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| * O FFICIAL TRANSCRIPT OF PROCEEDINGS UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
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| ==Title:==
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| 10 CFR PART 70 PUBLIC MEETING d b Case No: s N
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| g O
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| v c5 m
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| Work Order No.: ASB-300-621
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| ] c)
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| LOCATION: Rockville, MD DATE: Wednesday, January 13,1999 PAGES: 1 - 149 f
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| N,Al
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| 't d'
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| ,[' ANN RILEY & ASSOCIATES, LTD.
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| "N r3 1025 Connecticut Avenue,NW, Suite 1014 j (3 Washington, D.C. 20036 (202) 842-0034
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| _J
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| 1 1 UNITED STATES OF AMERICA 2 NUCLEAR REGULATORY COMMISSION 3
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| 4 ***
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| 5 10 CFR PART 70 6 ***
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| 7 PUBLIC MEETING 8 l 9
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| 10 11 12 U.S. Nuclear Regulatory Commission 13 2 White Flint North, Rm. T10-Al 14 Rockville, MD 16 Wednesday, January 13, 1999 17 18 'im_above-entitled meeting commenced, pursuant to 19 notice, at 9:00 a.m.
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| 20 21 22 23 24 25 ANN RILEY & ASSOCIATES, LTD.
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| Court Reporters O' 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
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| 2 1 PROCEEDINGS 2
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| [9:00 a.m.]
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| O(m/ 3 MR. SHERR: Good morning, and welcome. My name is 4 Ted Sherr. I'm Chief of the Regulatory and International 5 Safeguards Branch in Fuel Cycle Safety and Safeguards.,
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| 6 The purpose of today's meeting is again to provide 7 an opportunity to discuss the amendments of 10 CFR Part 70 8 in the interest of making the regulations -- or putting the 9 regulations on a more risk-informed basis, and the specific 10 focus of today's meeting is on the nuclear criticality 11 safety aspects of the rule and also corresponding standard 12 review plan chapters.
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| 13 You should have all gotten the blue folder, and in 14 that folder is the agenda of the meeting, three pieces of f- 15 correspondence relating to comments that have been received
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| '- 16 on nuclear criticality safety issues, and finally, a 17 discussion draft of rule changes that are under '
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| 18 consideration that is staff's attempt to address the 19 comments as we understand them at this point, and that will 20 be further discussed under item 3(b) in the agenda.
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| 21 Before we begin, it may be useful to just go 22 around the room and let everybody introduce themselves and 23 identify the organization that they're with.
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| 24 MR. GEE: Frank Gee, Inspections.
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| 25 MR. EDGAR: Jim Edgar, Siemens Power Corporation.
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| ANN RILEY & ASSOCIATES, LTD.
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| Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 )
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| x
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| E 3
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| 1 MR. VESCOVI: Peter Vescovi, GE Nuclear.
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| .l 2 MR. MANNING: Calvin Manning, Siemens Power C 3 Corporation.
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| 4 MR. WILLIAMS: Don Williams of the Oak Ridge 5 National Laboratory.
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| 6 MR. BADWAN: Faris Badwan, Los Alamos National 7 Lab.
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| 8 MR. MOTLEY: Frank Motley, Los Alamos.
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| 9 MR. FELSHER: Harry Felsher, Licensing Branch.
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| 10 MR. TORSKOSKI: Bill Torskoski, Operations 11 Branch.
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| 12 MR. PIERSON: Bob Pierson, Special Projects 13 Branch.
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| 14 MR. DAVIS: Jack Davis, Special Projects Branch.
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| / 15 MR. GOODWIN: Wilbur Goodwin, Westinghouse, 16 Columbia.
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| 17 MR. BIDINGER: George Bidinger, consultant.
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| 18 MR. FREEMAF: Bob Freeman, ABB Nuclear.
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| 19 MR. SANDERS: Charlie Sanders, Framatome.
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| I 20 MR. HOPPER: Calvin Hopper, Oak Ridge National 21 ' Lab, i
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| : 22. MR. LEWIS: I'm Rob Lewis. I'm in the Special 23 Projects Branch.
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| 24 MR. ROTHLEDER: Burt Rothleder, DOE.
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| 25 MR. CASTLEMAN: Pat Castleman, technical assistant ANN RILEY & ASSOCIATES, LTD.
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| ('-) Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 i l
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| x
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| 4 1 to Commissioner Diaz.
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| 2 MS. WINSBERG: Kathryn Winsberg, NRC, General 3 Counsel's office.
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| 4 MR. DAMON: Dennis Damon,; Fuel Cycle Licensing.
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| 5 MR. SHARKEY: Bill Sharkey, ABB Combustion 6 Engineering.
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| 7 MR. SCHILTHELM: Steve Schilthelm, BWX Technology.
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| 8 MR. VAUGHAN: Charlie Vaughan, GE Nuclear Energy.
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| 9 MR. BRACH: Bill Brach, Division of Fuel Cycle 10 Safety and Safeguards, NRC.
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| 11 MR. PERSINKO: Persinko, NRC, Special Projects 12 Branch.
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| 13 MR. COMFORT: Gary Comfort, NRC Special Projects 14 Branch.
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| 15 MR. ELLIOTT: Mark Elliott, BWX Technologies.
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| 16 MR. KENT: Norman Kent, Westinghouse.
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| 17 MR. KILLAR: Felix Killar, NRC.
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| 18 MR. SHERR: Before we begin, on the agenda, one of 19 our thoughts was that perhaps we would try to get through 20 this morning, before breaking, item 3 (b) , which is the NRC's 21 presentation on the draf t rule language that we put ,
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| I 22 together. )
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| 23 At that time, I think the request is that we break 24 for an hour-and-a-half, which would allow some internal 25 discussions.
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| ANN RILEY & ASSOCIATES, LTD.
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| 5 1 Then we would reconvene this afternoon, and 2 depending on how the time goes, perhaps we would conclude
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| \~- 3 the meeting today af ter item 4 (a) , which is the industry I 4 briefing on the comments on the SRP guidance and then we'd 5 continue our discussions tomorrow and perhaps wrap up 6 tomorrow morning or, if needed, we can go the whole day 7 tomorrow.
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| 8 That's kind of a tentative plan. We'll see how 9 that works out in practice as we go along. We don't have to 10 be fixed on that. Does that seem reasonable?
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| 11 MR. KILLAR: We have one suggestion. In the l 12 morning session hereafter, when we start talking about the i 13 rule language after the NEI presentation, we would like to 14 have some opportunity for the ANS to give a presentation on 15 the letters and information that they submitted prior to the 16 NRC providing their response.
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| 17 MR. SHERR: That's fine.
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| 18 MR. KILLAR: Just to have some background for it.
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| 19 MR. SHERR: Okay. Good. Thanks.
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| 20 I probably don't need to mention this, but the 21 usual restrictions on -- no smoking or other terrible things 22 like that.
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| 23 The rest-rooms are -- the ladies' room is right 24 over here, and the men's room is -- I don't know if you can 25 get through that way or not, but it's on the other side of ANN RILEY & ASSOCIATES, LTD.
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| O Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
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| 6 1 the hallway, and unfortunately, we need to work out some 2 kind of escort system, so we'll figure that out.
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| 3 And Mark Mahoney is recording the meeting for us 4 today, and to help him, anytime anybody's making a 5 statement, if they can use the microphone, and the first 6 time they make a statement, if they can mention their name 7 so he can associate the right name with the right person.
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| 8 At the last meeting, Drew Persinko apparently not i
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| 9 only was the head of the task force but was a spokesman for 10 the industry, as well.
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| 11 If there aren't any other questions or comments at 12 this point, we can proceed with the second item of the 13 agenda, which is just a brief update of where we are and the 14 status of things.
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| 15 I think as you know, we had a public meeting on 16 December 3rd and 4th, and at that meeting, we discussed the 17 performance requirements, the chemical hazards, some 18 discussion of criticality, ISA, standard review plan, and 19 preliminary ISA, 20 We received a number of written comments, NEI 21 letters on chemical hazards dated November 4th, the ISA 22 criticality and SRP issues, and we're going to be focusing 23 today on the Decenber 17th criticality letter as well as the j 24 ANS and Los Alamos letters, and that we still expect to 25 receive written comments on the balance of the rule, as well ANN RILEY & ASSOCIATES, LTD.
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| O Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
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| 7 1 as the balance of SRP issues.
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| I f- 2 Next slide, please.
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| t 3 We've set up a web-site in response to the 4 Commission direction to have public participation through 5 the internet, and this was established and formally 6 announced in the Federal Register on December 24th, and we 1
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| 7 have already placed on the web-site, in addition to 8 transcripts of meetings and the original SECY paper on the 9 staff-proposed rule-making, revised language that took into 10 consideration the comments that were provided on the 11 chemical hazards.
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| 12 As far as schedule goes, as we indicated at the 13 last meeting, we plan to be posting on the web draft rule 14 changes in the December to February timeframe, and we also 15 expect to be receiving comments from NEI and others in the 16 December to March timeframe.
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| 17 We have a meeting today and tomorrow on nuclear 18 criticality safety that we -- to meet the schedule, we are 19 setting a deadline for comments on the rule in the 20 mid-February time-frame and on the SRP and the early March 21 time-frame, and we'll discuss that later in the meeting, 22 more details on that, and finally, the rule package is due 23 to the Commission June 1, 1999.
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| 24 That's just kind of an overview of the status. If 25 there's no other comments before we get into the meat of the ANN RILEY & ASSOCIATES, LTD.
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| O Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
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| r 8
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| 1 m: sting, I guess, Felix, you can take over in terms of the 2 rule language.
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| \ 3 MR. KILLAR: Okay.
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| 4 Just'before we start that, I do want to mention 5 that we have looked at what you put on the web-site and 6 think that.you're moving in the right direction. You know, 7 nothing is always perfect, but it was pretty close, and so, 8 we will have some additional comments on what you provided, 9' but we certainly think that you've taken into account our 10 comments and reflected very well in what you've done to 11 date.
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| 12 We are concerned with the timing, as we continue 13 to indicate, that this is a rather short time period to get 14 all this in, and while we're focusing on the rule and we 15 feel we should be able to get everything done by 16 mid-February on the rule, we feel the SRP is going to be a 17 major challenge, and we don't think there's any way we'll be 18 able to make the.mid-March for the SRP, and we do intend to 19 continue working with you past mid-March on the SRP and
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| . 20 providing what we can to try and get the SRP the best it can 21 within the time constraints we have.
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| 22 We did -- we would like to set up or talk about -- 1 23 and I don't know if it's appropriate to do it now or maybe 24 at the end of the meeting -- another meeting, public 25 meeting, either the end of January or maybe the very first ANN RILEY & ASSOCIATES, LTD.
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| Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
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| 9 1 of February to talk about the next -- the iterations of the 7- g 2 rule itself here, because right now, we've only seen 60.62..
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| d 3 We've since gotten additional letters in, as you have 4 l mentioned, and we've got another one that we're in a final 5 process of review this week, which we hope to get into 6 sometime next week, dealing with the balance of our comments 7 on the rule, and so, what we'd like to do is see about 8 scheduling a meeting after you've had those letters, had 9 some time to consider them and what impact th'ey would have 10 on the rule and how the rule would change.
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| 11 So, I realize that we can't necessarily set 12 something up at this point in time, but we would like to put 13 that on the thoughts for the agenda in the future.
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| 14 With that, I think we will move into the nuclear :
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| 15 criticality section, and Norm Kent from Westinghouse is
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| ('']/
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| \_
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| 16 going to present an overview of the paper that -- the letter 17 we've sent in. So, I'll turn it over to Norm.
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| 4 18 MR. KENT: Thank you very much. I have one 19 view-graph, and I apologize to Jack and everybody on that 20 side. I will read it to you word by word.
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| 21 Again, I'm Norman Kent, and I am a nuclear 22 criticality safety engineer with Westinghouse Electric 23 Company in Columbia, South Carolina, and I am glad to be 24 here today to be able to present an overview of our response I 25 to the proposed re-wording of 10 CFR 70.60 and 62, and I am
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| -s ANN RILEY & ASSOCIATES, LTD.
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| Court Reporters ,
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| 1025 Connecticut Avenue, NW, Suite 1014 !
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| 2 84 - b3
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| .)
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| 10 1 going to address the December 17th letter that the industry 2 submitted to the NRC, and I did not make copies for 3 everyone, but the agenda that was given lists five separate 4 sections that I intend to address one by one.
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| 5 These are risk-informed regulation, double-6 contingency graded level of protection of items relied on 7- for safety, nuclear criticality quality assurance, and 8 historical nuclear criticality data.
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| 9 So, my talk will be brief, I think, but as I go 10 through it, I'll be reading both from my letter and 11 referencing different parts of the proposed' rewrite to the 12 rule.
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| 13 The first item in our letter on proposed changes 14 to the draft language has to do with risk-informed 15 regulation, and rather than read that entire first page, the 16 essence, I believe, was that we saw that proposed revisions 17 continue to address the consequences and likelihood of an 18 accident sequence, whereas they should focus regulatory 19 attention on the risk, and then I would like to jump to the 20 bottom of the page, where we have three bullets listed where 21 we would suggest that -- we had suggested that the NRC give 22 consideration to these three items. So, I would like to deal 23 with those one at a time.
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| 24 The first item was evaluate the risk -- that is, 25 the consequence and likelihood of patential nuclear l
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| I ANN RILEY & ASSOCIATES, LTD.
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| F-11 1 criticality accidents, whether initiated by external events, 2 process deviations, or internal events.
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| 3 .In the proposed rule, part 70.60, paragraph (d), a 4 nuclear criticality event is defined as a high-consequence 5 event,-and it is noted from a reading of the listing in that 6 rule that the other four high-consequence events are really 7 functions of radiation or' chemical exposures to the worker 8 or to the public, rather than just an event.
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| 9 Now, from a purely technical standpoint -- i.e.,
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| 10 the impact on health and safety -- an inadvertent i 11 criticality may not necessarily be a high-consequence event.
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| 12 Now, clearly, criticality is undesirable, and it's 13 the job of the nuclear criticality safety function of the 14 different licensees to ensure that an inadvertent I i
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| : 15. criticality remains highly unlikely, but it seems
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| }
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| 16 inconsistent to include in event -- that is, criticality --
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| 17' in the l'isting with resulting consequences from other events 18 that are not named.
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| 19 What is obvious, as a footnote, is that a 20 criticality would have resulting dose and exposure 21 consequences which are included in items 2.2 through 4 of
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| .22 that list.
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| 23 So, we request that the Commission consider 24 deleting the term " nuclear criticality" from paragraph (d) ,
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| l 25 as a high-consequence event, and I believe, Calvin, you will '
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| 12 1
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| -- or somsons will discuss the ANS letter which deals with 2 that same topic.
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| 3 Also, in the rule -- forgive me for thumbing 4- through pages to find it -- we would request that the 5
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| parenthetical note of the paragraph (d) which says "except 6 for nuclear criticality" would also be deleted from that 7 sentence.
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| 8 The second bullet under risk-informed regulation 9 that.we addressed was to establish appropriate risk base 10 graded levels of protection to prevent nuclear criticality 11 accidents.
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| 12 The intent of this request was to communicate to 13 the Commission that the levels of protection which are 14 applied to different criticality accident sequences should 15 be commensurate with the likelihood of occurrence of that O 16 accident sequence.
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| 17 That follows from the first bullet that not all 18 inadvertent criticalities are high-consequence or equally 19 likely. So, we want to include the likelihood and the '
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| 20 consequence of a criticality accident.
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| 21 So, we would request that the Commission would 22 consider acknowledging that the level of protection against 23 a criticality accident sequence be commensurate with the 24 likelihood of occurrence of that sequence.
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| 25 I counted four areas in 10 CFR 70.60 and .62 where ANN RILEY & ASSOCIATES, LTD.
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| 13 1 we are unsure of the intent of the meaning of these
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| ~
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| 2 references to controls, and I'd like to point them out.
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| O 3 The first one is' paragraph (d), part 70.60, 4 paragraph (d), which says "Each engineered or administrative 5 control necessary to comply with subsection (b) or (c) of 6 this section shall be designated as an item relied on for 7 safety."
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| 8 Paragraph 62 (a) (1) , second sentence, "The safety 9 program may be graded such that management measures applied 10 are commensurate with the item's reduction of the risk.
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| 11 Requirements for the safety program, including process 12 safety information, integrated safety analysis, management 13 measures are described in sections (b) through (d) of this 14 section." I believe that was the right sentence.
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| 15 70.62, paragraph (d), "To ensure that each item O,-
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| 16 relied on for safety will perform its intended function when 17 needed, integrated safety analysis shall be used by 18 licensees to establish safety program management measures.
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| 19 The safety program management measures shall ensure that 20 ..., " and then, subparagraph (1), right below that, where it 21 discusses engineered control, " Engineered controls that are 22 identified as relied on for safety pursuant to section 23 70.60(d) of this part are designed, constructed, inspected, 24 calibrated, tested, and maintained as necessary to ensure 25 the ability to perform their intended functions when needed.
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| 14 1 Items subject to this requirement include but are not 2 limited to principle structures of the plant, passive 3 barriers relied on for safety -- for example, piping, glove 4 boxes, containers, tanks, columns, vessels -- active 5 systems, equipment, and components relied on for safety, 6 sampling and measurement systems used to convey information 7 about the safety of plant operations, instrumentation, 8 control systems used to monitor and control the behavior of 9 systems relied on for safety, and utility service systems 10 relied on for safety."
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| 11 We believe that these four references in other 12 areas where levels of protection are addressed mean that an 13 item relied on for safety can be graded commensurate with 14 its ability to reduce the risk -- that is, the likelihood 15 and consequence of the criticality.
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| 16 Therefore, we request that the Commission would 17 consider revising 70.62, paragraph (d), management measures, 18 to state that the performance but not to include the 19 prescription associated with performance of the controls, 20 and we'll address that later on in this presentation.
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| 21 The third bullet is to establish appropriate risk 22 base -- that is, graded levels of assurance for items relied 23 on for safety to ensure their availability and reliability.
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| 24 There seems to be some overlap, and yet, I could 25 see a subtle difference between bullet number two and bullet i
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| 15 1 number three, and I will try to explain that the intent of
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| . 2 this item was to communicate to the Commission that,
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| \
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| 3 distinct from the levels of protection applied to a system 4 be graded, that an item, a specific item relied on for 5 safety also must have assurance of availability and 6 reliability that's appropriate to the risk that is designed 7 to prevent or mitigate.
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| 8 Now, we believe the words of the proposed revision 9 to part 70.60 and 62 provide the latitude, bu't again, 10 looking at those four references that I read to you earlier, 11 it does not seem clear from the reading that that, in fact, '
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| 12 is the case.
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| 13 So, we would request that the Commission would i
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| 14 consider clarifying that point.
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| 15 As a footnote, from the standard review plan --
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| 16 and I recognize that we're talking about the rule --
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| 17 paragraph 5.4.4.1.1 on page 9 states that the highest t 18_ quality assurance level is provided for all criticality 19 controls used to ensure double contingency.
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| 20 I think that seems to conflict with the need to 21 apply appropriate levels of assurance commensurate with the 22 risk involved.
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| 23 I'm ready to move on to paragraph (b) now, double 24 contingencies. I assume presentation means there are no 25 questions as I go through that, so I'll just continue until ANN RILEY & ASSOCIATES, LTD.
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| r 16 1 som2 body stops me.
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| 2 As the rule reads, we believe that it is f
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| 3 acceptable with respect to double contingency. However, we 4 do note or did note that double contingency is included in 5 10 CFR 70.64 -- I believe it's item number 9, which deals 6 with baseline design criteria.
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| 7 The licensees, the industry agrees with the 8 provisions of ANSI 8.1 as they apply to design criteria, but 9 we do not believe that baseline design criteria for any 10 regulatory discipline should be in the rule, and while we 11 find 70.60 and 70.62 acceptable, we do take issue with the 12 application of the double contingency principle as it's 13 found in the standard review plan.
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| 14 Specifically, the attempt to apply probability >
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| 15 criteria to double-contingency protection is totally 16 inappropriate and not in keeping with ANSI 8.1.
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| 17 Item number (c) or letter (c), graded level of i 18 protection: In reading through the introductory paragraph, 19 we restated what was in the earlier version of paragraph 60, 20 where it states that a nuclear criticality event is a 21 high-consequence event, and so, I am proceeding with that 22 pre-supposition, and this actually overlaps from the second 23 bullet on the first page.
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| 24 Therefore, if an event is high consequence, as 25 earlier stated in the rule, then risk becomes solely a ANN RILEY & ASSOCIATES, LTD.
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| 17 1 function of an accident's. likelihood of occurrence, and O
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| 2 again, we see that the choice of control should depend on
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| \- / 3 the risk or the likelihood, and though individual controls 4 may vary in their level of importance, the aggregate must 5 make a criticality highly unlikely.
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| 6 The second paragraph of the double contingency 7 section of our letter notes that 70.60(c) requires the 8 licensee to ensure that safety controls or items relied on 9 for safety to prevent a nuclear criticality accident are 10 continuously available and reliable.
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| 11 In practice, specific safety controls will not be 12 operational during periods of maintenance or calibration and 13 testing and will not be required to function, and they will 14 r.ot be required to function when SNM is not present.
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| (~g 15 Therefore, the wording of that paragraph should be V 16 modified to address the risk of a nuclear criticality 17 accident and to ensure that items relied on for safety are 18 available and reliable when required to perform their safety 19 functions, and in looking at the re-write based on this 20 letter, we see that those words were incorporated into your 21 re-write of the rule, and we appreciate that.
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| 22 The third paragraph in the section on graded level 23 of protection says -- that same paragraph, 70.60(c),
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| 24 incorrectly identifies only the likelihood of external 25 events as an element of risk from a nuclear criticality ANN RILEY & ASSOCIATES, LTD. l O Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 i
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| l l
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| 18 1 accident, thereby excluding the likelihood of process 2
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| deviations or other internal events as an element of the
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| -- 3 risk evaluation.
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| 4 In actuality, likelihood of a process deviation or 5
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| other internal event initiating an accident sequence leading 6 to a potential nuclear criticality is probably far greater 7 than that posed by an external event, and we suggested that 8 tl language be changed to include provisions for an 9 internal event, and that, too, was incorporated into your 10 re-write, and I'm finished with letter (c), I'm moving on to 11 (d) now, nuclear criticality quality assurance.
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| 12 Your script never looked so good the day after you 13 wrote it.
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| 14 The industry -- we support the concept of 15 employing management measures as introduced in Part 16 70. 62 (d) , but we do find the prescriptive level of detail 17 that's in the rule inappropriate for a rule and therefore 18 unnecessary, and I'm referring specifically to items (1) 19 through (8) of 70.62.
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| 20 And out of mercy to everybody, I will not read all 21 eight of those.
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| 22 We would like to suggest, rather, that the rule 23 concerning management me'asures read something like this, 24 which is, in essence, a modification of the introductory 25 paragraph that you have in subparagraph (d), management ANN RILEY & ASSOCIATES, LTD.
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| 1 19 1 maasures, and it reads something like this.
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| 2 " Management should establish appropriate measures 3 to ensure that all items relied on for safety perform their 4 safety function when needed," period.
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| 5 Again, the intent here, I believe, is that l
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| 6 ' " management measures" is a better way of describing the type 7 of assurance that the licensee needs to apply to the 8 controls or the control systems to ensure that a criticality 9 is highly unlikely.
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| 10 The phrase " quality assurance" connotes an 11 established 18-point QA program. In fact, NQA 1 is 12 referenced in the SRP, and we would rather not see that 13 become the standard QA program of choice.
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| 14 We believe that licensees should be able to 15 establish these management measures and document them, 16 perhaps in the application but I think preferably in a 17 lower-level document, and not necessarily be required to 18 employ all 18 points every time against every item relied on 19 for safety.
