I don't remember producing the document.
13 Going to the second page, where we. ' talk C 14 about bubble te stin g the OTSG lo o kin g for the leak, 15 absolutely. That is, the bubble testing of the OTSG would 16 have been done by operational forces in a shutdown. Then, l 17 the leak test, the helium leak test, would have been done 16 .by operations.
18 The sta te ments concerning the se c o nd ar y l
20 system g e n e r ally flo ws to the condensate pit sump, 1
21 an ything related to the pathway -- if you will -- wo uld 22 have been operations. Sure.
23 Those type of thing s definitely would have 24 been coordinated with the Department.
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1 25 Q So the pathway of effluents would fall under
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1 your re spon sibility?
2 A The pathway of e f flu e n ts fell under the 3 de partment's responsibility, naturally through me, as the department manager, yes. ,
5 Q Would you describe to us who you would have 6 interf aced with on the subject matter with whomever would 7 have put this memo together, and how?
8 A I would have interfaced with Ron Columbo.
8 Because, menos, that are sent off to the NRC, go through a 10 c hec k-o f f, if you wo uld, p ro ce ss. And, in the case of 11 Ron, since Ron was coming out of my organization to Region 12 V, it would have gone through our o rg aniz ation for 13 r e ading . And that is what I would have done: read 4t, to C 14 see if I see any errors, if I co uld pic k up any errors. -
15 Then, it would have gone to the next individual, probably 16 gone into the engineering arm, then to Rodrig ue z for 17 signature.
18 so, my interf ace wo uld have been with Ron 18 Columbo.
20 MR. MARSH: Off the record.
21 (Discussion off the record.)
22 MR. M ARSH: On the record.
23 BY MR. MARSH:
24 Q In this memo, it specifically states:
25 "The District a c k n o wle d g e s that the 1 l
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' I c alc ula te d r a diolog ic al exposure to the maximum 2 individual exceeds the limits of 10 CFR 20.405(c),
3 and 40 CFR 190 during the calendar year 1984 through
- But the liquid ef fluen t rele a se Aug ust 31, 1984.
I 5 c ondition s re sulting in these violations have now 6
been corrected."
7 what was Can you state sp e cific ally 8 corrected in that regard?
8 What we had done, I remember having a A
10 I thin k it was, a g ain , at a staff m e e tin g in town.
11 meeting, where I was given unlis.ited monies to prevent the 12 release of radioactivity from the site, e x c e ptin g for.
13 There was a r ecog nition that Tritium wo uld be Tritium.
G 14 one of the isotopes that we could not totally prevent from 15 leaving the site.
16 I remember coming back to the site and 17 pulling together a :,tairly large aroup of peo ple in the 18 meeting room, giving both George Coward and the Bechtel 18 organJ zation -- through a gentleman by the name of Dough 20 Wood -- thresponsibility to come up with modifications 21 to our systems that would keep us from not releasing any 22 asetetty.' specifically the isotope was cesium. There was 23 some short period of time, in a short week, two wee k time 24 frame s and, in that time frame, they came up with a design 25 and started implemen ting the in stshhHm= et a illtering i I ,
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unit, whereby liquid, would be in Regenerate Holdup Tanks 2 -- I believe it probably should be explained in that memo; 3 I did not see it, but it probably is.
4 There are several ac tio n s, Q near-term and 5 long-term, that were spelled out.
6 A Okay.
7 Let me ask you each of these c o r r e c tiv e Q
8 actions that are spelled out, and maybe you can comment as 8 to whether they are complete, or what their current status 1C g,,
11 A Sure.
12 The attachment to the special report, near-Q 13 term, entitled " Hear-term Corrective Action," states:
( '- 14 "The Distric t has implemented, and/or is in 15 the process of im ple m e n tin g , ne v e r al near-term 16 correc tive actions. These are: .
17 1. A d di tio n of the p u r ific a tio n 18 d o min e r aliz e r bypass valve to the locked v alv e 18 list."
20 A Very simple thing to de,i it must have been 21 done.
22 Q But you can't state for certain that you 23 know at this time--
74 A I can't state for certain, but that is a ,
25 very simplistic thing to do. I imagine it was.
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Q Number 2:
2 " Revision of the operating procedure for the 3 purific a tion d emin e r aliz er to re q uire notation of the valv e status." ,
5 A Same thing: it is a simplistic thing, and 6 probably known at the time. If I was making my best I
guess, I would say it was the same thing as the pre vio us 8 thing. But, without looking at the paperwork, I couldn't 8 .say for sure.
10 g . Locate and plug the small lea k in the 11 SOTSG.*
12 A When I lef t the Ranc h in August 1985, we 13 still had not found that leak.
C 14 Q Number 4: "Tighte n the secondary system--
15 A If it is the same leak we are talking about:
16 the little tin y le a k.
17 Number 4: .
Q 18 .Tig hten the secondary system to eliminate 18 leaks.'
20 A I remember going through, I remember the m 21 reporting on a process to t!ghten the secondary leak. To 22 say that we eliminated each and every leak, that would be 23 -- I dos.'t believe we eliminated each and every leak. ,
24 Q Number 5:
25 " Adopt a throwaway resin polic y for I ,
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I 1 polishing d o min e r ali z e r s, including solidific a tion 2 and burial at approved disposal sig ht s or 3 r adio ac tive 'r e sin s. "
4 A That was one of the designs I , was talking 5 about early that came out of the Bechtel r e co mm e nd a tion .
6 We implemented that.
7 Q Number 6:
8 "Interin use of portable d o min s with 8 temporary piping to allow a clesnup of RHUTs."
10 A That is e x a c tly, that is also the design I 11 was talking about earlier that we had implemented.
12 Q Number 7:
13 . Administrative policy to control releases G 14 such that monthly calculated dose is under technical 15 sp e cific a tio n s is under Te c hnic al Spe cific a tio n 16 3.17.2 u nits. " .
17 A That is where it refers to, that 'is ag ain 18 where I was referring to the polic y of not r ele a sin g 18 radio ac tivity other than Tritium. The policy related to 20 not doing it. Just changing the way we released liquids.
21 Number 8:
Q 22 " Processing of the PDS, or REUT, back to the 23 Rad Waste System in case activity is too high to 24 release."
25 A That was implemented.
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I Number 9:
Q "Use of the NRC Verified Computer Code to 3 calculate a running total of the dose f rom releases.
- l Data vill be processed as releases a r.e made to 5
insure r ele ase s meet limits."
6 I am assuming that.was done. But the code, A
7 Nuclear and the running of the code, was done by 8 Engineering.
8 There is a myriad of long-term corr e c tiv e Q
10 actions that are listed, also. And, rather than belcbor 11 the in te r vie w with each of those areas, wo uld it be 12 appropriate to say that, at the time that this submittal 13 was made to the NRC, that you were intending to do all of C 'd the things that are listed here?
15 A Yes, sir.
16 Are there any, is there an y, r e a so n for you Q
17 to b elie ve , at this time, that some of the conditions that 18 were described as corrective actions were never intended 18 to be done?
20 A There is nothing, to my mind, you know not 21 intending to put it down, not saying we are not going to 22 do it. The only -- and I don't even know if it is on that 23 listing ; I did not look that far back in the memo. But 24 the only thing that comes to mind that may be on there, if l 25 it said an ything about an ev apor a tion pond. We didn't I ,
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build an evaporation pond, but there was a lot of effort 2 in attempting to b uild an e va pora tio n pond. I was not 3 associated with Nuclear at the time when they were going 4 thro ugh the environmental impact statements, an'd whatever.
5 So, they truly intended to b uild an e v a p o r a tio n pond.
6 That is the only thing that comes to mind.
7 You are saying that they intended to, or did Q
8 not intend to?
8 A The y intended to, and that is the only thing 10 that I am -- that is the bigge st thing that I remember l
11 from the long-ters action. Even if it is addressed -- I _
12 am not even sure it is addressed. But they made a study, 13 on the long-terms was: what do we do, make a study; and C 14 what do we do long-ters? I remember the evaporation pond; 15 I remember b uilding a pipeline all the way from the Ranch 16 down to the San Joaquin River. That was eliminated due to 17 c o st , as part of the environmental impact report.-
18 But, to answer your question directly: is 18 there any ites in there that I as aware of that we did not 20 in tend to do? No; I am not aware of any item that we 21 didn't intend to do.
22 Q But, to your kn o wle d g e , the e v a p or a tio n 23 ponds were not constructed; is that correct?
24 A To my knowledge, the evaporation ponds were 25 not constructed; that is correc.t. In fact, I know they
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were not constructed.
2 one of the items in the long-term corrective Q
3 ac tion s is:
- Rancho "The e xistin g Seco .te c hnic al 5 specifications on failed f u el and steam generator f i
I 6 le a kag e win limit the radioactivity released to the 7 Hence, pond activities will never reach ponds.
8 le v els . who r e control win be required."
8 It so unds from this that so me of the other 10 long-term actions were depended upon the ponds themselves.
11 So, without the ponds, all the things that were 12 represented as being related to ponds may not be 13 functioning as represented.
14 That c o uld be. Wit h o ut lo o kin g at, you A
13 know, going through each item, that could very won be. I 16 bring up the ponds only from the way,you asked the 17 que stion. I know the ponds weren't, because they did not is
,get through the environmental impact report.
18 In the report 84-07, it states that the Q
20 major source of contamination of the secondary system was 21 the result of leakage from the steam generation tubes. Is 22 that the way you recan the problem?
23 A- Yes, that is correct.
24 Q Can you describe to us, in your words, what
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25 you believe was transpiring to cause these leaks, and what I
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the problems were?
2 A
If I knew how to prevent the m, I wo uld be 3
very w e al t h y rig ht now, 4 b e ca use that is the biggest problems with the tubes.
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6 There is a lot of research that has gone on in the tubes that were causing leaks. Now, what tubes I 7
8 am speaking about are the steam generator tubes which keep 8
the primary system separate from the secondary system It .
wasni t until about 1980, 10 when Rancho Seco had their first steam generator tube leak.
11 Prior to that, utilities, such as Duke, with Aconi ( pho n e tic ) P, 12 and Arkansas with Arkansas Nuclear I, were having lea ks also.
13 And we, for C 14 whatever reason, wished we knew why we went a much longer period of time 15 before we had our first leak. So we 10 didn't start handling reactivity -- activity, I should sa y
-- in the secondary system until about 17 that time frame.
~18 The cause was due to various cracking modes, stress corrosion cracking, a wear phenomenon 18 and I don't remember what they called it; 20 B & W had a name for it differently than the U-tube steam generators. It was a 21 wearing phenomenon.
22 We starting inspecting those tubes a lot; We had a lot of inspections.
23 And I thin k in the letter that you read e a rlie r ,
24 where we did helium lea k te sting ; we did air-bubble testing, and a whole bunch of 25 diff e ren t testing modes.
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' 1 the problems were?
2 If I knew how to prevent them, I wo uld be A
3 very we alt hy rig ht now, b e c a use that is the biggest d
problems with the tubes. ,-
5 There is a lot of research that has gone on
,6 in the tubes that were causing leaks. Now, what tubes I 7 am speaking about are the steam generator tubes which keep
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8 the primary system separate from the secondary system. It s
9 wasn i t u ntil ab o ut 1980, when Rancho Seco had their first 10 steam generator tube leak. Prior to that, utilitie s, such 11 as Duke, with Aconi ( ph o n e tic ) P, and Arkansas wit h 12 Arkansas Nuclear I, were having leaks also. And we, for 13 whatever reason, wished we knew why we went a much longer C 14 period of time before we had our first leak. So we 15 didn't start handling re a c tivity -- ac tivity, I should say ;
16 -- in the secondary system until about that time frame.
17 The cause was due to various cracking modes, I
'18 stress c orrosio n cracking, a wear phenomenon -- and I l
18 don't remember what they called it; B & W had a name for 20 it differently than the U-tube steam generators. It was a 21 wearing phenomenon. We starting inspecting those tubes a l
22 We had a lot of in spec tion s. And I thin k in the lott 23 letter that you read e a rlier , where we did helium leak 24 te sting ; we did air-bubble testing, and a whole bunch of 25 , different te sting modes.
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I 22 l I It is w ell documanted why the steam 2
generator tubes -- I should why; but what was causing the 3 cracking phenomenon that was causing the leaks.
Is that answering your question?, .
5 g y,,,
6 I am not trying to be evasive.
A Q No. I 8 A That is the best I--
8 Were there any other sources of radioactive Q
10 contamination in the secondary system, other than caused 11 by the tube leaks?
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12 g I can't remember any right now. That is the 13 one that stic ks in mind, b e c a use it was the major 14 If there were others, it must have been of a con trib utor .
15 s m alle r nature. I remember we had some minor 16 contamination of cooling water system.s, component cooling 17 water system, as an example, which cooled things.like the 18 spent fuel pools and we wo uld get co n taminatio n through 18 But, if we are talking about the feed those coolers.
20 water system, off the top of my head, I can't thin k of 21 anything else that can contaminate the feed water system.
22 Q Well, the way the system was designed -- and 23 correct me if I state this incorrectly -- but the primary 24 system was b a sic ally a closed lopp, and intending to 25 main tain its own in teg rity ; and the s.econdary loop was I ,
23 I 1 intended to main tain its own in teg rity separate from the 2 u n pla n n e d primary loop, w hic h, except for un usu al, 3 circumstances, the secondary system wo uld not become t 4 But, because of the c o n ta min a te d with r a dio a c tivit y.
5 leeks in the steam generation tubes, the secondary system 6 became contaminated with radioactive effluents.
7 Now, at that time, were you aware of a 8 modific ation to the plan t which allowed movement of water 8 from the tank known as the DRCST, Do min er aliz ed Reactor 10 Coolant Storage Tank--
11 A I am just trying to remember these tanks, go 12 ahead.
13 That that tan k was part of the primary Q
C 14 system, and, that the s vilification allowed movement of 15 water from that tank to the RHUT tanks.
16 A What tank, a g ain , are you talking about? I 17 am tr ying to remember. What did you call it, a'Dr. Cist 16 tankt 19 The acronym that peo ple are calling , that Q
20 seems to be familiar th*re, is the DRCST, Do min er aliz ed 21 Reactor coolant--
22 A Acronyms change, and I kind of need a piece 23 of paper that I can could put the acronym down so that I 24 can make sure I understand what you are talking about.
25 The Do min e r alized Reactor Coolant Storage
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'- ----_---__--________m _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
24
' Tank?
2 Q Right.
3 The De min e r alized Reactor Coolant Storage A
Tank, that is the big 500,000 g allon -- that is the large 5 tank on the o utside , the 250,000 g allon tan k tha t yo u ar e 6 The only modification, if you want to c all t alking about.
it a m o dific a tio n , was the abilit y to release T riti u m.
8 Because that is what was supposed to be in this tan k, is 8
Tritium. But that wasn't c o nn e c te d to the feed water 10 system.
Il No; I don't know of any design--
12 That wo uld move the water from the Q
13 Do min eralized Reactor Coolant Storage Tank to the REUT ,
14 tanks?
15 A Regenerative Holdup Tanks.
16 Right.
Q 17 A Tan ks on the outside. .
'O Right.
Q
'9 Yes; that was done -- that was done in the A
20 auch earlier '80s, the ability to get rid of Tritiated 21 Because we were getting rid water -- if not before that.
22 of Tritiated water from the site by shipment when we were l 23 allowed .to ship Tritiated water in tr uc ks. And, now you 24 are re ally pulling my memory. I can't remember the '77-l 1
25 78 time frame -- I am apparently off a year or so on that 1
(
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-- where the industry . -- we were the la st plant that was
]l 2 shipping Tritiated water. In f act, we . were shipping it to 3 G alve ston , Texas, to -- I can't remember the company --
4 for release down in Galveston, Texas.
That was removed by S the ICC.
6 And, after that, Tritia ted water b uild up 7 con tin ue d, and the method we used to get rid of Tritiated 8 water was to go from the Do miner alized Reactor . Coolant' I Storage Tank to the Regen Holdup Tanks; sample it there, 10 be within specs, and put it to the waste ponds, the two 11 ponds on the o ut flo w, then release it under a release 12 permit to the environment, yes.
13 But, 'in those, the only thing that was in k 14 that tank is Tritiated water.
1 15 Bow did you know that the only wate r in Q
16 there was Tritiated water? .
17 A sampling. The sample results -- sampling i 18 that's what I have to say.
18 so, if there were other passa emittin g Q
20 @siin that tank, what- would you have had to do with 21 it, then?
22 A Well, prior to the commitment the Distric t 23 made of not releasing anything but Tritiuse.it depended on 24 the level. If it was under -- what was the tera you used 25 for releasing it to the atmosphere?
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26 1
Q Lower levels of detection.
2 No, not even lower levels of detection; that A
3 is not it. 10 CFR 20 has got, if it is b elow a c er t ain 1e v el, you can release it to the e n vir o n men t , - A minimum 5
-- this is what happens when you get out of the industry 6 Anyway, it was for the tim e I have gotten out o f it.
7 looking 10 CFR 20. If the isotope was under the limit of 8 10 CFR 20 to release to the environment, then we released 9 Be ca use , under the lic ense, that is what we were it.
10 But the intent was, and I would like to be allowed to do.
11 able to say that the only thing we saw was Tritium. I 12 guess you can always be proven wrong, you know, one out of 13 a thousand; but, whatever. But the in tent was that we 14 always had Tritium within the RCS; and, if we had other 15 than that, we followed 10 CFR 20 release permits.
16 You got me confused when I was trying to get 17 it over to the feed water system, but there ain't no way 18
.you get over to the feed water system.
'8 Q But you were aware of that modification?
20 A Yes; everybody was aware of that.
- That was 21 all the way up to the general manager was aware of that.
22 Was there any tim e that the N u cle a r Q
23 Regulatory Commission informed of that modification?
24 A I co uld, no doubt in my mind say that, for a 25 fact, all of the resident inspectors were aware of it. I i
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1 Whether or not it was put down on a piece of paper and 2 f ormally -- I don't know. I can't say one way or the 3 other. I don't know about that one, b ut I k'now all the 4 resident inspectors were aware of it. Becauseps when that 5 was being done, when the release, or the reduction of that 6 water from this tank would occur, would have occ urred, it 7 was typic ally disc ussed about that in our o pe r a tio n al 8 meetings we had. We had daily operational meetings.
8 So, to say that we, in some way, did not 10 tell, I can't say that. I know there was no intent not to 11 t ell. I know that, hell, the pipe was right there. You 12 wo uld have to -- you stepped over it. So I know the 13 re sid ent inspectors knew about the ' transferring of,that-14 liquid.
15 Q Could you tell me who the NRC re sid e n t 16 inspectord were at that time? ,
17 A In 19847 Let me tell you, I will ' tell you 18 that the ones that I can remember, and you would have to 18 go look up -- the ones I can remember is Nervey Canter, 20 and I can't remember the other individual's name. He van 21 tr ansf e rr ed to Re gion V. He wis the assistan t r e siden t, 22 or whatever they call the second individ ual. And the 23 g en tle man that took Harvey Canter's plac e was Jim -- he 24 has since left the NRC. I don't remember the name.
25 Q All rig ht. W ell, I c an 'g e t those.
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A And, then, his assistant is the present 2 The g en tle man that is now the resident inspector r e sid e n t.
3 at the Ranch was this Jim, whatever his name's, a ssist an t.
1
- l That is the only four I can think of. ,
5 Tony D' Ang elo?
Q 6 No, no. Tony D' Ang elo was out at San Onofre A
at the time. He is used to come up f or construction. He 8 is a tall, thin f ello w. That is all I can remember, a 8 t all, thin f ello w . Anyway, I don't remember the guys 10 ,,,,,
11 But you are quite certain that the y wara Q
12 aware aware of it, and wo uld have been engaged in 13 conversations in the morning meetings? ,
i 14 Yes, sir.
A 15 Q Can you tell me what type of dialogue wo uld 16 have taken place when you were getting ready to move this 17 water?
18 A The dialogue that we were moving it; that ws 18 had to install the e quipment, or take the e quipm ent out, 20 or pick the equipment up; or we did a pressure test on the 21 equipment, and we found the leak so that we couldn't 22 transfer, or we prepared the le a k. You know, it was 23 operational oriented discussions in the morning mee tings. .
24 The idea being : the tank level got so high, we had to get 25- rid of some water. That's the kind of disc ussions. It I ,
. 29 4
1 may have been more de tailed than that. But that is the 2
kind I would have expected to hear; and that is the kind I 3 remembered: those kind of subject matters that you would 4
c want to get a job done, and these are the things that are 5 holding you up to get the job done. This is what you i.
6 wanted to do, go from here to here ; that type of thing.
7 W ell, that modification that was in st alle d, Q
8 it munds like it. was there for quite some time, then?
8 A blo g it was put in and taken out. Usu ally, 10 it was put in to drop the tank down in the 100,000 g allon 11 range prior to a refueling outage. Because', . during the 12 re f ueling outage, is when you would do a lot of draining 13 of 'other systems, running the evaporators, producing, that O 14 vater which would then, the evaporators discharged to this !
1 15 tank. So you needed the room to handle the re f ueling 16 outage. So, it would be put down and taken up. So, the
'17 length of times, I would have to go back to the data, you 18 -know, to look at the work request, to see when it was put 18 down or picked up. But it wasn't, I don't remember it 20 being put down and a year later you come back and it is I 21 -
still there. It wasn't that type of installation, if that 22 is what you mean, that I remember. Q I wo uld 23 like for you to read the last paragraph of page 2 of the 24 -
report 84-07, and tell me if that inf ormation . tha t is 25 contained in there would have come from you or someone in 1 < .
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your organization?
2 (The document was proffered to the witness.)
3 A (After a pa u se . ) I am sorry, what was the question? ,-
5 Q W o uld the in f o r m a tio n contained in that ,
6 paragraph come from you or someone in your org anization?
7 A The paragraph is -- it could have come from 8 not n ece ssarily my self or .-- I don't thin k it anybody, 8 came from me, because I don't believe I wrote this. But, 10 have come from someone in my or g aniz a tion ,or it could 11 somebody within the engineering or g aniz atio n. Because it 12 is not -- it is a known process. So, where it came from, 13 c o uld n ' t tell you. It c o uld have been Ron Columbo ; I 14 don't know.
15 Q Can you tell me why the modification to move 16 the water from the DRCST to the Regenerate Hold up Tanks 17 was not reported to NRC in the 84-07 report?
- 18 A I don't know why it was. But, if I read, l' like the paragraph I just read, the paragraph disc usse d l
20 the secondary system, that is, the feed water system. It 21 disc usse d the regenerate ways to p olis h e r s, those 22 processes. The Dominer alized Reactor Coolant Storage Tank 23 had nothing to do with that process.
24 Why? No; I don't know why. But, just that 25 paragraph alone was talking the secondary system.
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' 1 Q I recognize that the paragraph . is t alking 2
about the secondary system, and you are sa ying that the 3 DRCST tanks were unrelated to the secondary system.-
4
- A Yes. ,
5 But once you run that line from the DRCST to Q
6 the RHUT, you then connected the two systems; and you are 7
moving water into the secondary' system from the primar y 8 system.
8 A I don't argue that point at all. I am not--
10 And I as asking why was that not reported?
Q 11 A Not dir e c tly from the Reactor Coolant 12 Storage, and not directly from Reactor Coolant System, no.
13 I mean, it is a process. In order to get the water.over 14 to the Reactor Coolant Storage Tank -- and I am t alkin'g 15 about something I am not really concerned about. But tne 16 process is such that it goes for many steps to get over 17 here. "over here" being the Reactor Coolant Storage Tank.
18 The intent, I b elieve , of the meno -- and I L
18 am saying I believe, with the few minutes I have looked at 20 it - was to e xplain where, in they eyes of the author, 21 where the majority of the activity is coming from. I sure
- 22 in heck would not have had any problem with ex plaining 23 within 'the memo, or the letter, this' process, too.
24 Because the p roce ss, it was a known process. It ' was not 25 so me thing that we were trying not to tell people ; it was
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' well known.
2 So, what -- if I ma y, loo k at the memo --
3 But, sho uld why it wasn't discussed in here, I can't say.
it have been disc usse d ? That is for others'.to decide.
5 You know, if you look at the tone of the memo, I wo uld --
6 they are hang ing their hat, if you would, on the majority, 7
the continuous leak in the OTSG, cracking up the polishing 8 The polishing do min e r aliz e r s, therefore, d o min er aliz e r s.
9 That you get a back wash and that produces high activity.
10 is really where a goodly amount, a very high amount of the l' To activity, which we discharged, came from that process.
12 not talk about the Reactor Coolant Storage Tank, can't 13 But the proce ss was known,, the answer why it wasn't.
14 connection was known.
15 Did you, at any time, dissuas with anyone Q
16 that it was, that this modification should be reported to
'I the NRC7 .
18 A Not that I can recollect. s
'I 1985, there were approximately Q During 20 787,500 gallons of water released from the Dominer alized 21 Reactor Coolant Storage Tank to the environment through U the Regenerate Holdup Tan ks. Moving that much water 23 through that modification, was there any disc ussions or 24 plans to wake that a permanent fixture, or what was the-- l 25 A I am sorry. You move that amount of water
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I the Reactor Coolant Storage Tank; is that what you from 2
said?
3 Right.
Q 4
A I was trying to kee p up with you,.-
0 From the DRCST?
Q .
6 A DRCST to the--
7 To the environment through the REUTs.
Q 8 A Was there an intent to make that a permanent I system? Yes; I remember disc ussin g m a kin g that a 10 man hours involved; permanent system, only because of the 11 the putting the pipe down and pic king the pipe up, and 12 that was not one of those what I remember coming back out, 13 from a design standpoint, recognizing that the design. came (t
14 out of Nuclear Engineering. That was not one of those 15 priority jobs that was at the top of the list.
16 Q Do you know what the cause was for having to 17 move so many gallons of water. '
18 A No; You know, right off the bat, I probably 18 could come up with 200,000 gallons of water right at the 20 be ginning of a re f ueling outage.- 'Because we started a 21 refueling outage in the middle of March, a f airly long --
22 No; I shouldn't say that; it was a 90-day outage. We were 23 coming back up out of that outage when we had the pipe 24 break on top of the pressurizer incident. We went back 25 into another outage, so I could probably pick up another
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34 I
200 -- assuming that you were going rig ht back in to the 2 and that is e q uiv ale n t of another r e f u elin g outa g e, 3 another 200,000 g allon s. So that is 400,000.
4 Then, I lef t in August bef ore the y. came bac k ,
5 up. They came back up and, boom!, they went back down 6 And, you know, if I just use again in October time frame.
7 the 200,000 as a round number, I can come up with 600,000 8 on every up-down. So I am g etting pretty close to the 9 700,000. So that is not an unreasonable.
10 What controls were there to insure that the Q
11 water moved from the DRCST to the REUTs were p ro pe rly 12 sampled for radioactivity?
13 A DRCST7 Okay. The DRCST, they would be C 14 sampled, and the quantity of water that you could transfer 15 was determined by the che mistry group. And, by that, ,I 16 mean to meet the 10 CFR 20 Criteria, you may only transf er 17 100,000; fill it the rest of the way with DI Water, mix it 18 an up; and, at that time , it became a releaseable water, 18 if yo u would, ' by 10 CFR 20. I am speaking now prior to 20 the commitment we made to dump activity.
21 Then, if you had a c ti vit y , and it their ;
22 process was to be -- the y cycled the water through the 23 filte r , whic h, i b elie v e , was diato m ac eo us earth. I 24 belie v e that is what the Folsom material was. Then a 25 sa mple was taken of that material, and, if that material,
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'l 35 1
or that liquid -- I should say -- liq uid sample was taken, 2 and if the liquid sample showed no ac tivit y, except for 3 Tritium, then a release permit was made out and signed.
4
, Q Who was r e sp o n sible to m airttain those 5 records that to insure that sampling took place that you 6 described?
7 A W ell, you go up, you wo uld still have the j 8 same c hain that we went e a rlier on, whereby you wo uld 8 start with the supe r viso r -- I think we c alle d him a 10 superintendent at the time --- of Chemistry and 11 Radiological support, Rog er Miller, then Fred Kellie ; then 12 up through the plant superin tend ent s and then up to the 13 Manager of Nuclear Operations, up to the AGM. .
14 Are you aware of any written procedures or Q
15 specific ations that these folks were to be following in 16 doing this? -
17 A Yes. I an aware that, everything we did, as 18 it related to that, had written procedures and they were 18 to follow those written procedures.
20 Q But you have -- in looking at it, moving 21 water from the DRCST to the REUTs was not part of the 22 original design of the plant.
23 A What are talking about; are we talking about 24 the sampling procedures, or are we talking about moving 25 the liquid?
( t d
36 I W ell, moving the liquid is not part of the Q
2 d esig n of the plant.
3 A okay.
Q So wo uld there be written prod:e d u re s for 5 sampling water bef ore it is moved-- ,
6 A Absolutely.
Q --whenever you are not even designe d to 8 move it that way? ;
I A I can en visio n that we wo uld have written 10 procedures. You see, you had me back at the RBUT. I was 11 back at the REUT sampling the REUT. The process I just 12 went through was the RHUT sampling. Now you want to go 13 back to the Demineralized Reactor Coolant Storage Tank?
'# I want to know how you knew what you were Q
15 moving into the RHUT7 W ell, the RBUT was on a re g ular sa mplin g t
l 16 A l
17 -
period.
18 Rig ht , I und e r st a n d that. But I am Q
18 concerned about what you were moving into the RBUT from 20 the Demineralized Reactor Coolant Storage Tank?
21 A Demineralized Reactor Coolant Storage Tank 22 was on a routine sampling program. You sampled the tan k.
23 If you put water into the Domin er aliz ed Reactor Coolant 24 Storage Tank sinc e the last time you sa mple d the tank, 25 they took another sa m ple on the tank. That's what ,
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37 1 determined how much of that they put into the RHUT tank.
2 They had a procedure to do tha t, i.e., sa mple it before 3 you put the liquid into the RBUT tank. B e c a use that 4 de te r mined a transfer, the amount of liquid tio transfer.
5 T he y told the operators to transfer it with a piece of 6 paper. Not just, you know, with the back of an envelope, 7 but with a par tic ular piece of paper. You tr an sf er X-8 number of ganons from this tank to that tank.
9 When that was done, it was sampled. Then, 10 the Demineralized Recctor Coolant Storage Tank--
11 Q W hat paper was that that had to be fined 12 out?
13 A I can't tell you that exact number, of the l
14 but that is how they tr ansmitted the inf orma tion form:
15 from one group to another group. What I as trying to say 16 is, that, you would know what the activity, of the RCST is.
17 They calculated -- the c he mistr y group wo uld have 18 c alc ula ted how much to move to the RBUT tank and stin 18 have room to fin it wit h do miner alized water, with the 20 e x pectation that, when you took the sample, the release ,
21 sample, it would have been under 10 CFR 20 limits to aHow 22 you to release it to the pond.
23 Did I explain that?
24 Q W ell, I guess I as a little bit uneasy 25 about, that, you have a system that is not designed to be
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- I moved to another spot, such as, moving water from the Do min e r alize d Reactor Coolan t Storage Tank to the RHUTs.
3 plant was not d e sig n e d that way. The way Your I understand it is: the water in the Demineralized Reactor 5
c o ola n t Storage Tank was c h e mic all y, clean, but 6
radioactive.
7 A True.
8 And, that that water was intended to be put Q
9 back into the primary system, as it was needed, because it 10 It was in the Rad Waste System was che mically clean.
11 because the Rad Waste was not chemically clean.
I2 A Okay.
13 And what the concern here is, is. for ypu to Q
b 14 move water from a known r adioactive storage tank to the 15 secondary system, via a modified pathway, and not report 16 it to the NRC, seems strange to me. Secondly, I am having 17 dif fic ulty u nd e r sta n din g that, if you have a system 18 designed to sample the effluents in the r adio ac tive tan k, II that was not supposed to be moved into the se c o nd ar y ,
20 system under your d e sign, why you would not make some 21 altered sche d ule d, or plan, or re quire d te sting , before 22 any of that water was moved to the secondary system?
23 A Should I respond?
24 Q Okay.
25 A If I got your points, the plant was not
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o?iginally designed to make that transfer -- and you are t
2 ab solutely rig ht -- because the method of getting rid of l 3 Tritium, of liquid activity that' was of a nature that you c o uld transport, was by truck. That was thie o rigin al 5 d e sig n. And, when the plant was designed and licensed,
,6 that was an ac c e ptable way. It was not until the pla n t, 7
it was years after the plant went into operation when that 8 ability was removed.
8 The b uild up of Tritia ted water was a fact.
10 And other power plants released that Tritiated water to 11 the bodies of water that they were located on. We didn't L ,
12 have that body of water. So, as a re sult, the transfer 13 was being made from this to the REUT tanks.9 -
(. 14 Q W ell, how is that diff e r en t--
15 A Let me tell you.
16 Q All rig ht. .
17 A I am really trying to ex plain it, hon e stly.
18 We have told the NRC, and I b elie ve it is 18 the FSAR, that I remember discussing the que stion : W her e 20 is the release point of r a dio a c tivity from the power 21 plants what is c o n side r e d the release point? And I O remember discussing it during the licensing phase, later 23 on after the lic en sing phase; and even to the mid- and 24 }1 ate '70s, .that the release point * -- and numbers of people 25 .te r e discussing this point -- was the RBUT tan ks. The t
4 m____________.__________________.____ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _
40 I
' I RHUT tanks were the release point f or Rancho Seco. That's 2
the tanks that you took the samples of to determine if you 3 That is your last r elease point.
met 10 CFR 20.
So, if you look at that as be(ng the last 5 r elease point, what occurs prior to the RHUT tank -- yes; 6
you could argue that point; it is a philo so phic al 7 But the whole point being, as we argument, if you would.
8 have consistently told -- and again, I remember discussing 8 t he poin t -- that the release for Rancho Seco was the RHUT 10 And that is where our last and fin al sa mple , and tan ks.
11 that is our liquid release to the environment, if you 12 wo uld . We ware tak.ing samples in tanks prior to the REUT 13 tank; we were not taking from the REUT tank directly, into
'# the environment because we were going from the REUT tank 15 to the diversion ponds on the discharge of the plant. Two 16 500,000 gallon tanks, or ponds, if you would.
17 Another sample was taken to insure that wo 18 did meet the environmental release limit s before we 18 released it from that pond into the environment. So to 20 say that, you know -- I feel confident that we knew what 21 we were doing. No doubt in my mind, considering the fact 22 that we have many, many ways said: the REUT tan kte are 23 rele ase' points. That is the point that we say we can 24 defend that we are meeting environmental, or the 10 CFR 20 25 criteria.
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_ - _ - - ~ _ _ . - _ - _ . _ - _ - - - _ - . - _ _ _ _ _ _ _ _ _ - . _ _ . . _. _
41
' 1 You keep saying secondary system, and that's 2 c on f using me. In my mind, the secondary system is the 3 steam, the main steam system, the feed water system, and 4 the condensate system. That is the secondary system. We 5 were not div e rtin g , or putting Do min e r aliz e d Reactor 6 Coolant Storage Tank water inte the seconds ey system. We 7 were putting Dominer aliz ed Reactor Coolant Storage Tank 8 water into the RHUT. Maybe that is a small poin t, b ut 8 that is where I am getting confused, at time, when you use 10 the term " secondary systems."
11 So, you know, to take the sa m ple of the O
12 De min e r aliz e d Reactor Coolant Storage Tank prior to 13 ti 06sferring to the REUT tank, was an over and above move.
14 Then, the RBUT tank was our release point; and, then, not 15 to release it to the environment from the REUT, but to put
?6 I it into another ponde if yo u wo uld -- I wi.sh I re me mb e re d 17 the title of the pond, but the pond, the two.500,000 18 g allon ponds, and taka another sample in that 500,000 18 g allon pond, we were doing ever ything within out ability 20 to insure that we met 10 CFR 20.
21 Q I recogniu that you are sa yin g that the 22 proper place to measure whe:bar you are making radioactive 23 e f flue n ts, or not, is at the point of release. There is 24 no question that that is where your appropriate point is.
25 But the re pre sentation to the NRC in the September 27
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' letter- was: you were not going to- be r ele a sin g 2
. r a dio a c tivit y, because you were only, . yo u major and 3
generally only source of contamination was from the steam g eneration tube lea ks. [.
5 And I am not arguing that point.
-A 6 W ell, let me finish. Because I want.you.to Q
respond to - this.
8- okay.
A 8 You represented that as being the only, Q
10 g en e r ally, the only source of r adioactive contamination
" that would be into the secondary system, and on. into the 12 RBUTs, that would need measuring. W hen, in fact, during 13 that very time, you had an unreported modific ation in 14 place moving r a dio a c tiv e water dir e ctly from the 15 Demineralized Reactor Coolant Storage Tank into the REUTs.
16 Which, during 1985, you moved over 700,000, g allons through II And, NRC was led to believe that.the only that system.
18 source of r a dio ac tivity could be through the steam generation tube leaks, whien have claimed to be going to 20 correct thtsugh this myriad of corrective actions.
21 I as curious as to . how that could slip your 22 mind, and the need to report that as a pathway?
23 A Well, now I am going to -- I am conjecturing 24 now, strictly conjecturing now. I remember the major 25 disc ussio ns about where is this a c tivity coming from,
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Ce sium. That was our primary problem that I remember Ed 2 Bradley reporting, Cesium. How do you remove Cesium? You 3 remove Cesium through demineralizers.
- W ell, the water that is going into the 5 Dominer alized . Re actor Coolant Storage Tank went through 6 d o min e r ali z e r s. Because that water coces from 7 The water that didn't, that was r e ally evaporators.
8 co ming in , was the Cesium that was being pic ked up on the 8 secondary system d o min e r aliz ers, the condensate system 10 demineralizers.
11 You can say that it was an error in not 12 reporting it in the letter. But I accept that, because it 13 is not there. But the mindset was: Where was that cesium C '#
coming from that was causing our indication of Appendix I, 15 40 CFR 1907 16 I remember being told -- and , a g ain , I was 17 an observer in this process, because I don't know'anything 18 .about body flow, or body uptake -- Cesium, that is our big 18 Where is the Cesium coming from? The Cesium is problem.
20 cdgning from' the domineralizers in the second ar y system 21 wgen they are back washed, that is, regenerated. That is 22 where that C e siu m is c o min g from, because the 23 d omine r aliz ers are pic king them up. If we remove that
. 24 problem -- which is why we went to the throwaway resin -- j 25 if we throw away the resins, and we d'on't back wash them,
(
w-______________ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _
44 e .' 1 we remove the concern of the Ce sium. And it was that 2
isotope that everybody was keyed on.
3 should we have reported in the Now, to say:
4 same letter that: we have another so u r c e , hy the way; 5 this other so urc e is, you are right. You.can come back 6 Yes; we should have reported it in a letter, or and say: !
I it sho uld have been added to the letter. No problem with 8 doing that, or saying: Yes, we should have done it. But 8 to say that we -- I get the impression that you are trying 10 to say that we in t e n tio n ally did not want to t ell 11 so mebod y, and that I disagree with. Because, it was well 12 known. I mean, inspectors would come out, and, over .the 13 years, since 1980 -- there was nothing that we did in.1984
(. ' 14 and 1985, this was not a new process in that time period.
15 We had been doing it sinc e abo ut the 1980 time frame.
16 Inspectors and/or the r e sid e nt inspectors, and/or 17 How do you report things? You know, 'it is not whatever.
18 a plant design that came out when the plant was licensed.
18 Why didn't we report it? Well, as you know, 20 there is a process of 10 CFR 50.59. And, one of those 21 thing s, if you take a look at it, to see if 50.59 does it 22 incre ase in any way? I have lost all these words, the 23 nice words to say. When you make a work request out, you 24 have to look at that. Is a 50.50 needed for this? And, 25 obvio usly, the decision wasn't. If i 50.59 is made, and
( .
45
( 1 the 50.59 d e ter min a tion is: Yes; you have changed the 2 safety, or maybe done som e t hin g , then that is a 3 r e por t able. Because you report all 50.59 determinations, d
the yes determination. You don't necessarily' . report the 5 no determinations.
6 So, witho ut going back and loo king at all 7 the records to see do we do a 50.59, and was that 8 c o nside r e d, And, if it was, in f act, consider e d ? I wo uld 8 have to say it was. And, if the answer is no t
- hen, that 10 is not r e por table , in the sense that, you know, you just 11 don't' write a le tter and say: by the way, we also did 12 this thing over here.
13 g,11__
14 Q In the Report 84-07, action number 7 states:
15 "The District has initiated a polic y that 16 all releases will be c o n t r olle d , such as the 17 tec hnical specifications 3.17.2 limits will not be 18 exceeded. All sampling of the REUT and releases of 18 liquids will be based on this o bj e c tiv e . The 20 c h e mist r y and r a dia tio n p r o te c tion pe r so n n el 21 respon sible for ev aluating the r elease s have been 22 instr uc te d concerning these o b j e c tiv e s. This 23 action, coupled with action 9, will provide a second 24 level of control beyond the other near-term actions 25 specified h e r ein . " '
I ,
e
. 46 1
The key point here is: the c he mistr y and 2
radiation p r o te c tio n pe r so n n el have been instr uc te d 3
concerning these objectives.
Can you tell me how those folks were 5 and how they would have known of this instructed 6
commitment to the NRC7 7 You know, I can't tell you since I did not A
8 do the in str uc ting, I can't tell you how it was done. I 8 can tell you the process that they would have been 10 If someb ody wo uld have been dir e c te d ,- they directed.
11 would have gone from the document to the plant 12 s upe rin te nd en t, to somebody like -- depending on who was 13 in charge at the time -- Rog er Miller or Fred Kellie, who 14 wo uld have person ally done it, or directed one of their 15 soniors .o do it, one or the other.
16 Q With making a commitment such as this to the 17 NRC, did you feel any resp on sibility to f olic'w up to 18 insure that this was done?
I8 A We have a QA org anization whic h does that 20 very thing, and I wouldn't be surprised, and my memory 21 g,yy, ,, ,, y ,,,,t re member if I did, or did not. But it 22 womadn't surprise me that, because of what was occurring 23 at the time , that is: it was almost like a crash project, 24 that I didn't per son ally 'ask the question of the plant 25 superin tend en t, but I sure can't say I did.
l .
47-1 Q W ell, the status on this action indicates 2 that it was imple mented, which is September 27, 1984, and 3 would indicate that you all' had already done this.
4 A Then that in f or matio n wo uld have.come th o m 5 the basic letter. One thing that he was good for, from 6 Ron Columbo, was: If he had, if he wrote this, and he is 7 he wo uld have verified. If he putting his name on it, 8 says something was done, he would have gone to someone and 8 asked them: did you do this? And, for me, that used'to 10 be also my chec kpoint.
11 12 13 14 15 16 ,
17 .
18 19 20 21 22 23 .
24 25
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- 48 2 bro 3 3,- 1 1 O What were your responsibilities, and how would 2 you have carried them out to assure that Action #7 was 1 3 implemented in conjunction with the commitment? l 4 A Action #7 is the one we were jus.t talking 5 about?
6 Q Ri*ght.
7 A How would I have ensured that it was 8 implemented?
9 Q Right.
10 A Okay. I would have asked either Ron Columbo 11 how he knew it, or I would have asked the plant superinten-dent if it had been done. One of those two individuals is 12 13 where I got my information. ,
14 Q Plant super was --
15 A George Coward.
16 Q -- George Coward. Can we go off the record 17 a second? .
18 (Discussion off the record.)
19 Q Action #8 of the near-term corrective actions 20 of .Special Report 84-07 states that the processing of water 21 from the polishing demineralized sumps and the RHUTs to the 22 rad waste system, . coupled with the dilution of the liquids 23 will help ensure compliance with the limits of radioactivity 5 24 released in liquid effluents.
25 Can you explain what is meant in this case, W
E L 49 I ..- l
- l I by the term dilution of the liquids?
- 2. A Well, dilution of the liquids would.be where.
3 you would add demineralized water to whatever the amount 4 of water you had so that you would have met, we would haveL ,
5 met 10 CFR 20 limits.
6 Q The basis in Technical Specification 3.17.1 7 limiting conditions for operations . and technical specifica-3 tions 4.21.1, surveillance standards , state that the 9 specifications do not assure compliance with 10 CFR 50 ,
10 Appendix I, dose objectives. Explain why these technical 11 specifications cannot assure compliance with the dose l 12 objectives of 10 CFR 50, Appendix I?
13 A I can't explain that. That's out of'my fields d 14 Q You mentioned Ed Bradley earlier. Did Mr. 4 15 Bradley ever discuss with you during the first part of 16 1985 the possibility that Rancho Seco's technical specificac 17 tions on lower limits of detection was not sufficient to 18 assure compliance with the objectives of 10 CFR 50, Appendid 19 I?
'10 A No, I don't ever reasaber talking to Ed' 11 Seedley on Appendix I.- My inforaktion on Appendix I was 12 -- if.not always Roger Miller, if not Roger Miller, Fred 13 Kellie, although Ron Columbo knew a lot about Appendix I 24 too, but I never talked to Ed Bradley that often so I 15 wouldn't think so.
o
,, 50 i
1 0 Did you learn of any concerns about the lower l 2 limits of detection not being sufficient during the first 3 part of 1985 from some other source, other than Bradley?
4 A No, I can' t -- I don' t remenber.. I don't 5 remember talking about the lower level of detection.
6 Q- So you did not take any management actions 7 to assure that such concerns were --
8 A I don't ever remember discussing concerns on -
9 that subject.
10 Q Mr. Bradley issued a draft report on lower 11 limits of detection study on October 29th, 1985. The study la indicated that Rancho Seco's technical specification on 13 lower limits of detection was not sufficient to assure J 14 compliance with the dose objectives of'10 CFR 50, Appendix 8 15 Are you aware of that report?
16 A No, I wasn't associated with Ranche Seco at 17 that time. '
18 Q During that period you were the Program 19 Manager f or Power --
l 10 A Operations Management Program, downtown.
11 Q Okay.
12 A Well, there was a period of time, let ne 13 break it up into little pieces, there was a period of time 14 that I was with -- I was working directly for Rodriguez as 15 a manager of special programs -- special programs? special
51
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1 projects at that time, and what I was involved with before,
_2 oh,.that was about a three month period, was developing a Wh 3 -- what's the word, what's the right word, commitment- list, !
4 commitment tracking system, that's what it was, a commitmen%
5 tracking system, and I was out of that chain, you know, the 4 transmittti chain, and getting that information.
7 Q Okay. Who was -- who had replaced you as --
8 A George Coward. 'So there is a small period of 9 time there that I worked directly for Rodriguez, but doing.
10 another subject.
Il Q Okay. So if the report went out on October
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11 29th, as indicated, George Coward would be the responsible la person for responding to that? .
f T- 14 A Yes. If it went to the Manager, Nuclear 15 operations Department, yes.
16 Q Were you aware of the circumstances around 17 how a sanple' test is taken to detect radioactive,-- radio-18 activity in the effluents?
19 A only from ny background. I can go through' 10 the process with you, and the fact that -- how it's done, 11 is that what you wanted?
12 Q Yes.
23 .A okay. Well, you take a sample, it varies in 24 size, a liquid sample, it varies in size of anywheres from l
15 milliliters up to maybe a liter, depending on the activity l
r i
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i .
5 51 Y
- I that you're looking at, then you evaporate it onto some 2 kind of a paper, usually a filter paper, and then you 3 count this filter paper, and in counting the filter paper, j I
4 you' re looking f or the emissions , that's what ,I was f amiliag e
5 with when I did it a long time ago, today they got, you 6 know, computers to peaks. I wasn't familiar with peaks.
7 But it tells you the peaks, and it tells you 8 the isotope associated with the peak energy, or the energiec 9 and from that, they can determine the quantity.
10 0 Can you tell ne what the different variable I! factors would be in conducting a sample, taking a test 12 such as the volume of water --
13 A Oh, yeah.
s 14 Q -- the time that you expose to --
15 A Right.
16 0 -- sensitivity of the instrument?
17 A Sure. You know, that's all your vpriables.
18 As you said, the higher the activity, the smaller the I
sample you have to take, because you've got much more 19 20 activity. The lower the activity in the sample that you're al expecting, the bigger the quantity, so when you evaporate 12 it you can condense it down. That's one of the items withia 23 the formula.'
24 Then exposure, I'm not -- that's not a 25 variable, because'it's already there. The equipment that 1
G
52 4
1 you use, you know, from what I was familiar with'back in e
2 the Navy, it was much more rudimentary type equipment than 3 we've got today,-in fact, I couldn' t -- I don't think I 4 could operate today's equipment. -
5 Drying , the type equipment you got is 6 absolutely -- you know, before it was a gross number, and 7 I was familiar with just coming up with a gross number, and a now you're looking at each individual isotope, so you go 9 right to 10 CFR 20 on an isotope basis. .
10 counting time is a variable, the shorter, the 33 more activity, you don't have to count it as long, or the
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12 greater the activity -- or the lesser the activity, you g3 count longer, that's a variable. ,
r; L. 34 What would be another variable, in f act, three 15 of them, you mentioned the same three.
gg Q Well, is it save to say that if you're looking 17 for low levels of activity, you would count for a longer 13 time than what you would. if you --
19 A At high levels, yes, that's correct. ,
20 Q Would it seem to you to be appropriate to 21 reduce the time, the counting time for measuring the samples ,,
22 testing releases from the RHUT?
s 23 .A -
Seem appropriate?
24 Q From the process that has been carried out 25 for several years, counting for 2,000 seconds, once a
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53 I
1 commitment is made to the NRC that you're going to be more ,
1
- 2. vigilant as you had committed, would it make sense to be s cutting the time to 1,000 seconds when you're looking for 4 lower levels of detection? .
A It doesn't seem like it, and again, you' re 5
g getting me out of my field, but if you say that you can be 7 more vigilant, and I used to take a liter sample before, an@
g now I take a half liter sample, or if I -- you said 2,000 y and --
10 Q Well, we're not tsiking about any variance, gy other than reducing the time, other factors --
12 A Well, all other factors remaining the same, 33 even for any one of those variables, if any variable would r change which would reduce the sensitivity, I would say, V g4 15 yeah, you're not doing what you should have been doing.
Q That would be less vigilant?
is A That would sound like it was less vigilant, 17 la that's right.
19 Q Are you aware of any time, any instructions 20 to Mr. Kellie, or to Mr. Miller to.redece'the' count time l 022 the analysis of the RHUT samples?
l 21 22 A Instructions from who?
23 -Q Anyone?
24 A No. No, because I wouldn't even have -- that' 15 not a subject that I would even be attenoting to get 9
4
54 1 involved with. They would be my_ experts, in fact, it's just 2 been within the last 60 days that I became aware we used 3 to count.to 2,000 seconds.
4 Q I'm going to show you a memo tha,t was prepared 5 by Fred Kellie concerning a telephone call between he and 6 Greg Yuhas. I'd like for you to look at it and see if you 7- recognize this.
8 (Pause to examine document.)
9 A I des't remember it, but I see my name on it, 10 so it must hav4 come across my desk, yes.
11 Q Can you decipher from what exists there what 12 ,
that pertains to?
13 A Well, if I just go from the reason for the 14 call, resolve the meaning or interpretation of a sentence 15 within some table, notation C, table -- page 4 of -- and 16 I don't even know what table they're talking about, and he 17 and -- he, Kellie, and Yuhas came to resolution for the 18 nuclide is below the minimum required, because a positive 19 value must be recorded and reported, fine.
20 You know, I see two individuals, and I used 21 to see a lot of these, either Kellie, or Roger Miller type 12 phone calls t'o people like Greg, or whoever, used to get 18 these kinds,'so I would have road it, they had a question, 24 they. got an answer, cJreat, and go on from there.
15 0 okay. Well, can you -- the date on that memo l
55 1 is? .
2 A June 6th of '85.
5 QL All right. Now, that was still during the -l l
4 period of time that you were Manager of Nuclear Operations?
5 A correct.
6 Q If the intent is to report any peaks that are 7 detected as we have learned that this agreement or defini-8 tion as interpreted by Yuhas and Kellie intended, would it 9 reduce the number of peaks that would be detected by 10 reducing the count time of the sagle, and keeping all othes 11 factors consistent?