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| 20 Part (e) of our letter, historical nuclear 21 criticality data -- my note says to read the entire section.
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| 22 I think I'll not do that.
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| 23 In essence, Part 70 license applications for 24 operating facilities are required by section 70.65(c) to 25 include a description of operational events within the last ANN RILEY & ASSOCIATES, LTD.
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| 1 20 1
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| 10 years that had a significant impact on the safety of the 2 facility.
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| ()
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| 3 Detailed incident reports of nuclear criticality
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| .4 deviations or violations, including corrective safety 5
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| measure that were implemented, are submitted to the NRC at 6
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| the time of the incident and are retained in the licensee's 7 records.
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| 8 These are, among others, bulletin 91-01 9
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| notification, 10 CFR 70.50 notifications, and 10 CFR 70 10 notifications.
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| 11 Therefore, we would request still that the 12 Commission would strike this clause from part (e) -- did I 13 say (c) earlier? -- part (c) that reads that a description 14 of operational events within the past 10 years that had a 15 significant impact on the safety of the facility -- we
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| ( 16 request that that be stricken.
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| 17 As a footnote, this type of information is not 18 appropriate in a license application in that it does not 19 represent safety performance commitments.
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| 20
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| == Conclusions:==
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| As I had said at the beginning --
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| 21 and I do mean it -- I am pleased to be here this morning to 22 discuss this with you on a technical matter, and it's clear 23 to me from reading 70.60 and 70.62 before and reading it 24 after that there seems to be convergence between the 25 Commission and the industry on the rule, and so, I look ANN RILEY & ASSOCIATES, LTD.
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| 21 1 forward'to further discussion to resolve the matters that I mentioned today and hopefully clearly communicated.
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| g-sg 2 l
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| '- 3 MR. KILLAR: That completes the presentation of l 4 the overview of the letter we sent in, as well as some of 5 the items we saw and have in the proposed re-write of the 6 rule in these areas. j i
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| 7 MR. HOPPER: With regard to the ANS letter that l 8 was submitted -- !
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| l 9 MR. SHERR: Give your name. l i
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| 10 MR. HOPPER: All right. The name is Hopper, I i
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| 11 Calvin Hopper, and I am here to discuss the letter that was 12 issued by the ANS chairman of the nuclear criticality safety l
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| 13 division, Cecil Parks.
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| 14 There's been a lot of discussion, I noticed -- I 15 reviewed the interchange that occurred December 3 and 4, and 16 it seems to me that there's been a lot of appropriate 17 observations made at that time, and I think that the issues l 18 that were brought up -- I know the issues that were brought 19 up have been considered and address in this revised rule 20 change.
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| 21 There are some minor things that are disturbing to 22 people who have been in the business for some years, and 23 that is this risk business, how risky is criticality and the 24 consequences of it, and I would like to at least point some 25 of this discussion in that direction.
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| 22 1 As was brought up by Tom McLaughlin in his letter 2 and by the letter from Cecil Parks, the consequences of 3 ~ criticality accidents can be fatal but are typically quite 4 limited, and in that regard, it would seem inappropriate, as 5 Norm has mentioned before, to consider it just arbitrarily a 6 high-risk event without consideration -- or a 7 high-consequence event without consideration of the actual 8 mechanism of the criticality accident.
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| 9 So far as the unlikeliness of it, this is an issue 10 that has bothered the nuclear criticality safety community 11 for ages.
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| 12 As soon as concerns about 10 to the minus 6 or 10 13 to the minus 5 events per year saw the light of day in the 14 criticality community, it's raised some substantial 15 concerns, and if a person tries to get to the source of that 16 number or those numbers of probability, it's sort of 17 interesting, and if you continue to drive the question home 18 to people, where do these numbers and probabilities that are 19 acceptable come from, which I've done, and I've pressed that 20 back to the national reactor licensing board reviewers for 21 licensing of reactors, and it turns out 10 to the minus 6 is 22 where that number sort of saw the first light of day, and if 23 you take a look at what that 10 to the minus 6 is applied 24 to, it's applied to dumping a core into a parking lot, a 25 reactor core, not a criticality accident.
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| 23 1 Ten to the minus 6 is applied to dumping a reactor 2 core -- I don't mean losing secondary containment, I mean f%
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| s_/ 3 losing all containment, whereas 10 to the minus 5 and other 4 lesser frequencies, greater frequencies, are acceptable for 5 things like secondary containment loss, secondary coolant 6 loss.
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| 7 So, I think that it would be real beneficial, and 8 in fact, if you take a look at unlikely and you're talking 9 10 to the minus 5, at least people are thinking in terms of 10 being a little more realistic about this frequency business, 11 and that's anotner issue I did want to touch upon.
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| 12 I think that the concerns that have been expressed 13 by industry have been well-considered and well-articulated 14 by everybody. I don't see that it's necessary to pursue f-m 15 that any further.
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| 16 So far as the SRP goes, we concur that that is 17 going to require substantial work. The issue of what people 18 understand is a double contingency, how it was written in 19 the ANSI 8.1 standard, that has been misinterpreted numerous 20 -- for years.
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| 21 The appendix is a good example. If you examine 22 what 8.1 says you should apply double contingency where 23 appropriate and then they give -- talk about controlling 24 process parameters and then you look at the example that's 25 run in Appendix A, it gives you a good picture of what ANN RILEY & ASSOCIATES, LTD.
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| o 24 1 double contingency has meant.
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| 2 So, if you read the standard, the body of the 3 standard, which is the official portion of the standard, it 4 doesn't give you a clear picture of what was intended by 5 double contingency.
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| 6 Observation of Appendix A and the example is 7
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| clear, and so, a person needs to be very careful when they 8 talk about double contingency, whether that means two 9 controls on a single parameter or if it means independent 10 parameters,.and that's been misinterpreted for years by 11 people, and when I say misinterpreted, I mean both ways, I 12 mean properly interpreted as well as inappropriately 13 interpreted.
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| 14 Those are the three basic things I have particular 15 concern about so far as the rule-making go. J 16 As I said, I think people have articulated the 17 concerns industry has very appropriately, and I have little 18 more to say.
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| 19 Does anyone have any questions?
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| 20 MR. BIDINGER: I might just make two supporting 21 observations.
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| 22 One, at the last meeting I mentioned that one 23 facility in this country used to use a 10 to the minus 2 s 24 number. They have abandoned that approach. That was 25 Savannah River. They finally dug it out and faxed it to me ANN RILEY & ASSOCIATES, LTD.
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| 25 1 this last week. But they have abandoned that approach.
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| 2 The other one is double contingency -- there's
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| (_ 3 very little background on it, but there is -- Hugh Paxon, 4 who is one of the authors, did write in document LA3366, and 5 in there, he says it's very important in applying the double 6 contingency that expert judgement and experience are the 7 only two factors that can be used to properly apply the 8 double contingency principle, that it's not a numerical 9 concept at all, 10 MR. HOPPER: That's an interesting comment that 11 you had, George, because in the first criticality safety )1 1
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| 12 short course that the University of New Mexico put on in 13 Taos, New Mexico, in 1971, Bob Stevenson of the NRC staff 14 had a very interesting discussion -- and that's available in 15 the document, the proceedings of that short course -- Bob 7-s
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| ' 16 Stevenson of the NRC staff had a very interesting and 1
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| 17 enlightening and useful discussion on double contingency and '
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| 18 its application in industry, and that would be a good source 19 for people to consider. There was a substantial discussion 20 on that issue.
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| 21 Thank you.
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| 22 MR. SHERR: Thank you.
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| 23 Okay. Now we'll turn to Gary.
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| 24 As I mentioned in the beginning, in your folder 25 there is a document titled " Discussion Draft Text," which is ANN RILEY & ASSOCIATES, LTD.
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| h 26 1 the NRC staff attempt to be responsive to the comments or 2 take into consideration, at least, the comments that were
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| () 3 provided in the three letters, and Gary Comfort is going to 4 give us an overview of that draft.
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| 5 MR. COMFORT: Although everybody has it in front 6 of them, I'm putting up on the view-graph the changes that 7 we're considering to try to address some of these comments 8 relating to the rule directly.
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| 9 I noted from Mr. Kent's discussion that some of 10 the suggestions were to remove nuclear criticality as a 11 high-consequence event and alsu the parenthetical expression 12 up in 70.60 (b) . We looked at your comments and decided that 13 that would probably be appropriate, also.
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| 14 What the intent originally for that was is that t
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| 15 NRC has a strategic goal, which I expect most of the 16 licensees also have, zero inadvertent criticalities, and the
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| .17 best way that we thought that we could implement it at the 18 time that we were writing this part of the rule was to make 19 criticality a high-consequence event which would force the 20 probability to be highly unlikely for such an event.
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| 21 After looking at it and getting the comments that 22 we got, you know, we also can agree based on fact that a 23 criticality in itself doesn't necessarily have to be a high 24 consequence event. As shown by history, you can have events.
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| 25 Now, part of that, you know, is circumstance, part i
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| 27 1 of it's protection that that has occurred. In general, I s 2 think a unmitigated criticality would still be considered a
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| \l 3 high-consequence event based on, you know, the 100-rem 4 definition.
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| 5 However, again, when you have mitigation or other 6 protections, that may not be the case, and that's what we 7 wanted to do, is keep it consistent with the other 8 industrial accidents and pull it out -- the specific 9 comments out of the definition of high consequence so that 10 people could use it more in common with the way the rule is.
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| 11 However, we still are looking for the idea that 12 criticality should be prevented and not mitigated. We 13 aren't aiming to, you know, allow even mitigated, you know, 14 criticalities behind shield walls where there's not a chance rs 15 because it's just the strategic goal, and there's also, you k
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| 16' know, included perception unique to the nuclear business of 17 a criticality and the concern about that.
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| 18 You know, if you just state the terms of a 19 criticality occurred, people get worried whether there was a 20 consequence or not.
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| 21 In order to take that into account, we included a 22 new sub-paragraph that we inserted (d), which basically says 23 you go ahead and look at -- you know, you implement 70.60(b) 24 and (c) as you would with any other accident.
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| 25 However, the intent of it is again to direct any
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| 28 1
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| reduction in risk primarily through the use of reducing the 2
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| , . , frequency of the accident, and basically, the last sentence i ,) 3 of it, which is prevention of the reaction, shall be the
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| '4 primary means of protection against consequences of nuclear 5
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| criticality accidents, is taken with one slight change in 6
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| the word, you know, including the word " primary" from ANS 7 guidance.
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| 8 We're hoping that this change will be, you know, 9
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| acceptable to the industry and will be addressing their 10 comment on that, that again we're looking at them to 11 evaluate the accident, and the result of the evaluation 12 should put you in the categories above.
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| 13 We also -- going along with the intent of the 14 grading of the protection and assurances, also we hope that 15 n by making this change it will make it more clear that
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| ( 16 criticality is expected to again be addressed, similarly the 17 other industrial accider.ts, and that there would be a graded 18 approach addressed to it.
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| 19 To go further on some of the more specific 20 comments that we had, I think the intent of the comment (c) l 21 I heard, if I'm correct, was felt to be addressed l I
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| 22 appropriately in the last revision, which is basically 23 similar to -- or the last revision posted on the web, which 24 shouldn't have been changed much in this latest revision, in 25 the terms of continuously versus when needed or continuously ANN RILEY & ASSOCIATES, LTD.
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| 29 1 1
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| 1 available versus needed and then also the idea of only l 2 addressing external events, that it sounded to me, based on 1
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| \-/# 3 discussion, that that was considered to be reason -- you 4 know, acceptable at least on the first look by industry to '
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| 5 address that comment.
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| 6 The idea of the graded quality assurance that was 7 indicated, particularly the concern that was in 5.4.4.1.1 8 and the conflict in 5.4.4.1.5, which one basically stated {
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| 9 that all criticality events should be considered -- or all 10 controls should have the highest quality assurance, five 11 basically addressed the idea of a graded approach.
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| 12 We looked at that and we noted internally also 13 that there was a conflict and the intent was really to go 14 towards the graded approach. l 15 So, that type of thing will be revised when we 16 revise the SRP.
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| l 17 The other big issue, of course, was the l 18 implementation of double contingency, particularly in the 19 SRP, the use of probability numbers.
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| 20 We looked back at it and one of the problems that 21 I think when the SRP was written was the idea the double l
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| 22 contingency principle has the term "unlikely" in it, and i 23 that was used, I think, by the author to try to relate back 24 to the definitions that were used for unlikely and highly 25 unlikely in the rule. l I
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| 30 l' I think the way it really needs to be looked at
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| ,, 2 .and I think NRC's opinion is that those are two separate
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| (/'
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| x- 3 definitions for "unlikely."
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| 4 I don't expect right now that we would be 5 implementing double contingency in any probabilistic way, 6 that experience, you know, with some guidance will be the 7 way that we will be looking at it.
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| 8 The overall impact, though, of such a review and 9 the implementation of those controls would be expected to 2 10 . meet the rule language, you know, if it fell under the high l
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| 11 consequence event or of an intermediate consequence event.
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| 12 However, at the same time, for criticality i
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| 13 accidents, you would still be applying double contingency, 14 and "unlikely" in double contingency would not -- you know, 15 something that was defined as unlikely in double contingency Cg 16 may end up being highly unlikely by itself or may end up 17 being less than unlikely in the rule, but there would be i
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| 18 other controls that would have to be applied to put it in 19 the appropriate aspect.
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| 20 But the intent of it again is that they would be 21 handled as two separate conditions that would be evaluated 22 separately and not in a probabilistic way, particularly with 23 a double contingency.
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| 24 Again, that would be going through and trying to-25 show that intent when we go through and revision the rule.
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| 31 1 What we're planning on doing in that revision is to remove 2 the probabilistic content that we had in there and better 3 define what we would expect our license reviewer or l 4 criticality reviewers to be doing in that section.
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| 5 The final area that we had -- and this is the one 6 where we probably had the most disagreement on with industry 7 -- is the inclusion of the historical data that's required j 8 in 70.65(c). I 9 We realize that most of the significant reports 10 will be on file either here at NRC or at the licensee. The I 11 real intent is for that data to be looked at to be used in 12 their development of the integrated safety analysis, and it 13 was stated in the statement of considerations for the rule 14 that was put out with the SECY paper that, under section i g- 15 70.65 on page 27, that " Finally, the license application for I
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| (_/ 16 an operating facility should include a description of i
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| 17 operational events that have occurred during the past 10 18 years and had a significant impact on the safety of the 19 facility. These events should be addressed in the 20 applicant's ISA to ensure that the range of accident ;
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| 21- sequences considered in the ISA encompasses actual events 22 that have occurred at the facility," and that's really what 23 the rationale for having that included in, is to make sure 24 that the licensee does go back when they're developing the 25 idea of, you know, unlikely, highly unlikely, has it really ANN RILEY & ASSOCIATES, LTD.
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| 32 1 occurred back in their facility, and by including that 2 information in the application for the significant events, 3 it would allow the license reviewer to go back and make sure 4 that he agrees that it's been included, and by doing a past 5 check of the other events that we have on site, we wouldn't 6 have to do as detailed to pull out the information, because 7 it would presumably be detailed enough but not very detailed 8 to give us an idea of what the accident was, what was done, 9 and what caused it.
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| 10 We would do a spot-check to make sure we felt that 11 all accidents that had been done in the past were covered in 12 the ISA, that they hadn't been overlooked, and by providing 13 it in the application, though, it would reduce the amount of 14 review that we'd have to go back in trying to re-compile
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| - 15 those accidents ourselves, and so, all the intent is really 16 to do is show -- and it would be expected that that 17 information should be readily available from the licensee, 18- as stated, because it should have been reviewed in the ISA 19 itself.
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| 20 So, right now, our feeling is that we probably 21 won't remove that requirement.
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| 22 There were a couple of other comments that were 23 made that gave more specifications on the language in the 24 new rule that was posted on the web with some specific 25 citations particularly related, you know, again to rating ANN RILEY & ASSOCIATES, LTD.
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| 33 1 the levels of protection and clarifying the use of those 2 terms that -- you know, that it looked like -- I believe the 3
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| comment indicated that it looked like the information was 4 there towards the way industry would like but they just 5 wanted a little bit more clarification to just -- you know, 6 written into the language that would make it more specific, 7 and we haven't looked at the rule or had those comments 8 before, so we'll basically re-look at those sections and, 9 you know, post whatever appropriate revisions that we 10 foresee on the web.
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| 11 The web is hopefully going to be one of our 12 primary means to communicate both back and forth in a very 13 public forum.
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| 14 On the list that we have over there, we've got a 15 space for e-mail addresses, and our intent is, when we do 16 post changes onto the web, that you will be e-mailed with l
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| 17 such information that you can go look at the specific 18 changes and then make comments.
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| 19 I don't think it's been exactly determined how 20 we're going to respond to those comments, if we'll go back 21 immediately and put a resolution to that comment or an 22 attempted resolution immediately to the comment or if we'll 23 just gather that in, and if we make another posting based on 24 those comments, you'll see that revision there. But by it 25 being on the web, you'll know that we've made aware of the ANN RILEY & ASSOCIATES, LTD.
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| 1 ' comment'and that we're.considering it.
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| r 2 Based on that, as I stated, we're planning on k
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| 3 doing a complete revision, particularly of SRP Chapter 5. j 4 The current schedule for that is our hope is to get that 1
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| 5 completed sometime in probably -- no later than 6 mid-February, that it would be able to be posted on the web, 7 again for further comment ac that point.
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| 8 That's one of the few SRP chapters that -- based 1 9 on the time schedule, that we're trying to work actively on 10- getting out for another look by industry, and again, it's 11 just based on that we've had a lot more interaction on this 12 subject, and we're hoping, you know, that based on that 13 interaction, we'll be able to show that our intent is to ;
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| 14 follow up on your comments and try to address them as is
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| : 15. appropriate.
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| (''))
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| \n.
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| 16 l
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| Sometimes we'll agree with a comment, sometimes we 17 won't, but you'll be able to see where we're slanted, you 18 know, where we're aiming for on it, and then hopefully 19 you'll get a comfortable feeling that, based on this process 20 you're seeing here, that other comments that you're 21 providing, even if we don't have time to post them before we 22 go back to the Commission, that you'll feel that we're at 23 least, you know, trying to address whatever those issues 24 are. ;
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| 25 Are there any questions?
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| y 35 1 MR. MANNING: Paragraph (b), last sentence, you 2 used the word " practicable" as opposed to " practical."
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| 3 MR. LEWIS: " Practicable" is the correct word, I 4 believe, because that means when possible. " Practical" does 5 not mean that.
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| 6 MR. MANNING: So, that precludes any economic i 7 justification for not going with the design feature, j 8 MR. LEWIS: No, I don't see that at all.
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| 9 MR. ELLIOTT: Could we say that the engineered 10 controls are preferred over administrative?
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| 11 MR. LEWIS: Well, we had a lot of discussion about 12 that. " Primary" has a stronger connotation than " preferred,"
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| 13 of course, and it'may be a better term for a regulation.
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| 14 MR. COMFORT: I think the comment was actually 15 towards the latter part, which is " Engineered controls shall 16 be used" rather than administrative controls. I think the 17 comment is more towards engineered controls are preferred 18 over administrative controls.
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| 19 MR. LEWIS: Are you talking about " primary" and 20 " preferred" or "used" and " preferred"?
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| 21 MR. ELLIOTT: I was following onto the question 22 about " practicable." It says, "where practicable," they 23 shall be used. That could be a pretty strong 24 interpretation, and I guess I was suggesting that it may say 25 that engineered controls are preferred over administrative ANN RILEY & ASSOCIATES, LTD.
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| 36 1 controls.
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| 2 MR. LEWIS: The words "where practicable" are 3 actually from the standard. "Where practicable, reliance 4 should be placed on equipment design in which dimensions are 5 limited rather than on administrative controls."
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| 6 MR. BIDINGER: There's a big difference when it 7 says "should."
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| 8 MR. LEWIS: I agree.
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| 9 MR. COMFORT: You don't quote the standard and than 10 put a "shall" in there.
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| 11 MR. DAMON: There's no such thing as "should" in a 12 regulation. There's no point in putting anything that says 13 "should" in a regulation.
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| 14 MR. ROTHLEDER: It's a highly subjective term, p 15 " practicable."
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| 16 MR. LEWIS: " Practicable" -- I view it as allowing 17 the flexibility that you're seeking. So, "shall" combined 18 with " practicable," in my opinion, was acceptable. "Should" 19 combined with " practicable" basically says the same thing.
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| 20 " Practicable" allows the flexibility.
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| 21 MR. KENT: If it's practicable in whose judgement, 22 though? I mean that's the subjective part of it, and that's 23 where we get into huge debates during inspections, those 24 kind of words.
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| 25 MR. DAMON: I don't think any single word would ANN RILEY & ASSOCIATES, LTD.
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| 37 1 cvsr capture all the subtleties of this concept. It has to
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| -s 2 be explained somewhere in consideration of length what's V 3 meant there.
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| 4 MR. SHARKEY: I guess what we're trying to 5 accomplish here is low risk, and if you're doing your 6 evaluations right and you have the proper controls, the risk 7 is going to be low.
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| 8 If it's administrative and the risk is still low, 9 then it shouldn't matter, but typically you don't give the 10 same weight to an administrative control as a passive or 11 active control.
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| 12 So, it's really kind of redundant, the whole last 13 sentence. We all prefer engineered controls.
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| 14 MR. DAMON: My name is Dennis Damon, with the NRC.
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| r-' 15 We accepted the industry's idea that criticality
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| \.
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| 16 should not be called out in a special sense as a 17 high-consequence event, even though one normally, in an 18 unprotected case, can't preclude that it would be, but the 19 idea was we realize that one of the major reasons for not 20 automatically categorizing an event -- criticality as a high
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| .21 consequence is that there are facilities, not at these 22 licensees, but there are facilities -- shielded facilities 23 exist where criticality could happen behind engineered 24~ shielding, and therefore, no one would get a dose exceeding 25 any of those limits, and we recognize that that's possible.
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| t 38 1 We also recognize that the Commission has 2 specifically issued a =trategic safety performance goal of 3 .not having inadvertent criticalities.
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| 4 Therefore, we recognize that the rule language as 5 it was originally structured had a regulatory gap, namely a 6 shielded criticality that did not produce a dose exceeding 7 any of the limits stated in the rule would be simply not 8 addressed at all by the rule, it would have been -- there 9 would have been no requirement, and we realize that that was 10 not what the Commission intended.
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| 11 The Commission intended that criticalities be l
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| 12 prevented, perhaps not with the same degree of stringency if l l
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| 13 they were behind shielding than if they were not but that 14 criticalities had to be addressed and covered no matter rg 15 where the dose came out in the rule.
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| U 16 So, that's why this whole section is really in 17 here, and I'm just saying that so that, when we get wrapped 18 up in this language, that's really the objective here, is 19 simply to have a statement in here that criticality should 20 be addressed or prevented and rely on prevention, not simply 21 -- shielding alone is not enough.
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| 22 If you were going to have a criticality every week 23 behind that shield, the Commission still would not like I 24 that, they would want you to prevent -- if you want to have 25 a criticality every week, go see NRR for a reactor license, I I
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| 39 1 bscause that's what you're building, you know.
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| 2 MR. SHARKEY: You're operating a reactor without a 3 license.
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| 4 MR. DAMON: Exactly. That's really all this is in 5 here for. It's not really to have all the verbiage. It's 6 simply to say, okay, even if a criticality does not produce
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| '7 a dose, we still want you to prevent it adequately.
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| 8 As long as I've got the microphone, I'd like to 9 make a couple other remarks on two technical points, because 10 they are things which -- we have a mixed audience here, and 11 some people might understand this and others not.