~
12 A Based on what I said earlier, and based on -
13 my knowledge of, you know, of the older type equipment, I 14 would say, you reduce the count time, you're reducing, as 15 I said earlier, the sensitivity. Don't know that for the 16 equipment that these guys use, so I don't know its effect 17 on the new equipment that they have. So if I use the same 18 analogy, assuming the same response, you reduce the count 19 time, you reduce the sensitivity, yes.
20 o well, wouldn't it be true, then, that if you 21 continued to count at a 2,000 second period and had peaks, 22 those would be reportable. If you reduced to 1,000 secondst 23 counting time, those peaks may not show up, and therefore Is that an accurate analysis of -c 24 would not be reportable.
25 A From what you're saying, it ' sounds reasonable <
a 66-
-) .-- )
x 1 But' I could not say I know that for a ' fact, because I don't-2 know enough about the processes or the equipment that they 3 use, and from what you're saying, yes, that sounds like l
4 that's true.-
5 Q Well, after a commitment has been made to the 1 6 NRC that Rancho Seco will not release radioactive --
7 radioactivity to the environment --
^
8 A Right.
9 . Q -- to reduce the count time, which reduces to the detectability of a radioactivity peak, wouldn't that II. be somewhat disguising the -- or manipulating the system?
12 A Well, I den't want to respond on that one, la because I just found out within the last 60 days that in -
14 fact reducing the count time, does affect the sensitivity, 15 and that's when I read a recent, or a report, an NBC report j
16 that I didn't even know we had until it was -- we had a 17 large gathering if you would, and they were talking about 18 this repset, and I didn't know what they were talking abouto
~
19 so I askes for a copy.
20 The report discusses sensitsukty versus count 21 rate, and so I've. learned that -- you know, I've learned 22 something in the last 30 days. If you had asked me this 23 question in November, I would have given you an answer to 24 the point of saying, it sounds like that's what it would doo 25 Reading the report, I' d say, yeah, that's what it does, you
-L 1
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l
{
1 know, because I've got information tnat I didn't have. If ,
l 2 you asked me if I would have known that on 6-6-85, without 3 asking somebody a question, I wouldn't have known that.
4 4 Q Are you aware that Mr. Kellie di,d,_in fact, 5 reduce the count time?
6 A I found that out within the last 60 days, yeso 7 Q You stated earlier that once you moved to 3 your new position in August of ' 85, that you no longer 9 received communication and information concerning Mr.
10 Bradley's findings about the lower limits of detection?
11 A well, my -- you know, when I got out of the 12 Manager- of Nuclear Operations Department -- as Manager of 13 Nuclear Operations Department, you were on together with a 14 Rodriguez and cthe Manager of Engineering, you got almost 15 everything as you can see from that piece of paper that 16 you just showed me.
17 Af ter I left the Manager of Nuclear Operationc la it took awhile, and the names -- the name came off of forms c 19 and you didn't get all the information, so about that time,
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20 it was hit and miss, and even if I received it, I don't 21 necessarily think I would have necessarily read it, because 22 I wasn' t doing that kind of work, and I may very well have 23 just discarded it.
24 O Did you take any actions based on the Yuhas/
25 Kellie telephone conversation?
58 3
4 I A I wouldn' t have -- even today, - reading the
- 2. memo, or the thing that you just showed'me, you know, I 3- read it today, and I don't see where I would have even --
4 wouldn't even came to mind to take any action; because I 5 don't see any action it talks about.
6 Q Specification 4.21.1, Footnote C, Technical 7 Specs,, gives instructions.to report all identifiable peaks 3 of radioactivity made during .the sample analysis. The --
9 would you say that the memo created by Mr. Kellie confirms to that technical spec, and that the interpretation is that It it is reportable if it's detected?
l
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12 A That's what it says, if the nuclide is below is minimum required LLD, but is a positive value, must be ,
. 14 recorded and reported.
15 Q In October 1985, Lawrence Livermore National 16 Laboratory reported detecting Cesium 137 in downstream L 17 sediment at levels not anticipated due to the fagt that no 18 releases of radioactivity products had been reported. Of 19 course, that report came out after you had moved to your-20 new position, but that analysis may have been going on 21 during your tenure. Here you aware of that effort? I 1
22 A The Lawrence Livermore effort? Oln, yeah. Ani 13 I was also aware of the NRC effort.
24 Q Can you tell us what predicated that examina-15 tion by Lawrence Livermore?
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59
., j
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.j g A Well, the predication was when Roger Miller 1 i
2 brought it to my attention, let me back off a little bit. I l
! 3 Roger and Ed, at the time, produced the annual report, one j 4_ of which being the liquid and gaseous releases from the ,
5 plant. When Roger brought it to my attention that Ed was I g reporting for the 1983 annual report in the May/ April time 7 frame, something like that, that Ed was saying we were g exceeding Appendix I, and that's when I took it up to 9 Rodriguez, and I said, hey, something's wrong, and it was 10 about mid-year, mid of ' 84, the problem if you would, was 33 given to the Nuclear Engineering Department, specifically Roger Powers who Ed reported to within the Engineering
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12 33 Department.
14 And then I knew they contracted with Lawrence 15 Livermore Labs to do their thing, if you would,utheir tg environmental detailed analysis. I got involved plant-wiseg 17 because both the NRC and Lawrence Livermore upset my 13 neighbors in the area, because they wished to tromp on the 39 land and do all these things.
20 My job was going behind the Lawrence people 21 and the NRC people and calming the neighbors down from that 22 standpoint. So I knew Lawrence was out there, and I knew 23 the NRC was out there taking samples and trying to analyze 24 what was going on.
25 0 okay. Well, if you were aware that that was 6 O
- 4 60
- I l
1 possible problem of -- and needing examination. Did you 2 take any action to assure that the lower limits of detectiek 5 were being properly, or accurately assessed?
4 A No, I don't ever remember talking to Lawrence' 5 Livermore -- I don't ever remember talking about the lower ,
6 limits of detection question. I do remember talking about 7 are we meeting our discharge, are we meeting 10 CFR 20, my I 8 interest was 10 CFR 20. I would not have gone and talked 9 about lower limits of detection.
10 0 So no matter how hard you try to get someone 11 to --
12 A You want me to finish that answer? I 13 Q Yes. .
14 A I can't see me going to verify the processes 15 other than I do see myself going to verify with the Plant 16 superintendent, Roger Miller or Fred Kellie, individuals 17 that we were follosing the processes that -- are we meeting 18 10 CFR 20 as opposed to saying have you changed the LLD 19 or any of those kind of things. That's not something I '
20 would have asked.
21 Q Well, when something suddenly shows up to be 12 a little peculiar or strange, or different than what you 23 expected --
24 A Well, you've got to recognize, we had an 25 answer at the time. When it came up, that's when they did 0
68 I a -- in that time frame , that early ' 04 tire frame , the 2 answer on the why we were in that' situation was because of l i
3 what Ed found related to a dilution factor of 25, and that i 4 if we were on a coastal, or a lake, or whatever, that when 5 you discharged the liquid into that body of water, that the 6 sof tware program shows a 25 to 1 dilution factor.
7 Whereas, we didn't have that 25 to 1 dilution a factor going down the creek, because all the water that 9 went down the creek, we put down the creek, we didn't have 10 this big body of water. So it was being explained on what 11 our problems were', was the fact we didn't have that big 11 body of water.
13 If you go back and look at after I found out 14 about this 25 to 1 and they took it out of the formula 15 that they were using to talk about the uptake in MAN, we 16 increased the dilution flow, this is a - ,we yet our water 17 from what's called the Folsous South Canal, and we can contr@
18 the amount of water going down that creek. i, s.
19 Well, we increased before, we just -- we only 10 put water down the creek when we had to, because we were 11 paying for the water, When that occurred, when'this 25 to 1 22 question came up,.we went up to something like 10,000 13 gallons a minute just all the time, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day to, if 14 you would, 25 to 1 question, if'you would. You say why?
15 No, '12e concern, getting back to this 25 to 1 S
7 62 k3 e
I factor, as it was explained to ne, not just to ne, but the i
2 group as a group, that without that 25 to 1, there was some J 3 questions on -- to be able to run on a dry land site.
(
4 So I didn' t -- when you say, did you go and ,
5 check all these things, well, I -- we were giving an answer, ,
6 you know, the answer was related around this 25 to 1 diluti@
7 number that Ed had pulled out of the formula that got us 3 into the trouble.
p Q When did you 'first learn that the count time 10 was being reduced on taking test samples?
13 A In January of 1987.
12 Q Are you familiar with a company named Controlc 13 for Environmental Protection? .
- 14 ,
A Is there an acronym like C something?
15 Q It would be CEP.
16 A CEP, yes. Yes, CEP was a contractor to us, 17 gee, I don't remember exactly how, but I suspect,since the 13 title may have been our environmental where we sent the to grass and the mud and the liquid samples from the environ-20 mental monitoring program.
21 Q Are you f amiliar what a composite sample 22 from the RHUT would be?
23 .A No, but I know what a composite sample is.
24 A composite sample is when you take a little bit over a if 25 period of time, put it all together and you analyze it,
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63
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. 1 you would. But what specifically it is for the RHUT, I I
2 don't know. l l
3- Q Are you aware of any deliverables, or requireo 4 ments by CEP to document their analysis? ,
5 A You're going to be bringing back me'mories on 6 me now. I knew we had a criteria within the contract to 7 the point of saying we had a baseline, and we did an 3 environmental study prior to the plant going on-line, and y there was some percentage above that level at which they to had to pick up the phone and call us immediately. I don't 11 remember what that percentage was, and it was on various --
12 all the different items that we sent them to get analyzed.
13 Q Who would be responsible for -- during your 14 tenure at the plant to maintain the deliverables, or 15 documentation from CEP?
16 A That would have been Roger Miller, he was 17 the contract administrator for the CEP contract.,
18 O Did r.nyone ever report to you that they were 19 missing documents from the CEP records?
! 20 A Not that I can rememberi no.
I 21 Q If you were taking your sample from the RHUT, 22 being the point of release, what would be the effect of 13 diluting the RHUTs with service water on any detectability?
24 A Well, the effect of diluting the RHUT is l 25 effectively reducing the activity per your measuring level, 1
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'C 1 for instance, you know, so many microcuries per milliliter, 2 so if you had -- if you diluted the RHUT by 50 percent, 3 or doubled the volume, if you would, then you'd have the 4 milliliters per -- half the microcuries per milliliter you 5 had when you started.
6 0 okay. So something that was below the lower 7 levels of detection prior to diluting -- or were above the 8 lower levels of detection prior to diluting could, in fact, 9 be below the levels of detection after dilution?
10 A That could be, that's true. That's true.
Il Q Who authorized the dilution of the RNUTs?
12 A Who authorized it?
IS Q Right. .
a) 14 A Dilution came out of the chemistry group
- 15. based on a sample, 16 Q Why would it be based on a sample?
17 A Well, that's what you -- if you're looking 18 at being less than 10 CFR 20, then you're looking at a 19 sanple of what do you have in the RHUT, and again, I'm l
l 20 speaking prior to the criteria of releasing no activity.
21 But if you're looking at 10 CFR 20, and you want to be less 12 than 10 CFR 20, then you're going to dilute whatever liquid 13 you have to be less than 10 CFR 20.
24 Q So recognizing that you had radioactive 15 contamination exceeding 10 CFR 20 requirements, you could i
- L _
1 .
65 l 1 add water to the tested volume and then ultimately become--
2 A Under 10 CFR 20.
3 0 Under the 10 CFR 20 criteria.
4 A Recognizing, you know, when you , asked me 5 the question he're , I'm talking about prior to our commitment 6 of not releasing any activity, except tritium. This is 7 how we operated the RHUT is to be under 10 CFR 20 from our a release point, and our release point being the RHUT. One 9 thing we didn't have that a coastal plant has.
10 Q Ok'ay. So you're sayine that after September 11 1984 --
12 A Thereabouts, yes.
13 0 -- that would be an inappropriate process?
b 14 A That would have been an inappropriate process 15 in the sense that our commitment was we would cycle the 16 water through the deminerali-zer and remove those isotopes, 17 other than tritium, namely the Cesium isotope was the one 18 that was being quoted all the time as having a problem.
19 Q Okay. But if they -- if the nuclide was 20 detected --
21 A Okay.
22 0 -- after September ' 84 --
13 .
A Okay.
24 0 -- you're saying that it should havre been 25 routed through the demineralizers to scrub it clean.
G e
0 66 l
- A ., That's correct.
2 Q , To scrub the nuclides out of it.
l 3 A That was exactly why the demineralizers were l 4 put there. -
I Okay.
5 Q So following that, it would then be 1
6 inappropriate to dilute the RHUT rather than run it through 7 the domineralizer?
g A Well, yeah -- can I paraphrase your questions, g or --
10 Q Sure.
33 ,
A I think getcing back to your question before, 12 it would be inappropriate if I was to pick up a cesium 15 isotopa peak in the RHUT and then dilute the RHUT so that
(! 34 I go below the lower level of detection of the Cesium 15 isotope peak., and therefore, I said, oh, I don't have any 16 Cesium, so therefore, I can dilute it.
I 17 The answer to that is yes, if we sgw Cesium, gg any isotope, and I keep picking Cesium, but that's true, g, with the exceptiorr of tritium, because you can't remove -
20 triti e but the direction was, and the whole process was 21 Put in so that we -- this is where the isotope, in the 22 smallidemineralizer is where the isotope belongs, and not g3 in the. environment.
24 0 Okay. Would rautacing the count ti % in 25 effect, do the same thing to your test results an diluting?
I l $ ,
i l
l
!1 r
s.
1 A Trom my understanding now, yes , that's true.
2 Reducing the count times would reduce the sensitivity, and 3 therefore, whereas, if you increase the count time, you 4 may see something, if you reduce the count tine, it may be l
.5 in the grass, and therefore the sensitivity is not there.
6 That's correct. l 7 0 Who was responsible for collecting and 8 maintaining the records for the movement of water from the 9 DRCST to the RHUTs?
10 A ,
Who within the organization, I don't know by 11 name. It would come under the ' Chemistry and Radiation 12 Division. Someone in that would have been responsible for 13 pulling together the records.
14 Q Can you explain why all but a few of the.
15 records on transfers of water in that circumstance were 1
16 not kept or maintained? ,
17 A I didn't know they were until you asked the 18 question.
19 Q Are you aware of any discussions that were' 20 pertaining to whether or not the modification of moving 21 water from the DRCST to the RHUT should be made known to 22 the NRC, or not be made known?
13 .A I'm aorry, could you repeat that because I --
24 Q Are you aware of any discussions in which the 25 modification allowing the transfer of water from the DRCST l
i .
l
- 1. 1 68
\ --
I to the RHUTs was decidsd to be made known, or not to be 2 made known to NRC?
3 A No.
1 l 4 0 You mentioned earlier that up to.a certain 5 point in time, that you were shipping radioactive water off 6 of the site for disposal elsewhere. Can you describe what ,
7 was transpiring in that regard, whether you were shipping is a A We were shipping it -- initially -- well, 9 when we first started, I'm pretty sure they were going to 10 -- the liquid was going to Washington, to the burial site 11 in Washington. I believe Washington State, and again, this
~
12 is all past memory, Washington State passed a law, whatever o 13 refused to take liquid, radioactive water into the s, tate.
b 14 So then we went to Todd Shipyards, that's 15 what it was, Todd Shipyards in Galveston, Texas, and we 16 shipped the liquid to Todd Shipyards, and they were also ,
17 getting rid of liquid, they were doing work with,the Navy, 18 and they were barging it out into the gulf, and releasing 19 it into the gulf, and then there was a -- we did that with 20 Todd for about three years, four years, and then I think 21 it was the ICC, I think the ICC -- it may have been NRC, 22 I don't believe it was NRC, I think it was under ICC 23 control, they stopped that s,hipment, you cc,uldn't ship 24 . liquid, you could only ship solid af ter that, then they 25 went into the big question of do you dewater' the resin, you
(
69 l
1 31 L -
l I know, what is dewatered resin, and I guess that argument is )
2 still going.
3 Q Was that an expensive process? l 4 A Yes. ,
5 Q Shipping the water? Was it also expensive l
6 to shio the resins after the --
7 A Oh, yes.
3 Q -- water had been taken out of it?
9 A Well, we never did do that, we never shipped 10 dewatered resins, that's why if you remember, we talked 33' about -- somewhere's in there you mentioned it this 12 afternoon of solidification, because there was always --
13 there was a philosophical or a technical-- can you truly ,
r bJ 14 dewater resins and some people were dewatering resins, and 15 shipping it that way, other people were saying, no, you 16 can' t dewa+?.r resins, you've got to solidfy them. We went 17 the solification route. .
13 So I don't remember ever shipping dewatered to resins, we always solidified them.
20 /
21 /
22 /
23 */
24 /
25 / (Nothing omitted. )
4 m
__ __ _ _ . _ _ ___..s_ _ _ _ . . _ . _ _ _ _ . . . . . . . . _ . _ . .
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'? 1 Q- Prior to - let me see if I have a question.
2 When you suddenly quit shipping for disposal, how did you.
3 address that new problem? What was the plan?
4 A The now problem was a increase -- an increase.;
, i 5 The new problem w:s the fact that we were essentially 6 holding on site the liquid that came into the black box, 7 which we called the box, the radioactive liquid processes. l l
4 You' d run evaporators. We had two liquid 4 9 evaporators then in the basement and the discharge, the 10 condansate discharge, the steam which was condensed, was !
l 11 put into the regen holdup tank --the demineralized reactor )
1 12 cooling storage tank, excuse me. The. bottom of that j l
la evaporator was solidified and shipped off-site as solidifieg 14 waste.
I 15 So we had a condition where the reactor 16 cooling storage tank -- the demineralized reactor coolant 17 storage was increasing and we didn' t have any way to get 18 rid of the water and it wasn't necessarily leakage, just 19 hazard leakage into the black box, it was operational. We ,
10 went through a large program of reducing any water going 11 into the black box, it was that type of program. We got to 12 the point where we were in a senso drowning, if you would, !
13 in water and'we tried to hold to the original commitment of 14 not -- even though the license, we had made a -- kind of 15 position within the environment to that -- hopefully that
~
i e
e a
c I we wouldn't release any activity in the environment. But 1
- 2. the license allowed us to, which is when we started releasinj 1
3 tlue tritiated water out of the demineralized reactor coolantI 4 storage tank, estimated about probably a year af ter we lost l 5- the ability to ship by truck.
6 Q About 19817 7 A No. That's -- if you -- don' t ' tie -- it a could have been, but don' t tie the leak of the steam 9 generators to that. That's two separate items.
10 I thought it was in the late seventies. It 11 would seem to me we lost the ability to ship radioactive
~
12 water in the ' 77, '78 time frame. That's one condition.
13 That's one situation which is separate from the casium and ,
' .all the other activity problems we got into the steam
- 14 15 system and condensate system as a result of the steam is generator tube leaks. The steam generator tube leaks , the 17 first one occurred in 1980. ,
is Q All right, so you had a new problem to deal 19 with because you couldn't ship with the trucking anymore?
20 A That's correct.
21 Q And you dealt with that by the releasing of 12 water that was building up from the DRCST tanks into the 13 RHUTs and during that time frame you were under no commit-24 ment to NRC now or through your. plant operating procedures 15 not to make certain releases as long as they didn't exceed 1 ,
i 1
l 4
72 3
A 1 parameters. But you had made some commitments to the public 2 I guess that you wera not. But you had to mcVe the water, 3 so you reoved it with this modification?
4 A That is correct. ,
, . j 5 Q Then in 1984 you made a commitment to NRC 6 that you were not going to make any releases of radio-7 activity and what did you de then to plan to deal with the 3 new problem now that you would have of you were always 9 able to get rid of your water as you were before, moving it 10 through the DRCST, now you have a new problem that you can't 11 make release and what was your plan to make that commitment?
~
A Well, the plan there is whether the water came f 12 13 from the demineralized reactor coolant storage tank, or it ,
14 came from the secondary system was isunatorial. It went to 15 the -- or RHUT. Sampling of the RHUT, you get a peak, 16 process the water through the demineralized before you '
\
17 release the water to the retention basins. ,
18 So it was immaterial where the water came 19 from. The ef fect of recycling the water through the 20 demineralizar would remain the same.
21 Q In 19 81, did you have a -- I can' t recall if 22 you' were told the specific date, but in 19 81 it became 23 necessary to move water from the DRCST into the RHUTs ,
24 because you were waterlogged as you said. Do you recall a 15 meeting in which you told Fred Kellie and another individual !
i
_ _ _ . _ . _ _ _ _ _ _ _ - _ . _ _ . _ . _ _ _ _ _ _ . _ _ _ _ . _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ . _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ U
l 73 4 ,
O' I o
1 named I believe Ron Alexander that they should construct
- 2. the modification?
I don't remember this Ron Alexander name.
3 A l 4 I'd have to see him to recognize him but I know Kellie of l 5 course. No, I -- it wouldn' t surprise me that in 19 81 we l
l 6 constructed this modification and it also wouldn't surprise 1
- 7 me that they weren' t the only ones there or that I didn't 8 make this decision on my own to --
9 Q Ron Lawrence, I'm sorry.
10 A mere you go, Ron Lawrence that I know. It It wouldn' t surprise me but then again it wouldn' t surprise 12 that there wasn't other people there too. Nor that other 13 people were involved with what do we do with this water?
f.
L I4 How do we get rid of the water? We had a process to go 15 thron;;h from the standpoint of writing a work request out.
16 Reviewing that work request to determine is it unsafe?
17 Does it meet -- does it violate any rules, laws, .or whateveg 18 and then build the process -- build 'the modification. So, 19 it wouldn't surprise me that I said that.
20 Q Can you describe what a release permit is?
11 A A release permit is one where the chemir try 22 group would have taken a sample out of the regen holdup 18 tank, the RHUT tank as you've been calling them. They
- 24. analyze the sample. m ey put the information that they goto 15 the peak -- this peak that you were talking about, what e
f 74 5' -
o 1 isotopes,- and also on there whether or not thc;r meet 2 10 CFR 20. That is an acronym and I can' t remember the 3- ecronym that used there. Now --
4 0 Well does it mean a release from,--
e 5 A Permit and it's got to be signed. At the 6 time it had to be signed by the senior chem rad or above, 7 the senior or above. His supervisor, which would _ have been 3 Kellie, or' his supervisor, which would have been Roger g Miller, and then he turned that over to the shif t supervisor 30 which then gives him authorization to release the water, gg with the idea it meets 10 CFR 20. That's what a release permit does. f 12 33 Q Are you aware of any responsibilities to
{
b- 34 report the releases under the ALARA Provisions?
15 A Well, yeah. We're required to report releases 16 through the annual report. That's the only -- of course, 17 if you know, if you exceed some limit, then you've got 18 reporting requirements for that. But I'm not speaking about 19 that, but just normal releases which is really what 20 Appendix I was all about. Appendix I is ALARA and the iden 21 being that you're going to following the releases, report 22 those releases though the annual report and then the three 13 millirem whole body is, diat ALARA guideline, if you would, 24 to do engineering or modifications, or whatever you can do 15 to keep it below three, which was the driving force to go e e
6 75 j
1 to the solidification of the resin and all the other things
- 2. we did, as pointed up, that letter.
3 Q If the test that was done at the RHUTs re-1 4 vealed the peak and that test was then reaccomplished using '
5 a lesser time period to count, and that peak was no longer I 6 detectable, and that water was then released, would that in 7 your opinion be a release of radioactivity?
3 A In the guidelines we were operating under, 9 as to that number, the answer would be yes.
10 Q Did you during the period of time of.1985 11 report to NRC, '84 and '85, that you did not make any
~
12 releases to the environment through the effluents?
13 A I'm assuming we did. I'm sure you could 14 probably show me some piece of paper. I don't remember 15 anything that I could hang my hat on. I -- it wouldn't 16 bother me that we did make that report. Because to my 17 knowledge we were not making any releases and I had -- in 18 fact, I publicly said that.
19 Q okay . . So to your 1.nowledge you weren' t making 20 any releases and therefore you reported that way to NRC?
21 A Sure.
l 22 Q But 12!in fact seneraalwas manipulating the l 23 tW$e Weg I described to you, in your opinion you
- 24 would have besa in f act makAng releases"4 15 A F way you jus [ described it', we would have 9
i 76 7
m 1 been. making releases, that's right, and that would have 1
- 2. gotten me very mad. ,
3 Q Do you have any other information that you 4 would like to provide to this investigation at,this point?
5 A No, I don't. I am -- well I' m not really sure 6 of the, you know, the point that you're going after for 7 the investigation. I think you made about -- I remember 3 three or four different areas that you went in. One being 9 why didn' t you talk about the transfer of the demineralized 10 reactor coolant storage tank to the RHUT? And there, if it that's a investigation point, I don' t have anything else to 12 add to that. Other than as I've said, it was knowledge.
13 From the standpoint of this other area,you V 14 covered, it was the 2,000 versus 1,000 counts and having 15 just been made aware of that, with in mind like I say, a 16 19 87 time frame, no I don' t have anything to add to that 17 either. .
Is I think that's the two areas of the 19 investigation, from I heard you this afternoon. If there 20 was another area, I didn't pick it up.
21 Q Okay. Then you don't have anything further 22 to add?
23 A Not to those, no.
24 MR. MARSH: Okay. We can show the interview 25 as concluding at 3 : 19 p . m. , March 24 th , 19 87.
o0o e
' e.
_ _ - _ _ _ _ _ _ _ . . ._. . _ . _ _ _ _ _ . . __.________.___________...______.______________m. .____ _ _
g '
s.
N Shic is to certify that tho.cttachcd precaddings b3foro tho ITED STATES NUCLEAR REGUIATORY COMMISSION in the. matter of:
(E OF PROCEEDING: INVESTIGATIVE INTERVIEW (CLOSED) l 1
l I
~ '
DOCEET NO.: NONE i
- PIACE:
Rancho Cordova, California . .