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| 12 There was a discussion by Calvin Hopper and George (
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| 13 Bidinger about frequency of occurrence of events and 10 to 14 the minus 6 per year and stuff like that. I 15- I You have to remember that the numbers tha.t may 16 have arisen in some other context with respect to reactors 17 are usually a number applied to a single reactor. So, it's 1 18 the likelihood of something happening at the reactor.
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| 19 The numbers as they are stated and the terms 20 "likely" and "unlikely" - " highly unlikely" and "unlikely" 21 in the rule are referring to individual accidents, not to 22- the whole plant, and it's on a per-accident basis, and as 23 has been mentioned many times before, the ISAs that are !
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| 24 being submitted may have hundreds or thousands of these l 25 potential accidents.
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| l ANN RILEY & ASSOCIATES, LTD.
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| y 40 1 So, unless you take a plant-level numerical goal g-~ 2 and divide it by those hundreds or thousands of accidents,
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| ( 3 that's the number you have to get to. That's why the number 4 would have to be a low number. It's because there's 5 hundreds and thousands of -- it's applied at each individual 6 level and you have to do it that way.
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| 7 Second point on the consequences of criticality --
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| 8 some people here are not really criticality engineers, and 9 the area of consequences of criticalities is an area that I 10 have worked in in the past.
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| 11 I wouldn't be considered the world's foremost 12 expert, but there is a reason why -- there is a fundamental 13 physical reason why it is that almost any criticality would 14 produce a dose sufficient to exceed 100 rem to someone 15 standing,close to it.
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| (''/)
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| \_
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| 16 The physical reason is because in order to turn 17 .that criticality around and shut it down, it has to be done 18 by inherent feedback mechanisms.
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| 19 In order to get most feedback mechanisms, you have 20 to do something macroscopic to that material. Normally what 21 you have to do is to heat it up or to radiolytically 22 generate bubbles in it or something like this.
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| 23 You cannot turn a criticality around with trivial 24 feedback effects. You have to use substantial. And that's 25 why you always get a number like 10 to the 17th fissions in ANN RILEY & ASSOCIATES, LTD.
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| 41
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| : 1. a criticality event. It's because of having to put enough f~s 2 energy into the system to get the negative feedback
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| ''' 3 sufficient to shut you down.
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| 4 So, it is generally a true statement that it's t
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| 5 very difficult to get a criticality to be small enough that 6 it would not give you a 100-rem dose if you're standing 7 right next to it.
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| 8 MR. VAUGHAN: Charles Vaughan.
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| 9 The question I have is, in 70.60(d),'the 10 next-to-the-last sentence, it has a statement that goes "and 11 approved administrative safety margin." Could somebody kind 12 of amplify that just a little bit about what that means and 13- what's involved?
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| 14 MR. LEWIS: The intent was to have that section of 15 this code reflect the current practice, and I was just 16 discussing with Dennis if having an improved administrative i
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| 17 safety margin is current margin, whether the margin is 18 approved by NRC during licensing or by the facility on a 19 process-specific basis.
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| 20 MR. DAMON: I would say that the NRC practice has 21 not been consistent in this area. It's an area where we 22 should be consistent. There is a reason for having what I 23 would call a minimum sub-critical margin.
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| 24 The idea of having an arbitrary administrative 25 margin -- the rationale there is a little bit less clear, l
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| 42 1 but the idea is that, despite the fact that you believe you 3 2 understand all the uncertainties in setting a margin, that
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| ''- 3 you should still allow some additional margin for things 4 that you have not been able to identify or for some factors 5 which you have identified but haven't quantified is, I 6 think, still valid.
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| 7 So, the term " administrative margin" may be a bad 8 choice of term. It's not really purely administrative.
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| l 9 It's a real margin that you need for unknown uncertainties. l l
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| 10 MS. TEN-EYCK: This is Liz Ten-Eyck. Let me try to 11 explain it in non-technical terms from someone that doesn't 12 know much about criticality.
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| 13 We look at this as the difference that would be 14 below a K-effective of 1, that would be negotiated with the
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| -~
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| 15 licensee on the particular process or system or whatever and
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| \_)
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| 16 whether you're working for a K-effective of .95 or .97 or i
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| 17 whatever, it's that goal that you work at for K-effective l 18 that would be less than 1. l 19 Does that help you?
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| 20 MR. VAUGHAN: So, the measure of the dimensions 1 21 would be K-effective.
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| 22 MS. TEN-EYCK: Pardon?
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| 23 MR. VAUGHAN: The measure or dimension of this l 24 particular safety margin would be expressed as K-effective?
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| 25 MS. TEN-EYCK: Right. In other words, it's the I
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| 43 1 K-effective that you all decide on that's appropriate for 2 the thing.'Say it's .95, and then you back off on that to
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| ) 3 have your biases and all the other things that go into it, 4
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| and there's a little formula that I've seen that's kind of a 5 sigma and whatever.
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| 6- This is what you work out as your K-effective 7
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| limit, and that would introduce this administrative margin 8 that we're talking about.
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| 9 MR. DAVIS: Jack Davis, NRC.
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| 10-Because we stated that the normal and credible 11 abnormal conditions that the process should be sub-critical, 12 we wanted to clarify what we meant by sub-critical there.
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| 13 It just couldn't be .99, for instance, and the 14 administrative safety margin, of course, would depend upon 15 your K-effective sensitivity and how quickly you would 16 ~ approach a critical condition.
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| '17 So, that's why it says there use as an approved 18 administrative margin. It depends on the facility.
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| 19 MR. DAMON: There was an extensive discussion of 20 this topic among engineers here at the NRC a few years ago, 21 and we drew the conclusion, as an example, that you could 22 not pick any particular administrative margin, some absolute 23 number that would be used by everybody, and an example that 24 was given is someone could -- can design what amounts to a 25 sub-critical facility where they have very tight control ANN RILEY & ASSOCIATES, LTD.
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| 1 44
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| .1 over tha reactivity of that system, and not only that, they 2-could have run it up to critical to know exactly where (n) 3 critical is, then backed off.
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| 4 So, they can design systems that go .99, whatever 5
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| you want, provided it's tightly controlled, and they know --
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| 6 see, they now know where critical is, because they've done 7 it, see?
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| 8 So, that's the extreme case, but you don't 9
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| 'normally have that, and the next level below that is the 10 case where someone has built an experiment that's exactly 11 like what you intend to build, or very, very, very close.
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| 12 In that case, the numbers I've normally seen is 13 about 2 percent, you know, K-effective of .98 is about as 14 close as you want to get.
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| 15 f%
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| MR. HOPPER: Not to be picky, but I would like to
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| ( ,) 16 suggest that people use the language something like approved 17 administrative margin of sub-criticality for safety.
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| 18 People confuse safety and sub-criticality. You've 19 got to have sub-criticality to achieve safety, but one does 20 not mean the other, and that gets muddled. So, I'd like to 21 suggest that you say approved administrative margin of 22 sub-criticality for safety.
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| 23 MR. BIDINGER: I'll go a step further. I think the 24 phrase after the comma in that first sentence should be 25 deleted from regulatory space.
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| 45 1 ;MR . SHERR: Which comma?
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| -~ 2 MR. BIDINGER: "All nuclear processes are 3 sub-critical," period.
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| 4 When you put " including use of appropriate bias 5 and uncertainty adjustment," you're skipping over the 6 fundamental approach to criticality which used critical mass 7 data, which doesn't include appropriate bias and uncertainty 8 adjustments.
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| 9 This stems from Monte Carlo calculations. It's an 10 appropriate subject in a standard review plan but not in a j 11 regulation. If you have to put it in there, I second 12 Calvin's comments on the use of the margin of 13 sub-criticality.
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| 14 MR. VAUGHAN: I think Calvin's words will do what
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| /~T 15 you were trying to do and actually better convey the U 16 message.
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| 17 MR. SCHILTHELM: I would ask one question, 18 non-technical. Is this current practice -- I heard that 19 once -- in everybody's view here? I'm asking the NRC. You 20 all know what our licenses say. Are our current licenses i
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| 21 compliant with this, in your viewt 1 l
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| 22 MS. TEN- ~CK I'm not sure we all know exactly 23 what your licenses say.
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| 24 MR. DAMON: I would say they are, with the 25 exception, possibly, of the Westinghouse thing about g ANN RILEY & ASSOCIATES, LTD.
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| F-46 1 K-effective of 1.0. In that case, it's not clear there is 2 any additional margin.
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| 3 MR. KENT: This is Norman Kent from Westinghouse.
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| 4 [ Laughter.]
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| 5 MR. KENT: That's true, Dennis. In recent 6 inspections, we have discussed that with the NRC. Wilbur's 7 not looking at me right now, but I suspect that we will be 8 introducing a margin that I also suspect is still under 9 negotiation.
| |
| 10 MR. BIDINGER: I'd like to go back to one other 11 subject that Gary discussed, and that is that the NRC is 12 going to continue to insist upon the re-submittal of the 13 bulletin for reports.
| |
| 14 I know, seven years ago, the NRC had the f e~g 15 capability to do a literature search of all documents in a V 16 document file and bring up those kind of things, and I think 17 that, in a sense that relieving industry of unnecessary 18 regulatory practices, this seems to be one of the areas 19 where re-submittal of information that's already in a 20 document is an unnecessary issue.
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| J 21 MR. SCHILTHELM: I guess, before you answer that, 22 let me add a little to that, because that was exactly what I 23 was going to comment on.
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| 24 It seems as though you're trying to -- you have a 25 performance objective in mind, and that's that we use this ANN RILEY & ASSOCIATES, LTD.
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| 47 1 information in doing the ISA, but simply providing the i,
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| 2 information as a historical summary doesn't directly 3 accomplish that performance objective.
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| 4 Is there some way that, if that's the objective, 5 that that should be the language in the rule, rather than a 6 seemingly arbitrary requirement to submit information in the 7 license application?
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| 8 MR. VAUGHAN: Can I add a little to that before you 9 respond, because that was my next one, too.
| |
| 10 I think the part about the performance that you're 11 trying to get out of a licensee seems clear, and that is you 12' want the licensee to use this information in their ISA work, 13 in their corrective action work, their whole management 14 system, as one of the elements that you need to make 15 decisions, and I don't think we disagree with that, but the 16 mee.anics are very difficult, because it actually requires 17 us ao keep this information in a different form and do one 18 additional task with it that we wouldn't normally do if we 19 were working in-house.
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| 20 If you think about it, though, the fact that the 21 NRC -- and we've asked this several times over the years --
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| 22 the NRC gets all of these' reports.
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| 23 So, they have even a much larger database than the 24 licensees do, but the NRC has'not chosen to use that to 25 publish back to industry information that could be really 7y ANN RILEY & ASSOCIATES, LTD.
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| 48 1 important to the whole industry in terms of improving our 2 safety programs where there is a propensity for some 1
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| 3 particular situation to happen, and you can -- the more data 4 you have, then the better your ability is to make those 5 kinds of decisions.
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| 6 So, one of the things is, I think, with the NRC 7 receiving that much information, they need to think about 8 routinely -- and I don't know what that frequency is but 9 routinely making some evaluation and, you know, identifying 10 some lessons learned out of the information that's been 11 published.
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| 12 The second thing is a similar kind of question of
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| '13 the licensee.
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| 14 You want them to use it in their ISA, you want it 15 kept up to date, you want it to have a positive impact on 16 their safety program, and that's what you ought to focus on, 17 because having them make up this list at the end of 10 years 18 or a 10-year-long list and then suddenly getting to the lith 19 hour and decide that an ISA or two have been messed up 20 because all of the information wasn't used or something like 21 that has the horse out of the barn.
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| 22 So, what we need to do is look at a way to 23' implement.this whole thing that requires it to be used 24 currently and, as you move on, to minimize the fact that we 25 make mistakes as we move through here and those aren't ANN RILEY & ASSOCIATES, LTD.
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| p L 49 l 1 recognized for a period of time.
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| I k~ 2 MS. TEN-EYCK: I just want to make one comment.
| |
| l 3 We have attempted to feed back lessons learned from these 1
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| 4 reports through information notices when there were a number 5 of activities that were focused on some particular issue or 6 problem, and we have tried to do that, but I think what 7 you're proposing is that it be done across the board on some 8 ' periodic basis, lessons learned on all the events that have 9 happened during that timeframe? Is that correct?
| |
| 10 MR. VAUGHAN: Right.
| |
| 1 11 MS. TEN-EYCK: And you're proposing that that 12 . might be something that the NRC would do. I might like to 13 suggest also that that might be something also that industry 14 'could do. We're now in this vein of getting industry to
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| ("N 15 also do things that will help the industry.
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| 16 So, I hear you and I think it's a good point, but 17 I did want to note that we have been providing those as we 18 see the things coming up through information notices in the 19 past anyway.
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| 20 Thank you.
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| ; 21 MR. PERSINKO: I was just going to say you imply l
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| 22 that there would be a fair amount of work to submit this l 23 information. It didn't appear to us that there would be if l
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| 24 the information was used in the ISA. Could you explain the 25 additional work that would be necessary?
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| 50 1 MR. VAUGHAN: The thing that I see is that using it
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| (,%
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| 2 in the ISA, it's in one particular form, it's in one set of
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| \- / 3 documentation, and is put together for internal consumption.
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| 4 I mean it communicates to us, but it might not necessarily 5 communicate what's outside.
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| 6 So, all of those particular items will have to be 7 extracted and put on a list that we can supply in this 8 application, and that's just extra work that, if there is a 9 problem, identifies a problem long after the problem should 10 have been known and addressed.
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| 11 MR. SHERR: We hear what you're saying, Charlie, 12 and just a matter of policy as well as the requirements of 13 the Paperwork Reduction Act, you know, we don't want to be 14 asking for something we don't need and, more importantly, 15 something that we have already.
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| 16 I think we need to re-look at the language, and I 17 think the idea was that, in fact, past events would be 18 considered and which past events were considered would be 19 identified in some documentation and perhaps it might be 20 just a matter of referring to code numbers or whatever it 21 is, but we can figure that out.
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| 22 But as a matter of policy, you know, we shouldn't 23 be asking you to provide information that you've already 24 provided. In this context, I think it was asking for 25 information about that information that had been earlier ANN RILEY & ASSOCIATES, LTD.
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| 51 1 provided, how it was considered.
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| - 2 MR. SCHILTHELM: I guess, to respond to you, Drew,
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| \-- 3 my concern is not necessarily for the amount of work it 4 would be to put it in the license application. It just 5 appears to me that, if there's a performance criteria, it
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| ~6 ought to be stated as such rather than simply listing 7 events, because you haven't stated the performance criteria.
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| 8 If it's there, it should be stated.
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| 9 MR. KILLAR: We recognize from the beginning 10 paragraph which laid out the purpose that this was what the 11 intent was, but from the reading the rule, it certainly did
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| -12 not come across that way, and I think it may be a matter of 13 trying to revise maybe the summary ISA contents or something 14 along that line to capture this concept in there rather than 15 having a separate list of the events over the past 10 years, 16 and I think that would be more meaningful, and it also, I 17 think, will resolve the problem of just we're not just ,
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| l 18 sending you another list. It shows you how the information 19 has actually been applied, and so, you have the benefit.
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| 20 It may require that, as part of our summary ISA, 21 we may have to have a table which identifies where each of 22 these items have been -- I see Charlie cringing -- just so 23 you know that they've all been captured or something along 24 that line, but I would try to minimize that.
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| 25 MR. VAUGHAN: You've got to step-back and think ANN RILEY & ASSOCIATES, LTD.
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| I 52 1 1 cbout tha most important thing, and it's really not a record
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| ; 2 or what it takes to generate, if that record is necessary t
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| \ 3 and is needed for the safety of the plant.
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| 4 Philosophically, the performance that is needed is 5 that all of these events and the investigation of those
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| ''t 6 events that produce information need to be used as they are 7 available when ISA work is done. That's what needs to 8 happen.
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| 9 It doesn't need to be done 10 years later or get 10 10 years down the road and find out, because some 11 information we had six or eight years ago said this, we've 12 got a problem with what we're doing today. I mean that's 13 really the bottom line.
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| 14 MR. KENT: This is Norman Kent from Westinghouse,
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| ~g 15. and I don't know if I'm going to belabor it or contribute, J 16 but from a crit safety specialist who's been at Westinghouse 17 now for eight years, it seems to me that significant events 18 at the plant -- and I'm thinking now of ones that I know of 19 that happened in 1991 and 1992 and 1993 -- became either 20 licensee-identified violations or violations by the NRC at 21 subsequent inspections.
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| 22 To me, that meant many hours doing root cause 23 analyses and modifying systems that were involved in the 24 accident to make sure that controls were applied so that 25 wouldn't happen again, revised drawings, etcetera.
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| 53 1 Now, if I do the ISA on that system subsequently 2 to that, those controls will have already been in place, 3 those violations will already have been agreed and closed i
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| 4 out by the NRC, and so, they are there, and I don't know 5 that I would even consciously be looking to make sure I took 1 6 care of that particular incident that happened 7 nine-and-a-half years ago, because the configuration, in ;
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| 8 fact, was modified and it exists and it's been there for 9 seven years.
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| 10 MS. TEN-EYCK: Yes, but I think one of our 11 interests is also that the lessons learned from what you --
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| 12 what you experienced during that situation is applied to 13 other systems as you do your ISA, so it isn't just 14 specifically what you would do differently in that 15 particular system but how you use the lessons learned, I L 16 think is what we're really more focused on.
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| 17 MR. KENT: I think that's a good point, and that 18 goes back to what Steve has said about having the rule read 19 what the performance measure needs to be rather than a way 20 that we might want to satisfy that performance or show that 21 we did.
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| 22 We need to be learning lessons from things that i l
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| 23 happen to us as well as those that happen to our colleagues. I 24 MR. KILLAR: There is one other drawback. Some of
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| . ;25 these things, especially when you go back 10 years, if you ANN RILEY & ASSOCIATES, LTD.
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| s-54 1
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| look at the way the industry has changed and the process has ,
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| 2 changed, some of the events that we've had 10 years ago have O
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| kls 3 no relevance to the way we're doing business today, 4 particularly if you start looking at some of the wet 5 processes that we've basically done away with.
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| 6 MR. ROTHLEDER: I'd just like to go back to item 7 (d). This is Burt Rothleder from DOE. I'd like to register 8
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| agreement with what George Bidinger said, putting a period 9 after "All nuclear processes are sub-critical."
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| }
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| 10 I agree with George that I don't think it's l l
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| 11 appropriate in a rule to specify how you achieve 12 sub-criticality. I just think it doesn't belong there.
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| t 13 That was my only comment.
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| 14 MR. LEWIS: I haven't really heard, other than that
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| -- 15 comment and also a comment regarding the flexibility of the
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| \- ' 16 last part of the second sentence, the general feeling for if 17 (b) resolves the stated concerns that you've given us, and 18 I'd like to get the industry's opinion on that and whether 19 .there's any highly objectionable language that we have 20 inserted into (d) that is just a no-go as far as you're 21- concerned.
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| 22 MR. KILLAR: I think we need some time to sit down 23 and caucus amongst ourselves rather than trying to give you 24 an answer right now, after reading it for the first time 25 this morning, even though we've given you some, you know, ANN RILEY & ASSOCIATES, LTD.
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| 55 1 top-of-heads response.
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| ,_ 2 I think those probably were sort of the no-go,
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| %- 3 particularly when we started getting concerns about 4 engineered controls shall be used rather than administrative ,
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| 5 controls.
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| 6 I think we need to sit down and just go through it 7 ourselves and discuss it amongst ourselves.
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| 8 I think you've got sort of the top-of-the-head 9 immediate reaction, but I think we need to spend some time 10 focusing in amongst ourselves before we can give you any 11 more detailed answers than that, unless someone has 12 something.
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| 13 MR. SCHILTHELM: I'd just add one thing to that.
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| 14 Steve Schilthelm from B&W.
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| 15 We keep slipping -- I think both the NRC and the 16 industry keep slipping into the mode of putting how-to 17 information into the rule rather than the performance 18 requirement, and we seem to have a lot of discussion about 19 the how-to information, not necessarily the performance 20 requirement.
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| 21 So, I think if we could step back and, where 22 possible, eliminate that how-to information and just stick 23 to the performance requirements in the rule, we might 24 eliminate a lot of the disagreement over the seemingly 25 trivial words.
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| 56
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| ,. 1 MR. DAMON: Well, in that regard, with respect to
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| .. 2 item (b), I think the suggestion was made that the language
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| \m 3 following the comma be stricken, and the trouble with that 4 is the language following the comma all deals with a margin 5 of sub-criticality, and the danger of striking all that is 6 that there is a performance requirement related to the term 7 sub-critical, and.that is you must be confidently 8 sub-critical.
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| 9 It has to do with whether -- with the state of 10 knowledge, and there does need to be a performance 11 requirement that there be sufficient margin that you have 12 full confidence that the system will be sub-critical.
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| 13 MR. KENT: This is Norman Kent from Westinghouse.
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| 14 The first sentence of paragraph (d), assuring that 15 under normal and credible abnormal conditions, does,that 16 mean that I have lost double-contingency protection? Does 17 that mean that I have had a contingency or two, or does that 18 mean that I've had one?
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| 19 MR. DAMON: This is Dennis Damon again.
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| 20 All I can tell you is my own understanding of 21 those words is that, among the conditions referred to as 22 abnormal, the standard or the practice is to consider the 23 failure of any individual criticality control to be a 24 credible abnormal event, so that, therefore, the system 25 should be sub-critical given the failure of any one ANN RILEY & ASSOCIATES, LTD.
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| 57 1
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| criticality control when you have a double contingency-type 2 situation.
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| f_3
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| - 3 MR. KENT: By criticality control, you mean an item 4 relied on-for safety, as defined elsewhere in the rule? I'm 5 trying to tie ANSI 8.1, where it says --
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| 6 MR. BIDINGER: Well --
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| 7 MR. HOPPER: Excuse me, George, but there's a 8 section in 8.1, a paragraph entitled " Process Analysis."
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| 9 It's 8.2.1 or something like that, paragraph '8.2.1, I think 10 it is, and anyhow, the statement is almost precisely this --
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| 11 it's not precisely this, but what it does say is, 12 essentially, no upset or normal -- no condition shall lead 13 to criticality.
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| 14 That is the "shall" in the standard. Double 15 contingency business is "should." But thou "shall" have no 16 criticality as the result of any single failure, and that's 17 what the standard says, and I think that's the conveyance 18 that was intended here.
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| 19 Now -- maybe it wasn't, but -- l 20 MR. KENT: Okay. If that's true, then -- I didn't 21 see the words " single failure," I see plural " conditions."
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| 22 MR. BIDINGER: This is independent of the double --
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| 23 this statement should be independent of the double -- any 24 application or consideration of the double contingency.
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| 25 This is a blanket statement in 8.1. You can't allow any ANN RILEY & ASSOCIATES, LTD.
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| 58 1 operation that will critical.
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| 3 2 -So, this is a good -- I mean going up to the -- as
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| \-- 3 far as you go in (d) up to some critical period, that's a 4 very good and necessary statement in regulatory space.
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| 5 MR. DAVIS: I'd like to make a comment -- this is l
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| 6 Jack Davis -- in relation to what Burt said earlier.
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| 7 We're really not describing how to do it, because 8 we say -- whatever methodology you're using. We're just 9 saying that you need to know, reiterating what Dennis said 10 earlier, what sub-critical is. It just can't be some 11 arbitrary number, you know, because some people, in their 12 minds, .999 is sub-critical, which might not be adequate for 13 regulatory space.
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| 14 MR. ROTHLEDER: So, in other words, that's a rs 15 definition of what you mean by sub-critical. You're really 16 defining it.
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| 17 MR. DAVIS: In regulatory space, I guess I would 18 say yes.
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| 19 MR. ROTHLEDER: I don't know if that belongs in 20 there or not. I'm not an expert in how one words 21 regulations.
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| 22 MR. DAVIS: I don't think I am an expert either.
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| 23 MR. ROTHLEDER: At least we've got that statement.
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| 24 That's really a definition.
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| 25 MR. KILLAR: Can I go back to Norm's point just for l
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| a
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| 59 1
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| a second here on the use of the term " credible abnormal 2 conditions," with the "S"? That can be read two ways.
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| 3 One way is that you're talking about multiple 4 breakdowns of barriers, so you have multiple abnormal 5 conditions because you have multiple barriers to break down, 6 so you've lost your single, your double, and maybe even 7 triple contingency if you read it that way. I don't think 8 it was intended that way, and I think that's where the 9 question came from.
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| 10 MR. LEWIS: I'm not so sure that I agree. If it's 11 a credible event -- for example, if you have a combination 12 of two credible events that could fail, the combination of i j
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| 13 those two could fail every 50 years. That seems credible, 14 and it should be sub-critical under that situation.
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| r'' 15 If you have a diesel that is important to safety, NJ 16 that's needed if you lose power, and the diesel is out for 17 maintenance one month a year or inoperable for six months a 18 year, it's credible that you will lose power while that 19 diesel is not operable.
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| 20 MR. SHERR: I think, in terms of the agenda, if it 21 still fits, we've actually creeped into 3 (c) and (d), and 22 based on what Felix said earlier, what we might want to do 23 is come back to this agenda item later to further discuss 24 this, after you've had a chance to discuss aspects among 25 yourselves.
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| l ANN RILEY & ASSOCIATES, LTD. I I Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 l Washington, D.C. 20036 (202) 842-0034 l
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| L 60 1 Is that the sense, Felix?
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| s 2 So, what we could do -- it's now 10:30. Maybe it C''g) l 3 would be most useful to go on to the SRP thing, and if we 4 ' could go on to 4 (a) , where the industry would provide 5 comments on the SRP guidance, and then perhaps break after
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| : 6 that time period, at which time caucus among ourselves, l i
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| 7 among our particular groups, and maybe even have time to get 8 something to eat and then get back, and we can see what time 9 that is.
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| 10- In the meantime, it's 10:30. Does anybody have an 1 11 urgent need to take a break at this point? I see some heads i 12 nodding. So, why don't we take a break now and come back at 13 10 till 11, and then we'll see how far we get.
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| 14 [ Recess.] l 15 MR. SHERR: We'll continue on with our agenda item 16 4 (a) , with the idea that, once we complete that, we'll break 17 for lunch, and at that point, we'll decide when we'll l l
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| 18 reconvene. '
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| 19 Felix?
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| 20 MR. KILLAR: Mark Elliott from BWXT is going to 21 -
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| coordinate this presentation, with the help from a lot of i
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| 22 the members.
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| 23 Mark?
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| 24 MR. ELLIOTT: Mark Elliott, BWXT. I'm going to 25 talk about the standard review plan for the criticality ANN RILEY & ASSOCIATES, LTD.
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| 61 1 safety. I'm not a criticality engineer. I'm just a manager 1
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| -)
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| 2 there. The only reason I'm giving this presentation is i 3 because I said heads and it was tails.
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| 4 (Laughter.] '
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| 1 5
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| MR. ELLIOTT: We identified some issues that we 6 wanted to talk about, and I'm sure whoever wrote this thing 7 did a wonderful job with the guidance they were given. I 8 don't want for them to be throwing tomatoes at me when we 9 starting talking on some of these things.
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| 10 The first one -- well, four general issues, some 11 philosophical, fundamental-type issues we want to talk l 12 about. 1 13 Then we give some examples of the level of i
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| 14 prescription that's in it that we think is inappropriate, 15 some redundancies that are there that we think that that may 16 be able to restructure some things and to make it a smaller, 17 less voluminous SRP, and then some definitions, locations, 18 and maybe some inconsistencies in those. I i
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| 19 On the first slide, on the philosophical issues, 20 these philosophical issues are things that could really 21 impact the way we operate the facility and are kind of 1
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| 22 really big changes in direction from the way we've been 23 doing business over the years.
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| 24 First of all is the criticality controls don't 25 need the highest level of quality assurance in all cases, 7g ANN RILEY & ASSOCIATES, LTD. ;
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| 62 1 and I think we've talked about that a little this morning, f-'s 2 and I think that you've agreed that that needs to be looked Q. 3 again.
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| 4 And as you notice, the second of the SRP we put to 5 the left of this item up here, where we've identified this 6 concern.
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| 7 We don't think that there's a real need for a 8 performance-based training program.
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| 9- We think that the assurances, especially around 10 administrative controls that involve people and things, 11 should determine what needs to be done to assure that those l 12 controls are reliable, so that establishing some big 13 administrative program such as performance-based training, 14 we thought, was unnecessary. ;
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| I 15 On the second slide, if you read in that section
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| (}
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| 16 5.4.5.1.5, it talks about -- it's talking about how you can j
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| i 17 change your facilities without getting a license amendment, '
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| 18 and that paragraph talks about, you know, if you don't 19 decrease the effectiveness of your safety, then you can go 20 ahead and change it.
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| 21 Well, if you look at that section -- and I've read 22 that part of the section in a previous meeting -- it gets 23 down into the details of no new accident scenarios, no new 24 types of procedural failures, and things that really gets 25 down to the finite detail, and the way that's written now ANN RILEY & ASSOCIATES, LTD.
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| 63 1 would result in just numerous amendments to the license
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| , 2 application that are really -- most of which are all C 3 unnecessary, and we talked about that in the meeting, I 4 guess, with the commissioners that got us in those large 5 numbers we cited in that meeting. l j
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| 6 So, that needs to be looked at in regard to not I t
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| 7 l trying to -- we're not trying to increase the number of 8 license amendments from the ones we have today.
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| 9 In the next one, the K-effective formulation, we 10 just don't see where that's necessary to be in a standard 11 review plan, and I guess there -- it's got a lot of details 12 to its 13 It's one way to do it. I think there are several 14 other ways to do it, and I think that we should leave that l''N 15 up to the industry experts to talk about things such as how b
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| 16 to derive K-effective and how it's applied and things like !
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| 17 _that, and we don't think it's necessary to be in the i 1
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| 18 standard review plan. ,
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| 19 MR. DAMON: This is Dennis Damon. Could 1 ask a 20 question about that? I'm not sure which section you're i 21 referring to there when you say K-effective.
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| 22 MR. ELLIOTT: It's in 5.4.5.2.5.
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| 23 On the next slide we looked at -- there are some 24 technical inaccuracies, we feel, that are in the standard 25 review plan, and this just identifies one of them, 5.4.5.3.
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| 64 l 1 I think it's having to do with reflection and some spacing, i 2 required spacing to achieve certain reflection criteria that O- 3 our experts, our criticality safety people indicatcd was l 4 maybe inaccurate.
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| l 5 The second one, of course, we've talked a lot 6 about is the use of probabilistic techniques in determining 7 double contingency, what's unlikely and what's highly 8 unlikely and things like that, and we don't think that any i 9 of those are appropriate.
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| 10 In reading through the SRP, just in general, there :
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| 11 are a lot of determinations that are required of the license 12 reviewer in the SRP that we don't think can be made from I 13 reviewing a license application. We would think that you 14 would have to -- from a criticality perspective anyway, you '
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| 15 would probably have to look at the criticality safety 16 evaluations at the facility and the processes at the 17 facility to make such determinations as they're worded in 18 the standard review plan.
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| 19 Going on from the philosophical issues to the 20 prescriptiveness, we notice that there are audit frequencies 21 listed in the standard review plan that a license reviewer 22 would be asking the applicant why they're not committing to 23 quarterly audits and weekly inspections and things like 24 that, and again, we think that the integrated safety 25 analysis results, the assurances that are applied to the ANN RILEY & ASSOCIATES, LTD.
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| F 65 1 controls would determine the appropriate surveillance 2 and f-- .
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| frequencies, maintenance frequencies, thipGs like that, 3 that frequencies shouldn't be spelled out just arbitrarily 4 in the review plan.
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| 5 The standard review plan expands on the 6 K-effective calculation requirements. I'll try to give an 7 example of that.
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| 8 MR. KENT: This is Norman Kent from Westinghouse.
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| 9 This is 5.4.5.2 again, the NTS limits, where the 10 K-effective calculations were spelled out, and it's an 11 indication that that was the only means, the only method.
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| 12 MR. ELLIOTT: Again, we would defer to the industry 13 experts in the nuclear criticality safety division and the 14 ANSI standards to give guidance on such things like that, 15 and to have one method in the standard review plan we 16 thought inappropriate.
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| 17 There's.also some language in the standard review 18 plan that talks about how to adhere to the double 19 contingency principle and what are acceptable exceptions to 20 the double contingency principle and how to make those 21 exceptions, and I think that should be process-specific and 22 left up to the criticality evaluators at the facility.
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| 23 On the next slide, redundancies, throughout 24 Chapter 5 and, I guess, throughout most of the safety 25 program chapters, there are training requirements and 7s ANN RILEY & ASSOCIATES, LTD.
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| I 66 1 organizational requirements and management controls specific l 2 to criticality, safety, or fire, chemical, whatever, and we i l
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| 3 thought it would be more concise to have the training i
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| 4 ' requirements in 11.4, where it talked about training, put i 5 the organizational requirements in 2.0, which is the chapter 6 on organization, and put management controls in 11, which 7 talks about management controls, and just, if necessary, in 8 Chapter 5 or 4 or whatever, refer to 11 or 2, refer the 9 reviewer to those sections for guidance on those topics.
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| 10 On definitions, we've got -- we see that some 11 definitions -- and this may be just different people writing l 12 different sections but inconsistent definitions in the 13 standard review plan from what's cited in the rule, and one i
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| 14 of them was items relied on for safety, is worded
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| ''i 15 differently in those two documents.
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| (U 16 There's another definition in there that talks 17 about double contingency and then double contingency 18 principle, and it's quite confusing to read those two, 19 MR. DAMON: You understand that there is a 20 difference.
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| 21 MR. ELLIOTT: It was not easily understood in the 22 SRP.
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| 23 MR. DAMON: Basically, double contingency principle 24 as it's stated in the ANSI standard says you should have 25 double contingency, but that's different from what double ANN RILEY & ASSOCIATES, LTD.
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| l 67 l i
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| 1 contingency is. I 2 MR. KENT: This is one page 6 of the standard l
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| 3 . review plan, where double contingency is followed by double 4 contingency principle, and the principle -- well, the words !
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| 5 in the SRP are not the words in 8.1, and then, for double 6 concingency above, when I read that, I saw that as being a 7 definition for double contingency protection of a system, so i
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| 8 that the protection exists if those criteria are met, and 9 that's where the confusion comes in. I 10 MR. ELLIOTT: We saw some words in the standard 11 review plan that weren't defined, and then we saw some I i
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| 12 definitions that we never found where they were used. One 13 was criticality control system is defined, but we never see 14 where it's used in 5. Then we also -- and we've talked 15 about it a little bit this morning, about safety margin is
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| [}
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| 16 used, but it's not defined.
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| 17 There are several other things, prescriptive 18 things, definition things. These were just an example of 19 some.of the things that we saw in the review plan when we i 20 were reading through it.
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| 21 MS. TEN-EYCK: Are you going to identify specific j 22 examples or areas where this occurs?
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| 23 MR. ELLIOTT: We've got a partial list.
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| 24 MS. TEN-EYCK: I'd like you to, where you can, 25 point out exactly where you feel there are problems, rather ANN RILEY & ASSOCIATES, LTD.
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| T 68 1 than this overview type of a thing, because it's very r- 2 difficult, when you look at a document the size of the
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| 3 standard review plan, to have some general comments and then 4 expect us to go through'and find all of the ones that you're 5 identifying.
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| 6 MR. VAUGHAN: The biggest single problem that 7 you'll find with virtually every chapter in the SRP is this 8 idea of repeating subjects chapter after chapter after 9 chapter -- for example, training, and training is a subject 10 and the requirements for training need to be integrated and 11 they need to be in in one place where the licensee and 12 everybody else can understand what the requirements are for 13 training for them to comply with their license, and there's 14 other subjects that are much the same.
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| /'' 15 They're cross-cutting subjects that basically go D) 16 across every element in the license, and they need to be --
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| 17 I mean you've got a good start, because you've got all of 18 the things written down, now you've just got to synthesize 19 what needs to go in the right place, and we've pointed out a i
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| 20 few of those, but let me say, we did not get through all of 21 the SRP or even all of that particular chapter.
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| 22 We spent most of the time working on the rule and 23 not so much time on the SRP, but -- and I apologize, because 24 our charts were supposed to have the precise references on 25 them, just like the earlier one, but somewhere between our ANN RILEY & ASSOCIATES, LTD.
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| F[T 69 1 draft and what we had today, they don't have them, and we
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| -x s 2 should be able to give those to you, I mean right quick,
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| '' 3 because we actually had them very specifically piece by 4 piece.
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| 5 Let me just say it seems to us, with some of the 6 examples that we've given, that it may be easier for you to 7 work with it at this stage, as opposed to us try to go 8 through everything and feed you more words. I mean, once 9 you get into it, the concept was not too bad.
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| 10 MS. TEN-EYCK: I guess my concern was just you 11 have these comments, and then the question is, if you have 12 identified where -- at least one example where it is we can 13 focus on exactly what it means, rather than just a general 14 comment, can you talk about just -- I mean audit frequency, 15 things like that, there could be multiple places, and maybe 16 you just meant a concern on one or something. So, that was 17 my concern.
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| 18 MR. VAUGHAN: Yes. We probably didn't do an 19 exhaustive review. We were trying to get examples, and we 20 were trying to find out how to get examples together that 21 would give us some indication of what we felt the root 22 problem was, not trying to be all exhaustive in terms of 23 finding every example.
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| 24 MR. ELLIOTT: Most of these comments that we've 25 given today are from Chapter 5, not the entire plan.
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| 70 1 MR. GOODWIN: This is Wilbur Goodwin with 2 Westinghouse. You mentioned audits, audit frequencies. That 3 specific example was under 5.4.4.3, under the heading 4 " Operational Inspections, Audits, Assessments, and 5 Investigations."
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| 6 If you look at item 2 and 3, it talks about 7 quarterly audits and weekly NCS inspections, and to us, that 8 seems somewhat arbitrary, because it depends on the -- you 9 know, the risk of the system, the type of system, the 10 complexity.
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| 11 You know, some things need frequent audits, other 12 things need less frequent. But that's just an example that 13 we're trying to make. I think there may be others in here.
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| 14 I think those are the only two references we made yesterday.
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| /"% 15 But just to add, this whole 5.4.4.3, operational U 16 inspections, audits, assessments, investigations, I believe 17 that could be moved to one of the SRP chapters or 18 subsections in Chapter 11, audits, inspections, 19 self-assessments, or whatever, same thing as training, 20 management organization, management controls, what have you, 21 get all of that into one chapter so it doesn't continue to 22 repeat.
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| 23 It confuses and maybe frustrates us a little bit n
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| 24 as we go through and review this, and I can imagine it might 25 be the same f.or the license reviewer, as well, just have it 1
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| 71 1 all in one place, g- 2 MS. TEN-EYCK: Let me ask you, in your 3 application, would you discuss all of your audits in one 4- section, or if you were talking about your criticality 5 program, would you include information regarding your 6 audits?
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| 7 We're trying to prepare the reviewer to be able to 8 address the different types of things as they would be 9 presented in your application.
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| 10 So, I agree with you. I think that we could 11 easily put all of that back in a section that deals just 12 with training or whatever, but is that the way that you 13 would normally propose it to us in your application?
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| 14 MR. GOODWIN: I can speak for Westinghouse. I
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| (''s 15 think that's the way we would do it.
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| b 16 We would have the core program, if you will, 17 training that applies, you know, to a large group of people, 18 and if there's something specific to radiation safety or 19 crit _ safety or something else, then we would add special, 20 you know, requirements, or requirements over and above the 21 core program, you know, for that, but still have it all 22 under one training section.
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| 23 MS. TEN-EYCK: A lot of this is in the SRP is kind 24 of to jog the licensing reviewer's memory or thought of 25 looking at it, and I think there may be times when there's E
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| 72 1
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| advanttgas to having some redundancy versus not having it 2
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| 7-~- there, and then, by the time they get back to look at your 3
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| training and they look at all that, they forget, gosh, did 4 we'look at the crit training.
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| 5 So, I think it's a good point, and we'll certainly 6 look at it.
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| 7 MR. VAUGHAN: What happens, I think, in the general 8
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| world is what Wilbur was saying, is we look at training.
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| 9 I mean we don't necessarily have separate looks at )
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| 10 criticality or radiation. We look at training, and yes, 11 there are some different elements of the program for 12 different things, but we look at training, we look at 13 ' reporting, we look at, you know, some fundamental elements, I 14 and so, if the guidance has the reviewer to expect to see it
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| V 15 in criticality safety, then it's going to come up 16 automatically with probt.bly the way that he's going to see 17 it, or if we write the license application in accordance 18 with this, it's in a disconnect between the way the work 19 really needs to happen at the facility, and that's not good j 20 either. '
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| 21 So, we'd like to get the licensing to follow a 22 flow which is as close as possible to the way we do our work ,
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| 23 or look at our work and also is clear to the reviewer then 24 so the reviewer doesn't get surprised when they get the 25 information submitted in that form.
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| 73 i 1 So, I think that's where we got to get.
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| 2 MR. SCHILTHELM: One of the things that we're U'' 3 looking at is the table of contents to this standard review 4 plan is, in a sense, the standard format and content guide 5 for a future license application, since we've sort of 6 combined the two now. l 7 '
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| Our license in some way -- our current licenses in 8 some ways mirror that table of contents, but in a lot of 9 ways, that's probably better and more concise that our 10 current license applications are, because like you said, 1 11 training is scattered about a bit, and if we could clean 12 that up, even in our current licenses, I think it would be i 13 to our advantage, from B&W's standpoint, anyway.
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| 14 MR. GOODWIN: It might be a relatively easy way of
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| () 15 16 reducing the volume of this document, you know, fairly simple.
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| 17 MS. TEN-EYCK: I totally agree. If we can get a 18 format that we all agree is a good way to go and you all 19 present your applications in that format, we review them in 20- that format, I think that that will help everybody and 21 streamline the process.
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| 22 MR. VAUGHAN: Felix, can we recover the references?
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| 23 MR. ELLIOTT: I've got them. If you want to talk 24 about the prescriptiveness, we can go to those chapters, if 25 you want to look at the now.
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| 74 1 MR. KILLAR: Can I just make one comment?
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| 2 One of the things, too, I feel, from a perspective 3 of a licensee, although it's been many years since I've been 4 a licensee, it's better to have all the training 5 requirements in one area, because when you make commitments, 6 usually you have one individual that's responsible for those 7 commitments, and if you make a commitment on training that's 8 over the criticality section, whoever is responsible for l
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| 9 training may think, well, that's criticality and I'll take 10 care of it, but they forget about it or what have you.
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| 11 So, it's much better to have all the commitments 12 in your license on an issue, whether it's training or 13 quality assurance or criticality or radiation protection, in 14 that section, rather than have those commitments scattered 15 throughout various sections, and that way, you don't have 16 the possibility of a commitment falling through the cracks 17 somewhere, and so, for that logic, I think it makes more 18 sense to have them in the various sections, specific 19 sections, rather than spread out.
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| 20 MR. DAMON: This is Dennis Damon. One thing occurs 21 to me. The SRP does different things. One thing is 22 acceptance criteria. Another one is instructions to the 23 reviewer as to how to proceed.
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| 24 One suggestion that occurred to me is, in the --
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| 25 you go to crit chapter, in the section where it tells the g ANN RILEY & ASSOCIATES, LTD .
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| 75 1 reviewer what he's supposed to do, it could just tell him, )
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| ("') 2 go to the chapter on training and review the following )
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| V 3 things that are in there regarding training, and the 4- acceptance criteria would all be in the training chapter, 5 something like that.
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| 6 MR. KENT: Cross-referencing, in other words.
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| 7 MS. TEN-EYCK: If you've got something to give us 8 that cites the specific areas, that's no problem.
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| 9 At least from my perspective, you a'ddressed it so 10 generally that my thought was, well, how are we going to 11 know specifically what the concern was. So, I think that, if 12 you can give us some references, I think that will surely I 13 suffice.
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| l 14 MR. SCHILTHELM: We agonized, Liz. We didn't feel b'
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| /'' 15 16 it was productive to come in here with a list of 200 specifics and just throw it all out on the table. We, on 17 the other hand, didn't know how productive this would be 18 either in generalizing. !
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| 19 MS. TEN-EYCK: Well, we'll look at it, and we may 1
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| 20 have some questions we can come back to you on to make sure l
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| )
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| 21 that we certainly understand the issues, but I think most of l
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| 22 them are -- I think we should be able to understand.
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| 23 MR. GOODWIN: We were trying to find enough 24 substantive examples to give you a flavor of what we see and 25 a pattern here, if you will, and then, obviously, we can 1
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| 76 1 give you more.
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| 2 I will say, with regard to training and probably 3 some of the other chapters within Westinghouse, we're not 4 totally clean. We still have some fragmentation.
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| 5 In fact, I think we have some licensing action 6 going on now'to try to get some of the training and maybe 7 some other things into the appropriate chapter, but that is 8 our objective, is to -- you know, to compartmentalize these 9 various disciplines.
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| 10 MS. TEN-EYCK: And I think that's a good 11 objective.
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| 12 I think what we're looking at is that we've gotten 13 all kinds of formats and applications and everything that 14 aren't always consistent, and so, we need to try to put 15 together something that will address all those
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| [V) 16 contingencies.
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| 17 I think that if we could come up with a standard 18 review plan that the table of contents would be used by 19 industry as a standard format and content guide, then I 20 think that we're making giant steps forward on streamlining 21 and making this process a lot more consistent across the 22 industry, so I don't have any problem with that.
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| 23 MR. VAUGHAN: It could sure cut down the pages, 24 too. Some guy over here at the NRC said it was too many 25 pages.
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| 77 1 MS. TEN-EYCK: What was too many pages?
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| 2 MR. VAUGHAN: The SRP.
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| V MS.
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| 3 TEN-EYCK: Oh. <
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| 4 MR. VAUGHAN: I mean you would take out a lot of 1 5 pages by just eliminating that redundancy.
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| 6 MS. TEN-EYCK: Well, we certainly would like to 7 streamline'it, too When you put all these pieces together, 8 sometimes you end up with a bigger pile than you would 9 normally like to have, that's for sure.
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| 10 I have one other question, though, regarding the 11 standard review plan. Did you all have other comments on 12 this, or is this your presentation on the standard review 13 plan?
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| 14 MR. ELLIOTT: Yes.
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| 15 MS. TEN-EYCK: Okay.
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| 16 Going back to our previous meeting, you talked 17 about the standard review plan, you know, in generalities, 18 and that's why we were trying to get some more specifics, 19 but one of the things was we went back to you and said, 20 well, can you give us a straw man on how you would re-write 21 the standard review plan, so -- but I didn't hear anything i 22 about how you would propose to rewrite it.
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| 23 Are you saying now that you've had a chance to 24 really study and look at it that these are the areas that 25 you felt that you would want us to address and that you're ANN RILEY & ASSOCIATES, LTD.
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| 78 1 not going to come with some type of a -- I mean I was kind
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| } 2 of left feeling that we're going to come with this whole new 3 ~ approach about how they would want to see the standard 4 review plan written that kind of left me with a degree of ,
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| l 5 uncertainty of what it was going to be and what would the 6 impact be on us of trying to totally restructure an SRP. !