'DATE: 2f March 1987 - .
were held as herein appears, and that this is the original transcript thereof for the file of the United States Nuclear Regulatory Commission." ,
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f . Reporter's Affiliation Jim Higgins and Associates
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. EXHIBIT '33
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( 10 MUDSACRAMENTO MUNICIPAL UTILITY DISTRICT 6201 S Street. Po Bos 15820. Sacramente CA 95852 1830.1916) 452 3211 i
RJR 86-087 AN ELECTAIC SYSTEM SERVING THE HEART OF CALIFORNIA March 3, 1986 j d
1 J B MARTIN ADMINISTRATOR U S NUCLEAR REGULATORY COMMISSION REGION V 0FFICE OF INSPECTION AND ENFORCEMENT 1450 MARIA LANE SUITE 210 I WALNUT ' CREEK CA 94596 I i
l DOCKET NO. 50-312 .
]
OPERATING LICENSE DPR-54 ]
SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT j The Sacramento Municipal: Utility District hereby transmits the Semiannual )
Radioactive Effluent Release Report for the period of July 1, 1985 through 1 December 31, 1985 in compliance with Technical Specification 6.9.2.3.
The format of this report follows that found in Appendix B of Regulatory Gutde 1.21. In addition. changes to the Offsite Dose Calculation Manual and to the Process Control Program are included as required by Technical Specification )
6.9.2.3. Corrections, if any to previous reports are also included. j Operational Events Rancho Seco was not operating during the period of July 1,1985 to September 28, 1985 for plant modification's.
On September 28, 1985, the reactor achieved criticality and was brought to 40%
power on September 30, 1985 for physics testing, q The power level was reduced to 15% on October 1,1985 to perform the turbine overspeed trip' test. The reactor trippad on high coolant pressure following the turbine trip. The plant was brought to hot shutdown, where it remained j until Novembtr 2, 1985.
Criticality was achieved on November 2,1985. The 220 KV breakers were closed i on November 3, 1985 and power was leveled at 42%. Power was reduced to 15% i and the turbine overspeed trip test performed on November 7, 1985. The 1
. breakers were reclosed on November 8, 1985 and the reactor power was brought to 42%. The power level was increased to 75% on November 17, 1985 and to 91%
L on November 19, 1985. Power level remained at approximately 91% to December 5, 1985. .
b*
.ygrg )
d 8 1 86-010 - EXHIBIT N -
! 5 Page k -ef_ Pages ,- i L _ --_--___ _ - _ _
, i RJR 86-087 .
l J B Martin March 3, 1986 I
i 1
The reactor tripped on reactor coolant system high pressure on December 5, 1985. The reactor was brought to hot shutdown for control system maintenance until December 12, 1985. The reactor was brought back on line on December 12, 1985. Power was gradually increased to near 100% on December T5, 1985 and remained at 100% until December 22, 1985. On December 22, 1985, the plant was brought to hot shutdown to repair a leaking pressurizer sampling isolation valve. The reactor was brought back on line on December 24, 1985.
The plant tripped on December 26, 1986 due to loss of power to the Integrated Control System. The plant was brought to cold shutdown for the remainder of December.
Environmentally Related Eventi, An abnormal release occurred on July 24, 1985 associated with recirculating the "A" steam generator. Approximately 1.6 pCi of Cs-134 and 3.5 pCi of C5-137 were released to a plant drain which discharges directly to the plant effluent. Samples of the plant effluent indicated no detectable activity.
The carbon adsorber. trays in the reactor building's exhaust air system were replaced in August 1985. The replacement became necessary due to spray painting in the reactor building. .
A sample for the regenerate hold-up tank release 85-137 was taken for composite but was not included in the August composite. A separate area in '.
the hot laboratory has been set aside for storing composites to reduce confusion. .\-
A release of noble gases occurred through the secondary steam relief' valves on the December 26, 1985 plant trip. A description of the event was transmitted to your office by District letter RJR 86-075 dated February 19, 1986.
Approximately 30 gallons of tritiated water were released to the ground on the south side of the protected area on December 28, 1985. No radiological impact is expected due to dispersion and evaporation in the soil. Samples taken at the fenceline indicated no detectable activity.
Water samples were collected in October 1985 from the regenerate hold-up tanks and the retention basins onsite at Rancho Seco for purposes of the ongoing Lawrence Livermore National Laboratory (LLNL) Rancho Seco Environmental Studies program. The results of the analysis of these samples are included in LLNL report UCID-20641 Part I, February 6, 1986 and Part II, January 23, 1986 which have been already transmitted to you. Radionuclides were detected in these samples but at concentrations less thin the Technical Specifications
~
EX- B 3' Page A of S Pa m
RJR 86-087 .
J B Martin March 3, 1986 .
l ..
4.21 Lower Limits of Detection requirements. However, the District will prepare a special report which will reevaluate the 1985 radiological liquid effluent source term and submit this report to you by August 31. 1986.
This report was prepared by R. G. Desterling and R. L. Gardner of United Energy Services Corporation under the review and direction of E. W. Bradley, Supervising Health Physicist in the Nuclear Engineering Department.
If there are any questions concerning this report, please contact Mr. Bradley a(the,Districtoffice. ,
k\$chp R. J.: RODRIGUEZ i ASSISTANT GENERAL MANAGER, NUCLEAR Attachment t o
1 EX- BIT 3' -
w a Lw
l SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JucY - DECEMBER 1985 ' ..
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I2 b" Rancho Seco Nuclear Generating Station Unit No.1 Clay Station, California License Number DPR-54 4)suun SACRAMENTO MUNICIPAL UTlLITY DISTRICT
~X -B 37
$' 5 Page H of N Pages d
SEMfANNUAL RA010 ACTIVE EFFLUENT RELEASE REPORT JULY-DECEMBER 1985 RANCHO SECO NUCLEAR GENERATING STATION ,
s
^
Prepared by: Richard G. Desterling, CHP*
Robert L. Gardner, CHP*
Reviewed by- l f C bbb,-
Edward W. Bradley Supervising Health Physicist Nuclear Engineering t.
- Contractors, United Energy Services Corporation EXHIBIT 37
- - _ - -- - _ e - c _ ;
l
- . . SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT l JULY-DECEMBER 1985 RANCHO SECO NUCLEAR GENERATING STATION .
EXECUTIVE
SUMMARY
The Rancho Seco Nuclear Generating Station Unit No.1 is located approximately 26 miles north-northeast from Stockton and 25 miles southeast from Sacramento in Sacramento County, California. Rancho Seco Unit No.1 began comercial operation on" April 17, 1975. The single unit on the Rancho Seco site is a pressurized water reactor supplied by Babcock & Wilcox. The rated capacity is 2,772 megawatts thermal and 963 gross megawatts electrical.
This report has been prepared by the Sacramento Municipal Utility District to meet the reporting requirements of Technical Specification 6.9.2.3, Operating License No. OPR-54. . It,is transmitted to the U. S. Nuclear Regulatory Commission Region V.0ffice of Inspection and Enforcement. Copies are provided to the California Energy Comission, California Department of Public Health, California State University at Sacramento, Central Valley Regional Water Quality Control Board, local libraries, and District offices as a public document.
This document reports the quantities of radioactive materials released as liquid, gaseous ef fluents and solid radwaste shipments. Estimates of the radiological impact to man associated with the gaseous and liquid effluent releases are presented. The report period is July 1, 1985 through December
- 31. 1985.
The format of this report follows U. S. NRC Regulatory Guide 1.21 as required by Technical Specification 6.9.2.3. There are six major sections: (1)
Gaseous Effluents; (2) Liquid Effluents; (3) Solid and Radwaste shipments; (4)
} Radiological Impact on Man: (5) Meteorological data; and (6) Supplemental Information. Each section contains, where appropriate, narrative or descriptive material to clarify the tables. Several items included in Regulatory Guide 1.21 have been retained in this report, although they are t
outmoded by the ' dose-based' Technical Specifications.
Additional information required by Technical Specification 6.9.2.3 is also l
presented. This includes changes to the Offsite Dose telculation Manual and the solid radwaste Process Control Program, assessments of radiation doses associated with radioactive effluents to unrestricted areas, and reports of abnormal releases.
In accordance with Technical Specifications, the District maintains an Offsite Dose calculation Manual. It contains the detailed methods and procedures to perform the calculations supporting this report.
The radiological liquid effluent control program that began in October 1984
- has continued during this report period. Normal liquid effluent releases of radionuclides other than tritium continue to be below the limits of detection required by Technical Specifications 4.21.
The estimates of radiation doses equivalent to the non-occupational maximally exposed individuals are one or more orders of magnitude smaller than the limits of 10 CFR Part 50, Appendix 1.
l
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SSINS No.: 6830
\ Accession No.:
8006190038 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 August 22, 1980 IE Circular No. 80-18: 10 CFR 50.59 SAFETY EVALUATIONS FOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS Description of Circumstances:
Recent inspection efforts at operating power reactors have revealed numerous instances in which licensees have failed to perfonn adequate safety evaluations
, to support changes made to the design and/or operation of facility radioactive waste treatment systems. These safety evaluations are required by the regula-tions of 10 CFR 50.59 whenever changes are made in the facility as described in the Safety Analysis Report (SAR).
The inadequacies of the evaluations have caused radiological safety hazards to occur unidentified and therefore to remain unevaluated and uncorrected. In
,f two particular cases, the inadequately evaluated system changes resulted in system failures that caused an uncontrolled release of radioactivity to the
. environment. In each of these situations, a proper 10 CFR 50.59 safety evalua-tion should have identified and corrected deficiencies in the system modifica-tion and/or operation and would have prevented the inadvertent release of radioactivity.
NRC followup examination of the situation indicates that the inconsistency and/or inadequacy of licensee safety evaluations may be widespread. A wide range of opinions seems to exist among licensees as to what constitutes an appropriate 10 CFR 50.59 safety evaluation, particularly for radwaste systems.
Therefore, the following discussion and/or guidance is provided for licensee use in preparing future 10 CFR 50.59 safety evaluations to support changes in the design and/or operation of the radioactive waste treatment systems of .
licensed facilities.
, Although the contents of this guidance are specifically directed to the
, '.' radioactive 10 CFR 50.59 waste safetysystems, evaluationthe general guidance areprinciples and tophilosophy also applicable the facility of the
? design and operation as a whole; thus, the application of 10 CFR 50.59 should ,
\ reflect a consistent approach.
Discussion:
, ; The requirements of 10 CFR 50.59 are composed of three essential parts.
' First, paragraph (a)(1) is pennissive in that it allows the licensee to make A i changes to the facility and its operation as described in the Safety Analysis MN Report without prior approval, provided that a change in Technical Specifica- ,
l \ tions is not involved or an "unreviewed safety question" does not exist.
Criteria for determinin whether an "unreviewed safety question" exists are 3
%@ defined in paragraph (a (2). Second, paragrap changes made under the authority of paragraph a requires that records of .
) be maintained. These l records are required to include a written safety evaluation that provides the Adnu nt fb{ _ _ _ . . . . . _ . _
y 4'e IEC 80-18 August 22, 1980
( Page 2 of 3 basis for determining whether an "unreviewed safety question" exists.
Paragraph (b) also requires a report (at least annually) of such changes to, the NRC. Third, paragraph (c) requires ~thatproposedchangesinTechnical Specifications be submitted to the NRC as an application for license amendmen.t.
L.ikewise, proposed changes to the facility or procedures and the proposed conduct of tests that involve an "unreviewed safety question" are required to be submitted to the NRC as an application for license amendment.
Any proposed change to a system or procedures described in the SAR, either by l text or drawings, should be reviewed by the licensee to determine whether it involves an "unreviewed safety question." Maintenance activities that do not' result in a change to a system (permanent or temporary), or that replace components with replacement parts procured with the same'(or equivalent).
purchase specific'ation, do not require a written safety evaluation to meet
- 10 CFR 50.59 requirements. However, a safety evaluation is required to meet
~
the provisions of 10 CFR 50.59 and any change must be reported to the NRC as required by 10 CFR ponents described 50.59(b) in the SAR are if the following) removed; circumstances (2 component occur:
functions (1)com-are altered; (3) substitute components are utilized; or (4) changes remain following comple '
tion of a maintenance activity.
Notice to Licensees:
i For all cases requiring a written safety evaluation, the safety evaluation must set forth the bases and criteria used to detemine that the proposed change does or does not involve an "unreviewed safety question." A simple statement of conclusion in itself is not sufficient. However, depending upon the significance of the change, the safety evaluation may be brief. Th.e scope of the evaluation must be commensurate with the potential safety significance of the proposed change or test. The depth of the evaluation must be sufficient to detemine whether or not an "unreviewed safety question" is involved.
These evaluations and analyses should be reviewed and approved by an appro-priate level of management before the proposed change is made.
An important part of the "unreviewed safety question" determination is the evaluation (1) potentialand analysis safety of the hazards areproposed identified,change and 2) ty(the licensee corrective to assure actions are that_
taken to eliminate, mitigate, or control the hazards to an acceptable level. .
All realistic failure modes and/or malfunctions must be considered and protec-
-tion provided commensurate with the potential consequences. All applicable regulatory requirements, including Technical Specifications, must be complied with so that the proposed change shall not represent an "unreviewed safety question." Also, the margin of safety as defined in the bases of the Technical Specifications shall not be reduced by the proposed change.
For radioactive waste systems, the appropriate portions of 10 CFR 20, 30, 50, 71, and 100, the facility Technical Specifications, and 40 CFR 190 (Environ-mental Dose Standard) are applicable, g Additional specific criteria th,at should be reviewed prior to the modification of radioactive waste systems are presented below:
(1) System modifications should be evaluated against the seismic, quality .
group and quality assurance criteria in Regulatory Guide 1.143. Design ,
y 1-C IEC 80-18
^ August 22, 1980 Page 3 of 3 M provisions for controlling releases of radioactive liquids, as presented in Regulatory Guide 1.143, should also be evaluated .
(2) Radiological controls shou m be evaluated agafnst the criteria in ' '
Regulatory Guide 1.21 and af1MitM ew Plan Section 11.5, " Process and Effluent Radiological Monitoring and Sampling Systems." *
(3) Systems involving potentially explosive mixtures should be evaluated against the criteria in Standard Review Plan Section 11.3, " Gaseous Waste Management System " subsection II, item 6.
(4) System design and operation should be evaluated to assure that the radiological consequences of unexpected and uncontrolled releases of radioactivity that is stored or transferred in a-waste system are a: small fraction of the 10 CFR 100 guidelines; i.e., less than 0.5 rem whole body dose, 1.5 rem thyroid from gaseous releases, and less than the radionuclides concentrations of 10 CFR 20, Appendix B. Table II, Column 2 from liquid releases at the nearest water supplies. (See Standard Review Plan Sections' 15.7.1,15.7.2,and15.7.3formoredetails.)
The evaluation must include an analysis encompassing the above criteria to the extent that the criteria are applicable to the proposed changes; i.e., if the modifications involve a change addressed by the above regulations and criteria, f then the modifications must be evaluated in tems of these regulations and criteria.
' (( In conclusion, for any change in a facility radioactive waste system as described in the SAR, a safety evaluation is required in accordance with 10 CFR 50.59. In this safety evaluation and the "unreviewed safety question" determination, the evaluation criteria in Items 1-4 above should be used. If the proposed modification (design, operation, or test) represents a departure from this evaluation criteria, one of the following actions should be taken:
(1) The proposal should be modified to meet the intent of the criteria; (2) The evaluation / determination must present sufficient analyses to demonstrate the acceptability of the departure; or. ,
(3) Comission approval must be received prior to implementing the .
modification (i.e., an unreviewed safety issue may be involved).
No written response to this circular is required. If additional information j
regarding this subject is required, contact the Director of this office. .
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.. E o NUCLEAR REGULATORY COMMISSION 3,p aj wAsHtNGTON, D. C. 20655
%, . . .. . / February 22, 1984 Docket No. 50-312 Mr. Ronald J. Rodriguez Executive Director, Nuclear Sacramento Municipal Utility District 6201 5 Street P. O. Box 15830 Sacramento, California 95813
Dear Mr. Rodriguez:
The Commission has issued the enclosed Amendment No. 53 to Facility Operating License No. DPR-54 for the Rancho Seco Nuclear Generating Station. The amendment censists of changes to the Technical Specifica-tions in response to your letter of July 13, 1979, as revised by letters dated March'11, 1982 and December 7, 1982.
The amendment to the Technical Specifications (1) implements the reovire-rents of Appendix I to 10 CFR Part 50, (2) establishes new limiting conditions for operation for the quarterly and annual average release rates, and (3) revises environmental monitoring programs to assure conformance with Comission regulations. -
A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Comission's Monthly Notice.
Sincerely, r
.. g e ,}
Sydney Minef, Project Manager Operating Reactors Branch #4 Division of Licensing
}
Enclosures:
E
- 1. Amendment No. 53 to DPR-54 i ~#
T.
- 2. Safety-Evaluation '
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cc w/ enclosures: T U E See next page r T.-
M dg - g N
A _
FILE 84-0 50 t
S i
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b Sacramento Municipal Utility District Rancho Seco. Docket No. 50-312 ccw/ enclosure (s):
David S. Kaplan, Secretary and Christopher.Ellison, Esq.
General Counsel Dian Grueuich Esq.
Sacramento Municipal Utility California Energy Comission District 1111 Howe Avenue 6201 S Street P. O. Box 15830 Sacramento, California 95825 Sacramento, California 95813 Ms. Eleanor Schwartz -
- California State Office Sacramento County 600 Pennsylvania Avenue, S.E., Rm. 201 Board of Supervisors Washington, D. C. 20003 827 7th Street Room 424 Sacramento, California 95814 Docketing and Service Section Office of the Secretary ,
Mr. John B. Martin, Regional U.S. Nuclear Regulatory Comission Administrator Washington, D. C. 20555 U.S. Nuclear Regulatory Commission Region V ResiIent. Inspector / Rancho Seco 1450 Maria Lane, Suite 210 c/o U. 5. N. R. C.
Walnut Creek, California 94596 14410 Twin Cities Road Herald, CA 95638 Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear Regulatory Comission Washington, D. C. 20555 Regional Radiation Representative EPA Recion IX Alan S. Rosenthal, Chairman 215 Fremont Street San Francisco, California 94111 Atomic Safety and Licensing Appeal Board Mr. Robert B. Borsum U. S. Nuclear Regulatory Comission Washington, D. C. 20555 Babcock & Wilcox Nuclear Power Generation Division Dr. John H. Buck Suite 220, 7910 Woodmont Avenue Bethesda, Maryland 20814 Atomic Safety and Licensing Appeal Board Thomas Baxter, Esq. U. S. Nuclear Regulatory Corrnission Washington, D. C. 20555 Shaw, Pittman, Potts & Trowbridge 1800 M Street, N.W.
Washington, D. C. 20036 Christine N. Kohl
_ Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Comission Washington, D. C. 20555 Joseph 0. Ward, Chief Radiological Health Branch ,
Helen Hubbard State Denartment of Health Services P. D. Box 63 714 P Street, Office Building #8 Sunol, California 94586 Sacramento, California 95814 I
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UMTED STATES
! a-NUCLEAR REGULATORY COhi..JISSION
- r. 1 wAsmaton. o. c. 20ssa SACRAMENTO MUNICIPAL UTILITY DIST9ICT
, DOCKET N0. 50-312 RANCHO SECO NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE l .
Amendment No. 53 License No. DPR-54
- 1. The Nuclear Regulatery Cosmission (the Commission) has found that:
A. The application for amendment by Sacramento Municipal Utility District (the licensee) dated July 13, 1979, as revised March 11,1982, and December 7,1982, complies with the standards and requirements of the Atomic. Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations af the Commission; C.
There is rea'sonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Consission's regulations; D.
The issuance of this amendant will not be inimical to the cosmon defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CPR Part
.51 of the Consission's regulations and all applicable requirements have been satisfied.
l
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating 1.icense No. DPR-54 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendices A and B. as revised through Amendment No. 53. are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment becomes effective _1_50. days after the date 6f its issuance.
FOR THE NUCLEAR REG TORY C0fEISSION F. Stolz, Chief ratingReactorsBra/ch#4 vision of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: February 22, 1984 f
I- .
DIVISION OF REACTOR AND HAS NEEDS SAFETY PROJECTS WTI COPY COPY V Bishop \V Gilbert Western HAS NFIDS REACTOR SAFETY BRANCH INITIAL COPY COPY REACTOR PROJECTS BRANCH Kirsch l Tolksdorf '
Bianchi l REACTOR PROJECTS SECTION 1 OPERATIONS SECTION i Dodds l Pate "
Maist Zwetzig Willett l ll Miller Elin RESIDENT INSPECTORS O'Brien I Toth 1 1 Gage Waite u Muscat '
1 -
Johnston l l Rfchards 1 REACTOR PROJECTS SECTION 2 ,,/ ENGINEERING SECTION lV Young l A/// l V Kirsch (Acting) ! // l V Johnson,P. I N Ve- 1 Wagner /VS I Narbut l V I l Burdoin "/ l i Prenderasist RESIDENT INSPECTORS l Eckhardt Perez l 1 11 Casella Hollenbach Kanow l l' I
/
/ / g\\ i Vorderbrueggen I Sorensen / n '
Fiorelli l Hon I/ \) \
Polich y Il Ziauneman REACTOR PROJECTS SECTION 3 l I Canter l!
DOCKET NO.
h'IN FILE b/M[,/ ,
V Q(j Morrill TOSS 'N '
/
Hernandez RESIDENT INSPECTORS
- l Chaffee 'I l 1 COPY CAN BE FOUND ON TERA Stewart D'Angelo PROJECT INSPECTOR: Toss or Retain Mendonca l Padovan l ! I ROUTE TO:
Ross l COMMENTS: ~
r Il C r an w
AT.nCHMENT TO LfCENSE AMENDMENT NO. 53 FACILITY OPERATING LICENSE NO. OPR-54 DOCKET NO. 50-312 Remove or replace the following pages of the Appendices "A" and "B" Technical Specifications with the enclosed pages as indicated. The rev'ised pages are identifled by Amendment number and contain vertical lines indicating the area of change.
Aeoendix A lemove Pace Insert Pace .
if 11 fia 11a iv iv v -
y vi vi vif vii viii viii fx fx x
- xi -
"..x 11 1-4 1-4 1-6 1-6
( -
1-7 3-60 thru 3-92 4-63 thru 4 90 6-9 6-9 6-10 6-10 6-11 6-11 6-12a -
6-12a 6-12b 6-12b 6-12e 6-12c 6-12d 6-12d 6-12e 6-12 f 6-14 6-14 6-15 6-15 6-16 6-16 6-17 6-17 6-18 6-18 6-19 thru 6-21 Accendix B R'emove Pa'ce Insert Page i 1 4 4 9 thru 24 -
31 31 32 thru 41 * -
44a 44a 45 45 46* 46*
50 -
l l m_a m _____ __- _
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a UNITED STATES '
- a NUCLEAR REGULATORY COMMISSION wAsMiworow, c. c.2osas
\; .eaj SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REAGTOR REGULATION SUPPORTING AMENDMENT NO. 53 TO FACILITY OPERATING LICENSE NO. OPR-54 '
SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DOCKET NO. 50-312
1.0 INTRODUCTION
To comply with Section V of Appendix I to 10 CFR 50, Sacramento Municipal Utility District (the Licensee) has filed with the Commission plans and proposed technical specifications developed for the purpose of keeping releases of radioactive materials to unrestricted areas during normal operations, including expected operational occurrences, as low as is reasonably achievable. The Licensee filed this information with the Comission by letter dated July 13,1979 (revised March 11, 1982 and December 7,1982) which requested changes to the Technical Specifications appended to Facility Operating License No. DPR-54 for the Rancho Seco Nuclear Generating Station. The proposed technical specifications update those portions of the technical specifications addressing radioactive waste management and make them consistent with the current NRC staff positions as expressed in NUREG-0472. These revised technical specifications would reasonably assure compliance, in radioactive waste management, with the provisions of 10 CFR 50.35a, as supplemented by Appendix ! to 10 CFR 50; 10 CFR 20$105(c),106(g), and 405(c); 10 CFR 50, Appendix A, General Design Criteria 60, 63, and 64; and 10 CFR 50, Appendix B.
T 3g 4
~
2.0 BACKGROUND
AND DISCUSSION 2.1 Regulations 10 CFR 50, " Domestic Licensing of Production and Utilization Facilities "
Section 50.36a, " Technical Specifications on Effluents from Nuclear Power Reactors " provides that each license authorizing operation of a nuclear power reactor will include technical specifications that (1) require compliance with applicable provisions of Subsection 20.106, " Radioactivity in Effluents to Unrestricted Areas." (2) require that operating procedures ,
developed for the control of effluents be established and followed. (3) require that equipment installed in the radioactive waste system be maintained and used, and (4) require the periodic submission of reports to the NRC specifying the quantity of each of the principal radionuclides released to unrestricted areas in liquid and gaseous effluents, any quantities
)
of radioactive materials released that are significantly above design objec-tives, and such other information as may be required by the Commission to estimate maximum potential radiation dose to the pbblic resulting from the l effluent releases.
10 CFR 20, " Standards for Protection Against Radiation " Para-graphs 20.105(c), 20.106(g), and 20.405(c), require that nuclear power plant and other licensees comply with 40 CFR 190, " Environmental Radiation Protection Standards for Nuclear Power Operations," and submit reports to the NRC when the 40 CFR 190 limits have been l or may be exceeded.
- 10. CFR 50, Appendix A - General Design Criteria for Nuclear Power
~
Plants, contains Criterion 60, Control of releases of radioactive materials to the environment, Criterion 63, Monitoring fuel and waste storage; and Criterion 64, Monitoring radioactivity releases. Criterion 60 requires that the nuclear power unit design include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced 2
. during normal reactor operation, including anticipated operational occurrences. Criterion 63 requires that appropriate systems be provided in radioactive waste systems and associated handling areas to detect conditions that may result in excessive radiation levels and to initiate appropriate safety actions. Criterion 64 requires that means be provided for monitoring effluent discharge paths and the plant I environs for radioactivity that may be released from normal operations, including anticipated operational occurrences. q l
10 CFR 50, Appendix B, establishes quality assurance requirements for nuclear power plants. 3 10 CFR 50, Appendix I, Section IV, provides guides on technical speci-fications for limiting conditions for operation for light-water-cooled nuclear power reactors licensed under 10 CFR 50.