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| 7 I guess I'm looking for some feedback. Am I 8 correct that these are the areas you want us to focus on and 1
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| 9 not the fact that it's going to have to be -- your approach 10 would be to totally reformat it or whatever? I mean there's 11 a lot of work involved in this document, we're working on a 12 very short time-frame, and I'd like to kind of get a good 13 feel on what are the areas that we need to focus on to get a 14 final product in place in time to meet the schedules that 15 are imposed upon us, which have a very short time-frame Os_-
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| 16 involved.
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| 17 MR. SCHILTHELM: I'll say something from our )
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| i 18 perspective. We talked a long time about this yesterday. j 19 The philosophical issues on Chapter 5, I think ;
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| 20 we're pretty concerned about, the things about PRA versus 21 deterministic methods in criticality and safety. I think 22 those are major concerns, and I think we've talked about all 23 those quite a bit.
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| 24 Drew said earlier you were rewriting SRP Chapter 5 25 and were on a schedule to do that by mid-February?
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| 79 1 MR. PERSINKO: Gary said that.
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| ('')
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| \_/-
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| 2 MR. SCHILTHELM: I'm sorry. That's probably 3 something we ought to discuss, because it doesn't seem 4 productive for us both to go off on a rewrite and then come 5 back and it likely won't look the same. So, we probably 6 ought to spend some time talking about that, you know, how 7 is it productive to go forward?
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| 8 MR. VAUGHAN: Yes, I think that's something that we 9 really need to discuss. We've talked about a number of 10 options, but it's not clear to us what's the best option.
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| 11 Let me repeat again, though, what we just shared 12 with you was not a complete review of Chapter 5. We started 13 at the beginning and, relatively quickly, given the scope of 14 the information, went part of the way through it. I can't
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| () 15 remember where we stopped off.
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| 16 MR. GOODWIN: At the technical practices, 17 basically.
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| 18 MR. VAUGHAN: We stopped before we went through 19 technical practice, and there are, as Steve said, several 20 philosophical things that really don't have anything to do 21 with how you write the SRP, they're just kind of 22 philosophical issues that need to be resolved.
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| 23 The second set of the presentation, with the 24 reference numbers, points to a number of things that deal 25 with how the standard is written, and the most key one of ANN RILEY & ASSOCIATES, LTD. l
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| 80 1 those, probably, is this degree of redundancy that goes 2 through there, which is counter to the way that it's done in 3 the facilities and counter to the way that the information 4 is collected and we know is going to cause a problem, and it 5 also creates a situation because different sections are 6 written by different people, and even though each one of 7 them writes about training, their words about training are 8 different in different places, and we need to get some 9 consistency and some integration and consolidation. That's ;
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| 10 the biggest point.
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| 11 But I think, from our references, you'll be able {
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| 12 to see a lot of that, and then, of course, the other level
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| : 13. of comments that are in there that we looked at is there's a 14 number of places in there where it seems like that there's
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| (; 15 too much prescriptive information, too much how-to that is 16 very specific, and in fact, as a result of that, there's 17 some places that there are probably technical errors in 18 there because of the attempted degree of specificness that 19 is tried to put in the -- you know, the how-to kind of 20 instruction, and it may overlook or not include a number of 21 things that you all would want included. ;
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| 22 So, it's move a number of the requirements and 23 consolidate them in the license and really go after some of 24 this prescriptiveness. I mean I know there's a certain 1
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| 25 amount of detail you've got to do, but it should be able to f"- ANN RILEY & ASSOCIATES, LTD.
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| y 81 1 be prea nted a little bit more broadly so you can encompass gs 2 more of the things that you may run across than what it 3 currently does.
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| 4 MR~. GOODWIN: The focus, I think, should be more on 1 5 performance criteria as opposed to methodology, 6 specification of the how-to's and methodology, and I think 7 that's where it deviates at different places in there, it i 8 gets into the how-to, rather than providing performance l
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| 9 criteria, you know, what are you really trying to achieve?
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| 10 You know, tell us what you want us to do but not how to do 11 it, in essence, if it's not necessary.
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| 12 MR. PERSINKO: The same comment you made on the 13 rule earlier, you said make the rule performance-based and 14 take the how-to's somewhere else. I think Steve said that, 15 and now we're hearing the same comment about the SRP, make
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| (~)T 16 it perform as base, too.
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| i 17 So, essentially, you're saying take the how-to's 1 l
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| 18 out. '
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| l 19 MR. GOODWIN: To a large extent, yes. ;
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| j 20 MR. SHERR: Going back to what Liz was saying, at 21 the meeting on December 4th, concerns were expressed about 22 the prescriptiveness of the SRP, and we discussed, well, 23 okay, I mean the purpose of the SRP is to help the reviewer 24 work from the more general aspects that are in the 25 regulations-and all those things, so some degree of 3 ANN RILEY & ASSOCIATES, LTD.
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| 82 1 prescriptiveness is appropriate, and we were trying to get a r''N 2- better' handle in terms of what that level meant to you, and 3 one of.the suggestions there was perhaps you could provide a 4 revised chapter.
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| 5 It wasn't necessarily criticality. In fact, I 6 think the example was training that we identified at the 7 meeting, and there wasn't any commitment to do it, it was 8 just that, you know, you might be able to do that.
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| 9 But I think that's the -- I mean we'll give our {
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| f 10 best shot in terms of revising the criticality chapter, and '
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| 11 any specific comments that you have would help us in doing 12 that.
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| 13 But there's still this concern about this 14 prescriptiveness, and it doesn't seem to us a basic sin that l
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| 15 the SRP would be prescriptive. I mean if that was a
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| (} ]
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| 16 criticism of the rule, then we would understand that.
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| 17 So, we're trying to get a better understanding of 18 what degree of detail that we're talking about that seems to 19 be appropriate.
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| 20 MR. GOODWIN: I think it's the level that we're 21 concerned with. We realize that there has to be some, and 22 you can certainly give examples of, you know, how to do 23 things, but I think it's just the level of prescriptiveness 24 that we've got a problem with.
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| 25 MR. DAMON: This is Dennis Damon again.
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| 83 1 The crit specialists here in the agency have met
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| : 2. and read the NEI comments that were submitted, and I.just 3 thought I should share with you -- and any other crit 4 specialist should speak up if I misstate this, but my 5 impression is that we feel we understand what you're saying, 6 which things are too prescriptive and which things are not.
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| 7 So, I think that whoever undertakes this task I 8 probably, if he solicits the opinions of his fellow crit 9 specialists here, will not be too far off from what you're 10 seeking. I don't see this as being that -- as difficult as 11 is being presented here.
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| 12 MR. DAMON: Do you agree, Harry?
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| i 13 MR. FELSHER: Yes, I agree.
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| 14 MR. KENT: Was that a commitment on your part to l 15 proceed with the standard review plan and not expecting 16 industry to provide you with a straw man?
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| 17 MR. SHERR: I think what Dennis is trying to say is 18 that, yes, we will proceed with the revised standard review 19 plan chapter on criticality, and the request for the 20 possibility of industry providing an example for a revised 21 chapter just to get a feeling for what degree of 22 prescriptiveness is considered appropriate was kind of a I 23 separate request from the criticality.
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| 24 At the same time, I think we're suggesting that 25 any other specific comments that can be provided relating to i
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| 84 I 1 the criticality SRP chapter would be helpful in our 2 revisions to that.
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| (~N 3 MR. SCHILTHELM: I think I understood earlier this l
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| 4 is going to go on the web, as well, in the same manner, so 5 that we could get somewhat interactive with the criticality 6 people?
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| 7 MR. COMFORT: Yes.
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| 8 MR. SCHILTHELM: I guess, then, if you've got the 9 lead, it's incumbent on us to participate and get on there 10 and work with you to try to get to the end point on this.
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| 11 MS. TEN-EYCK: Definitely. We have a very short 12 time-frame. We're going to be moving to be responsive to 13 your comments the best we can, and we put it on the web, and 14 we're going to be looking for your input in a timely manner.
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| () 15 When we get to the point -- at certain cut-off 16 points, we're going to have to move with what we have, and 17 then we can, you know, address it through the public comment 18 forum that we will have when the proposed rule goes out, but 19 we're looking for your input, we're going to put it on the 20 web, we're going to try to address your comments the best we 21 can, and we need your feedback on whether we missed the mark 22 or we've addressed your concerns.
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| 23 So, it's a very interactive process at this point, 24 and timeliness, I can't stress it enough.
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| 25 MR. GOODWIN: In terms of this public rule-making, ANN RILEY & ASSOCIATES, LTD.
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| 85 I
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| 1 the_ eventual public rulc-making, what will be the comment ]
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| 7e ') 2 period duration? Will that be a 90-day comment period, or V 3 do you know at this point? I'm just curious. J 4 MS. TEN-EYCK: We would go out with what we would 5 normally consider an adequate public comment period, l 6 particularly since the interaction that we've had on this 7 rule -- it isn't like we're dropping something on you that 8 you've never seen before.
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| 9 I think that we're going to be going forward, and 10 there may be some issues that we still haven't resolved, and 11 we will try to identify those in the commission paper that :
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| 12 goes up, and the commission can make the decision on whether 13 they want to publish it or they want to send it back to us 14 to resolve, but I would think that a 90-day public comment
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| () 15 16 period would be adequate based on the fact that we have had so much interaction on this activity over a period of -- I 17 hate to think of it -- five years or more, and so, I would 18 think that 90 days would be adequate, but as I say, that 19 would be something that would go up and we would look at it 20 through our -- how we would address it through our normal 21 process.
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| 22 If you feel that 90 days is not adequate, then I 23 would like to know that, so at least we consider that in our 24 determination of what would be an adequate comment period.
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| 25 MR. GOODWIN: I think our main concern is being ANN RILEY & ASSOCIATES, LTD.
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| able to do as much as we can before it is noticed or
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| -~ 2 published for rule-making, but I know -- at least I don't
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| 3 think we'll be able to get through the entire SRP.
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| 11 So, hopefully, we'll have adequate time, then, to 5 complete the effort, you know, during the public comment 6 period.
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| 7 MR. SHERR: I think, if I recall correctly, the 8 standard comment period is 75 days, and I think that's what 9 we had in our draft proposed rule, and I can check that, but 10 also, whatever we go out with in terms of the comment 11 period, there's always -- a press can always be made for 12 extensions.
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| 13 MR GOODWIN: Right.
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| 1 14 MR. KILLAR: Liz, if I could go back briefly to a
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| ''N 15 discussion of the overall philosophy of the SRPs, I think (d 16 that we've captured quite a bit already in some of our 17 . general concerns -- that being the redundancy, the 18 consistency of definitions.
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| 19 One of the things that we'd recommend is that you 20 have a final editor, so to speak, someone who reads it from l 21 front to back for consistencies and things.
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| i 22 I know -- or it appears to us that the chapters 23 have been written by different individuals and so you have a 24 different tone sometimes in different chapters and different 25 verbiage and what have you, and so, while it may not be !
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| 87 l 1 intended, but you sometimes read the same thing in different l
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| ,3 2 chapters and it comes out with two different intentions, and
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| 3 so, I think one of the things that we're concerned about is '
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| 4 consistency throughout the standard review plan.
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| 5 The other item that I mentioned at the December 6 workshop that I'd like to reiterate here -- and that is 7
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| reflected in Chapter 5 -- is that the integrated safety 8 assessment drives a program, and so, when you get into 9 criticality safety, it's criticality safety as identified 10 from the ISA.
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| 11 When you read the section reflecting to the ISA 12 summary in here, it's along the lines of looking at what --
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| 13 d<>es the ISA capture this, capture that, or capture this, 14 and it should not -- from our perspective, not be that way.
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| (~') 15 It should be the ISA has identified this, this,
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| %.J 16 and this, how is that reflected in the safety programs, and 17 so, that's the nuance, I think, that we need to really 18 capture somewhere, and that was part of what we were trying 19 to say back in December, is that all the various programs, 20 whether it's radiation protection, quality assurance, 21 training, human factors, what have you, is dependent upon 22 the ISA and how the ISA determines the need for those 23 various programs, and so, the need for the program, the 24 depth of the program is generated by the ISA, and I'm not 25 sure that's being captured in the individual chapters.
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| 88 1 MS. TEN-EYCK: We've had that comment before, and 2 .that's certainly something we're going to take into b
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| \~ l 3 consideration as we rewrite these chapters, yes.
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| 4 MR. ROTHLEDER: Burt Rothleder from DOE.
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| 5 I'd like to make a suggestion that, where you have 6 parts of the SRP -- many parts, probably -- that read 7 prescrptively and need to be prescriptive, you can use 8 diffusing language to point out that there are other ways of 9 doing this and perhaps give some examples or references. If 10 you do this, you can make it less onerous and less 11 apparently prescriptive. This has to be done, of course, 12 ' carefully, but I think this would help.
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| 13 MR. DAMON: This is Dennis Damon again. '
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| 14 I'd like to make a comment about your -- Felix's
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| ('% 15 remarks about consistencies of definitions, because I've
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| : k. !
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| 16 sometimes detected a misunderstanding of what the SRP is 17 trying to do in some cases.
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| 18 I totally agree, obviously, that definitions
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| ~
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| 19 should be consistent. It's also true that a strong attempt 20 has been made to be completely consistent with the language 21 in the ANSI standards.
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| l 22 However, you have to understand that what an SRP 23 is doing is giving guidance to a license reviewer on how to 24 apply those definitions to a particular case, and what we 25 have found in the past is that things like the double !
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| I ANN RILEY & ASSOCIATES, LTD.
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| 89 i
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| i contingency definition and even a term like items relied on l 2 for safety, if you hand that to a license reviewer -- two O\2 3 different license reviewers and apply it to the same system, 4 you may get very different results because of some things 5 about the definition that are not clear.
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| 6 For example, the double contingency definition is 7 only about a sentence long, and there are many situations 8 that require interpretation.
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| 9 So, the goal of this standard review plan, in my 10- view, should be to make interpretations of definitions.
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| 11' Now, anytime you make an interpretation of a 12 definition, someone will disagree with you and say no, 13 that's not what it meant and you are redefining the term. I 14 do not regard it as redefining the term. I regard it as 15 this is an interpretation of that definition.
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| 16 So, any definition is a broad idea, because it 17 uses a small number of words, and then, when you interpret 18 it, yes, it becomes more specific, and some people might say 19 that you've redefined it,-but that's not what we're doing.
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| 20 We're interpreting it in the process of applying it, and in 21 fact, you are forced to do that. That's what a license 22 reviewer does.
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| 23 If someone says he's got double contingency and 24 he's reviewing it, he's making the judgement, does he or 25 does he not have double contingency, and in order to do
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| 90 1 that, he has to apply certain interpretations of certain gsg 2 things, in particularly how unlikely is good enough, and 3 what we would like to see happen is that each license 4 reviewer that does this does it in the same way, and we have ;
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| 5 found~in the past that license reviewers, because they did 6 not have an SRP, were doing completely different things 7 without any justification, and so, we do need to clarify 8 definitions, and there will be clarifications and 9 interpretations of definitions in the standard review plan.
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| 10 MR. SHERR: Anymore comments at this point? l 11 [:No response.]
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| 12 MR. SHERR: Well, this may be a good breaking 13 point, and what we talked about earlier is that, when we 14 reconvene, we can go back to agenda item, I guess, 3 (c) and ,
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| l
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| /''N 15 3 (d) , and any further comments that there are on the O 16 discussion draft rule language we can discuss at that time, 1
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| 17 as well as, on reflection, any further discussion, or if 18 -there are anymore specific comments you want to provide on 19 the SRP, the criticality chapter in the SRP at that time. j 20 Does that make sense?
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| 21 Okay. It's now 20 to 12. I think the suggestion 22 was that we break for about an hour-and-a-half. So, maybe if 23 we reconvene at 1:15?
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| 24 My suggestion is, since a lot of people need 25 escorts to get up here, that maybe about 1:10, somebody ANN RILEY & ASSOCIATES, LTD.
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| 91
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| 'I would meet the group downstairs and bring everybody up at
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| -g-, 2 one time, if that's okay.
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| 3 Thank you.
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| 4 [Whereupon, at 11:50 a.m., the meeting was 5 recessed, to reconvene at 1:15 p.m., this same day.)
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| 6 7
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| 8 9
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| 10 11 12 13 14 16 17 18-19 20 21 22 23 24 25 ANN RILEY & ASSOCIATES, LTD.
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| E 92 1 AFTERNOON SESSION 2
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| 7-s [1:20 p.m.]
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| -> 3 MR. SHERR: If we can reconvene, I think we should 4 probably have just one person talking at a time so -- Mark 5 was able to get it all down. He probably is trying to 6 record everybody talking at one time here but -- okay.
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| 7 I think what we said before lunch is that we would 8 go back to agenda item 3 (c) and (d), which is essentially 9 any comments on the discussion draft rule language, whether 10 there's any additional comments on that at this point or 11 whether we have exhausted our discussion of that.
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| 12 MR. KILLAR: From the industry's perspective we 13 think we have fairly well covered that. I wouldn't think 14 that we need to spend any more time on that unless you had 15 some questions based on any of the discussion we have had 16 since that earlier this morning.
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| 17 MR. SHERR: I guess one question I have is there 18 seemed to be some -- there was one suggestion for a change 19 of language and another suggestion for putting a period and ,
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| 20 deleting the rest. l 21 Was there a unified view with regard to either of 22 those alternatives or is that something --
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| 23 MR. KILLAR: The unified view is that we would 24 like to see what the NRC's consensus is --
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| I 25 [ Laughter.] I ANN RILEY & ASSOCIATES, LTD.
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| 93 1 MR. SHERR: Okay. Okay, so we will consider both 2 those alternatives.
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| iO 3 MR. BIDINGER: Ted, I think that second sentence, 4 in that same sentence, that word " practicable" is a much 5 bigger issue than running a tutorial in the first sentence.
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| 6 MR. SHERR: Okay. Well, if that is -- we have 7 basically exhausted that one. )
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| 8 MR. LEWIS: There is something -- when Mr. Kent 9 was speaking he mentioned there's four areas in the rule --
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| 10 MR. KENT: Can you speak up?
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| 11 MR. LEWIS: Yes. When Mr. Kent was speaking, he 12 mentioned there was four areas in the rule where the 13 industry was unsure of how we were allowing graded levels of 14 protection to be applied and I didn't really understand that 15 explanation, so I thought that at least one of them 16 everybody has a copy of. It's now this (e) here that was 17 one of theirs you mentioned.
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| 18 MR. KENT: Let me find my copy of the rule.
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| 19 MR. LEWIS: Okay. I guess I wrote down that you l 20 thought it was more an issue of clarity than content.
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| 21 MR. GOODWIN: I think that's right. I think it 22 was the SRP that appeared -- the graded level of protection 23 possibly but we interpreted those four sections as allowing 24 that, yet it looked like it was subject to 25 misinterpretation, okay? Just needed further clarification ANN RILEY & ASSOCIATES, LTD.
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| l 94 1 in those four areas that Norm mentioned.
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| 2 MR. LEWIS: Okay. With that clarification, it
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| (-'-
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| 3 doesn't necessarily need to be a change to the rule 4 . language -- it's a change to the accompanying language and i 5 the statements of consideration are to the SRP.
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| 15 MR. GOODWIN: Well, particularly the SRP. Maybe 7 I'll let Norm speak to that, since he was the one that l
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| l 8 addressed that. !
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| 9 MR. KENT: I will speak but others are welcome to 10 help.
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| 11 Graded levels of protection for the controls, for 12 the items relied upon for safety, the first reference I made 13 was 70.60, paragraph (d) and the sentence in the rule says 14 that each engineered or administrative control necessary 15 shall be designated as an item relied on for safety, and 16 from a practicing standpoint we use controls the aggregate j 17 of which we say together may form an item of safety or an' l 18 item relied upon for safety but each separate one may be of 19 a lower level.
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| 20 If you look at your definition of criticality i 21 control system, it seems to me you introduce the opportunity 22 to do that where you say you can clump several controls l
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| 23 together to have the same overall function as an item relied '
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| 24 on for safety but yet this sentence seems to say that each 25 control I use which to support (b) and (c) shall be an item ANN RILEY & JSSOCIATES, LTD.
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| 95 1 relied on for safety. l 2 MR. LEWIS: It says each engineered or l f
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| 3 administrative control necessary to comply with (b) or (c) 4 or (d) shall be designated as item relied on for safety. i 1
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| 5 I guess I don't see a problem with that statement. !
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| l 6 If it is necessary.-- it has to be important to safety. I 7 MR. EDGAR: Norm, would it work to say each 8 engineered or administrative control or family of controls? ,
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| 9 Would that get us where we want to be?
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| 10 MR. LEWIS: Or set of controls or control system.
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| 11 MR. KENT: Yes, I think that I may want to use 12 some controls which of themselves may not be safety 13 significant as we call them or items relied upon for safety 14 in that the failure of that single control is not going to ,
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| l g 15 cause (b) or (c) to happen and so I don't know that I am J 16 going to designate that particular control as an item relied 17 on for safety, but I may have a few of those family would 18 be. Is that still unclear?
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| 19 MR. LEWIS: A little bit. Maybe if there is an 20 example.
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| 21 I think I see what you are saying but to me either 22 it is required to meet (b) and (c) or it is not. If it 23 isn't, it is not important to safety and it is outside the 24 scope of the rule but if it is then we are not preventing yo 25 from putting other things under the management measures or ANN RILEY & ASSOCIATES, LTD.
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| 96 1 safety program but we are specifying that these that are 2 necessary for (b) and (c) do have to be under the safety 3 program.
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| 4 MR. DAMON: This is Dennis Damon. I think what 5 the source of the difficulty here is is the fact that in an 6 individual plant very often a safety system, a safety 7 program or whatever is set up using terminology to identify 8 certain pieces of hardware, and the terminology used might 9 be safety related equipment or something or other, some kind !
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| i 10 of .tenan like that, and certain pieces of equipment have that i I
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| 11 label put on them and the plant, once the label is put on 12 that piece of equipment, then a bunch of specific things are 13 done by the plant to handle that particular thing.
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| 14 My own personal understanding of the intent of the
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| /"N 15 rule language is that it is not trying to describe such a
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| (. 16 system. It is simply saying -- it is describing that same 17 idea at the higher level of generality, which is if 18 something is something you rely on for safety, you need to i
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| 19 do whatever you need to do to make that thing work.
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| 20 It's really a very benign statement. It's a very 21 harmless statement in one sense. It simply says it is not i
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| 22 good enough that you installed the thing originally. You 23- have to maintain it and it has to continue to be operational 24 and so on, but it is not trying to describe that you need to 25 label everything " item relied on for safety" and put a
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| 97 1 sticker on there and then everything that has that sticker
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| ,- 2. has to have exactly the same thing done to it. That would
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| (
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| 3 be a separate system.
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| 4 That would be'one part of the management systems 5 that'are applied to achieve this more generic concept, which 6 is you have to do something to make sure things work, so 7 that is my perception of it, but I think many people don't 8 have that perception. I think people have a perception we 9 are trying to map the rule language -- the rule language is 10 trying to create an arbitrary management system like that 11 and then you have got to map your system and your plant onto 12 that scheme, and I think that is absolutely impossible 13 because each plant has their own scheme. They identify 14 things and they apply different criteria to what hardware 15 goes in their particular categorization scheme, and any map 16 we created here would not fit everybody. It would be lucky 17 if it fit anybody.
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| 18 MR. GOODWIN: I think where Norm is coming from is 19 we may have a barrier, if you will, which is a set of 20 controls that can preclude or should preclude criticality 21 and maybe one or two of those controls may not be what we 22 designate as safety significant but in the aggregate, you 23 know, they all or would be defined by us as an item relied 24 on for safety but an individual control might not be.