2.2 Standa_rd Radiological Effluent Technical Specifications NUREG-0472 provides standard radiological effluent technical speciff-cations for pressurized water reactors which we find acceptable.
( Further clarification of these acceptable methods is provided in NUREG-0133,
" preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants". NUREG-0133 describes methods found acceptable to the staff of the NRC for the calculation of certain key values required in the prep-aration of proposed radiological effluent technical specifications for light-water-cooled nuclear power plants. NUREG-0133 also provides guidance to licensees in preparing requests for changes to existing radiological effluent technical specifications for operating reactors. It also describes our current positions on the methodology for estimating radiation exposure due to the release of radioactive materials in effluents and on the adminis-trative control of radioactive waste treatment systems.
The above NUREG documents address all of the radiological effluent technical specifications needed to assure compliance with the guidance and requirements provided by the regulations previously cited. However, alternative approaches to the preparation of radiological effluent technical specifications and alternative radiological effluent tech-nical specifications may be acceptable if we determine that the alternatives are in compliance with the regulations and with the intent of the regulatory guidance.
3
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The standard radiological effluent technical specifications can be grouped under the following categories:
(1) Instrumentation (2) Radioactive affluents (3) Radiological environmental monitoring (4) Lisign features *
(5) Administrative controls Each of the specifications under the first three categories are com- .
prised of two parts: the limiting condition for operation and the surveillance requirements. The limiting condition for operation provides a statement of the limiting condition, the times when it is applicable, and the act 4ns to be taken in the event that the limiting condition is not met.
In general, the specifications established to asrure compliance with
( 10 CFR Part 20 standards prov1de, in the event the limiting conditions for operation are exceeded, that without delay conditions are restored to within the limiting conditions. In general, the specifications estab-lished to assure compliance with 10 CFR Part 50 provide, in the event the limiting conditions foroperation are exceeded, that within specified times corrective actions are to be taken, alternative means of operation are to be employed, and certain reports are to be submitted to the NRC describing these conditions and actions.
The specifications concerning design features and administrative con-trols contain no limiting conditions foroperation or surveillance require-ments.
Table 1 indicates .the standard radiological effluent technical specifications that are needed to assure empliance with the particular provisions of the regulations described in Section 1.0.
'4
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3.0 E"ALUATION 3.1 General Description of Radiological Effluen't System This section briefly describes the radwaste liquid And gaseous effluent treatment and control systems . installed at the Rancho Seco plant.
3.1.1 Lieufd Effluents -
The water required for the operation of the Rancho Seco plant is recycled, with makeup water being taken from the Folsom South Canal. The only liquid discharge pathway of.potentially contaminated water is via the regenerant holdup tank which normally receives water only from secondary side sources. I the event of a primary to secondary leak, this pathway would be the only way contaminated liquids would be released to the '
environment. See Figure 1. There are no direct connections from the radioactive liquid radwaste treatment systems to the envi onment. Excess liquids from these systems are solidified. Figure 2 shows ' diagram of the liquid and solid radwaste systems for radioactive liquids.
3.1.2 Gaseous Effluents Building locations ter the gaseous effluent discharge points are shown in Figure 1. The three radioactive gaseous effluent discharge pathways are shown in Figure 3 with the auxiliary building ventilation shown in more detail in Figure 4 The turbine building ventilation l
exhaust is not shown as this system is normally not a release point for radioactive material. The turbine level is open to the environment with the lower levels enclosed.
3.2 Radiological Effluent Technical Specifications (RETS)
The evaluation of the Licensee's proposed specifications against the requirements of Appendix I to 10 CFR 50 included the following:
(1) a review of information provided in the Licensee's July 13, 1979 9
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- i11 ,g{ Q ;l P.c - gg,,
- (.X a=
system 40,000 cfm
!><3 ---* g t 2;3 I
3 7 Winter 3 3 g.
18100 3wmmer eacn t , i I
I .
I #
FIGURE 4 1 l
AUXILIARY BUILDING STACK VENTILATION SYSTEM i
)
Auxiitary building stack .
ik l'
l Exhaust fans I
l and filters -
l i
l 4 i < i P H C W.4:,te Steam jet a i
n E
air ojector ' r P a 9y" t A r i S YM""
- l . . .
F 656 filter-Dank Gland scal '
cxnaust t r -
i Fuel handling Rsdeochemics!
- l
, ,,,, and service i ;
j atca
. 4 Auxiliary building ;
l
-._.4 4
_ _ _ _ . . _ _ _ _ _ . ___ _ __ __ _.____.___._________._________m__ _ _ _ _ _ _ _ _
- ,y submittal [1,2], which included copies of the Offsite Dese Calculation Manual
-(00CM) and the Process Control Program (PCP), (2) the resolution of problem areas in that submittal by means of a site visit E33, (3) a. review of the Licensee's March 11, 1982' submittal E43, (4) telephone conferences on LSeptainber 20, October 1, 4, and October S, 1982[5,6,7] to discuss the review, and (5) review of the Licensee's revision.E83. A conference call was then held on November 4, 1982 and November 30, 1982 with the EG&G review team to
~
discuss deviations from the model requirements. All open questions were resolved.in the December 7, 1982~ final submission of the licensee, and a tech-nical evaluation report was finalized for transmittal 93 3.2.1 Effluent Instrumentation-The primary objective of the RETS with regard to affluent instrumentation is to ensure that'a111significant liquid and gaseous releases of radioactivity are monitored. . The Licensee's information documents that the liquid effluent
- ' release point is monitored. Liquio radioactive wastes are nonnally solidified or stored on site. . Normally only water from the regenerant holdup tank, which is the waste water from regeneration of resins on the secondary side, is
[k released to the environment. This release point is monitored and will give automatic termination of the release if predetermined concentrations of radioactive material is detected. Rancho Seco uses once through steam generators. Therefore, they have no steam generator blowdown. Service water is not used to cool components on radioactive systems and therefore j requires no conitoring instrumentation. The component cooling water system at Rancho Seco does not require monitoring as this is a closed system and cannot be released to the environment. Also service drains cannot be released to the environment.
Gaseous radioactive affluent releases from Rancho Seco are monitored and have alarm functions. All release points have provisions for automatic @
termination of. release with the exception of the radwaste service area vent.
~'
This building is not considered to have the potential for high level gaseous releases. Therefore, the automatic termination of release function is considered unnecessary. The functions of the waste gas processing system monitoring instrumentation.are performed by the Auxiliary Building stack-mo'nitor. All other inplant' systems, including the condenser air ejector, 11 1:
1
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9 are monitored at the~ effluent release points.
3.2.2 Concentration and Dose Rates of Effluents The objective of the RETS with regard to concentration and dose ,
rates _of effluents is to ensure that offsite effluent concentrations do not exceed the maximum permissible concentrations (MPC's) established by 10 CFR 20, Appendix 8. Table II, Columns 1 and 2. The Licensee has -
stated that the concentration of radioactive material will be monitored "at all times," or "during releases" for batch releases. The setpoints ,
of the monitors at each release point are pre-established to prevent exceeding the release concentrations or corresponding dose rates of 10 CFR 20 in unrestricted areas. The concentration of liquid effluents and the dose rate due to gaseous effluents will be determined.tn accordance with the 00CM.
The liquid effluent. release pathway is the Regenerant Holdup Tank
( line. This effluent line has all the sampling art analysis, and instrv-mentation requirements as a liquid radwaste effluent line. Adequate assurance is therefore present that the 10 CFR 20 objectives will not be exceeded and any releases will be monitored.
The gaseous monitcring systems, with the exception of the radwaste ,
building which has an alarm function only, are equipped with automatic ~~T~ '
termination of effluents. Should concentrations be found to exceed the MPC specified in 10 CFR 20, based on monitoring setpoint values, release rates will immediately be decreased. The Lower Limit of Detection (LLD) for noble gas monitors at Rancho Seco is specified as 10~4 pCi/mi as Xe-133 equivalent. This is considered acceptable as existing instru-mentation is unable to meet the LLD listed in the model RETS (i.e.,10-6 uC.1/ml . ) . _
The concentration of radioactive materials in releases will be determined as required by the model RETS. Sampling requirements for startups, shutdowns, and 15% power changes are worded more conservatively 12 I
9
t
)-
L by the licensee.
I 1
3.2.3 Offsite Doses From Effluents The objective of the RETS with regard to offsite doses from effluents is to ensure that offsite doses are kept As Low As Reasonably Achievable (ALARA), are kept to a small fraction of the 10 CFR 20 limits, and are in accordance with 10 CFR 50, Appendix !. The Licensee has committed to meet the quarterly and yearly dose criteria for liquid effluents, and to -
use the ODCM methodology for determining the cumulative gaseous dose to individuals, thus meeting the intent of NUREG-0472. The Licensee has committed to maintain the air doses in unrestricted areas, for noble gases, to those specified in Secion 3.11.2.2 of the model RETS. The Licensee has also made a commitment to maintain the dose to an individual from release of Iodine-131, tritium and radioactive particulate with half lives greater than eight days at the values listed in Section 3.11.2.3 l of the model RETS, thus satisfying the intent of NUREG-0472.
3.2.4 Effluent Treatment The objectives of the RETS with regard to effluent treatment are to ensure that wastes are treated to keep releases ALARA and to satisfy the requirement for technical specifier.tions governing the maintenance and use of radwaste treatment equipment. Technical specifications for liquid IdE radwaste treatment are not required as no pathways exist for release to the environs from the liquid radwaste treatment systems. The Licensee has committed to use the gaseous radwaste treatment system when the projected doses averaged over 31 days exceed 25% of the annual dose design objectives prorated monthly. This meets the intent of 10 CFR 50, Appendix-I,Section II.D. The Licensee has also canmitted that the gaseous radwaste system components shall be operable when required to process waste. Also, a commitment has been made to make necessary dose projections in accordance with the ODCM at least once per month.
Therefore, the Licensee has met the intent of NUREG-0472.
13
3.2.5 Tank Inventory Limits The objective of the RETS with regard to tank inventory limits is to ensure that the rupture of a radwaste tank would not cause offsite doses greater than the limit: set in 10 CFR 20 for non-occupational exposure. The Licensee has put a curie limit on all temporary outside liquid tanks that are not diked and has consnitted to surveillance in accordance with NUREG-0472. For liquid holdup tanks, this limit (i.e.,
j,10 curies) excludes tritium and dissolved or entrained noble gases.
For waste gas storage tanks which are in constant use, a limit of 135,800 -
curies for noble gases has been set. Surveillance to detennine gas storage tank inventory will be done via daily grab samples when the primary coolant exceeds 436, for greater than thirty minute half-life radionuclides. This reactor coolant activity would result in storing a small fraction of the total curie limit for noble gases in any waste gas decay tank, initiating sampling in a timely fashion. This is considered an acceptable surveillance method for determining that an unplanned k release from a waste gas decay tank could not exceed effluent release limits.
3.2.6 Exelosive Gas Mixtures The objective of the RETS with regard to explosive gas mixtures is to prevent hydrogen explosions in the waste gas treatment system. The l
Licensee has committed to maintain a safe concentration of oxygen in this system as hydrogen is present in excess. The oxygen concentration willbemaintainedat14%. If the concentration increases above this limit, addition of waste gases will be halted and the concentration will be reduced to the acceptable limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The Licensee will l maintain constant monitoring of 02 in the waste gas holdup Jysten.
l The system will be in use only during system operation, which is adequate.
l The requirmaents of a non-explosion proof system (Section 3.11.2.5.S of i the model RETS) are being met.
3.2.7 Solid Radwaste System 14 i
i - _ __ __________________________________a
. 1 l
The objective of the RETS with regard to the solid radwa'ste system is to ensure that radwaste will be properly processed and packaged before it is shipped to the burial site. The Licensee has comitted to use the methods prescribed in the process control program (PCP) to ensure that the requirements of 10 CFR 20 and 10 r.FR 71 are met prior to shipment of radwaste from the site. The plant will use the Chem-Nuclear waste solidification system.
3.2.8 Environmental Monitoring The objectives of the RETS with regard to environmental monitoring are to ensure that an adequate and full-area-coverage environmental monitoring program exists and that the 10 CFR 50, Appendix I requirements for technical specifications on environmental monitoring are satisfied.
The Licensee has explicitly followed NUREG-0472, where applicable, including the Branch Position statement dated November, 1979. The Licensee's
( methods of analysis and maintaining yearly records satisfy the requirements and meet the intent of 10 CFR 50, Appendix 1. The specification for the land-use census satisfies the requirements of Section 3.12.2 of NUREG-0472 by providing for the census once a year in the areas specified.
The specification for interlaboratory comparison satisfies the requirement of Section 3.12.3 of NUREG-0472 by stating they will participate in an NRC-approved program.
3.2.9 Audits and Reviews The objective of the RETS with regard to audits and reviews is to ensure that audits and reviews of the radwaste and environmental monitoring programs are properly conducted. The Licensee's administrative struct are designates the Plant Review Comittee (PRC) and the Management Safety Review Cemittee (MSRC) as the two groups responsible for the review and audit of the radiological environmental :nonitoring program, the ODCM, and the PCP. The MSRC is responsible for auditing those three programs and a Quality Assurance (QA) program, with the frequency of review to be 15
1
. ~
equal to or greater than that required by NUREG-0472. The PRC is responsible for reviewing every unplanned release of radioactiv'e material; the review is to include an event description, remedial action to prevent recurrence, and corrective action. The PRC also reviews any changes in the ODCM and the pCP.
. 3.2.10 Precedures The objective of the RETS with regard to procedures is to establish a requirement for implementing the ODCM, the PCP, and the QA program.
The Licensee has conmitted to establish, implement, and maintain written procedures for the PCP, ODCM, and QA program.
3.2.11 Reoorts The objective of the RETS with regard to reports is to ensure that
( appropriate periodic and special reports are submitted to the NRC, and that these reports meet the requirements of 10 CFR 50.36a. The Licensee has made commitments to issue annual and semi-annual reports as required l under Sections 6.9.1.12, and 6.9.1.9, respectively, of NUREG-0472.
3.3 Offsite Dose Calculation Manual (00CM1, A brief discussion of the methodology and approach used by the Licensee to calculate offs:ite dose and to maintain the operability of the effluent system is provided in this section. The methodology used by the Licensee is evaluated for consistency against the methodology and guidelines set by the NRC staff. As a minimum, it is required that the ODCM provide equations and methodology for the following topics:
e - alarm and trip setpoint on effluent instrumentation e liquid effluent concentration in unrestricted areas e
gaseous effluent dose rate at or beyond the site boundary J
16
q
.. 1 e liquid and gaseous effluent dose contributions e liquid and gaseous affluent dose projections e description and location of samples for the environmental monitoring program In addition, it has been suggested, but not required, that flow diagrams defining the treatment paths and the components of the radio-active liquid, gaseous, and solid waste management systems be' included and reviewed for consistency against the system being used at the ~
station. Rancho Seco has not provided diagrams of the radwaste treatment systems. A description and location of samples in support of the environ-mental monitoring program has been provided in the ODCM.
3.3.1 Evaluation The Licensee has followed the methodology of NUREG-0133 and_ Regulatory Guide 1.109 to determine the alarm and trip setpoints for_ the liquid and gaseous effluent monitors. A conservative f actor is used for the setpoints, which ensures that maximum permissible concentration (MPC) will not be exceeded.
The dose rate at or beyond the site boundary due to gaseous effluent release is in compliance with 10 CFR 20. Gaseous effluents are released from three release points for which conservative values of relative concentration and relative deposition for the average atmospheric dispersion conditions are used by the Licensee. !
The dose evaluation of pathways associated with the release of radioactive material in liquid effluents is stated to be in compliance with 10 {fR 50. The dose contributions are calculated once per 31 days for all applicable pathways.
i Evaluation of noble gases released to the atmosphere include both beta and gamma air doses at the off-site location with the highest long 17
. l term vent X/Q. The critical location is based on the external air dose pathway only. .
For radiciodine, tritium, and particulate, the Licensee has stated that the method used in the ODCM for calculating releases to unrestricted j.
areas will meet the design objective values of maintaining an annual \,
dose or dose commitment not-to exceed 15 mram to any organ of the maximum /.
exposed individual. The Licensee has rown the methods of calculating '. l the dose using X/Q and D/Q values fm ,1 appropriate pathways. !
The Licensee has'comitted to performing dose projections for gaseous effluent releases once every 31 days to determine the use of appropriate partions of the radwaste system except where systems are in operation at all times.
The Licensee has provided a complete description of sample locations
- in the ODCM Figs. 5.1-1 through 5.1-3 and Table 5.1-1. This description is consistent with the sampling locations specified in the Licensee's RETS. Table 5.1-1 in the Licensee's ODCM tabulates the site number identification, sector location, distance from station center, and sample point description. Table 5.1-1 covers all of the Licensee's committed sampling exposure pathways in accordance with Table 3.12-1, RETS Environmental Monitoring Program.
3.4 S,ummary of Technical Evaluation Table 2 contains a correspondence of major sections of NUREG-0472, the current technical specifications, and the Licensee's proposal. The Licensee's proposal was evaluated and the following conclusions were reached:
- 1. The Licensee's proposed RETS meets the intent of the NRC staff's current standard, " Radiological Effluent Technical Specifica-tions," NUREG-0472, Rev. 2, February 1,1980.
18
. e.[
- 2. The Licensee's Offsite Dose Calculation Manual (ODCM) uses documented and approved methods that are consistent with the NRC's methodology in NUREG-0133. The ODCM is also consistent with the Technical Specifications.
- 3. The Cicensee is presently operating under a PCP that is available for review by NRC at any time.
O 19 1
g- :
TABLE 2. CORRESPONDENCE OF PROVISIONS OF NUREG-0472 THE CURRENT TECHNICAL SPECIFICATIONS AND THE LICENSEE'S PROPOSAL Current NUREG- Technical Licensee RETS 0472 Specifications Proposal Requirement (Section) (Section)* (Section)
Effluent 3.3.3.9 2.6.1.D&E, 2.6.2.0-F 3.19 & 4.19 Instrumentation _ 3.3.3.10 2.6.3.0, 2.6.4.B-E 3.20 & 4.20 Concentrations 3.11.1.1 2.6.1.A. 2.6.2.B&C 3.21.1 & 4.21.1 3.11.2.1 2.6.3.A 3.22.1 & 4.22.1 Offsite Doses 3.11.1.2 2.6.1.B&C 3.21.2 & 4.21.2- -
3.11.2.2 2.6.3.8 3.22.2 & 4.22.2 3.11.2.3 2.6.3.8 3.22.3 & 4.22.3 3.11.4 3.29 & 4.29 Effluent 3.11.1.3 2. 6.1. F ---
Treatment 3.11.2.4 ---
3.23 & 4.23 Tank Inventory 3.11.1.4 2.6.1.G 3.21.3 & 4.21.3 Limits 3.11.2.6 2.6.3.E 3.24 & 4.24 Explosive Gas 3.11.2.5 ---
3.28 & 4.28 Mixtures Solid Radwaste 3.11.3 2.6.5 3.25 & 4.25 Environmental 3.12.1 4.0 3.26 & 4.26 Monitoring Audit and Review 6.5.1 5.3.A 6.5.1.6 6.5.2 5.3.B 6.5.2.7 Procedures 6.8 5.5 6.8 Reports 6.9.1.6 5.6.1.A 6.D.2 6.9.1.8&9 5.6.1.8 6.9.3
- Being Revised or Deleted 20 l l i
l \
L i l _ _______________1_O
j
)
ll' -
~
3.5 Res ul ts ,
The proposed changes to the radiological effluent technical specifi-l cations for the Rancho Seco Nuclear Generating Station have been evaluated, reviewed, and found to be in compliance with the requirements of the NRC regulations and with the intent of NUREG-0133 and NUREG-0472 (Rancho Seco
~
Station is comprised of a single pressurized water reactor). and thereby fulfill all the requirements of the regulations related to radiological effluent technical specifji:ations. ,, .
The proposed changes will not remove or relax any existing requirement related to the probability or consequences of accidents previously consider-ed.
The proposed changes will not remove or relax any existing requirement needed to provide reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner.
4.0 ENVIRONMENTAL CONSIDERATION
We have determined that issuance of the proposed amendment to the Technical Specifications appended to Facility Operating License No.
DpR-54 for Rancho Seco will not authorize a significant change in the types or a significant increase in the amounts of effluents or in the authorized power level, and that the amendment will not result in any significant environmental impact. Having made these determinations, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR Sl.5(d)(4), that an environmental impact statement, negative declarati.on, or environmental impact appraisal need not be prepared in f connection with the issuance of this amerdment.
5.0 CONCLUSION
We have concluded, based on the considerations discussed above,.that:
21 j
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: February 22, 1984 The following NRC personnel have contributed to this Safety Evaluation:
W. Heinke. -
S 22
l; ..
6.0 REFERENCES
- 1. Sacramento Municipal Utility Df strict, Letter of Tra'nsmittal, Rancho Seco RETS, July 13, 1979 and revised December 7,1982. .
- 2. F. B. Simpson, Letter of Transmittal, Transmittal of RETS Criteria and Rancho Seco ~ RETS - SIM-27-81, October 21, 1981.
- 3. Rancho Seco Plant Visit, Review of Rancho Seco Radiological Effluent Technical Specifications, November 4-6, 1981.
- 4. Sacramento Municipal Utility District, Letter of Transmittal, Proposed Technical Specification Amendment No. 62. Revision 1. March 11 1982.
- 5. S. Miner (NRC), C. A. Willis (NRC), R. Colombo '(SMUD), R. Miller.
(SMUD), S. W. Duce (EG&G).- J. W. Mandler (EG&G), W. Serranc (EG&G).
Telephone Conference, September 20, 1982.
- 6. R. Colombo (SMUD),_ R. Miller (SMUD), S. W. Duce (EG&G), Telephone Conference, October 1 and 4,1982.
- 7. R. Colombo (SMUD), R. Miller (SMUD), S. W. Duce (EG&G), Telephone Conference October 8,1982.
- 8. J. J. Mattimoe, General Manager, Sacramento Municipal Utility District, k- to S. Miner, Project Manager, Letter, November 1982.
- 9. B. F. Saffell, Letter of Transmittal, Rancho Seco Nuclear Generating Plant - Technical Evaluation Report, SAFF-24-83 Jantary 24, 1983.
3 23
,4 ig.x .
,]d C! MUD.SACRAMENTO MUNICIPAL UTIUTY DISTNCT "" $201 s Street. P.o. Som 15a30. sacramemo. CA 95813: (914) 452 3211 '
b AN ELsCTRIC SYSTEM SERVING THE HEART OF CAUFoRNIA
,4
.RJR 84-425 g.
3 September 27, 1984 J B' MARTIN, REGIONAL ADMINISTRATOR REGION V 0FF. ICE OF INSPECTION AND ENFORCEMENT U S NUCLEAR REGULATORY COMMISSION .
1450 MARIA LANE, SUITE 210 -
.' WALNUT CREEK CA 94596
' DOCKET NO. 50-312. '
LICENSE NO. DPR-54 SPECIAL REPORT NUMBER 84-07 .
In accordance with the requirements of Rancho Seco Nuclear Generating Station Technical Specifications. sections 3.17.2 and 3.25, the Sacramento Municipal Utility District hereby submits the following Special Report.
These specifications.are part of the Radiological Effluent Technical Specifications (RET 5), wtrtch were implemented on July 21, 1984 Tecnnical' Spec.ifications section 4.21.2 requires that the cumulative calcu-lated radiological exposure resulting from liquid effluents be determined in' ~
accoroance with the Offsite Dose Calculation Manual.(ODCM) at 1 east monthly.
Technical Specification section 3.17.2 requires that the District prepare and summit to' the Connission within 30 days a Speciah, Report if the calculated radiological 1' exposure resulting from liquid effluents exceeds the calendar cuarter limits of'l.5 mrem total body and 5 mrem to any organ,.and the calendar year. limits of 3 mrem total body and 10 mrem to any organ:- Techni-cal Specifications section 3.25 requires the_ District to. prepare and submit to tne Commission within 30 days a Special Report, as defined i.n 10 CFR -
20.405(c), if the calculated radiological exposure resulting from a uranium fuel cycle source excereds the annual limit of 25 mrem total body or any k . organ or 75 mrem to the thyroid. ,
The District acknowledges that the calculated radiological exposure to the
" Maximum Individual" exceeds the limits of 10 CFR 20.405(c) and 40 CFR 190 during the calendar year 1984 through 8/31/84, but the liquid effluent release conditions resulting in this violation have now been corrected.
Near term and long term corrective actions are detailed in the attachment.
Based on these facts, the District believes that a request for. variance is not recuired at this time.
~
Details of the calculated radiological exposures to the " Maximum Inoividual" th IIf resulting from liquid effluents is also presented in the attachment to this The calculated radiological exposure for calendar year 1984 through J
l
~report.
S/31/84 indicates a total body exposure to a " Maximum Adult" of 185 A '
v %;ps yy_
, EXHIBIT ,
,u,
( ,
~e -
J. B. Martin -2 September'27, 1984
.c liver exposure to a " Maximum Child" of 302 mrem. The District fully expects that the corrective measures now in effect will result in a calculated radio--
logical exposure, from liquid effluents, to the " Maximum Individual", less than the limits of Technical Specification 3.17.2 for the fourth quarter 1984 and for all subsequent reporting periods.
Following is-'a discussion of the circumstances that led up to the exceeding of these limits. -
The root cause of this unusual condition was a small but continuous leak in
.the "B" Once Through Steam Generator (OTSG). This small leak apparently originated on or about September 17, 1983, when the plant was shutdown to repair a tube leak in the "A" OTSG. The leaking tube in the "A" OTSG was stabilized and plugged. When the plant was returned to service the small continuous leak was detected in the "B" OTSG. The leak rate was confirmed to be on the order of'O.01 to 0.09 gallons per minute. Typically, the leak rate of such leaks, if they are tube leaks, will increase as the size of the tube flaw increases (crack opens). Since the leak was of such a small size, it was considered to be. extremely difficult to detect once the unit was shutdown.
Therefore, the decision was made to continue operating. The leak rate subse-quently increased and the District attempted to locate the leak again on an l outage consnencing July 3,1984 This time a " bubble test" was attempted which involved pressurizing the secondary side of the OTSG with nitrogen and visually scanning for buboles above the tube sheet. This' test, which was more sensitive
-than the conventional Ettdy Current test that the District has used in the p,ast, eadily found a tube'with a large crack, but f ailed to locate the residual small leak. The plant was restarted on August 16, 1984, and following un-usual increases in primary iodine levels, was shut down on August 31, 1984 This time a helium leak test, never used by the District before, .was emolayed and a leaking tube was'identifieo. The tube was suosecuently stabilizeo and plugged. This action may eliminate the major source of secondary system contamination which lead.to the exceeding of the above stated limits.
The path that the radioactive material takes to get from the secondary system to the general public is as follows. Backflush water, regenerant waste and flush water from the polishing demineralizers flow to the Polishing Deminera- -
lizer Sump (PDS). Also, leakage from the secondary system generally flows to the Condensate Pit Sump where it is transferred either to the PDS or directly to one of the Regenerant Hold Up Tanks (RHUTs). The PDS is typically.pumoed !
to one of the RHUTs, which, when full, are agitated, neutralized, and sampled.
The results of the sample are used to determine the total release activity i for each isotope. Based on this data, a Liquid Waste Release Permit is
! generated, then the tank is pumped to one of the Retention Basins. When the Basin is full, it.is recirculated and sampled to determine a dilutior, rate that would conform to appropriate limits. Samoles are also taken during tne discharge to provide assurance that regulations are met. Previously, 10 CFR 20, Appendix B requirements were applied as limits as the discharge left the site. The District now is limiting its disenarges so that 10 CFR 50 ,
Appendix ! limits will not be exceeded.
c .
EXHIBIT
- a a *A s
.- 'J. '. Martin B -
Septameer 27, 1984 s
'he design for Rancho Seco is such that liquid radioactive discharges should
- at be required. Due to the difficulty in locating the small OTSG 1eak, the releases became necessary. The attachment considers in detail the near term and long term corrective actions that the District is taking to Ansure that this condition is corrected. Also, the attachment details the Dis-trict's assessment of the radiological aspect of this condition.
Following startup from the current outage, due to those near term corrective actions already implemented, it is anticipated that there will be no radio-active discharges that will exceed the limits of Technical Specification 3.17.2. '.
If-th *e are any questions concerning this report, please contact Mr. Ron o bo,a the Rancho Seco Nuclear Generating Station.
Tv '7 R. J. riguez j
ExecutiveDirec$g, Nuclear S
9 9 O
i w
- EX .B T Page of 6 Pages
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, ATAC*-MENTTOSPECIALREPORT84-Of ~I DISCUSSION OF CORRECTIVE ACTIONS AND RADIOLOGICAL EXPO 5URE CALCULATIONS .
4 '
- ar Te-m Corrective Actions .
The District has implemented and/or is in the process of implementing several near term corrective actions. These are:
- I. Addition of the purification demineralized bypass valve to
- the locked valve list. ,
II. Revision of the operating procedure for the purification deminera-lizer to require notation of the valve status.
. III. Locate and plug the small leak in the "B" OTSG. .
IV. " Tighten" the secondary system to eliminate leaks.
V. Adopt a " throw away resin policy" for the polishing dem'ineralizers, including solidification and burial at approved disposal site of i radioactive resin.
VI.
Interim use of portable demins with temporary piping to allow o+-e$
cleanup of RHUTs.
A
/ VII. ' Administrative policy to control releases such that monthly calculateql dese is under Technical Specification 3.17.2 limits.
VIII. Processing of the PDS or RHUTs back to the radwaste system in case activity is too high to release.
IX. Use of the NRC verified computer code to calculate a " running" total of the dose from releases. Data wil'1
- be processed as releases are made to ensure releases meet limits.
Following is a discussion detailing the above actions.
Action I: The District has taken action to assure that the bypass valve for the, Pur1fication Demineralizers fe a proper position by adding this valve to the locked valve list, controlled by procedure SP 214.03, " Locked Valve List",
approved on 9/27/84 This prevents unintentional mispositioning of the valve, as was the case during August 1984, when iodine levels in the primary system rose to unusually high levels, although they remained within the Tech Spec limits. Action II, coupled with this action, should ensure satisf actory operational alignment of these demineralizers.
Status 2 Complete.
Action II: The district has taken action to revise the operating procedure A.lz, " Reactor Coolant Chemical and Hydrogen Aedition System", accroved on 9/20/84 The revision recuires that tne purification demineralized bypass valve position be recorced,in the proteoure to ensure proper alignment.
.atus: Complete.
. EXEBIT - 1
_ m - m
[ -
- Action III: On Septa'mber 16, 1984, the District identified a leaking tube in s se "B" OT5G, lane tube 74-2. The leak was located using a. helium leak de-ection method and verified using Eddy Current techniques. The leak was at the 15th tube support plate. The tube was removed from service by insta)11ng a stabilizer red and explosive plugging. The leak rate increased just as the unit was being shut down, which makes it impossible to be certain if the leak repaired was the residual small leak that has been present since September 1984, er is a new leak that opened up as the stresses on the tubes increased during shutdown. In an attempt to determine if this was the case, the District con-tinued its inspection of the OTSG after this tube was identified. No other leaks were found, leaving the issue unresolved until startup when leak rate determinations can bq made to verify the presence or absence of the small leak.
. If the small leak has not been corrected, the proper operation of the polishing and purification domineralizers, along with Actions IV through VIII, should Gnsure that releases meet the requirements referenced above.
Status: Complete (tentative pending leak tests during operation). cg-Action IV: The District has taken steps to " tighten" the secondary system to eliminate small leaks / drips from components that would be collected typically in the. Condensate Pit Sump and eventually be transferred to the RHUTs. Since the secondary system has some activity present, mainly as a result of the re-sidual small tube leak, these small leaks or drips constitute a source of radioactive liquids to the RHUTs. The Operatipns Department has undertaken a program to tighten Y1anges, fittings, packings and gaskets. Also, specifi,c 1 pairs of known leaking valves has been initiated by the Maintenance De-artment. This program will be continued to complement the other corrective cetions listed herein.
Status: Complete / Continuing. -
Action V: The District is implementing a program to use disposable resins for tne Concensate Polishing Demineralizers. This will include the solidification and burial at an approved site of radioactive resins. This avoids the necessity of regenerating the resins which produces considerable amounts of radioactive liquid wastes'which ultimately reach the RHUTs. Under this new program, resins-would be dewatered and then solidified. Any liquids from this process would be returned to the radwaste system. Then new resins would be added to replace the spent resins. The District has already implemented this program by reasing to regenerate resins. Instead, they are being dewatered and stored, awaiting the arrival,and installation of solidification equipment. It is anti-cipated that this equipment will be on site and operable by October 15, 1984 This measure shoulde , nsure that all liquid releases will be maintained low enough to satisfy Technical Specification 3.17.2. ,
Status: Disposable resins - Comolete.
Status: Solidification - Projected implementation by October 15, 1984 Action VI: The District is utili:ing, on an interim basis, sortable deminerali:ers
~itn temporary piping connected to the two RHUTs to allow cleanuo of the RHUTs
.d to provide a means to transfer from one RHUT to the other. The use of onese demins may be terminated when the remaining near term corrective acticns are' completed, although the cross-c:nnecting piping may :e retainec.
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Status: Implemented.
ction VII: The District has initiated a policy that all releases will be - '
<::ntrolleo such that Technical Specification 3.17.2 limits will .not be exceeded.
All sampling of the RHUTs The Chemistry and releases and Radiation of liquids Protection will be based personnel on }forhis objec-responsible i tive.
evaluating the releases have been instructed concerning these objectives. .
This Action, coupled with Action IX will provide a second level of control beyond the other near term actions specified herein. ,
Staths: Implemented. .
Action VIII: The District has implemented a program that permits the PDS .
or RHUTs to be processed back to the Radweste System to provide a reduction 7' in the amount of activity released. This option can be exercised at the option of the Chemistry and Radiation Protection personnel. This method, coupled with dilution of the liquids, will help ensure that any licuid ~
wastes released meet the objectives of this program.
Status: Implemen'ted.
Action IX;~ The District in the past used the Offsite Dese Calculation Manual (ODCM) to control its releases. This manual provides calculational tech- .
nicues to evaluate the offsite dose based on the sampled activity of a release. To provide greater assurance that limits are not exceeded, the District has implemented a policy of using the NRC verified computer codes '
'
- hat are normally used to generate the Annual and Semi-Annual Effluent. .
.elease Reports to calculate a " running total" of the dose from releases.
Data will be processed as releases are made to ensure that limits are met.
Status: Implemented (see Radiological Exposure Calcu2ations section).
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t.ono Term Cerrective Actions ,
e As part of its long term control measures for the disposal of excess regenerant waste water, the District intends to design an.d construct two 8 acre evaporat. ion ponds at Rancho Seco. Figure 1 shows the proposed location of these evapora-tien ponds.
The District will release only the polisher demineralized regenerants to the pond. It is anticipated that no significant residues will originate from the polisher regenerants while salt concentrates will not reach levels where
.they will precipitate.
The existino Rancho Seco technical specifications on failed fuel and steam generator leakage will limit the radioactivity released to the ponds. Hence,
. pond activities will never reach levels where control will be required.
i Although the District has not fully developed the specific design criteria, the District will draw upon the experience of other utilities who use evapora-tion ponds. In any case, the evaporation ponds will contain the following design features:" l
. The District will censtruct two ponds for maintenance purposes.
The ponds will incorporate a leak trace system.
~
To minimize seepage and to resist problems by rodents and plants, the*
ponds will be fully lined.
~
Structurally, the ponds will have seismically designed raised embank-ments of sufficient height to account for wind and wave action.
Design consideration will include provisions s'uch that surface drainage
- of all magnitudes, up to and including the probable maximum flood will bypass the ponds.
The ponds will be of sufficient depth to contain all residues and dust loadings over a 25-year life without the need for disposal. .
The design will enable residues collected on the sides to be washed back into the pond.
The District will completely enclose the ponds by a 6' high chain link fence with 3-strand barted wire as a protective measure. The perimeter of the ponds will also be adequately marked with "NO TRESPASSING" signs.
The final design will meet all local, state, and federal agency guice-lines and standards, such as:
I Water Quality Control Board I Division of Dam Safety I EPA, etc.
EXi B T 2 l
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3 */ *k, UNITED STATES 28 i .n NUCLEAR REGULATORY COMMISSION 1 wAsmNGTON, D. C. 20555 E- 10 f.y D h.e l * .,d ,
November 15, 1984 G :Ci yJ..:
.[. Docket No. 50-312 0 f , ,
- y Mr ' Ronald ~J. Rodriguez E.
A Ex$cuthe~ Director, Nuclear * ,
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Sacramento Municipal Utility DJs rict
' 6201 5 Street j l I DY. ".
I P. O. Box 15830' V J
- Sacramento, California 95813 .
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Dear. Mr. Rodriguez:
i SUB.1ECT: ' RANCHO SECO NUCLEAR GEN RATING STATION - CALCULATED DOSES'
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.IN EXCESS OF 10 CFR 20.4 5(c) AND 40 CFR 190
. ' By letter" dated September 27, 1984, you submitted Special Report Number 84-07
. as required by the Rancho Seco Tech ical Specifications Sections 3.17.2 and-3.2.5. In your letter, you acknowl ged that the calculated radiological exposure to the " Maximum Individual" exceeds the limits of 10 CFR 20.405(c) and 40 CFR 190 during the' calendar y ar 1984 through 8/31/84. However, you noted that you have' taken near term c rrective action to cDrrect the '
release condition which resulted.in t e violation. .You further stated that
- - c ' you. fully expect that the.near term co rective actions.now in effect will result in.~ calculated radiological expo res, from future liquid effluents' to
~ the " Maximum Individual" to be less tha the limits of Technical Specification "
^
' Section 3.17.2 for the fourth quarter o 1984 avid for all subsequent ;
reporting' periods. Based on the_near te corrective actions you.have taken j:.
and the long tern corrective actions you lan.to take, you have concluded.y.. , s f.
that a variance in accordance with 40 CFR 90.11 is not. required at this % :h:,. ' ).
.- time. As part of your near. term actions t reduce. radioactive liquid W 4.:
. I'- releases,iyou have implemented a program to use disposable resins.for the ' " 'YT(
- condensate polishing demineralizers. This oids the necessity of ..
y.
.. regenerating the resins which produces consi erable amounts of, radioactive
- liquid wastes. Your long tem actions will b to design.and construct two. ~
. large evaporative ponds for disposal of the r enerative wastes. . Until,such' -
time as the evaporative ponds have been design , approved, installed and
. made operational, you will continue to utilize he disposable resins.
We have reviewed the actions you have taken and ince you have alread Y -implemented the actions that are expected to redu e the " calculated" radio y '
- logical exposure from liquid effluents to within t e 40 CFR 190 limits, we agree
.'p. ,
that a variance in accordance with 40 CFR 190.11 i not needed at this time.
' k'.) N Si erely, ,
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Gus C. Lainas, Ass tant Director N ,
for Operating Rea ors
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v . Division of Licensing i cc: See next page .
FYdlRIT 3.
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L l-Sacramento Municipal Utility Rancho Seco. Docket No. 50-312 i District e
cc w/ enclosure (s):
Branch ner$1Cens1 doo e a Sacramento Municipal Utility State Department of Health Services Distric. 714 P 5tiset. Office Sufidin . #8 6201 S street Sacramento, California N _
~
P. O. Box 15830 Sacramento, California 95813 Ms. Eleanor Schuertz .'
California State Office Sacramento County 444 North Capitol Street, N.W. Suite 305 >
Board of Supervisors Washington. 0.C. 20001 l4 827 7th Street, Room 424 4 Sacramento, California 95814, Docketi med Service Section .
Office the Secretary Mr. John 8. Martin, Regicnal U.S. Nuclear Regulatory Cammission -
Washington. D. C. 20555 e Administrator U.S. Nuclear Regulatory Comstssion Res17entInspector/RanchoSeco Region V 1450 Maria Lane, Suite 210 c/oU.S.N.R.C. '*-
v 14410 hrin Cities had Walnut Creek, California M596 Herald, CA 95638 5
, . .u . .
[;
, .u Di e r Energy Fact 11ttes Siting Division T Enery Renoirrces Conservation & 4*
Development Commission Regional Radiation F. representative 1516 9th 5 f EPA Reaton IX 215 Fremont Street Sacramento, rata 95814 s
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San Francisco, California .94111 - -
Mr. Robert 8. Borsum Babcock & Wilcox p k lear Power Generation O' vision Suite E20, 7910 Woodmont Avenue -
Bethesda, Maryland 20814 Thomas Baxter Esq.
Shaw, Pittman, Potts & Trowbridge 1800 M Street, N.W.
Washington, D. C. 20036
.. e 21en Hubbard P. O. Box 63 Sunol, California 94586
- _ .~ _ a __ 6) _
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-/fenem%, - UNITED STATES q * 'Qn Q
-[
4 g NUCLEAR REGULATORY COMMISSION T 8 REGION Y
] 1450 MARIA LANE, SUITE 210 WALNUT CREEK, cALIFoRNI A s4596
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JUN 061986 .
Docket No. 50-312 Sacramento Municipal Utility District P. 0. Box 15830 Sacramento, California 95813
~
' Attention: Mr. John E. Ward Assistant General Mansger, Nuclear Gentlemen:
Subject:
NRC Inspection Report .
This refers to the special inspection conducted by Messrs. G. Eamada and G. Yuhas of this office on April 1, 2, 29, and May 15, 1986, of activities authorized by NRC License No. DPR-54, and to the discussions of our findings held by Mr. Yuhas with Messrs. R. Colombo and R. Rodriguez and other asabers
. of your staff at the conclusion of each site visit during the inspection.
This inspection was conducted to evaluate your management of' liquid radioactive affluents during 1985. The inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observations by the inspectors.
Enforcement action related to the enclosed inspection report will be addressed in separate corres-&.dence.
In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure wig be placed in the NRC Public Document Room.
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- Sacramento Municipal Utility District JUN 06.1986 Should you have any questions concerning this inspection, we vill be glad to df.scuss them with you.
Sincerely, ,
f .
Ross A. Scarano, Director
[/ Division of Radiation Safety and Safeguards
Enclosure:
Inspection Report No. 50-312/86-15 cc'w/ enclosure:
L. G. Schwieger, SMUD G. Coward, SMUD State of CA FEMA, Region II e
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U. S. NUCLEAR REGULATORY COMMISSION REGION V )
I Report No. 50-312/86-15 Docket No. 50-312 License No. DPR-54 - ,-
Licensee: ' Sacramento Municipal Utility District P. O. Box 15830 Sacramento, California 95813 Facility Name: Rancho Seco Nuclear Generating Station
~
Inspection at: Clay Station and Sacramento, California Inspection conducted: April 1, 2, 29, May 15 and subsequent
- telephone discussions through May 23, 1986 Inspectors: A h'(A M _ jo/6/85 G.'Hama adiation Laboratory Specialist Dat's Signed 0@Uk-G.P.Y a 6/A N6 -
, Chief, Facilities Radiological Date Signed ction S ction Approved by: .[ )* g/4/f/
.F.'A. Wenslawski, Chief Emergency Preparedness Dat's S'igned
. and Radiological Protection Branch Summary: '
Inspection on April 1, 2, 29, and May 15, 1986 and subsequent telephone discussions through May 23, 1986 (Report No. 50-312/86-15)-
Areas Inspected: Special unannounced inspection by two regionally based NRC specialists to close previously identified Unresolved Item 50-312/84-06-01, and to review the licensee's' management of radioactive materials released in liquid effluents during 1985. The following Inspection Procedures were
~
utilized: 30703, 84523, 84723, 84725, 92700, 92701, and 90713.
Results: Of the three areas inspected, apparent violations involving failure to develop procedures to implement 10 CFR 50, Appendix I criteria and failure to report.the results of radioactivity measured in liquid effluent
'(Paragraph 4); failure to comply with T.S. 3.17.2 liquid affluent dose limits for 1985 (Paragraph 5); failure to perform safety evaluations required by-10 CFR 50.59 and failure to establish, implement and maintain procedures required by T.S. 6.8-(Paragraph 6) were identified. -
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EXHIBIT
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Details
- 1. Persons Contacted A. Licensee Personnel
+R. Rodriguez, Assistant General Manager, Nuclear
- R. Powers, Manager, Nuclear Engineering
- J. McColligan, Assistant Manager, Nuclear Plant
- S. Redeker, Manager, Nuclear Operations-
- F. Kellie,' Radiation Protection Superintendent
!*R. Colombo, Regulatory Compliance Superviser
+*E. Bradley, Supervising Bealth Physicist
- C. Stephenson, Principle Regulatory Compliance Engineer D. Mixa, Cost Analyst
- B. Wilson, Senior Chemistry and Radiation Assistant- (SCRA)
- S. Manofsky, SCRA W. Bampton, Chemistry and Radiation Protection Technician (CRPT)
D. Kaarl, CRPT M. Leiwander, CRPT W. Partridge, CRPT B. Non-Licensee Personnel R. Miller, Acting Chemistry Supervisor, Sierra Technology R. Gardner, Certified Health Physicist, United Energy Services Corp. ..
R-. Oesterling, Certified Health Physicist, United Energy Services Corp.
C. Nuclear Regulatory Commission (NRC)
- +G. Peres, Acting Senior Resident Inspector
- Denotes attendance at exit interview conducted on April-2, 1986.
+ Denotes attendanca at exit interview conducted on April 29, 1986.
- Denotes attendance at exit interview conducted on May 15, 1986.
In addition to the individuals identified above, the inspectors met with contractors and.other members of the licensee's staff.
- 2. Unresolved Item (50-312/84-06-01)
NRC Inspection Report 50-312/84-06, dated May 31, 1984, describes an NRC Region V concern that members of the public may have received a dose from ionizing radiation in excess of the values presented in 10 CFR 50 Appendix 1. Technical Specification objectives and 40 CFR 190 as a result if radioactive materials contained in liquid effluents released from the Kancho Seco Nuclear Generating Station (RSNGS).
EXH B T 9 Page 4 of 30 pages
- 4 2
In a special report dated May 14, 1984, the licensee provided the results of calculations which indicated that these values had been exceeded for 1981 1983, and 1984 for a hypothetical " maximum adult" exposed via the liquid-fish-man pathway. The licensee stated that based on the-concentrations measured in fish flesh and a whole body count of the
" maximum" individual, the actual calculated dose to a real member of the-public was 12 arem and therefore they had not exceeded the 25.arem per year standard of 40 CFR 190.
- Region.V requested the NRC Office of Nuclear Reactor Regulation (NRR) to establish the validity of the licensee's calculations and to determine if the values presented in 40 CFR 190 had'been exceeded.
NRR's evaluation included review of numerous' licensee reports, an extensive environmental survey performed by Oak Ridge National Laboratory (NUREG/CR-4298) and serial measurements of radioactive materials in the vicinity of the'RSNGS performed by EG&G Energy Measurements, Inc. NRR's evaluation was completed in the spring of 1986 and the results -
transmitted by separate correspondence to d e licensee and Region V. ..
, NRR found that during 1984 (the most limiting year) it could not be concluded that the whole body dose to the maximally exposed member of the public as determined from environmental measurements exceeded the 25 arem standard of 40 CFR 190 in view of whole body count data. The calculated dose to this real person based on measured radionuclides concentration,s in fish flesh and recalled ingestion rates was about 50 arem. However, a whole body count performed on the individual failed to detect any radioactivity associated with releases from RSNGS. The whole body count had a minimum detectable activity which would have confirmed a dose of' about 7 mram.
Accordingly, since.it has not been reasonably established that a real member of.the public received a dose in excess of the 40 CFR 190 standard, no violation of 10 CFR 20. 106(g) has been identified. This matter is closed.
- 3. Radioactive Liquid Effluents During 1985 A. Background On July 21, 1984, the licensee implemented Amendment No. 53 to the RSNGS Technical Specifications (T.S.). This amendment incorporated 10 CFR 50 Appendix I, Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low As Is Reasonably Achievable" for RadiU ctive Material in Light-Wate E S oled Nuclear Power Reactor Effluents, requirements into the T.S.
The licensee submitted Special Report No. 84-07, on September 27, 1984 (RJR 84-425) as required by T.S. 3.17.2 and 3.25 to report that the cumulative calculated radiological exposure resulting from liquid effluents exceeded the calendar quarter and calendar year dose limits of T.S. 3.1752 and fuel cycle dose limit of T.S. 3.25 for calendar year 1984 through August 31, 1984. In their September 27, 1984 letter, they stated that, "The District now is limiting its discharges so that 10 CFR 50 Appe will no w Ch __ A_ _ $_Cncm_-__
.t
- 3 be exceeded." In their " Attachment to Special Report 84-07" the I l
licensee presented their near and long term corrective actions. In response to this licensee submittal NRC issued a letter dated l i
November 15, 1984 which concluded, based on the corrective actions taken and planned, that a variance for continued operation pursuant to 40 CFR 190.11 was not needed.
During follow-up inspections conducted in November 1984.a'nd October 1985 Region V inspectors found the licensee was implementing the ;
near term corrective action involving the Polishing Demineralized System and the Regenerant Holdup Tanks (RHUT) (Inspection Report )
No. 50-312/84-27) and that review of the liquid effluent release records confirmed no detectable concentrations of fission or I a'ctivation products were apparently released as described in the j licensee's Semiannual Effluent Radioactive Release Report dated 1 September 26, 1985 (Inspection Report No. 50-312/85-28).
The licensee's contractor Lawrence Livermore National Laboratory (LLNL) collected two vater samples from the Rancho Seco RHUTs on October 14-15, 1985, for isotopic analysis. On November 22, 1985, the licensee's Supervising Health Physicist discussed the adequacy of Rancho Seco's lower limit of detection (LLD) capability in terms of the 10 CFR 50 Appendix I criteria with the NRC Region V, Chief, Facilities Radiological Protection Section. The licensee representative stated that he had initiated a study of the matter and samples had been taken. The Supervising Health Physicist -
advised the Chief. Facilities Radiological Protection Section, on November 26, 1986, that based on verbal results from LLNL; radioisotopes of cesium and cobalt had been detected at concentrations about a factor of two below the onsite Rancho Seco laboratory capability. Since extrapolation of this one data set for
, all liquid releases made during 1985 could call into question compliance with the 3 mram per year total body and 10 mrem per year organ dose commitment of Appendix I and T.S. 3.17.2, other possible sources of data were discussed. The licensee representatives indicated that composite samples of each liquid batch release collected pursuant to T.S. 4.21.1 and sent to another contractor, Controls For Environmental Pollution. Inc. (CEP), could be analyzed for gamma emitting' isotopes with an LLD better than the onsite capability. The licensee pointed out that for cesium and cobalt T.S. Table 4.21.1 requires an LLD of SE-7 uCi/m1, the onsite laboratory reported typical LLDs of SE-8 uCi/m1, CEP ranged from 2E-9 to IE-8 uCi/ml while LLNL reported values of 2E-11 uCi/mi for their LLD.
~
The licensee defines LLD in their Offsite Dose Calculation Manual (ODCM) as:
"The rmallest concentration of radioactivity in a sample which will be detected and reported as a positive value approximately 95% of the time. Conversely, a sample with no real net activity above background will be reported as a positive value about 5% of the time'."
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.In addition ,the licensee states in the 0DCM that:
"LLD is predictive estimate (a priori) representing the
- capability of a measuring system.-not after the fact (a posteriori) estimate of a particular sample. As such, typical values of E. V, Y and T should be used. Stated LLD's
.may not always be achievable due to background fluct,uations, interfering radionuclides or other. conditions affecting the normal measurement process."
This definition is consistent with the standard NRC definition presented in NUREG-0472.
EnDecember5,1985,theSupervisingBealthPhysicistdiscussedthe LLNL results of the October 1985 sampling with Region V. The Cs-134 activity was re' ported- at 8.6 E-9 uCi/ml; Cs-137 at 2.17 E-8 uC1/ml; Co-60 at 1.3 E-9.uci/ml and Mn 54 at 5 E-10 uC1/al. Although these activities were all less than the onsite'LLD, a dose projection would place the 1985 exposure to the " hypothetical maximum .
individual" in close proximity to the T.S. limit if one assumed that activity was' representative of the entire year's releases.