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| 25 I don't know if I explained it very well. We may ANN RILEY & ASSOCIATES, LTD.
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| 98 1 hnv0 to think about it.
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| ,_ 2 MR. DAMON: All I am saying is even if you have
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| ( )
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| N/ 3 got something like a nut or bolt in a thing, you know, and 4 you are relying on that to hold the thing together -- if the 5 nut falls out the thing doesn't work -- even at that level 6 you have got to use a good enough nut. That is what the 7 rule is saying. You have got to use a good enough one. If 8 they are breaking every week and they don't work, they're 9 always falling out, you have got to get better ones.
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| 10 So it is not trying to create this map and say 11 every nut and bolt has to be an especially identified item.
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| 12 It is not -- that's my view -- that it should not try to do 13 that because we cannot succeed in creating a scheme like 14 that and saying you have got to replicate this scheme in r3 15 your plant and make you completely reorganize the way you do
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| ('-) 16 this function at your plants, and I don't think -- that is 17 just my personal view -- I don't think that is practicable 18 to make all the licensee follow some NRC generated, 19 prescriptive system like that.
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| 20 I think over in NRR where the systems they work 21 with are more similar, they can kind of do that, but I don't 22 think it works here.
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| 23 MR. KILLAR: Yes. I guess maybe an example --
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| 24 this is purely hypothetical -- is that you have a density 25 gauge which have two probes on it feeding into one meter.
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| 99 1 You put the high concentration of quality assurance on the
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| -s 2 meter because that is your reliance, but you don't put as 3 much a reliance on the probes because they have two 4 different probes providing the impact and so the concern is 5 that as you read this in a literal interpretation you have 6 got to put as much quality control or management on each of 7 those individual probes as you with the meter because that's 8 part of the same system, and so I think that is kind of, you 9 know, a nuance here is a difference of interpretation.
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| 10 MR. DAMON: Yes, I think the thing that is more 11 germane to this thing, that actually needs to be listed or 12 clarified in the SRP is what gets submitted that constitutes 13 a list of safety controls. In other words, the degree of 14 specificity in that, because in that list nobody wants to
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| (''g 15 see a list of every nut and bolt in the plant.
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| O 16 In other words, you could take everything in the 17 plant that is a hardware item that is part of a safety 18 system, break it down into every single component and list 19 it. That is not what is intended.
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| 20 It was intended to be listed at the system level.
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| 21 There would be one or two line items for each process in the 22 plant, you know. That is the level of specificity there.
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| 23 It comes out of the ISA. You do the ISA, identify the 24 accidents. Each accident will have like a couple items on 25 it relied on for safety and that it failed -- you list i
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| 100 1 thosa. Thny are in the description of event there. That is 2
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| what is wanted in the.way of being a list but in this other O
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| \_ ,/ 3 sense, see, that's why I say the language can be very 4 confusing.
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| 5 In the sense of the rule, each nut and bolt in 6 there is an item relied on for safety -- you know what I mean? It's something relied on -- literally it meets the i
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| 7 8 words of what the rule says. It is something you are 9 relying on for safety.
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| 10 (
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| MR. MENDELSOHN: If I could add one thing to that. l 11 This is Barry Mendelsohn.
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| 12 If you go back to when we first started to 13 identify, a number of years ago, one of the things that we 14 were trying to do in the industry was supportive of this was 15 to find some way to say what parts of the plant the NRC was
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| \- ' 16 not interest in, and you could do whatever you wanted to- !
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| 1 17 there, that it was not something that would concern us. I 18 remember there was something special about the fire 19 protection systems and the fact that we only had interest in 20 certain -- fire protection over certain parts of the plant, 21 not the whole plant,-and I think that is what we are trying 22 to do here when we say items relied on for safety. Those 23 are the things that we are going to be interested in.
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| 24 Now the level of interest -- I mean there's some 25 things that may be part of the safety system that, as Dennis ANN RILEY & ASSOCIATES, LTD.
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| 101 1
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| was saying, may b2 a fairly insignificant piece, but it is
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| ,im 2 still part of the safety system. So, you know, I am not 3 talking at what level quality assurance we need. That is 4 not my issue. I am just saying that there are certain 5
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| things that we say yes, these are safety -- part of the 6
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| safety system therefore NRC has some interest in it. The !
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| 7 other stuff we are not interested in.
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| 8 MR. SCHILTHELM: Let me try to put this in -- I 9
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| don't know -- in perspective and ask the NRC when we do an 1
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| 10 ISA we are going through systematically and identifying all 11 the items relied on for safety, okay? They may be nuts and 12 bolts or they may be very elaborate engineered controls in 13 my monitoring systems, but we are identifying all of them 14 essentially and some of the other licensees might not agree 15 with me, but this is what we are doing at B&W.
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| 7 t
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| V) 16 Now when we go into the risk part of this equation 17 and the graded approach, we have to somehow identify the 18 importance of each of those items and the words in the rule 19 and in fact the example of the ISA that is in the Standard 20 Review Plan has a classification scheme -- Class A, B and C, 21 okay?
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| 22 I think what we are struggling with is we want to 23 hear clearly from the NRC that all these items relied on for 24 safety because they relate to criticality safety don't fall !
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| i 25 into a Class A, that you acknowledge that there is a whole i
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| 102 1 spectrum of items relied on for safety as they relate to 2 protection, criticality protection of a given system.
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| 3 If you apply double contingency properly, we don't 4 think there's probably a Clase A criticality control. There 5 are probably a lot of Bs and there may be some Cs. Now 6 whether it is A, B, C, 1, 2, 3, whether it is A, B, C and D, 7 that whole issue is pretty confusing and I don't think any 8 two of us would agree on it, but that is the concept I think 9 that we are concerned about, because this whole graded 10 scheme lends itself to a classification scheme that we are 11 going to have to develop and implement.
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| 12 I guess I would like to hear the NRC's view on 13 that, or if.I have misspoken for the licensees, some of 14 their views, because there's a lot of confusion in that 15 point for us.
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| 16 MR. LEWIS: I can't say what NRC's view is but I 17 agree with everything you said. As a matter of fact, I was 18 just leafing through the next section that you referenced 19 and we included language ia the rule, SOP aside for the 20 moment, because it hasn't been updated yet, but in the 21 update that we put on the Web for the rule we had a phrase 22 in 70.62(a) that said the safety program may be graded such 23 that management measures applied are commensurate with the 24 items for reduction in risk - " item" meaning each item 25 relied on for safety.
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| 103 1 The reason I asked the initial question was I 2 couldn't see why that was inconsistent with what --
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| 3 MR. KENT: With what I said.
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| 4 MR. LEWIS: -- what industry's initial problems 5 were.
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| 6 MR. SCHILTHELM: I think what Norm said this l I
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| 7 morning was, when he read those four points, we think that's 8 what those four points say but we are nervous about it and 9 we are just trying to make absolutely certain that that is 10 what we are reading I guess was the point.
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| 11 MR. GOODWIN: I was just saying you may want to go 12 back and look at the words to see that they are as clear as 13 possible to eliminate any possible misinterpretation. At 14 least that's the way we interpret it. We hope the 15 inspectors and license reviewers will interpret them the 16 same way.
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| 17 MR. FREEMAN: I think the suggestion -- the 18 confusion that we talked about was that that paragraph that 19 Rob just read under 70.62 (a) (1) , the only mention of graded 20 approach is there. It is not in 70.60(d). It is not in 21 70. 60 (2) (d) where you thoroughly get into item relied upon 22 for safety.
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| 23 It's somewhat disconnected with the mention of j 24 item relied upon for safety, which would lend itself to 25 confusion -- is that accurate?
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| c 104
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| 'l MR. KENT: Yes.
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| s .2 MR. FREEMAN: We felt it's all there. We felt we 3 needed to interpret and wanted to make sure we interpret it 4 correctly and the clarification request is just that.
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| 5 MR. SHERR: 70.60(e) cross-references to 70.62.
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| 6 MR. FREEMAN: Correct.
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| 7 MR. SHERR: Is that correct?
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| 8 MR. FREEMAN: It is there, yes.
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| 9 MR. SHERR: I guess the question, going back to 10 Rob's additional question, was in light of the words that's ,
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| i 11 in 70.62, is there still concern with the opening phrase in 12 paragraph (e) where it says "each engineered or 13 administrative control" -- does something need to be added 14 to that or not?
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| 1
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| % 15 MR. FREEMAN: I would suggest the addition would j
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| ~-l 16 go on 70.62(d) under management measures -- repetition of 17 what is in 70.62 (1) -- some link to the graded approach --
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| 18 maybe it is redundant.
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| 19 MR. LEWIS: I think I understand.
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| 20 MR. SHERR: Okay. Any other questions, comments?
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| 21 George?
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| 22 MR. BIDINGER: Just one word of caution. In 70.60 23 every time you use the word " control" it's clear. In the 24 nuclear fuel cycle where there is only one control, where 25 the double contingency doesn't apply. I think you need to ANN RILEY & ASSOCIATES, LTD.
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| 105 1 be very. careful in editing this document to make it clear l 2 that it is one or more engineered controls.
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| O 3 When you just say controls that means more than one, and that will put 4 your five-inch cylinder out of business in a hurry -- I mean 5 your 30-inch cylinder out of business in a hurry. Controls 6 are more than one.
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| 7 MR. LEWIS: Controls will be defined in the SRP.
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| 8 MR. BIDINGER: I was talking about the rule 9 though.
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| 10 MR. LEWIS: Right. I understand your comment.
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| 11 MR. GOODWIN: I couple of minor questions or 12 comments regarding the rule itself.
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| 13 In the purpose, and maybe I just need 14 clarification, you have included uranium enrichment and 15 enriched uranium hexafloride conversion and I thought of O' 16 course the GDPs are licensed under Part 76. Why were they 17 included in that section?
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| \
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| 18 MR. BIDINGER: LES would have been under Part 70.
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| 19 MR. GOODWIN: Okay. I should have remembered 20 that.
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| 21 The other one -- in the critical mass of SAM under 22 70.4, it would basically reference the critical mass numbers 23 at the 4 weight percent'and I think every one is licensed at 1
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| 24 up to 5 or 5.1. Is there any reason to stick with the 4 25 weight percent as opposed to, say, a 5 weight percentage, ANN RILEY & ASSOCIAT2S, LTD.
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| 106 1 may be more representative or better representative?
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| g-wg 2: MR. VAUGHAN: I think there's been some new data d 3 generated that helps fill in the gap at 6 percent actually 4 up to around 10.
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| 5 MR. GOODWIN: I think you're right.
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| '6 MR. LEWIS: I might be wrong but I think the 7 origin of those numbers was Part 150, which is our Agreement 8 . State regulation where it says if you possess less than that 9 material, that amount of material, then you are not subject 10 to NRC licensing in an Agreement State.
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| 11 MR. SHERR: One of the subobjectives of the rule, 12 when we tried to parse it this way, was trying to not impact 13 the Agreement States -- trying to make sure the rule didn't 14 affect them.
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| /' 15 MR. LEWIS: No, I think you are correct. I think i'
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| 16 that is where it does come from. Why the 4 percent is 17 called out -- the only reason I can think of is that there 11 8 - was analysis available to permit them to determine that 19 number and it gave additional flexibility to people who were 20 trying to determine whether they complied or not, but I 21 really don't know why that was called out specifically.
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| 22 MR. GOODWIN: Just curious.
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| 23 MR. SHERR: Okay. So I guess then we can shift 24 back to Agenda Item 4 and I guess -- are there any other 25 comments industry wanted to make on the Standard Review Plan 1
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| I l 107 1 or specific references or did we accomplish that already?
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| 2 MR. KILLAR: I think we gave you all the 3 references -- we passed that information on as far as the 4 references from the slides.
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| 5 MR. ELLIOTT: Yes, we did.
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| 6 MR. SHERR: I assume the question is whether there 7 is any clarifications we want -- one reaction that I took 8 from your comments this morning were of two varieties, some 9 that were specific to the Chapter 5, but then some of them 10 were ones that would apply to other chapters as well, like 11 the notion of cross-referencing and that type of thing.
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| 12 MR. KILLAR: I might point out that at this point 13 in time what we have decided is it's apparent that we are 14 going to wrap up today. We won't need tomorrow and so what m
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| ) 15 the industry is going to do is we are going to get together (G
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| 16 back at our offices tomorrow and go through Chapter 5 in 17 detail, and annotate the Chapter 5 SRP, and provide you the 18 annotated Chapter 5 with all our comments in detail.
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| 19 We anticipate we can probably get that in to you 20 hopefully early next week to try to help you as you are 21 developing the rewrite of Chapter 5.
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| 22 MR. SHERR: Okay, good. That would be helpful for 23 us to consider those changes.
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| 24 Okay -- well, maybe we are at the closing part of 25 the meeting.
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| 108 1 MR. PERSINKO: Let me ask one question on r'% 2 historical _ events of significance that we talked about 3 earlier.
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| 4 What would be your thoughts if -- part of the 5 problem as I understand it are, you know, trying to make 6 it -- it will take effort to reformat it and put it in a 7 form that would be able to be sent in. What would you think 8 of saying "The following events were considered significant 9 and were factored into our ISA and they are as follows" and 10 just list the reference numbers to ones you looked at, and 11 then you say "The following events were not reportable but 12 we factored them in anyway since internally we think they 13 are of significance for another reason."
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| 14 You just give references. You don't have to
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| /''\ 15 report them.
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| d 16 MR. VAUGHAN: It doesn't do a thing for safety.
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| 17 MR. PERSINKO: No, but it lets us know that you 18 have sorted out which ones you think are significant and 19 which ones you have looked at in your ISA.
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| 20 MR. ELLIOTT: Drew, the AICHE Handbook requires 21 you to go back and look at all that.
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| 22 MR. PERSINKO: Yes.
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| 23 MR. ELLIOTT: So we look at internal events, those 24 that are reported to you, and those that are not. We also 25 look at industry-wide events, probably through the DOE gs ANN RILEY & ASSOCIATES, LTD.
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| 109 1 weekly summaries and things like that, so it's very 2 difficult to go back and -- I mean I can ask you for a list 3 of all my 9101s and 7050s and then hand it back to you.
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| 4 MR. PERSINKO: Yes, but not all of them are viewed 5 as significant by you. Maybe only a subset are. I don't 6 know.
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| 7 MR. BIDINGER: But the process should be that they l 8 look at them -- look at all of them for significance and 9 incorporate lessons learned from their looking.
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| 10 MR. PERSINKO: Right and I am saying we looked at 11 them and we consider the following ones to be significant 12 and we have factored those into our ISA.
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| 13 It would eliminate the work of reformatting or 14 rewriting --
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| 15 MR. VAUGHAN: Why don't we just commit to include
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| )
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| 16 those in the conduct of the ISA.
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| 17 I mean I don't know what everybody else is doing, 18 but our ISAs that are done on operating processes, as 19 opposed to something new generally list all of the unusual 20 incidents that have gone on in the near term, at least 21 probably within the last five years that deal with a 22 particular process that is being studied, and that is where 23 it needs to be incorporated and that is where it needs to be 24 -factored in and it really doesn't need to be factored in ,
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| I 25 whether it is significant or not. It needs to be factored ,
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| l (gj f- ANN RILEY & ASSOCIATES, LTD.
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| i 110 1 .in and the ISA needs to consider that information, 2
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| 7 It is going to be very hard to write for example a 3 definition of what-the significant ones are, and on the 4 other hand, that information ought to be available to the 5 team that is doing the ISA. That is where it is valuable in 6 terms of making the improvement or potentially making the 7- improvement into the safety, and the list is way after the 8 fact and the list is a list, and 10 years is a long time to 9 go back and recompile a list.
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| 10 But at least in our case the ISAs for operating 11 process generally list all of the unusual incidents that 12 were considered by the team and what impact that might have 13 on the safety of the process.
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| 14 MR. SHARKEY: I also think that it needs to be
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| ~}
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| 15 taken in context with the ISA. The team spends a lot of
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| %/
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| 16 time talking about them and not everything they discuss is 17 captured in the written portion of the ISA and they may get l
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| 18 down to a little detail where turning a nut or a bolt causes l l
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| 19 people to scrape the skin off their knuckles and suggestions 20 in those areas.
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| 21 I think you have got to take it in context with 22' the ISA and it is part of the ISA for most of us anyway.
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| 23 MR. GOODWIN: Well, it's like Mark said, the 24 standard practice I think with the AICHE as a process is to 25 consider all the events, process upsets, unusual, whatever, ANN RILEY & ASSOCIATES, LTD.
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| 111 1 and the other thing is 10 years old -- the circumstances may i t
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| 2 be different enough, you know, with changes to your process, !
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| \ -
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| 3 some things that happened 10 years ago, even five years ago I 4 may not even be relevant anymore. )
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| 5 So I think it is probably best just to follow the 6 standard process and really try to take that in zero base 7 and consider everything that can happen that is unique to 8 that particular process.
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| 9 I think what we are saying is we just don't really 10 see any value added to that step and it does -- l 11 MR. ELLIOTT: If the goal is to ensure that we 4
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| 12 look at it, we can commit to look at them. We can also tell 13 you that we have looked at all of them, but it doesn't seem i
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| 14 logistically prudent to send you a list of significant
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| ("'
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| 15 issues as your stated goal was to make sure that we have l i
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| 16 looked at them. '
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| i 17 MR. DAMON: See, I'd like to confirm what Charlie I 18 Vaughan said. He said something that may have slipped by 19 people, which is that in doing the PHA you do consider all 20 events that have happened in that system and it is not just 21 the significant ones. I mean significant in the sense that, 22 boy, that was close -- it was very close to a criticality.
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| 23 You consider failures of relatively minor things 24 th:+.t would be anything that would contribute to anything you 25 would want to prevent and so you don't want that list and ANN RILEY & ASSOCIATES, LTD.
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| 112 1 the other list will be only the really big ticket items and fs 2 so I kind of tend to agree with the industry that the thing 3 that really gets them to the safety is to make sure that in 4 the process of doing the ISA that they actually have the 5 operations staff there and the engineer for that system 6 there, the guys that have been in the plant for the last 7 five years.
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| 8 That is where I think you get the knowledge from 9 is requiring the team to be made up of people who actually i
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| 10 have been there before who would actually have this 11 knowledge. I 12 MR. VAUGHAN: Right, and the performance -- a 13 clear performance requirement that that information and 14 knowledge will be used in the ISA. A list just doesn't get 15 it.
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| s 16 MR. DAMON: I mean as an example you could 17 require -- we only require that the operations staff -- that 18 there be a representative operations staff on the ISA team 19 and it occurs to me in this context that that is not quite ;
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| 20 good enough, is it, if that guy has only been at the plant 21 for one year?
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| 22 There's something sort of related to this that 23 Norm Kent said that I want to make a comment on. I was 24 looking back over my notes here and he made a comment about 25 taking credit for various things in assessing the likelihood ANN RILEY & ASSOCIATES, LTD.
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| 113 1 of failures that would get you to accidents, and mentioned 2 taking credit for external events and the low likelihood of 3 that, and you mentioned taking credit for process deviations I 4 and the fact that those deviations might be unlikely.
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| 5 I just want to say that's a trick area because 6 that way we would view that, it's perfectly acceptable to do 7 that, but by doing so, by saying the reason this sequence of 8 events is unlikely is partly because the process upset that 9 initiated it is an unlikely condition -- the minute you do 10 that, that means that process upset -- whatever it is that 11 makes that process upset unlikely is an item relied on for 12 safety and it should be in the list of things, and the 13 reason I say that is because unless you do that, the next 14 step that should happen is it's in the list.
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| r''g 15 The next step that should happen is it gets put V 16 into-some part of your safety management system -- say it's 17 either listed in an NCSA or it's put in a configuration 18 management system, probably in an NCSA as something. This 19 process should have this certain upset be a rare event and 4
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| 20 the reason it has to be done that way is if you don't do 21 that the next time somebody does a change to that process, 22 they can change that process in such a way that that process 23 upset is no longer an unlikely event, and the reason I am 24 sensitive to this is that is the kind of thing -- it's 25 really quite common that there are processes that have quite ANN RILEY & ASSOCIATES, LTD.
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| i 114 1 frequent process upsets like solvent extraction processes.
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| 2 Some of them, the way the controllers work, they 3 have process upsets relatively frequently, and so if you are 4 actually relying on that low likelihood of that, you have 5 got to declare it and then write it down somewhere so that 6 some guy doesn't come in and change the controller valve and 7 put in something that now that process upset is a much more 8 frequent event.
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| 9 MR. FELSHER: Harry Felsher, NRC.
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| 10 I would also like to point out another reason why 11 the historical information is very important not just to the 12 NRC but to the licensees is exactly the same reasons said 13 before. New people come in. They need to know what is 14 going on in that system and they may have arrived after the
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| ( 15 ISA was completed and yet they are going to be running the U) 16 operation or overseeing the operation and still would need 17 to go back and know what are the historical events related 18 to that operation, and then for the license reviewer to know 1 l
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| 19 perhaps they have only been on the job a short period of 20 time and had not been involved with that licensee -- they 21 may not know what has been going on for the last five years.
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| 22 So I think there's two reasons for the historical 23 information. One is to develop the ISA and the other is 24 ongoing training of personnel associated with the facility, 1
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| 25 both the NRC and the licensee.
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| I f"N ANN RILEY & ASSOCIATES, LTD.
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| 115 1 MR. MANNING: Harry, this is Cal Manning. I'd 2 comment on that that generally the people in the plant are 3 going to be trained to internal documents in the ISA itself.
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| 4 Very rarely will the process operator or a new manager be 5 referring to the NRC license with that information.
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| 6 MR. VAUGHAN: In fact, the ISA drives training in
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| .7 a lot of respects and if you get to a point that the ISA has 8 to be redone, _then you are going to pull out the old one, 9 which as we have said automatically includes the events that 10 have been related to those particular processes, where you 11 are dealing with operational events, and so the people, new, 12 old or indifferent, that are involved in the update or 1
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| 13 review of the ISA are going to have that information plus 14 any information that has been generated since then to O 15 consider.
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| N-) l 16 MR. BIDINGER: Harry, there is nothing sacred 17 about 10 years. You were in our short course a year or two j 18 ago and we were talking about accidents that happened 40 19 years, 50 years ago -- there is nothing sacred in terms of 20 training in the last 10 years. The training -- if you have 21 a good ISA it wraps up all of the lessons learned from 22 ancient history into the document and there is nothing 23 sacred about 10 years in terms of training, j 24 MR. FELSHER: I think, perhaps it's a mistake, but 25 I thought the 10 year was because we are currently going to ANN RILEY & ASSOCIATES, LTD.
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| 116 1 10 year license renewals, and that was just to capture what 2 has gone on --
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| 3 MR. BIDINGER: In terms of training and 4 informati'on being in the ISA, it's real old. I mean there 5 is nothing sacred about 10 years.
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| 6 MR. SHERR: George, we'll assume that you are 7 suggesting we change the rule language from 10 years to 40 8 years?
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| 9 [ Laughter.]
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| 10 MR. BIDINGER: No.
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| 11 MR. SHERR: I am just teasing.
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| 12 MR. BIDINGER: A good ISA incorporates lessons 13 learned. This could mean going back and if you have an 14 event you go back and do a portion of the ISA all over. You
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| (''
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| 15 don't wait 10 years to do it.
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| 16 MR. SCHILTHELM: We are not suggesting that the 17 staff doesn't need a list, whether it is a 10 year list or a 18 40 year list, doesn't need a comprehensive list to 19 understand their business. What we are suggesting is the 20 license application is not the place for that list. We are 21 almost maybe bridging over into this discussion about where 22 the ISA summary sits, but it is not the license application.
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| 23 Our view of the license application is those 24 safety commitments we make that we can't change or deviate 25 from without a license amendment.
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| 117 1 A simple list of those things that have occurred 2 over the last 10 years doesn't fit that mold.
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| 3 MR, SHARKEY: You have got the list already, 4 Harry, in your docket file. l 5 MR. ELLIOTT: I guess it appears to us that I am 6 going to give you a list of events that you already have and j 1
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| 7 that in no way is going to assure you that I have put it 8 into my ISA, that action.
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| 9 If we tell you and commit in our license 10 application that part of the ISA process includes these 11 types of things, I think that gets you there and you can 12 look at the PHA documentation and find it, but for a 13 licensee to give you a list of significant events that have 14 occurred in the facility in the past 10 years, a list which
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| /''i 15 you can get right off of your computer doesn't seem to do b
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| 16 anything to anybody but cause a little extra burden and put 17 a little superfluous information in a license application.
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| 18 MR. FELSHER: Well, let me ask you, at your site 19 would you have that type of document?
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| 20 MR. SCHILTHELM: If you can't find it, we'll send 21 it to you. It just doesn't belong in the license 22 application.
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| 23 MR. FELSHER: There's two questions here. One is 24 do you have that information? And I think I am gathering 25 here that you do have that information at the site and it is q' ANN RILEY & ASSOCIATES, LTD.
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| 118 1 relatively kept up to dato?
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| ~~ : 2 MR. SHARKEY: When you say up to date, I mean you
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| ~
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| 3 . don't go back and update whole event reports. I don't j 4 understand.
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| 5 MR. FELSHER: No, no, not the event report, call 6 it a summary of the significant events in the last "x" i
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| 7 number of years. l 8 MR. SHARKEY: No, I don't have a summary of it, 9 nor do I have a file with all events. I have no 10 record-keeping requirements for things that happened 20 11 years ago or 10 years ago. There's no record-keeping 12 requirement for those events.
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| 13 MR. PERSINKO: So how would you factor it then !
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| 14 into your ISA?
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| i
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| . 15 MR. SHARKEY: A lot of it is -- it may not go back 16 that far. It relies on people's institutional memory.
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| 17 I would say going forward from a couple of years 1 18 ago, yes, there's a pretty good document trail.
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| 19 MR. PERSINKO: Well, what happens if the person 20 retires? I mean you are talking institutional memory now.
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| 21 MR. SHARKEY: To go back in history it's what we 22 got.
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| 23 MR. VAUGHAN: The problem I suspect all of us 24 have, if you go back far enough in history the expectation 25 and need was not as well appreciated as it is today, so if ANN RILEY & ASSOCIATES, LTD.
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| 119 1 you go far enough back~into history of these plants we 2 didn't understand the importance and we didn't have systems 3 to go some of the things we do today.
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| '4 In the last five or six years, you'll find the 5 quality of information that we have, and even though there 6 may not even be long-term retention requirements that we 7 pretty much have available all of the time, that is of the 8 quality that we are talking about of being a benefit.
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| 9 Actually, if you took our plant right now and 10 asked us for a list of events 10 years back, significant 11 events, we might not have the wherewithal to go quite that
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| '12 far back and be accurate at it because of the lack of 13 records and the lack of systems that were in place, but as 14 move toward this integrated safety business, and industry 15 has been to a certain degree moving in that direction before 16 all of the i's are dotted and t's are crossed, we have 17 recognized a number of pieces of information that are 18 important to that process and important to the safety at our 19 facilities and while our systems may not be perfect yet, I 20 think most everybody is moving in that direction, so that is 21 one that is hard to talk about.
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| 22 We have to remember that this industry has evolved 23 over a number of years and what we say about today and what 24 we want today is maybe significantly different than the way 25 it was 10 or 15 years ago.
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| fg ANN RILEY & ASSOCIATES, LTD.
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| 120 1 MR. FELSHER: Well, let me ask you, do they
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| (~)
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| %J 2 have--- does your plant -- more recently, can you go back 3 for two or three years now and pull out the most recent 4 reports.--
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| 5 MR. VAUGHAN: Yes.
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| 1 6 MR. FELSHER: --
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| and factor those in? j i
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| 7 MR. VAUGHAN: Yes. j l
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| 8 MR. FELSHER: So you have put that into effect for j 9 more recent events, okay. You can't go back 10 years but 10 you can go back whatever -- three, five maybe, something
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| -1 11 like that. )
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| l 12 MR. SHARKEY: Yes. !
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| 13 MR. GOODWIN: I'd say since the introduction of 14 the Bulletin 9101 -- I could certainly speak for rN l Westinghouse.
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| Q 15 i '
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| Our records are pretty darn good and we keep 16 active files. !
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| l 17 MR. SHARKEY: There's also things that occur that i 18 may not cross the threshold of 9101 reporting that I would 19 consider significant and do in an ISA and accident reports 20 is one of those. We fill out a lot of accident reports 21 during the course of a year. They may be very minor issues.
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| 22 Those are reviewed during the ISA process.
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| 23 MR. PERSINKO: Yes.
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| 24 MR. SHARKEY: In the context of the ISA.
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| 25 MR. SCHILTHELM: It seems to me we are almost
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| 121 1 offering more by suggesting that if there is a performance
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| (~} 2 requirement that we be reviewing historical information and C
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| 3 maybe 70. 62 (c) , as you have written it, needs something to 4 say that we do that as part of the ISA, that seems to offer l 5 more meat to the regulatory and safety performance than 6- having a requirement to put a list in a license application.
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| 7 I am a little confused why we are arguing --
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| 8 because it seems like we are offering more substance than we 9 are asking to take out of the license application.
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| 10 MR. FELSHER: Speaking as a licensing reviewer, if 11 I were unsure that you had taken that into account -- let's 12 say you make that commitment that you will do it in the ISA, 13 but how can I be sure if you are doing that without knowing 14 what those events are? Would I have to go to your site to
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| ( ) 15 take a look at this handbook of previous accidents to 16 determine that yes, you did, take those into account?
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| 17 MR. VAUGHAN: We'd verify that you need to come 18 look at the ISA and see how it was constructed and see what l 19 information was used and then go back to the unusual event 20 logs and --
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| 21 MR. SCHILTHELM: How do you know that we have had 22 the right team members do the ISA? I am not giving you a 23 list of those team members and there is an inspection 24 process --
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| 25 MR. FELSHER: Right.
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| 122 1 MR. SCHILTHELM: Trust and verify is the name of I I
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| 2 the game.
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| 3 MR. FELSHER: Trust possibly --
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| 4 MR. SCHILTHELM: You can't possibly verify every 5- single thing.
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| 6 MR. FELSHER: I am trying to understand, it seems 7 to me that you have some kind of document at the site. How 8 long it goes back_to or not, it is site-specific. What you 9 are talking about here is you don't want to put it in the 10 license application and I thought that you not only send the 11 license application but you also send other documents to 12 supplement the license application during the license 13 renewal. ,
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| 14 MR. GOODWIN: Harry, I think we are not 15 necessarily talking about a document, We are talking about
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| (. )
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| 16 a set of files that we have, okay? It's not like a single 17 document -- just to make sure that you understand that --
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| 18 but we do have a set of files that we maintain for, for the 19 most part, not only the 91.01 reportables but there may be 20 -others that are of somewhat lesser significance that may be 21 near-misses, precursors, things like that, that we choose to 22 set up a file for. That is the way we do it.
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| 23 MR. KENT: This is Norman Kent from Westinghouse.
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| 24 I agree with what Wilbur says. Though I can't 25 show you a piece of paper that shows what lessons I have ANN RILEY & ASSOCIATES, LTD.
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| i 123 1 learned from these incidents and how I applied it to the
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| /~} 2 different areas in the plant because a particular 3 significant accident in the uranium processing part may have 4 application elsewhere and I need to make sure those lessons 5 learned get applied, I think we should be able to show you 6 how we applied those lessons learned, not during the license 7 application review but during subsequent inspections.
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| 8 MR. KILLAR: If I could just kind of sum up here, 9 I don't think that we have a problem with wha't you are 10 looking for as far as the end-product is that you want to 11 make sure that we have incorporated lessons learned in our 12 Integrated Safety Assessment.
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| 13 What we have a problem with is the way you are 14 asking for that information we don't feel is going to be
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| ) 15 very helpful to you or us and so we don't feel it is a very 16 productive process, the way you are asking for that 17 information.
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| l 18 What I suggest you do is maybe go back and look at '
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| 19 the ISA section of the rule and try and capture this concept 20 of the_ISA section of the rule, and so you can get what you 21 are looking for and we can provide it to where it is more 22 meaningful and more productive for both parties. '
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| 23 MR. LEWIS: And I agree with you. We do owe you a 24 posting of 70.65, isn't it? -- yes. We are in the process 25 of trying to revise that, although I wouldn't characterize ANN RILEY & ASSOCIATES, LTD.
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| 124 1 that as the ISA portion of the rule, because that's the 2 contents of application.
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| 3 I kind of took what you just.said to mean t'en 4 stuff we have already posted but I don't think we wanted to 5 include the history in that portion. Not clear?
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| 6 MR. KILLAR: Let me -- what you are looking for is 7 to be comfortable that when someone has done their ISA they 8 have taken the lessons they have learned from past events, 9 upset, abnormal conditions, thinks along that line, and 10 factored that in so that they would potentially not have 11 that occur in the future. What you are looking for is a 12 list of those events so you can compare that list against 13 something, possibly -- or I would think the ISA -- to see 14 that they were factored in there, and what I am saying is
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| /~N 15 just giving you a list doesn't necessarily help you because b 16- you still have to compare it against the ISA.
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| 17 So my thinking is that if you look at how you can 18 incorporate that requirement into the ISA portion, you get 19 what you want and we are doing it already, so we get to the 20 same end but with less paperwork and generation of less 21 lists.
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| 22 MR. BIDINGER: I think for every required report 23 from the licensee there is a regulatory process for closing 24 out, seeing that corrective action has been taken, and that 25 is proper. For the things that are not reported, whether ANN RILEY & ASSOCIATES, LTD.
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| 125
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| : 1. they are. closed out or not is not the NRC's business. If it 9
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| 2 were.the NRC's business, it would be a reportable item, so I 3 think that the emphasis here on these little items at this 4 point in time is not a proper matter for discussion, but 5 since you have, the regulatory staff itself has a closecut 6 mechanism for the reports that do come in, this concern.that 7 they have been looked at every 10 years is not a realistic 8 issue. I think it should be dealt with accordingly.
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| 9- MR. PERSINKO: Let me ask a question. When you do 10 your ISA would the team go through each of the reports? Do 11 they go through those files one by one to see how it is 12 incorporated or would the ISA be done based on the memory of 13 the team members?
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| 14 MR. GOODWIN: I'll defer to Norm because he is our
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| / 15 ISA team member for the criticality program.
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| 16 MR. ELLIOTT: Let me speak to that for a minute.
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| 17 [ Laughter.)
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| 18 MR. ELLIOTT: When we do the ISA process you've 19 got a list of events that have occurred on that process and 20 you have got a list of events that have occurred throughout 21 the facility and industry that may have generic applications 22 to the process you are looking at.
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| 23 You are not going to go into that PHA and say, 24 well, incident da-da-da was looked at here in this accident 25 sequence. You are going to have an accident sequence that ANN RILEY & ASSOCIATES, LTD.
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| 126 1 was similar maybe to that incident.
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| 2 It may give you -- it is going to require you to
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| (~%
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| 3 look at it in that process and it is going to also give you 4 some better idea of likelihood of the failures of those 5 things that you are looking at in that process then you 6 would have had had you not had that information, but I don't 7 think, and I am just speaking for our paperwork, that you 8 are not going to be able to go back there and say, well, the 9 incident that was in weekly summary, DOE, and so on, was 10 covered under the accident scenario, and so on -- it's not 11 there.
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| 12 MR. PERSINKO: Would the team members review each 13 of the reports as a team to say, okay, this is a significant 14 one or not?
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| I~) 15 MR. ELLIOTT: Yes.
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| V 16 MR. PERSINKO: I mean you want to see if they are 17 doing it systematically or is it just done based on the 18 memory of team members.
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| 19 MR. ELLIOTT: We print out a list on the database 20 and I print out a list for the PHA -- we are looking at this 21 system, here are all the events that occurred in that system 22 and here are all the events that the different disciplines 23 think that have generic applicability to that type of 24 process from industry standards and we give them to the team 25 members.
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| 127 1 MR. VAUGHAN: We do the same thing. It comes out
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| (] 2 of our configuration management system.
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| 'V 3 Now the second piece of that question is do the 4 team members accept those at face value or do they go back 5 and dig into lower, you know, additional levels of detail, 6 and the answer to that is they do both, l
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| 7 I mean you can't write a formula that works for i 8 every single solitary thing because it depends on the team 9 members. It depends on the nature of the event -- a lot of a
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| 10 things -- and so we look at those lists and when you look at 11 some teams you may have people there that were intimately 12 involved in this particular thing that went on and they 13 don't have to go back and read the history.
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| 14 On the other hand, if you have a newer person that D 15 (d 16 wasn't there, then as part of their work with the team, they are going to have to look at the event and the summary and l
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| 17 closecut of that, and come up to speed in terms of what does '
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| 18 that mean for the overall process, so there is not a single 19 equation but that information has to be factored into the 20 ISA by the team. What that means is you have to take 21 whatever action is necessary to make that happen.
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| 22 MR. PERSINKO: Is that the way everybody has been 23 doing it?
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| 24 MR. KENT: This is Norman Kent from Westinghouse.
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| 25 When the Commission asks the industry one em ANN RILEY & ASSOCIATES, LTD.
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| 128 1 question, you will probably get seven answers, so this is my
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| >- 2 turn.
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| [
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| \'
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| 3 Being a member of the ISA team, and we were doing 4 CSEs before the acronym changed, we at Westinghouse pride 5 ourselves on having as many filing systems for tracking 6 things as you have rules for allowing us to report things, 7 so I look at datapacks where we record significant items and 8 close them out with a root cause analysis. I look at the 9 PHA reports. I look at our computerized tracking systems.
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| 10 I look at all the red book items which are a low level --
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| 11 these things happen in the plant.
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| 12 Now I don't have a piece of paper that says today 13 I looked at these six, but I read through them. I read the I 14 NRC inspection reports because if I am doing an ISA on a 15 system that has a nonfavorable geometry, I think, oh, yes, 16 in 1995 in a different area of the plant we had a problem 17 with duct tape and NFGs -- well, I am going to go look at ]
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| 18 that.
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| 19 So the answer is yes, those things have to get 20 looked at in order for me to do my job and have it 21 independent reviewed by someone who will say yes, I agree 22 with the evaluation you did, but I didn't keep a list.
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| 23 MR. SCHILTHELM: I think we have made the 24 transition a little bit on criticality safety with this 25 list, but we are starting to talk about contents of a
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| 129 1 license application, and as an industry we have a fairly 7 2 clear picture in our mind of what the contents, of what we 3 believe the contents of a license application should be in 4 that scenario where we have had some differences, so maybe S those differences are starting to show up here in that in 6 our mind the contents of the license application are those 7 safety commitments that we are making that we plan to live 8 by and that is why we see this information as out of line 9 with the license application.
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| 10 On the other hand, you maybe haven't quite bought 11 into that concept yet, so that might be the problem here, 12 that we are kind of jumping into another ballfield.
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| 13 MR. SHERR: Well, I think we understand the views.
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| 14 I think the notion of --
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| ('')
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| %)
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| 15 MR. LEWIS: If I could just get one last 16 clarification.
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| 17 There seem to be two issues here. One is whether 18 anything at all is submitted to NRC in terms of the previous 19 events, and the second issue is whether the backward look 20 should go back 10 years, so there are two things. The 10
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| : 21. years is an issue. Am I correct?
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| 22 MR. GOODWIN: I am not sure 10 years is the issue.
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| 23 MR. VAUGHAN: Well, if you go back 10 years, you 24 are not going to get good information, but I wouldn't say 25 that 10 years versus five years is significant. What is ANN RILEY & ASSOCIATES, LTD.
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| 130 1 significant'is the problem that this 10 year list or a five 2 year list is kind of an after-the-fact look at safety and 3 what the NRC needs to be driving if they want safety at the 4 plants is safety while the plant is operating.
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| 5 You don't want to try to drive safety after the 6 horse has gotten out of the barn. Make sure you keep things 7 healthy as you go and that is a combination of commitments 8 in the license that the licensee has to live with and an 9 inspection program that periodically looks at that.
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| 10 MR. LEWIS: Well, I guess my point was if we just 11 put in the rule that the applicant in performing the ISA 12 should look at previous events at the plant that are 13 significant then that begs the question of how long you have 14 to go back and it begs the question of what significant is 15 and something in the application is going to have to show
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| (}
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| 16 that that was done.
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| 17 MR. SHARKEY: You are getting back into the 18 prescription again and how you perform an ISA and the detail 19 and there's a good guidance document, the A7CHE book, the 20 SRP will have more guidance. I don't think it is really 21 something that needs to be in the rule.
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| 22 I think it is something we pretty much do already.
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| 23 It's part of doing a proper ISA so to speak.
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| 24 MR. SCHILTHELM: One of the chapters is not 25 currently in any of our licenses at least in the commitments r~s, ANN RILEY & ASSOCIATES, LTD.
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| 131 1 portion is a chapter on how we do ISA and we are all going 2 to have to write that chapter and put it in our license.
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| 3 MR. VAUGRAN: One or two of us have it in our 4 license --
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| 5 MR. SCHILTHELM: And we have it in Part 2 of our 6 license, but clearly that is the area in the license 7 application where we will be making commitments as to the 8 quality and the performance of the ISA. There is a standard 9 review chapter on what goes in that license application.
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| 10 Also it would seem like those ISA commitments and 11 that commitment to~ review historical data, whether it is in 12 the regulation or ends up in the license application, is 13 going to be a commitment that we are going to have to make.
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| 14 MR. BIDINGER: You want to be careful about that
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| () 15 16 commitment for historical data.
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| overly retroactive --
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| You don't want to make it 17 MR. SCHILTHELM: Well, you can only review what 18 you have. I can't recreate history. We can review what we 19 have so, you know, if you specify 10 years and I have only 20 got 8 years' worth --
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| 21 MR. BIDINGER: But it's only when the rule becomes 22 effective that you start looking.
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| 23 MR. SCHILTHELM: Yes. You make your best effort 24 given the information you have.
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| 25 MR. SHERR: Okay. Well, I think we can digest all ANN RILEY & ASSOCIATES, LTD.
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| r-132 1 that.
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| g 2 If there are no other comments, the question is CI 3 where we go from here. We talked a little bit about that.
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| 4 One thing is this meeting is being recorded and a 5 transcript of the meeting will be put on the website. I 6 have asked Felix to give us an electronic version of the 7 viewgraphs so that they can be conveniently included with 8 the transcript.
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| 9 We understand -- I am just not sure whether it is 10 on Friday or early next week we will be receiving the 11 comments on the balance of issues -- the reporting 12 requirements, baseline criteria, and change process, so we 13 look forward to those comments.
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| 14 Also we indicated you all will be working hard 15 tomorrow and giving us annotated comments on Chapter 5 and 16 that will be helpful in our considerations of the changes, 17 and I am sure Harry will appreciate that.
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| 18 As mentioned earlier of course, we have already 19 put one version of 70.60 and 70.62, the basic performance 20 requirements of the rule, on the Web in response to the KIM 21 safety comments. We will be posting a revision of those i l
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| 22 sections considering the comments we have discussed today 23 and then the criticality comments in the letters as well as 24 comments that we have received on the earlier posting, and 25 we are shooting to have that posted on the web in one to two ANN RILEY & ASSOCIATES, LTD.
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| 133
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| .1 weeks from now.
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| f-s 2 As far'as the Chapter 5, our target is to get a
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| ~
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| t-) 13 revised Chapter 5 and to post that on the Web and with a 4 target of somewhere in two to four weeks, depending on how 5 many issues we have to resolve.
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| 6 As Rob mentioned earlier, we are currently 7 reviewing the so-called ISA comments, which gets into some 8 basic issues in terms of what is in the license application 9 and all that, and we are hoping to post something on the 10 Web, proposed draft rule language on that, in the next 11 couple, two to three weeks.
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| 12 In that same timeframe, hopefully we get the 13 balance of rule comments early enough to the point that we 14 can post rule changes in relationship to those, in a two to
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| /''}
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| V 15 three timeframe too. Of course, we haven't seen any yet so 16 we don't know how complicated that is.
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| 17 We are in the finishing touches of revising the 18 decommissioning SAP chapter, which has been revised. Quite 19 expansive too, I think, the comments that have been received 20 on that, and we will be posting that chapter within the next 21 couple of weeks.
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| 22 So those are the postings that we are 23 anticipating. When we post those, we will be asking for 24 comments, if possible, within a couple weeks, to keep the 25 process moving, and those timeframes are kind of consistent ANN RILEY & ASSOCIATES, LTD.
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| t 134 1
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| with allowing another round of reactions to what is posted m 2 within the timeframes we have set up for in terms of
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| - C-I 3 mid-February and early March for the rule and SRP comments.
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| 4 Felix had mentioned in his opening comments about 5 future public meetings. I think in this context it is best 6 to see how this plays out. In other words, when we post 7 these things, I think if it becomes evident that there is a 8' need to discuss some issues, that we have missed some 9 serious points or something like that, and it would useful l i
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| 10 to go over that, probably in the early February to 11 mid-February timeframe would be the time to do that, but at 12 this point I would say let's just kind of see how it goes l
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| 13 and we can always arrange that meeting on a week's notice or 14 something. l l
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| f 15 At the last meeting we understood that we would be
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| \s' 16 receiving some additional comments on the Standard Review 17 Plan in late January to mid-February timeframe. I think, 18 Felix, in your opening comments you were suggesting that 19 that may be delayed? So we will see. We can only consider 20 what we get.
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| 21 MR. KILLAR: Well, at lunch we talked some about 22 the morning session and what we heard, and you all probably 23 need to confirm, but what we heard is you are going to 24 rewrite the SRP and that the things that might be more 25 helpful for you at this stage is like we talked about doing ANN RILEY & ASSOCIATES, LTD.
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| g 135 1 with Chapter 5, where we go through and not rewrite the
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| - 2 chapter for you but just annotate certain things that we
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| 'l 3 have pointed out.
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| 4 Now, you know, if that is not helpful, then maybe 5 we ought to do something else but it seemed like that would 6 be the most helpful thing to do.
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| 7 MR. SHERR: Any comments we receive will be 8 helpful. I mean if the most convenient way to do it is in 9 terms of the annotated comments, then that is fine.
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| 10 I mean the SRP -- it is easy to talk about putting 11 the rule out and we are putting out two SRP chapters in 12 terms of revision. We don't intend to do the whole SRP in 13 this way. It just isn't possible.
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| 14 There are two other things we talked about at the r 15 December 4th meeting is -- they weren't commitments, they N-]J 16 were just possibilities.
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| 17 One had to do with the matter that Liz raised 18 earlier was an example chapter that would identify the level 19 of prescriptiveness that the industry considered appropriate 20 and that was just identified as a possibility, not a 21 commitment. I don't know if we have grown stronger or 22 weaker at this point or whether it is still just viewed as a 23 possibility.
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| 24 MR. KILLAR: I think it is safe to view it as a 25 possibility but as the process goes on, it is an iterative ANN RILEY & ASSOCIATES, LTD.
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| o 136 1 process, and I think we are capturing some of the generic
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| <- 2 type issues that we have identified and we have talked about
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| \'' 3 Chapter 5 today, and I think as we go through tomorrow we 4 may be able to pull up on that some more, so it's still a 5 possibility. It's just a function of timing and resources 6 to work on it.