In a telephone discussion between the Supervising Health. Physicist and the Region V Chief, Facilities Radiological Protection Section, it was agreed that the licensee's Semiannual Effluent Release Report for July through December 1985 would address the LLNL water sample results, the LLD issue and the licensee's plans to submit the results of their evaluation.
The Semiannual Effluent Release Report was transmitted by letter (RJR 86-087) dated March 3, 1986, and contained the above o information and a commitment to prepara and submit a special report by August 31 '1986, of the 1985 liquid radiological effluent release source term.
In the course of preparing for a meeting to finalize the NRC response to the 1L84 liquid effluent issues (Paragraph 2 of this report) an NRC Licensing Project Manager became svare that the licensee may have changed their onsite LLD to facilitate the-release of potentially contaminated liquid to the environment.
As a result of this information, Region V contacted individuals within the licensee's organization by telephone on March 21, 27, and 28, 1986. Frem these telephone discussions, Region V was informed of at least one instance when the routine three liter effluent water sample analysed according to the norma 1' procedure of samma counting for 2000 seconds showed identifiable and measurable concentrations of cesium and the technician was directed by a management representative to recount the sample for 1000 seconds. Decreasing the counting time by a factor of two has the effect of reducing the sensitivity considering all other parameters remain constant.
The 1000 second recount did not show any identifiable or measurable concentrations of cesium so the volume of liquid was released to the environment.
EXHIBIT Page 4-of M Pages
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5 Region V was told this matter had been brought to the attention of licensee management during December 1985 in the context of a j
violation of T.S. and that management had concluded no violation had ;
occurred. '
The purpose of this inspection was to address the following issues: i
-Did the licensee change the onsite laboratory LLD'to facilitate the release of potentially contaminated liquid to the environment?
i
- c. -Did the. release of radioactive material in liquid effluent 3 e during 1985 exceed the criteria in T.S. 3.17.27
-Has the licensee's management of liquid radioactive effluent been effective?
- 4. Lower Limit of Detection -
~~
Technical Specification 4.21.1, " Liquid Effluents" Concentration, reads
, in part:
"The radioactivity content of each batch of radioactive liquid waste to be discharged shall be determined prior to release by sampling and analysis in accordance with Table 4.21-1.- The results of ,
pre-release analyses shall be used with the calculational methods in the ODCM to assure that the concentration at the point of release'is limited to the values in Specification 3.17.1.
" Post-release analyses of samples from batch releases shall be performed in accordance with Table 4.21-1. The results of the post-release analyses shall be used with the. calculational methods in the ODCM to assure that the concentrations at the point of release are limited to the values in Specification 3.17.1."
Table 4.21-1, Radioactive Liquid Waste Sampling and Analysis Program, requires that each batch of liquid, waste to be released be sampled and analyzed for various radioisotopes. The minimum required LLD for mixed fission and activation products including Co-58, Co-60, Cs-134 and Cs-137 is stated as 5 E-7 uCi/ml.
The Bases of this specification reads in part:
"This specification is provided to ensure that the concentration of radioactive materials released in liquid waste affluents from the site to areas beyond the site boundary will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures within: (1) the Section II.A Design Objectives of Appendix I, 10 CFR Part 50, to an individual, and (2) the limits of 10 CFR Part.20.106(e) to the population."
E B~ 9 Page E of __
Y Pages
6 The inspector notes that the word "not" has been included in the second sentence.
)
The saae sentence from NUREG-0472, Revision 3, Standard Radiological . !
Effluent Technical Specifications For Pressurized Water Reactors, reads:
"This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design o'bjectives of Appendix I, 10 CFR Part 50, to a HEKBER OF THE PUBLIC and (2) the limits of 10.CFR Part 20.106(e) to the population."
The Regulatory Compliance Supervisor stated to the inspector that the word "not" had been deliberately inserted into the " Bases" during development of this specification because members of the licensee's organization recognized that the LLD values for their site might not be adequate to demonstrate compliance with the design objectives of Appendix I,Section II.A. Section II.A. limits the dose due to liquid affluents to 3 mrem per year to the total body. T.S. 3.17.1, Dose, implementsSection II.A. of Appendix I. --
The. inspector found that several individuals within the licensee's organization were not aware that the LLD values presented in Table 4.21-1 were not intended to provide assurance that the T.S. 3.17.2 dose limits would be met. Specifically, the Radiation Protection Superintendent, Acting Chemistry Supervisor, and two Senior Chemistry and Radiation .
Assistants all stated to the inspector that they believed if their onsite laboratory capability had an LLD of at least the value in Table 4.21-1 and they did not identify measurable radioisotopes in the liquid affluent releases, the dose limits of T.S. 3.17.2 and Appendix I would not be exceeded. All four CRPT interviewed confirmed.that they had been told
, this was the case by their chemistry and radiation protection supervisors.
In a December 16, 1985, siemorandumfromtheSupervisingHealthPhysicist to the Manager, Nuclear Engineering, the Supervising Health Physicist presented the October 1985 LLNL sample respits and described his awareness beginning in January 1985 that the onsite LLD's may not be adequate to assure compliance. The Memorandum described his efforts to evaluate the LLD issue, the lack of management support, his awareness of NRC interest, and proposed six specific actions to be accomplished. The issue of communications is discussed in Paragraph 6 of this report.
10 CFR 50, Appendix I, Section IV.A. reads in part:
"A. If the quantity of radioactive material actually released in effluents to unrestricted areas from a light-water-cooled nuclear power reactor dur'Ing any calendar quarter is such that the resulting radiation exposure, calculated on the same basis.
as the respective design objective exposure, would exceed one-half the design objective annual exposure derived pursuant to Sections II and III, the licensee shall:
"1. Makeanthivestigationtoidentifythecausesforsuch release rates:
EXHIBIT -- 9 Page- 3 -of M Pages
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L 2.
I Define and initiate a program of corrective actions:
and'...."
Paragraph B. adds: *
"The licensee shall establish an appropriate surveillance and monitoring program to:
"1. Provide data on quantities of radioactive material i
released in liquid and gaseous effluents to assure that the provisions of paragraph A of this section are met:"
As of, April 1, 1986, since the Radiation Protection Superintendent apparently believed that the LLD values presented in T.S. Table 4.21-1.
were adequate tesmeet Appendix I, no other appropriate monitoring program had been established to provide data.on quantities of radioactive material released in liquid affluents to assure the dose criteria of Appendix I were met.
Failure to establish appropriate surveillance and monitoring procedures ~'
represents an apparent violation of 10 CFR 50 Appendix I, (50-312/86-15-01).
The licensee's Semiannual Effluents Release Reports dated September 26, 1985, (RJR 85-491) and March 3, 1986. (RJR 86-087) presented in Table 2C the Rancho Seco onsite " Liquid Effluent-Lower Limit of Detection" for' Co-134 as less than 4.82 E-8 uCi/mi and Cs-137 as less than 5.92 E-8 uCi/ml. The licensee representative stated that the LLDs presented in Table 2C were based on a normal three liter liquid effluent sample counted for 2000 seconds using the average background counting rate on the gamma counting system.
The LLDs presented in this table were meant to show that the onsite capability clearly exceeded the values in T.S. Table 4.21-1 for the normal measurement procedure.
Since the licensee's counting system calculates an LLD for each' measurement, the NRC Radiation Laboratory Specialist reviewed the licensee's methodology.
Rancho Seco uses " machine" generated "LLD" values to determine whether or not the LLD limits for specified nuclides are being met. The procedure used for calculating LLD is contained .in the Canberra software associated.
with the Canberra Spectran F gamma spectroscopy system. The document that addresses'the software is Canberra Technical Reference Manual for Spectran F Version 2. This document, however, does not contain sufficient i@mnatf m to determine exactly how the LLD is calculated.
Furthermore s.everal key equations in this document appear to contain typographical errors, and the discussions on LLD and LD (detection limit) seem to indicate that a procedure is being used to calculate LLD which is not consistent with the NRC definition of LLD.
In an attempt to resolve some 6f these issues, a telephone call was made to Mr. Markku Koskel of Canberra on Wednesday April 2, 1986. On the EX- B T 9 Page W of- M Pages
8 basis of this discussion with Mr. Koskel, it appeared that appropriate LLDs were being generated. On the other hand, because a simple test involving a manual LLD calculation using raw spectral data would readily resolve the issue, it was requested that Rancho Seco perform manual LLD calculations and compare these to the software generated LLD values.
This was done, and good agreement was obtained for the energy range tested. It can be concluded, therefore, that the software gen,arated LLDs are consistent with the NRC definition of this term.
- l Accordingly, by having established confidence that the LLDs presented in computer printouts for liquid effluent analysis were credible, the (
inspectors reviewed records of liquid effluent releases made during June 1985 and the Chemical / Radiation Log for the first calendar half of 1985.
On March 20, 1985, an entry at 1730 in the Chemical / Radiation Log reads:
"Name deleted and name deleted concurred that we should count ARHUT for release for 1500 sec.to preclude obtaining a Cs peak which could prevent the RHUT's release to the basin. 1500 see would meet CE and LLD on Canberra." Enclosure 4.1 " Rancho Seco Radioactive Liquid Waste Release ~' '
Permit Regenerant Holdup Tank to Retention Basin" No. 85-76 indicates the "A" Regenerant Holdup Tank (RHUT) containing 85950 gallons of liquid was released to the basin for discharge to the environment.- The permit contains the comment "No Peaks" and lists the gross beta activity as 9.15 E-8 uC1/ml and E-3 as 1.42 E-5 uC1/ml.
Based on standard practice, all REUT liquid release samples were normally counted for 2000 seconds at this time; however, the licensee only had a record associated with a 1501 second count at 1720 on March 20, 1985, of a three liter sample from A RHUT. This printout did not show any gamma isotopes greater than LLD. Cs-134 had an LLD listed for this analysis of less than 8.23 E-8 and Cs-137 of less than 1.17 E-7 uCi/al.
'The Chemical / Radiation Log dated June 4, 1985, contains the following entry:
"85-98 3 REUT -
Scan Cs-137 2.33 E-7 2 4.91 E-8 E-3 4 4.26 E-6 T Scan Repeat Os-137 2.59 E-7 i Scan Repeat 1000 See No Peaks i 1638 B RHUT 85-98 E-3 4 4.34 E-6 Gross On 2.93'E-7 2 4.04 E-8
$ Scan No Peaks (minor, but some peaks wit'h LLD's )P 5 E-7)"
Enclosure 4.1 for. release 85-98 indicates 150,767 gallons were transferred from the B REUT to the Retention Basin for release. The form reads, "No i peaks" for mixed fission and activation products and E-3 less than 4.26 E+6 uCi/ml. No entry is made regarding the gross beta activity.
EXHBIT 9 -
Page_ lI of Pages
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l4 9
The following table summarizes the results of gamma scans performed on three liter samples from the B REUT tank on June 4, 1985.
Sample Time (PST) Counting Time Result (uC1/ml) -
1020 2000 sec. Cs-134 4 1.2 E-7 Cs-137 2.72 E-7 1313 1000 sec. Cs-134 dC 1.25 E-7 Cs-137 < 1.67 E-7 1347 2000 sec. Co-134 4 7.79 E-8
,. Cs-137 2.33 E-7 1701 1000 sec. Cs-134 4 1.51 E-7 Cs-137 4 1.88 E-7 In discussions with several chemistry and radiation technicians, the, inspector was. informed that when a 2000 second count showed identifiable peaks, the matter was brought to the SCRA's attention. The SCRA told --
them to recount the sample for 1000 seconds. If the 1000 second count
- did not show the presence of identifiable peaks and the LLD for this measurement as indicated on the printout was less than the 5 E-7 uC1/mi value in T.S. 4.21-1 then the releace could be made. Some technicians stated that they did not believe this was the correct action, however, they did what they were told. Several individuals stated the motivation for the change in counting time stemmed from the excessive inventory of plant water and the licensee's public statement that they would not release any additional liquid radioactive affluents.
The Radiation Protection Superintendent and the SCRA told the inspector that their concern was the need to release water and that they believed that as long as the LLD for a given sample analysis was less than the T.S. number they would not exceed the Appendix I dose objectives. The Radiation Protection Superintendent stated that he did not become aware that the T.S. number (5 E-7 uCi/ml) might not be adequate to assure compliance with Appendix I until he received a copy of the Supervising Health Physicist " Draft LLD Study" on October 29, 1985.
The technicians recalled that the practice of recounts occurred on other occasions.
The table below summarizes other examples noted in June 1985:
Release Initial Initial Final Final.
Counting Cs-137 Counting Cs-137 Date No. Tank Time Activity Time Activity 6/6/85 85-99 ARHUT 2000 sec. 2.11E-7 1000 sec. 41.53E-7 uCi/mi 6/16/85 85-109 ARHUT No record 1700 sec. 41.03E-7 uCi/mi 6/17/85 85-110 BREUT 2000 sec. 1.2E-7 1700 sec. 4 9. 29E-8 uC1/ml Licensee procedure AP.306 V-13, " Lover Limit of Detection Count Time Determination," issued June 26, 1984, captures the essence of NRC's 9
1 Exmar ,
h A@ rA 3U lP/ARfE3 !
- O 10
~
definition of LLD as presented in NUREG-0472, Revision 3 Table 4.11-1.
The procedure is designed to calculate the optimum counting time to meet the LLD minimum requirement specified in T.S. Table 4.21-1. The procedure does not indicate that the author realized that a far more -
sensitive LLD may be necessary to meet the dose limits of T.S. 3.17.2.
From discussions with the SCRA and review of data, it appears the first time AP.306 V-13 was fully implemented for detector 1 of the,Caaberra system was on July 30, 1985. At that time, a three liter liquid background sample was counted for 1000 seconds, five times. This test demonstrated that the LLD for Cs-134 was SE-8 uC1/mi and Cs-137 was 6E-8 uCi/ml. As a result of this test the licensee posted Enclosure 7.3,
" Liquid and Gaseous Effluent Release Recommended LLD Counting Time," on the Hot Laboratory bulletin board recommending a 1000 second count time for three liter effluent samples on Canberra detector 1.
Because technicians continued to believe a 2000 second count was more appropriate, during the remainder of 1985, of the 111 samples analyzed.
69 were counted for 2000 seconds.
During the initial phase of the inspection the licensee was unable to locate the Canberra printouts for several of the initial 2000 second counts.
T.S. 6.10.2 reads: "The following records shall be retained for the duration of the Facility Operating License:... .
"c. Records of gaseous and liquid radioactive material released to the environs."
During the subsequent inspection visits, the licensee was able to locate
, records except in two instances:
Release Permit No. Date Tank 85-203 10/29/85 ARHUT 85-213 11/13/85 ARHUT Failure to maintain records of liquid radioactive material released represents an example of failure to comply with T.S. 6.10.2 (50-312/86-15-02).
Technical Specification 4.21.1 reads in part:
"The radioactivity content of each batch of radioactive liquid waste to be discharged shall be determined prior to release by
- sampling and analysis in accordance with Table 4.21-1:
" Post-release analyses of samples from batch releases shall be performed in accordance with Table 4.21-1..." .
EXiB T Page N DI - Pages
11 T.S. Table 4.21-1 footnote c. reads:
"Other peaks which are measurable and identifiable, together with-the listed nuclides, shall also be identified and reported.
Nuclides which are below the LLD for the analysis should not be reported as being present at the LLD level.
T.S. 6.9.2.3, Semiannual Radioactive Effluent Release Report,' reads in part 6.9.2.3.1:
"The radio.ctive effluent release reports shall include a summary of the quantities of radioactive liquid and gaseous affluents and solid waste released from the unit as outlined in Regulatory Guide 1.21,
'Heasuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants,' with data summarized on a quarterly basis, following the format of Appendix B thereof."
Regulatory Guide 1.21 reads in Paragraph B.2:
"In many cases the criteria for sensitivity of effluent measurements have been modified to refleet as low as practicable dose considerations in the offsite environs; i.e., the sensitivity of effluent measurements should be sufficient to detect concentrations which, when dispersed in the offsite environs, would result in a dose to individuals of a small fraction of natural background radiation."
and Paragraph C.10 "The sensitivity limits given for radioactivity Analyses in Appendix A of this guide are based on the potential significance in l the environment of the quantities of radioactive materials released.
For some radionuclides, lower detection limits than those given herein may be readily achievable and when measurements below the stated sensitivity limits are attained, the results should be recorded and reported."
The licensee's Semiannual Radioactive Effluent Release Report for the first calendar half of 1985, dated September 26, 1985, stated in Section B. Paragraph C.
" LIQUID EFFLUENTS "As a result of steam generator tube failures in May 1981 November 1982. Septeuber 1983, and July, August, and September 1984, a significant . quantity of radioactive primary fluid has been '
circulated through the steam generation cycle. After the September 1983 occurrence, a small leakage path appeared to remain, on the order of 0.07 gym, which could not be located even after extensive investigating and testing.
" Residual gamma emitters from the secondary system have not been released in the waste water stream during thi ac : od. g so IM ua __ WLee
f
.e ,
-12 "There were 199 batch releases from Regenerant Holdup Tanks with the material ultimately released by 46 Retention Basin discharges.
Liquid releases are summarized in Table 2A and the isotopic contents I are detailed in Table 2B." -
Table 2A contained the fo13owing statement for fission and activation products:
"N/A - no releases containing detectable fission or activation products were made.in the period of January through June 1985" On June 6, 1985, the Radiation Protection Superintendent called the NRC Region V Chief, Facilities Radiological Protection Section, to discuss footnote c. The licensee documented the call as follows:
" Reason for Call:
Resolve meaning or interpretation of second sentence Table Notation 'C' Table 4.21-1 page 4-71. .
l'
" Resolution Reached: ..
If a nuclide is below minimally required LLD (SE-7 uCi/cc) but is a positive value it must be recorded and reported."
Failure to report positive results for Cs-137 activity which was identified and measured on June 4, 6 and 17, 1985, in liquid effluent releases 85-98, 85-99 and 85-110 in the Semiannual Radioactive Effluent Release Report dated September 26, 1985, is considered an apparent violation of T.S. 4.21 (50-312/86-15-03).
l T.S. Table 4.21-1 requires that a monthly composite be collected from each Batch Waste Release Tank for quarterly analysis of Sr-89 and Sr-90.
From January to November 1985 the licensee interpreted the requirement such that samples were only taken from releases which showed gross beta, samma or tritium activity in excess of their respective LLDs. This interpretation is considered to be inconsistent with the NRC intrepretation of the T.S. (50-312/86-15-04).
In November 1985, after distribution of the " Draft LLD Study," composite samples were collected from all batches of liquid released from the REUTs to the basins.
As previously discussed in Paragraph 3 of this report, the licensee representative was aware on November 26, 1985, that the composite samples could be analyzed for gamma emitting isotopes to aid in better determining the 1985 liquid radioactive affluent release source term.
A December 31, 1985 Memorandum from the Manager, Nuclear Engineering, to the Radiation Protection S' superintendent requested that the composite samples be analyzed by their contractor for radioisotopes of cesium.
On April 1, 1986, the inspector inquired as to the results of the analyses. The Radiation Protection Superintendent stated that they had 1
EXHIBIT 9 pggg IC5 of M Pages
O 13 received the results but they were not prepared to accept the_ data provided by CEP.
The inspector was allowed to review a letter dated February 24, 1986,.
from CEP to the licensee representative. The results transmitted indicated unrealistically high concentrations during the months of ;
i February, March and April; insufficient volumes of liquid to make the ;
measurement for May, August and September; November ~and December samples had not yet arrived; January was below their detection limit; and June, July and October showed measured concentrations well above CEP's LLD for Cs-134 and Cs-137.
The licensee had decided that the high activities observed for February, March'and April were the result of using contaminated glassware. No other explanations were offered regarding the remaining months and no one had initiated dbse calculations to determine compliance with T.S. 3.17.2 as required by T.S. 4.21.2 Doses.
T.S. 4.21.2 reads: '
e -
" Dose Calculations
" Cumulative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calculation Manual (ODCM) at least monthly."
ODCM Section 2.3, compliance With 10 CFR 50 Liquid Radioactive Effluents, reads in part:
"It is necessary to demonstrate compliance with 10 CFR 50 Appendix I only if liquid affluents contain measurable quantities of
, radionuclides. The point of liquid effluent radionuclides quantification is defined as the regenerate holdup tanka. The liquid effluent is to be analyzed in accordance with Technical Specifications 4.21.1."
As of April 1, 1986, the licensee had maintained that since they had not measured Cs-134 and Cs-137 in liquid effluent, they were not required to perform the dose calculations for these isotopes. On April 2, 1986, the inspector requested that the licensee expeditiously resolve their 1985 liquid effluent source term, inform Region V of their conclusions, perform the required dose calculations, and submit the required reports J
if the results indicate the limits of T.S. 3.17.2 had been exceeded.
At the conclusion of Paragraph 3, the following question was presented:
"-Did the licensee change the onsite laboratory LLD to facilitate the release of potentially contaminated liquid to the environment?"
This paragraph do'cuments that the Rancho Seco onsite organization altered the counting times of liquid effluent samples to facilitate the release of liquid as necessary to relieve operational restraints. The individuals involved stated that they believed that as long as the concentration was less than SE-7 uCi/m1, the design objectives of 10 CFR 50AppendixIanddoselimitsofT.S.3.17.2wouldnotbe$xypded. EXm 't Page__ E of 30 Pages
4 14
- 5. Compliance With Liquid Effluent Dose Objectives 10 CFR 50 section 50.36a contains provisions designed to assure that releases of radioactive material-from nuclear power reactors to -
unrestricted areas during normal reactor operations, including expected operational occurrences, are kept as low as practicable. In July 1984, T.S. 3.17.2 became effective. This specification reads: ,
"The dose or dose commitment to a member of the public from radioactive materials in liquid effluents released beyond the site boundary shall be limited:
"p. During any calendar quarter to 1.5 area to the total body and to 5 arem to any organ; and i
"b. During any calendar year to 3 arem to the total body and to 10 arem to any organ."
The action statement requires that:
~
I "With the calculated dose or dose commitment from the release of
~
radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days a Special Report. Tais Report will identify the cause(s) for .'
exceeding the limit.and define the corrective actions to be taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits."
As a result of a licenses commitment on April 2, 1986, to provide their position with respect to the activity reported by CEP from the analysis
. of composite samples, the licenses submitted a letter to Region V on April 17, 1986 (RJR 86-135).
The' April 17, 1986, letter contained three enclosures and five immediate actions to preclude noncompliance with the 10 CFR 50 Appendix I design objectives in 1986. .
{.
Enclosure I - A letter dated April 8, 1986, from CEP to the licensee summarizing the composite sample results for 1985.
Enclosure II - An arithmetic composite of LLDs for 1985.
Enclosure III - A summary of liquid waste released, and total volume released during 1985.
The immediate actions included:
"1) All liquid samples for affluent release are being counted for 2000 seconds; 2) the average plant effluent release rate has been increased to 5000 gym to more closely represent a 'non-dry site' power station; 3) all documentation relating to liquid effluent releases are placed in a ' separate folder which will contain all the ;
paperwork associated with the release (i.e., gamma scans, beta j results, tritium results, chemical data sheets (Enclosure 4.2) and '
(Enclosure 4.1)ofAdr.inistrativeProcedure(AP)3] % C}
Gnm ll_ __ dL___A O Sun
t 15 ,
stated LLD values from the gamma scan, gross beta, or tritium analyses will be written on Enclosure 4.1 of AP 305-13; 5) a change is in progress to clarify compositing requirements and require compositing be performed in the secondary lab with only clean -
glassware to preclude contamination of samples."
Review of Enclosure I indicated that four monthly composites had positive results for cesium isotopes, three months showed no detectable' activity and five months ware not of use due to either not enough sample or contaminated glassware. Based on the licensee's evaluation of the four months of clearly indicated cesium activity, they initiated development of the 30-day Special Report required by T.S. 3.17.2.
The inspector noted that Enclosure I did not contain: a result for that composite classified as a "non-radioactive" release volume during December 1985; results of alpha, Sr-89 and Sr-90; and an explanation of why sample results were not available for May, August, and September.
The licensee responded by a memorandum on April 21, 1986, which indicated the December "non-radioactive" releases contained 2.8E-8 uCi/ml 1 8E-9 uCi/mi of Cs-137. The alpha, Sr-89, and Sr-90 activities were all -
less than LLD, and sample volumes for May, August, and September were not available due to repeat analyses for gross alpha and strontium during those months.
Based on an inoffice review of the potential liquid radioactive release source term and the licensee's September 27, 1984, commitments to reduce liquid affluents, the inspector concluded that an additional site visit would be appropriate to determine the origin of released activity and the potential that T.S. 3.17.2 might have been exceeded.
On April 29, 1986, the inspector returned to the site and corporate
. office. This visit found that since 1983 the licenses has engaged in a water management practice inconsistent with the description in the Final Safety Analysis Report (FSAR).
10 CFR 50, Appendix A, 1.5.51 CRITERION 60 - CONTROL g, RELEASES g RADIOACTIVE MATERIALS y THE ENVIRONMENT reads:
"The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operations, including anticipated operational occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions ~can be expected to impose unusual operation limitations upon the release of such effluents to the environment."
Section 1.5.51 of the FSAR reads:
"The radioactive waste system collects, segregates, processes, and i disposes of radioactive solids, liquids, and gases in such a manner that suitable control is ' provide: over releases in order that ,
EX'r B T 9 Page @ of M Pages i
p .
16 numerical guidelines can be met for as low as practicable as defined in 10 CFR 50, Appendix I."
The updated FSAR submitted July 22, 1982, and subsequent amendments -
through July 1985, provide information in Section II., Radioactive Waste and Radiation Protection ,that: ~
"The radioactive vaste disposal systems provide for the.c'ontrolled handling and disposal of liquid, gaseous, and solid wastes. The systems are designed to ensure that plant personnel and the general public are protected against excessive exposure to radiation from vastes, in accord with limits defined in 10 CFR 20, 10 CFR 50, and 40 CFR 190. i "The systems minimize or preclude discharge of radioactive liquids, gases, and solids of station origin to the surrounding environment.._.__ = _ , _
Liquids are not discharged to the environment during normal operation but are processed and held for reuse or for solidification and shipment offsite by an NRC-licensed contractor."
The licensee has, as a result of limited radioactive water storage capacity, routinely transferred water from the Demineralized Reactor Coolant Storage Tank (T-621) to the REUTs (7 950 A and B) for discharge to the environment.
In response to a request by the inspector, the licensee determined that during 1985 787,500 gallons were transferred from T-621 to the RHUTs and released to the environment. Management issues surrounding this transfer are discussed in Paragraph 6 of this report. ;
The Demineralized Reactor Coolant Storage Tank is a 450,000 gallon,
. quality class 1, seismic, category 1 tank which receives water from the coolant radwaste system. The licensee representative stated that due to chronic steam generator tube leaks and plant operational configurations, water had to be transferred to the RHUTs.
No specific gamma activity analyses were made each time the transfers took place. The licensee was concerned that the tritium concentration could be limiting; therefore, the tant was sampled for tritium concentration 26 times in 1985. The average tritium activity was 2.46 E-2 uC1/ml. The inspector was told that the standard practice was to use a temporary piping system to pump between 10,000 to 40,000 gallons to the REUT as a function of tritium activity. The REUT would then be filled from the normal secondary system sources or the service water system to '
dilute the tritium as necessary to assure that the concentration limits of 10 CFR 20 were not exceeded.
Based on the limited sample data available, the inspector prepared a "best" estimate of activity released during 1985. The inspector used four gamma scan results made available by the licensee to prepare the following source term estimate of cesium activity. It must be noted that other isotopes including iodine-131, antimony 124 and 125, silver 110M, niobium 97, and cobalt-58 were observed in low concentrations in some samples. The inspector selected the highest. cesium isotopic activity EX- BIT 9 Page M of % Pages
. J 17 from either the CEP composite for that month or the T-621 sample as the best estimate. This appears justified since at times additional sources of activity could have been added to the RHUT resulting in the CEP composite calculated activity exceeding the T-621 source term. For e example, in December 1985, the 5000 gallons of water from T-621 transferred to the RHUT for discharge contained 37 uCi of Cs-137, but the activity calculated from composite samples of the 1,560,000 gallons of "non-radioactive" REUT discharges using the measured activity"of 2.8 E-8 uC1/ml amounted to 165 uCi of Cs-137.
The data below summarizes the 1985 "best" estimate cesium source term:
Month CEP Activity Best Estimate (uci)
T-621 Activity Cs-134 Cs-137 January Not measured Not measured February Contaminated samples Not measured March Cs-137 = 713 uCi Cs-137 = 267 uCi , 713 April Contaminated samples No water transferred May No measurement made No water transferred -'
June Cs-134 = 176, = 511 Cs-137 = 171 176 511 July Cs-134 = 302 Cs-137 = 465 No data available 302 465 August No measurement made Cs-134 = 177, Cs-137 = 225 177 225 September No measurement made Cs-134 = 1137,Cs-137 - 1443 1137 1443 October Cs-137 - 153 No data available 153 November Not measured No data available -
i December Cs-137 = 165 Cs-134 = 29, Cs-137 = 37 29 165 TOIAL 1821 3678
. Note -The cesium estimate only includes releases made during the six months of 1985 for which there were sample data that indicated activity greater than the cesium LLD.
-Large volumes of water released from the RHUTs classified as "non-radioactive" from January through October 1985 were not composited for gamma isotopic analysis.
I
-Other gamma isotopes were not considered to simplify the presentation.
In order to establish the credibility of this estimate, the inspector compared the 1985 tritium activity released as reported by the licensee in their Semiannual Radioactive Effluent Release Reports to the tritium released from T-621 to the REUT for discharge using the average activity from the 26 tritium samples.
The licensee had reported 89.86 curies had been released. The inspector calculated 73.23 curies originated from T-621. From this, the inspector I
concluded that the "best" estimated cesium activity most probably underestimates the actual release source tern.
t EX -l B 9
-nsm-
18 Technical Specification 4.21-2, Doses, reads:
" Dose Calculations
" Cumulative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calculation Manual (ODCM) at least monthly."
The Bases reads in part:
"The Dose Calculations Methodology in the ODCM implements the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures-b'ased on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated."
From Revision 3, effective September 23, 1985, of the licensee's ODCM, the calculated dose from liquid effluents released dt. ring 1985 using the nonconservative best estimate of activity is 3.89 arts to the total body ..
of the hypothetical maximally exposed member of the public.
At the conclusion of Paragraph 3, the following question was presented:
"-Did the: release of radioactive material in liquid effluent during 1985 exceed the criteria in T.S. 3.17.27" ,
This paragraph documented that radioactive material in liquid effluents exceeded the criteria in T.S. 3.17.2.
Although this appears to represent an apparent violation of T.S. 3.17.2 b (50-312/86-15-05), it is reasonable to expect that no real member of the public actually received a dose greater than this value as a result of i the liquid releases made during 1985.
In the course of developing the source term, the inspector found several additional deficiencies including erroneous data in the licensee's Semiannual Radioactive Effluent Release Reports, failure to complete the land use census required by T.S. 4.27, failure to revise the ODCM consistent with T.S. 6.16, failure to follow procedures required by T.S. 6.8, failure to perform safety evaluations required by 10 CFR 50.59, and failure to update the Final Safety Analysis Report as required by 10 CFR 50.71(e). These findings involve management issues which are described in the next paragraph.
(
- 6. Management Issues A. Changes ,
As previous 1Iy noted in the introduction to. Paragraph 5 of this )
report, the action statement associated with exceeding the dose limit of T.S. 3.17.2 recognizes the limited safety significance of the Appendix I dose values and requires a special report that: q 9
EXFIBIT Pages
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A 19 3
"will identify the cause(s) for exceeding the limit and define the corrective actions to be taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above -
limits."
- The licensee's Special Report No.' 84-07 submitted in response to this requirement on Septemoer 27, 1984, described the cause for
~
I exceeding the dose limits for 1984 as a small but continuous leak in the "B" Once Through Steam Generator. The report reads:
"The path that the radioactive material takes to get from the-
, secondary. system to the generci public is as follows.
~
Backflush water, regenerant waste and flush water from the polishing domineralizers flow to the Polishing Domineralizer Sump CPDS). Also, leakage from the secondary system generally flows to-the Condensate Pit Sump where it is transferred either to the PDS or directly to one of the Regenerant Hold Up Tanks.
(RHUTs). The PDS is typically pumped to one of the RHUTs, which, when full, are agitated, neutralized, and sampled. The -
results of the sample are used to determine the total release activity for each isotope. Based on this data, a Liquid Waste Release Permit is generated, then the tank is pumped to one of the Retention' Basins. When the Basin is full, it is recirculated and sampled to determine a dilution rate that would conform to. appropriate limits.- Samples are also taken during the discharge to provide assurance that regulations are met. Previously. 10 CFR 20, Appendix B requirements were applied as limits as the discharge left the site. The District now is limiting its discharges so that 10 CFR 50 Appendix I limits will not be exceeded."
Based on this inspection, it appears that the cause and pathway were not entirely correct. Specifically, beginning in 1983, the licensee initiated a procedure which allowed the frequent transfer of water recovered from the liquid radioactive waste treatment systems to the REUTs for release to the environment.
10 CFR Part 50.59(a)(1) reads:
"The holder of a license authorizing operation of a production or utilization facility may (1) make changes in the facility as described in the safety analysis report, (ii) make changes in the procedures as described in the safety analysis report, and (iii) conduct' tests or experiments not described in'the safety analysis report, without prior Commission approval, unless the proposed change, test or experiment involves a change in the technical specifications incorporated in the license or en unreviewed safety question." -
The updated FSAR submitted July 22, 1982,.and subsequent amendments through July 1985, provide the following it. formation in Section 11.,
Radioactive Waste and Radiation Protection, that:
EX J ,B~ 9 Page N of- 30 Pages
L l
o ,
20 "The liquid waste systems are designed to permit plant operation without discharging radioactive liquids to the environment under normal operating conditions. The boric acid concentrator and miscellaneous waste evaporator can each -
process vaste liquids in excess of the maximum expected waste i
l generation rates. The coolant waste receiver and holdup tanks
) are sized to store one reactor coolant system volume of vaste during an evaporator outage or during maintenance..'
"The coolant waste system is a closed loop water system with the recovered water and boric acid stored onsite for reuse.
< "The miscellaneous liquid radwate system, through the use of the miscellaneous water holdup tank and the shipment of concentrated wastes offsite by an NRC-licensed contractor, allows normal operation without requiring the discharge of liquids from the system. The entire liquid waste processing system is contained within the Aur.111ary Building. Therefore, any leaks will be retained within the building, collected in the sumps, and reprocessed through the miscellaneous liquids radwaste system.
"All vents, drains, and secondary flow paths in the liquidi radwaste system are shown in Figures II.1-4 and 11.1-5. Thef system is designed so that no liquid radwaste will be released to the environment." -
Section 11.1.2.2.2., Miscellaneous Liquid Radwaste System, reads:
"In addition, spent regenerant vastes from the polishing demineralizers can be processed if they contain radioactivity as the result of operation with a small steam generator tube leak."
Based on review of licensee records, it appears that on December 7, 1982, a temporary change to Procedure A.29, " Waste Water Disposal System," was implemented which allowed radioactive water to be pumped from the Demineralized' Reactor Coolant Storage Tank (T-621) through a temporary conduit to either Regeneration Hold-up Tank (T-950 A or B) for ultimate release to the environment and the Principle Regulatory Compliance Engineer was unable to provide any indication that an evaluation had been performed to determine if a change in the T.S. was required or if an unreviewed safety question was involved. The temporary change expired on January 7, 1983, and was reestablished on February 8,1983, and then expired on March 30, 1983.
Again, on January 6,1986, a temporary change to Procedure A.10,
" Demineralized Reactor Coolant Storage System," was implemented which allowed radioactive water to be pumped from T-621 through a plastic pipe to either T-950 A or B for ultimate release offsite and the Principle Regulatory Compliance Engineer was unable to provide any indication that an evaluation had been performed to determine if a change to T.S. was required or an unreviewed safety question was involved. "
EX , B "7
- m a so m
L 21 From January 1983 through March 13, 1986, the licensee routinely transferred liquid through various conduits including firehose and plastic pipe from T-621 located within the tank farm to either T-950 A or B which are located in an uncontrolled area such that failure of the temporary conduit might have resulted in an uncontrolled release of radioactive material to the surface waters.
On May 15, 1986, the inspector physically observed that.the plastic pipe which had been connected to T-621 drain line had been removed leaving the exposed open pipe in close proximity to the tank. The
, licensee representative stated that the temporary pump which had been installed in the system had a flow rate of 166 gallons per minute.
T.S. 3.17.3, Liquid Holdup Tanks, limits the quantity of radioactive material which can be contained in the REUTs and outside temporary storage tanks to 10 Curies. T.S. 4.21-3 contains the following comment: ,
" Tanks included in this specification are those outdoor tanks --
~
that are not surrounded by liners, dikes, or walls capable of bolding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system."
The connection of a non-quality class temporary piping system, with no automatic isolation capability, to T-621 raised the question as
~
to whether the licensee had performed the weekly surveillance on T-621 to determine that the activity was less than the 10 Curie limit while the temporary system was in operation. The licensee indicated that the surveillance had not been performed. The
. inspector attempted to review the accident analysis for failure of either T-621 or the Borated Water Storage Tank (450,000 gallon) since both are outdoors, not surrounded by liners, dikes, or walls capable of holding their contents, and the licensee does not perform the weekly surveillance. Neither the FSAR nor the licensee presented a safety analysis which would bound these tank failures.
This matter has been brought to the attention of NRR (50-312/86-15-06)4 The installation of a piping system specifically intended to transfer water from the liquid radioactive trcatment system to the RHUTs for release to the environment without first performing 'a safety evaluation is considered an apparent violation of 10 CFR 50.59 @ 0-312/86-15-07).
10 CFR 50.71(e) requires in part that each person licensed to operate a nuclear power reactor shall annually update the final safety analysis report (FSAR) to assure that the information included in the FSAR contains the latest material developed. The update must be submitted to the NRC and shall contain all the changes necessary to reflect information and analyses submitted te the Commission by the licensee or prepared by the licensee pursuant to Commission requirement since the submission of the original FSAR or, as appropriate, the last updated FSAR.
EXHIBIT 9 Pap M Of - "3 D Pags
. +
22 4
Based on discussions with licensee representatives and review of records, including Control Room Logs, it appears that the licensee has discharged liquid radioactive effluents from T-651 to the REUTs for release to the environment from early 1983 through March 13,.
1986 and did not update the FSAR to reflect this information. This represents an example of failure to comply with 10 CFR 50.71(e)
(50-312/86-15-08).
Failure to perform the safety evaluations and update the' FSAR is I considered an example of failure to properly manage changes at the facility.
B. P,rocedures
- 1. T.S. 6.8, " Procedures," reads in part that, " Written procedures shall be established, implemented and maintained covering the activities referenced belows a. The applicable procedures recommended in Appendix 'A' of Regulatory Guide 1.33, November 1972." Regulatory Guide 1.33, November 1972, recommends in G.,
" Procedure for Control of Radioactivity (For Limiting Materials -
Released to Environment and Limiting Personnel Exposure)," that procedures be developed for liquid radioactive vaste systems including discharging of effluents.
Based on discussions with licensee representatives and review of records, it appears that from March 30, 1983, to January.6, 1986, to procedure was maintained which controlled the transfer of radioactively contaminated water from the Demineralized Reactor Coolant Storage Tank (T-621) to the Regenerate Hold-Up Tanks (T-950 A and B) for ultimate release to the environment.
During 1985, about 787,500 gallons were transferred from T-621 to T-950 A and B and released to the environment.
Based on review of the Control Room Logs for March 1986 and document' control records, it. appears that on March 6, 1986, the te=porary change to Procedurs A.10. " Demineralized Reactor Coolant Storage System," which authorized transfer of water from T-621 to T-950 A and B was not maintained in that the procedure expired and a transfer of 6,000 gallons was made to T-950 A on March 10, 1986, and 15,000 gallons were transferred to T-950 B on March 13, 1986.
- 2. T.S. 6.8.3 reads: " Temporary changes to procedures of 6.8.1 above may be made provided:
"a. The intent of the original procedure is not altered.
"b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's Litense on the unit affected.
"c. The change is documented, reviewed by the PRC and approved by the Plant Superintendent within seven (7) days of implementation."
EXhiBT e P:ge M _ot '50 __ p, )
s
.o 23
~
Based on review of the Procedural Change Approval Form and l discussions with the Principle Regulatory Compliance Engineer, I it appears that on January 6, 1986, a temporary change to l Procedure A.10, " Demineralized Reactor Coolant Storage Syste'," m l was approved and implemented which allowed pumping water from {
T-621 to T-950 A and B for offsite release without review by '
the Plant Review Committee (PRC). From January 6, 1986, to March 6, 1986, the licensee estimates that about 350,000 gallons of water were transferred.
In addition, the inspector noted that AP.2 Revision 21, Review, Approval and Maintenance o_f, Procedures, had not been developed consistent with this T.S. in that it does not require temporary changes to be reviewed by the PRC. The Principle Regulatory Compliance Engineer informed the inspector on May 21, 1986, that this issue had been previously addressed by the PRC and that they believed the previous NRC Senior Resident had agreed that the review of non intent changes to procedures could be delegated to a Group Supervisor, reviewed by the PRC Chairman and approved by the Plant Superintendent. The inspector ~
commented that if T.S. 6.8.3.c. were revised, their technique would be considered acceptable. In any case, the inspector considered the revision to A.10 to be an intent change in view l of the FSAR information.
Failure to implement and maintain procedures is considered an apparent violation of T.S. 6.8 (50-312/86-15-09).
The establishment, implementation, and maintenance of procedures is a management function. It is the inspector's conclusion that the proper establishment of these procedures considering the guidance provided in IE Circular No. 80-18:
10 CFR 50.59, " Safety Evaluations for Changes to Radioactive Waste Treatment Systems," could have resulted in recognition of the need to perform a 50.59 review, update the FSAR, and assure proper sampling of T-621 prior to transfer such that compliance with I.S. 3.17.2 could ha,ve been achieved.
C. Quality of Technical Work and Reports
- 1. The licenree's reports involving liquid radioactive effluents have frequently contained inaccurate information and have not been submitted in a timely manner,
- a. Semiannual Radioactive Effluent Release Report, dated September 26, 1985, was required pursuant to T.S. 6.9.2.3 to be submitted within 60 days after July 1, 1985.
The statement in Table 2A that "no releases containing detectable fission or activation products were made in the period of July through June 1985" is incorrect as previously described in Paragraph 3 of I this report.
EXHIBIT 9 Page 2 6 of_ M Pages
._ ___ - . _ - _ _ _ - _ _ - _ _ - _ _ _ _ _ _ _ - m
1 l.
.= ,
24 Table 2C, Liquid Effluents Lower _ Limit of Detection. l may be in error since the licensee changed the ( laboratory capability by altering the sample counting time. -
- Section H., Estimation of Error, presents an inaccurate evaluation of the error associated with the reported releases. Since this same de'at was L presented again in the second half 1985 report, the Radiation Laboratory Specialist performed the following review of the licensee's error analysis for liquid affluents.
Reportedly, for liquid releases, the error analysis includes error contributions due to sampling, volume measurements, and counting statistics. The error. formula Rancho Seeg usesa f fissi n and activation products is 1. ( U" +10)if2 (page 26). The 10 term is apparently the 10 (%) error they list for volume of water. This leaves OP as due to a
~
combined sampling error and counting statistics ters; or either sampling error or counting statistics alone with the other term being zero. This does not calgulate to 2%, the value given on Page 27, even if c~ is zero. ' For low level samples with concentrations near the LLD, the counting error term would normally be 5-10 percent. Sampling error would probably also be in this range. 8 8 8
* ( (7.5) + (7.5) + (10) )1/2 = 14.6 (1)
For 95% confidence interval (21r"), the 14.6% value has to be multiplied by 2. The licensee intends to revalidate their entire error analysis. The licensee's corrective action regarding this matter will be reviewed in a subsequent inspection (50-312/86-15-10).
- The licensee included a copy of Revison 3 of the ODCM. Since Revison 3 had an effective date of September 23, 1985, its adequacy will be addressed with the evaluation of the second half 1985 report.
The inspector observed that T.S. 6.9.2.3.1
. incorrectly refers to T.S. 6.14 in describing what information must be included with revisions of the ODCM. The licenses was encouraged to correct this error (50-312/86-15-11).
- Section I, Table-1, Page 29, items 11 and 12 appear to be in error.
EXHBT 9 Page M l of M Pagas
O 25
- b. Semiannual Radioactive Effluent Release Report, dated
( March 3, 1986, contained the following errors.
"The estimates of radiation dose equivalent to the' non-occupational maximally exposed individuals are one or more orders of magnitude smaller than the limits of 10 CFR Part 50, Appendix I."
- l By virtue of the December 16, 1985, memorandum from
- the Supervising Health Physicist to the Manager Nuclear Engineering; the fact that affluent counting manipulation had been brought to the attention of all
< 1evels of the facility management including the Assistant General Manager Nuclear in December 1985 and January 1986; and that CEP data had been received which raised obvious questions if Appendix I had been met, the inspector considers chie statement to be mishading. -
The report reads on Page 7:
"The only gaseous abnormal release was associated with a reactor transient on December 26, 1985.
Radioactivity was released via primary-to-secondary leakage hence to atmosphere from secondary safety, relief and dump valves." This statement is incorrect. Nearly all of the 32.7 Curies released originated from the make-up pump failure and were discharged via the plant vent stack.
- - Table III-C contains the same LLD values presented in the previous report.
- Table IV-A, Waste Disposal Summary, incorrectly reports the solid radioactive waste data for the period July through December 1985. The Supervising Health Physicists stated the data reported is for the entire year.
- The Estimation of Error is again incorrect.
- Table VI-B, Page 31, is in error. It appears to be a reprint of the data contained in Section I. Table 2 Page 30, of the previous report.
- Section VIII, Page 123 reads:
"No changes were made to the ODCM during this period." This is incorrect. Revision 3 to the ODCM became effective September 23, 1985.
On April 29, 1986, the Assistant General Manager agreed that the reports needed to be corrected (50-312/86-15-12). Page 0% of D Pages
26
- 2. T.S. 6.9.2.2, Annual Radiological Environmental Operating Report, reads in part that.: " Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year."
Based on discussions with the Supervising Realth Physicist on April 29, 1986, the inspector learned that the annual land use census required pursuant to T.S. 4.27 had not been completed due to ongoing litigation with the near site residents. Since the T.S. requires the results to be included in the Annual Radiological Environmental Operating Report, it did not seem likely that the report could be submitted on time. On April 30, 1986, the licensee advised that tha required report would be submitted by May 30, 1986.
- 3. T.S. 6.16, Offsite Dose Calculational Manual (ODCM) reads in 6.16.2: "Any changes to the ODCM shall be made as follows:
"A. Licensee-initiated changes:
"1. Shall be submitted to the Commission by inclusion in the Semiannual Radioactive Effluent Release Report and shall contain:
"a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental
. information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations
, justifying the change.
"b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and..."
Revision 3 of the ODCM effective September 23, 1985, was srpplied with the Semiannual Radioactive Effluent Release Esport dated September 26, 1985. That report and the subsequent report did not contain information to totally l support the rationale for the change. In addition, the change included a revision of the bioaccumulation factor for cesium from 2000 pCi/kg per pCi/1 to 1500 pCi/kg per pCi/1 without a determination that the change will not reduce the accuracy of L the dose determination. Failure to provide the required supporting data represents noncompliance with T.S. 6.16 (50-312/86-15-13).
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pg M ofJO pgu i 1 . t - - - - - _ - - -_ - - - _ _ _ - ---. -------_--------.-- ----- _ _ _ _ _ _ _ _
28 1 l with the dose limits of T.S. 3.17.2. This information apparently was not provided to the 1985 operating chemistry and radiation protection personnel responsible for evaluating the releases. ~ The Supervising Health Physicist was aware of the LLD/ Appendix 1 issue in early 1985 but again this concern was not translated into action. ' Technicians expressed their concern to their supervisors that the adjustment of counting time obscured the presence of
, radioactive material. Again, an opportunity to resolve the
< issue in a more favorable manner was not realized. '
At the conclusion of Paragraph 3, the following question was presented:,'
"-Has the licensee's management of liquid radioactive effluents been effective?"
~
This paragraph documents instances observed by the inspectors which indicate a lack of management effectiveness that appears to have resulted in a failure to operate the facility consistent with the As Low As is Reasonably Achievable (ALARA) criteria during 1985.
- 7. Exit Interview
- The inspector met with the licensee representatives denoted in Paragraph 1 at the conclusion of each site visit. The scope and findings of the inspection were summarized. The licensee representatives were informed of the apparent violations of NRC requirements discussed in this report.
The licensee indicated that the matters would be evaluated and appropriate actions would be taken as indicated in this report. f IXHIB T 9 - page 38 of_.,36 Pages
[- UNUT.D STATES [ [ NUCLEAR REGULATORY COMMISSIDN
! REGON V ,
4 $ 180 MARIA LANE, SulTE 210 p' WALNUT CREEK, CAUFORNIA sm06-535
""* DEC 231987 Docket No. 50-312 EA 86-110 ,
Sacramento Municipal Utility District Rancho Seco Nuclear Generating Station ATTN: G. Carl Aadognini Chief Executive Officer Nuclear 14440 Twin Cities Royd Herald, Californian 95638-9799 Gentlemen: .
SUBJECT:
REQUEST FOR INFORMATION This refers to a special NRC inspection conducted at Rancho Seco by members of the NRC staff during the period of April 1, 1986 through May 15, 1986, of activities authorized by NRC License No. DPR-54. The subject inspection report was transmitted to you by separate correspondence dated June 6, 1986. An enforcement conference related to the inspection findings was held at the Region V office on June 20, 1986, with members of the District staff. Based on the results of this inspection and discussions during the enforcement conference, it appears that certain of your licensed activities were not conducted in full compliance with NRC requirements. ., The findings set forth in the inspection report demonstrate a failure of the District to: (1) implement its technical specification requirements to preclude release of liquid affluents containing radioactivity in amounts exceeding the dose criteria set forth in 10 CFR 50, Appendix I; (2) evaluate changes to the facility as required by 10 CFR 50.59; and (3) maintain and implement procedures and report the release of radioactive material in effluents as prescribed in the facility Technical Specifications. The inspection findings reveal that during 1985, management personnel permitted piping to be connected between the primary and secondary water storage systems creating an unevaluated and unreported pathway for discharge of radioactivity to the environment. In addition, technicians were directed to decrease the counting times for effluent analyses apparently without regard that such an action could preclude detection of radioactivity at levels that could exceed Appendix I dose objectives. These actions occurred in spite of the District's commitment to limit effluents to Appendix I dose objectives in a letter dated September 24, 1984. We have received an.1 are reviewing your July 3,1986 and October 8,1987 letters describing your corrective actions in response to the inspection findings. We also acknowledge-your continuing efforts to improve your ; performance in the area of radioactive effluentsThe as well two as ,many other referenced areas letters, since the plant was shut down two years ago. however, limit'themselvas.largely to corrective actions of k technical and programmatic nature. In' order'to enable the NRC to determine the extent of the enforcement action to be taken, including what action, if any, should be pg m pk e ?n _ - _ _ - _ _ - _ - -__m
i . p *, ""'041' 4 taken with respect to your license, you are to submit within 30 days, under the provisions of 10 CFR 50.54(f), your assessment of the management and personnel considerations surrounding the inspection findings. Your submittal should include your views on, and the basis thereof, whether you have full faith and confidence in the present managemer,t personnel who were, or may have been, involved in these matters such that similar performance is precluded in the future. In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosure will be placed in the NRC's Public Document Room. The responses directed by this le'tter and the enclosed Notice are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511. nc rely, John B. Martin Regional Administrator cc: State of California S S
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