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| 7 MR. SHERR: Okay. The other thing that we talked 8 about in December was feedback on -- if you have 9 attachment -- was the example of the ISA summary that was in 10' the rule package, to give us an idea of if we are looking 11 for things in there that are more detailed than considered 12 appropriate, what are those things in your mind.
| |
| 13 That would be helpful. One thing that I want to 14 mention -- if there is anybody here who is not currently
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| ,/ 15 receiving E-mails telling that something has been put on the 16 website who wants to receive such E-mails, probably the best 17 thing to do is to send an E-mail to Barry Mendelsohn, which 18 is BTM1#NRC. gov and just let him know that you want to be 19 added to the list.
| |
| 20 Again, I would like to thank the industry for your 21 helpful participation and the written comments that were 22 provided. All this helps us in terms of advancing our 23 development of the rulemaking package, and we also look 24 forward to receiving the annotated chapter and I think that 25 will help a lot too.
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| ANN RILEY & ASSOCIATES, LTD.
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| r 137 1 She is not here, but again I want to thank Carrie 2 Brown for all the work she did to make the arrangements for 3 this meeting, including a lot of twisting a lot of arms so 4 we got a decent-sized room, even though we didn't make use I
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| 5 of the whole room, but these things are important. ]
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| i 6 Finally, I would like to thank Mark Mahoney, who l 7 is the recorder for the meeting, and the transcript is very 8 important to all of us, and we thank you for your efforts --
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| 9 and thank you --
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| 10 MR. KILLAR: Wait a minute. There's a few things 1
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| 11 that we want to discuss.
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| 12 MR. SHERR: Okay.
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| I 13 [ Laughter.) I 14 MR. KILLAR: While we've got you. On the 15 information that you have been putting up on the web, we 16 find it very helpful, but one of the things that would be 17 even more helpful, if you would put the entire rule up on 18 the web and then identify those sections that are changed.
| |
| 19 One of the concerns we have with reading the first 20 iteration, it says this changes this section but it doesn't 21 mean because thiE isn't here it's not included, so it left a 22 question of, well, wait a minute, they didn't have this 23 here. Does that mean it's not included or does that mean 24 that will be included but they just didn't put it up here?
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| 25 So it would be helpful if you put the entire rule ,
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| 138 1 up on the web, and so when you do make the changes we know
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| , 2 what changes have specifically occurred and what sections
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| )
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| 3 you specifically have not changed, so that it is clear to us 4 what has changed and what hasn't changed.
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| 5- MR. SHERR: Okay. I think one problem is that the 6 fact that we haven't changed something doesn't mean that we 7 don't intend to change it. We are looking at different 8 sections --
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| 9 MR. KILLAR: Well, that's fine as long as -- you 10 know, our concern is that when you are changing one section 11 that may have reference into anocher section, and we want to 12 know how that section is changed.
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| 13 Now if you don't change it, then we are going to 14 assume that it's still the same way it has been, rather than
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| () 15 16 assuming that section is going to change until we see that change.
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| 17 The concern is that we want to make sure that when 1
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| 18 we look at this we are looking at it in its entirety and so 19 we' understand how maybe one section impacts another section 20 of the rule and when its left with just the specific thing 21 that's changed, it is not clear to us what impact it is 22 having on the other sections, so we would like to have the 23 whole thing up there and as sections change to the rule you 24 make those changes in any other associated sections that 25 change accordingly, so it is clear to us what has changed ANN RILEY & ASSOCIATES, LTD.
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| 139 1 and what hasn't changed, understanding that maybe it hasn't 2 changed today but maybe next week or two weeks from now as 3 you go through another iteration, you may change that.
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| 4 MR. FREEMAN: But then you put it up there.
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| 5 MR. GOODWIN: We had to work out of two or three 6 copies of the rule -- at least two copies, the last version 7 of 70.60 and .62, but we had a prior version and then we 8 ended up getting the whole rule, and it is kind of hard 9 going back and forth, and I think it would be much easier if 10 you could put the whole rule and then, you know, use the 11 vertical margins or line-outs, or annotations, whatever you 12 might do, to indicate what has changes, but that way at 13 least we can review it within the total context.
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| 14 MR. SHERR: The changes would be from the original 15 rulemaking package?
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| 16 MR. GOODWIN: Right.
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| 17 MR. SCHILTHELM: How many more subchanges do you 18 have to make, are you planning to make before you get to 19 Version 2?
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| 20 MR. PERSINKO: Of the entire rule?
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| 21 MR. SCHILTHELM: I mean Version 1 was presented to 22 the Commission and we saw 1-A on the web. We saw 1-B. We 23 saw a portion of 1-A on the web. We saw 1-B handed out 24 today. There is a 1-C up that -- do you have D and E 25 coming, or do you have D through H coming?
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| 140 1 MR. PERSINKO: Well, we know some of the changes 2 based on some comments we are working on, but we also don't tO 3 know how many more comments we are going to receive.
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| 4 MR. SHERR: I think one of the things I wanted to 5 clarify is the fact that we are not going to -- every time 6 we receive an individual comment, we are not going to go 7 back and respond to it. We are doing it kind of in a batch 8 process. What I was suggesting earlier was that we do plan 9 on what I consider the basic performance requirements of the 10 rule, 70.60 and 70.62, we are going to come out with another l 11 version of that in light of the comments in the letters and 12 the comments of this meeting as well as the comments on the 13 earlier posting of that, so that is kind of one thing that 14 we plan to do.
| |
| 15 Then we plan to address the license application 16 type issues -- the ISA comments type thing in another 17 posting and in the balance of the rule comments, so when we 18 do the balance of the rule presumable we have covered the 19 entire rule at that point.
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| 20 MR. SCHILTHELM: At some point we have got to 21 clear'our heads and get the whole thing in front of us.
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| 22 MR. SHERR: Yes.
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| 23 MR. SCHILTHELM: Maybe it is not the next time, i
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| 24 but -- i 25 MR. SHERR: Yes. If we meet our schedules, in l
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| ANN RILEY & ASSOCIATES, LTD.
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| 7 141 1 three weeks' timeframe we will have all that out there, and 2 I guess I think -- I mean my feeling is that the first thing
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| \- 3 we had before, where we are dealing with the performance 4 requirements of_the rule and focusing on that to nave that 5 by itself seemed reasonable to me, but okay -- maybe because 6 I see what is going on -- but of course I think as we move 7 on these things, maybe we include these things in the 8 context of the overall rule, as we roll these things over 9 kind of thing.
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| 10 MR. SCHILTHELM: At some point though --
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| 11 MR. SHERR: But these are the chunks that we are 12 focusing on.
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| 13 MR. KILLAR: You see, to give you an example, when 14 you posted the 70.60 and 70.62 you included in here the
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| ~
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| 15 related definitions to 70.4 and you did away with the acute 16 exposure guideline level -- the EGL. You say annotation --
| |
| 17 this term is not used in the rule anymore, which implies you 18 are going to delete all those tables and things along that 19 line, yet you didn't bother putting that on the website to 20 let us know that, yes, that in fact is what you are doing, 21 where if you had the full rule up there and then you had 22 annotated it at this section - "This appendix is 23 deleted" -- we know in fact that is what the intent is.
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| I 24 MR. SHERR: YEs. l 1
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| 25 MR. KILLAR: That is the point we are trying to l l
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| m
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| r 142
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| 'l make.
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| 7- 2 MR. SHERR: Okay.
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| 3 MR. KILLAR: And then as the other sections get 4 updated, those sections will be revised accordingly.
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| 5 MR. SHERR: Yes. I think, especially as we 6 continue on now and where we are changing other sections and 7 there's interdependencies with the earlier sections, it is 8 useful to have it all in context anyway, I agree.
| |
| 9 Okay. That's the first one?
| |
| 10 MR. KILLAR: Constructive criticism. I am just 11 trying to make it an easier process for all parties 12 concerned. i 13 The other thing I wanted to spend a little time 14 on, we had sent in our letter December the 22nd dealing with 15 the ISA process. You have indicated you have started work 16 on that and are looking at some changes to the proposed rule 17 accordingly.
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| 18 Can you possibly give us some insights and maybe 19 any questions you may have on anything we supplied?
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| 20 MR. PERSINKO: Right now we are waiting for your 21 balance of that. You have a section on 70.72 that you were 22 going to submit, so that ties to the ISA and we are kind of 23 waiting to see that letter.
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| 24 MR. KILLAR: Our December 22nd letter is clear? j 25 You don't have any problems or questions or anything?
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| 143 1 MR. PERSINKO: No.
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| s 2 MR. KILLAR: You agree with everything we said?
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| '- 3 [ Laughter.)
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| 4 MR. PERSINKO: It's under consideration.
| |
| 5 MR. SHERR: We agree with everything they agree 6 with.
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| 7 [ Laughter.]
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| 8 MR. KILLAR: The other thing, I guess maybe the 9 last thing, is we right now do have a good group together.
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| 10 We have some time. Is there any other issue that you have 11 some concern with that we raised previously that you may 12 want to discuss -- like.70.72, the change process, you know?
| |
| 13 We are concerned with that and we want to make sure that 14 comes out correctly, and I think there are some mixed r~N 15 emotions from the industry.
| |
| b 16 Some people feel that, gee, that's no big deal, 17 and other people think that may be a big deal, so if you can 18 give us some of the insights as to how you view it, it may 19 - help.
| |
| 20 Similarly with the ISA summary, there was some 21 concern when reading at 70.60 and 70.62 that you have put on 22 the website here, talking about the ISA summary I believe 23 submitted with the application gave us the impression that 24 it may be more than on the docket, and so we want to once 25 again clarify that the ISA summary is only on the docket and ANN RILEY & ASSOCIATES, LTD.
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| 144 ;
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| 1 not part of the license itself, so if you have got anything 7% 2 - along that line that you want to share with us, we would be 3 glad to hear it.
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| 4 MR. PERSINKO: As far as the ISA summary goes, we 5 are seeing that as not in the license but on the docket.
| |
| 6 We're real interested in 70.72. I mean we haven't i
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| 7 changed anything really from our previous views at the 8 December 3rd meeting yet, where we gave up ideas about 9 70.72, so that really hasn't changed, that part, although we l 10 have been working.on it. We have some internal things we l 11 are doing, trying to just get a better handle on 12 possibilities on how we could change -- some possibilities j 13 on addressing changes in 70.72 but we are still waiting for
| |
| '14 your letter.
| |
| / 15 MR. KILLAR: I think then in that case I think
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| (]_/
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| 16 that's pretty well wrapped up. Anything from anybody else?
| |
| 17 Any questions any members of the industry have for the NRC?
| |
| 18 MR. DAMON: I have been thinking about -- I am 19 Dennis Damon again.
| |
| 20 There's a group of us here who have been looking 21 at the pieces of ISA summary or ISA submittals that have 22 been coming in, and I feel I should share some feedback on 23 this. I mean there have been enough of us look at them that 24 one thing that is clear is that it is very -- it is possible 25 to do a good description of what the safety control scheme ANN RILEY & ASSOCIATES, LTD.
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| 145 1 is for a process and it is possible to do a poor job and (g 2 that is really the key to the ISA summary.
| |
| V 3 It doesn't do any good if all the ISA summary is 4 is basically a set of notes to the guy who did it and it's 5 only meaningful to him. It is an attempt to communicate to 6 us here on the Staff, and what we find is some of the things 7 that are submitted we can't figure out what in heck this is, 8 you know? We can't understand what the nature of the event 9 was or the safety control or whatever it is, so I am just l
| |
| 10 trying to feed that back to you, that that is where the
| |
| ]
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| 11 rubber meets the road. All this other stuff is about what 12 is committed to and it is very important to the legal {
| |
| 13 structure of what goes on, but in terms of what the staff 14 was going to actually do, it is the quality of the writing 15 and the communication that is going to be the actual
| |
| [}
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| 16 important thing.
| |
| 17 MR. SHERR: True -- based on discussion earlier, 18 we were saying there are some good examples.
| |
| 19 MR. DAMON: Yes, there are some good ones that 20 have been submitted and B&W -- the combination of what B&W 21 submitted and what GE submitted would be the ideal thing.
| |
| 22 [ Laughter.)
| |
| 23 MR. SCHILTHELM: You don't have to do it here, but i
| |
| 24 if you guys could get back with us specifically and talk .
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| J
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| ~
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| 25 through some of the things you view aren't adequate, that I i
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| i l
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| 146 1 sure would help,,because the wheel is cranking at home as we g3 2 speak and we are making more of them so --
| |
| 3 MR. GOODWIN: Yes, we are too. On the schedule 4 the '98 versions are about ready to come.
| |
| 5 MR. DAMON: Well, let me describe what it is about 6 what B&W did and what GE did that's good. '
| |
| 7 To a certain extent the tables in the back of the l 8 ISA chapter of the SRP tried to reach in this direction, but i k
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| 9 I mean there is so much other stuff in there it may have i 10 gotten lost and it's hard, it's very difficult -- what I i
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| 11 discovered from that is that it is very difficult for me to, 12 or anyone else I think, to generate an example of something.
| |
| i 13 The reason is you have to have a real physical system that 14 you thoroughly understand yourself because you are 15 talking -- you start to generate an example and pretty soon 16 you are making up stuff that you don't know that it's 17 physically realistic or not even, so you really have to have 18 a real system, and someone who understands it, to generate a 19 good example, so it is much easier for us to look at things 20 that you have done as good examples.
| |
| 21 The thing about what B&W did that was good is it's 22 a structured table. It's in the chapter on the SRP chapter 23 on ISA. There's a table with headings and it tells you what 24 the control parameter is, what the control limits you are 25 trying to control it to, what you are using to control it, ANN RILEY & ASSOCIATES, LTD.
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| 147 1 and what you are doing to those' controls to make sure they 2 are reliable.
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| %) 3 The good thing about GE's is that it was an 4 attempt, a narrative attempt, to describe how the control j 5 scheme works for a whole process -- you know, just narrate j l
| |
| 6 it, try'to explain it to somebody, so that whoever was doing j 7 that had very clearly in his mind an attempt to communicate 8 and so he succeeded. He succeeded in communicating because 9 he was trying to do that, and by narrative I mean it was 10 maybe four sentences about the safety control scheme for a i
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| 11 process. j l
| |
| 12 Now the thing that Westinghouse -- I should give 13 Westinghouse a pat on the back here -- Westinghouse submits 14 fault trees for doubly contingent situations like
| |
| (~'\ 15 criticality. That is ideal for communicating how the V
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| 16- different things are supporting one another in terms of 17 redundancy and we have seen other attempts to do that, and 18 it is very -- we took one example which we thought was a 19 nice redundant system and then we tried to replicate and !
| |
| 20 draw the fault tree, and then we called the licensee and 21 they said, no, you're wrong, completely wrong -- it isn't l
| |
| 22 anything like that, you know --
| |
| 23 (Laughter.)
| |
| 24 MR. DAMON: So that is what I am saying is fault i
| |
| 25 trees are an unambiguous statement of how you see the thing l p)
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| ('_,
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| 7 148 1 as being redundant, and just talking about a qualitative 73 ) 2 fault tree, you know, and so that was the -- we learned that
| |
| '' 3 lesson from that is that if we review, if we don't have a 4 fault tree -- if it is a redundant thing like that with 5 multiple different ways that you get the redundancy and you 6 try to describe it verbally, we are likely to screw it up in 7 understanding it. We just won't get it.
| |
| 8 MR. KILLAR: So in the future if all of our 9 facilities sound like GE and they are mapped out like B&W 10 with fault trees like Westinghouse --
| |
| 11 [ Laughter.)
| |
| 12 MR. BIDINGER: That fault tree process works 13 wonders for reporting as well -- just laying out the logic 14 of your control system. I have been exposed to it, worked
| |
| (~T 15 under it, and I recommend it to everybody -- except the
| |
| ! /
| |
| 16 regulators. It is not a rule.
| |
| 17 [ Laughter.)
| |
| 18 MR. GOODWIN: Just a final comment in regard to I i
| |
| 19 some earlier comments or remarks that Dennis made regarding l 20 the nuclear criticality when we were discussing the removal 21 of that from the list of high consequence events.
| |
| 22 I kind of got the feeling that maybe there was 23 some reluctance in moving that, maybe because it might l
| |
| 24 detract from the importance or the priority that was given 25 by the licensees, but given that we all understand that if '
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| 1 I
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| 7s ANN RILEY & ASSOCIATES, LTD.
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| | |
| l 149 1 we had an accidental criticality we most likely would have
| |
| ?
| |
| - 2 or certainly would have potential health and safety effects, i 3 it would certainly be a political disaster to have a 4 criticality at one of our sites and not to mention the fact 5- that it would, if it didn't destroy it, it would really set 6 back the industry, but for the record I just wanted to let 7 you know and to affirm and to assure you -- and if I am not 8 speaking on behalf on the industry, someone speak up -- but 9 it is our strategic initiative as well not to have a nuclear 10 criticality, accidental that is, at our sites, and that is 11 first and foremost, so Dennis mentioned that a couple of 12 times and I thought I just had to respond to that --
| |
| 13 always -- it may be unwritten but it is there.
| |
| 14 MR. SHERR: We never doubted that.
| |
| 15 MR. GOODWIN: Just wanted to make sure.
| |
| 16 MR. SHERR: Thank you very much. Thank you.
| |
| 17 [Whereupon, at 2:52 p.m., the meeting was 18 concluded.]
| |
| 19 20 21 22 23 24 25 ANN RILEY & ASSOCIATES, LTD.
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| I Court Reporters
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| REPORTER'S CERTIFICATE This is to certify that the attached proceedings i
| |
| ds / before the United States Nuclear Regulatory Commission in the matter of:
| |
| NAME OF PROCEEDING: 10 CFR PART 70 PUBLIC MEETING DOCKET NUMBER:
| |
| PLACE OF PROCEEDING: Rockville, MD O)
| |
| \'~
| |
| were held as herein appears, and that this is the original transcript thereof for the file of the United States Nuclear Regulatory Commission taken by me and thereafter reduced to typewriting by me or under the direction of the court reporting company, and that the transcript is a true and accurate record of the foregoing proceedings.
| |
| b d9 '\
| |
| Mark Mahoney Official Reporter Ann Riley & Associates, Ltd.
| |
| (h w) l l
| |
| l
| |
| | |
| Aaenda for January 13-14.1999. Public Meetino on Revision to 10 CFR Part 70 Meeting Objective: The purpose of the meeting is to discuss industry comments on SECY-
| |
| . 098-185, draft revision to 10 CFR Part 70, that relate to nuclear criticality safety (NCS). i Specific and general comments related to Chapter 5.0 (NCS) of the associated standard i review plan (SRP) and other SRP chapters related to NCS (e.g., Chapter 3 - Integrated Safety Analysis) will also be considered.
| |
| | |
| ==Background:==
| |
| At the December 3-4,1998, NRC Public Meeting on Amendment to 10 CFR Part l 70, the Nuclear Energy Institute (NEI) and interested members of the public expressed l concem over NRC proposals addressing NCS in the proposed revisions to 10 CFR Part 70. J As the December 3-4,1998 meeting primarily addressed rule language, NRC and NEl agreed to postpone discussion, until January 1999, related to detailed concems with rule language in the proposed revision to 10 CFR Part 70 and the guidance language found in the associated draft SRP.
| |
| Meeting Agenda:
| |
| : 1. Introductions
| |
| : 2. Brief Update on Current NRC activities (e.g. web postings) and industry feedback
| |
| : 3. NCS related rule language
| |
| : a. Industry - NEl presentation on rule language concems
| |
| - Risk informed regulation
| |
| - Double contingency
| |
| - Graded level of protection of items relied on for safety
| |
| - Nuclear criticality: quality assurance
| |
| - Historical nuclear criticality data b NRC presentation on draft rule language to address industry /NEI items above and in ANS, LANL letters
| |
| : c. Industry-NEl/NRC discussion on rule language
| |
| : d. Comments by other attendees
| |
| : 4. NCS comments related to SRP
| |
| : a. Industry- NEl NCS comments on SRP guidance l
| |
| : b. Industry - NEl/NRC discussion on NCS comments on SRP guidance
| |
| - c. Comments by other attendees
| |
| : 5. Closing Remarks NOTE: THE MEETING ROOM HAS CHANGED TO ROOM T10A1 (it was previously to be held in Room OB11). Lunch on both days willlast approximately 90 minutes and the meeting will adjoum at approximately 3:30pm on January 13,1999 in order to allow NRC and industry to discuss and consider, among themselves, the information that was presented.. The meeting on January 14,1999 will begin with a summary of those discussions.
| |
| Website address: http://techconf.llnl. gov /cgi-bin / messages? dom _lic d
| |
| | |
| o A
| |
| Los Alamos National Laboratory Nuclear Criticality Safety Group (ESil-6)
| |
| P O. Bos 1663, Mail Stop F691 Date: 2 December 1998 Los Alamos. New Mexico 81545 s.s mbol: ESil-6-98-ADM-05 (5n5) 667-4789 / FAX: (5n5) 665-497n Dr. Carl A. Paperiello, Director Office of Nuclear Material Safety and Safeguards U S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
| |
| | |
| ==Dear Dr. Paperiello:==
| |
| | |
| SUHJECT: Draft Proposed 10CFR70 and Associated SPR 1 have recently become aware of the subject document and, as an employee of one of the National Laboratories, the following statement therein caught my eye: "A /yd/ cants operating existmgfacdities that could become neu ly subject to the C<nnminion's authority, such as DOEfacilities, u onld he expected to " l have read drails af the NRC's review of the REDC facility at ORNL, and the criticality safety implications on Los Alamos portend to be enormous in cost and could be detrimental to worker safety if scarce criticality statT resources are required to respond to additional documentation requirements and prevented from spending time on the process floor.
| |
| Permit me to introduce myself so that it will be apparent where my interests lie. I am the Group Leader of the Nuclear Criticality Safety Group at Los Alamos National Laboratory Nuclear Criticality Safety Group (ESH-6). After ten years spent performing critical experiments, reactor design, and theoretical reactor safety research, I migrated into criticality safety fulltime and have been a practitioner for the last twenty years. Thus my interests and comments are limited to nuclear criticality safety. In the comments that follow, I am speaking as the Group Leader of the Nuclear Criticality Safety Group at Los Alamos National Laboratory. l Since there was little time to review this drafl and its supporting documents, such as the Standard Review Plan (SRP), prior to the December 3-4 meeting, these comments are brief. Nevertheless, my quick review has convinced me that, while the intent of the revision, as stated and expressed in general terms such as ' performance-based' and ' risk-informed,' is completely reasonable, as the saying goes: "the devil is in the details." I will highlight my concerns with a few examples, and provide a more thorough, documented review at a later date should that be appropriate. Also, I am aware of comments provided by Dr. Cecil Parks, representing the Nuclear Criticality Safety Division of the American Nuclear Society, and will not cover issues and points previously addressed.
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| " Risk informed and performance based regulation"(words found in the Description of Proposed Action) operated by the l'nnersity of cahrorma for the Department ofI.nergs j An 1: qual opporturuly I mployer
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| \*
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| While these words imply a consistency in philosophy with the ANS-8 standards , they are followed by numerous uses of the term consequence criteria' and discussions and examples using ' quantified likelihoodsJ Basing requirements for documentation, QA, training, etc. on the worst-case (of a criticality accident) consequence alone is certainly not risk-informed regulation. Further, the attempt to have PRA or any other form of quantified risk assessment become a major pan of the safety basis of nuclear criticality safety at any facility would be inappropriate at best. The dau on which to base failure rates simply do not exist. Considering the track record of criticality safety in the U.S. and worldwide, as evidenced by the accident rate, one can only conclude that the absolute performance and its trend have been very positive compared to other worksite hazards
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| -70.62 ISA Requirements As a direct result of a criticality accident being labeled a 'high-consequence' event, there ;
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| are potentially severe implications in the rule on required actions and documentation !
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| j compared to how the DOE regulates criticality safety, the latter being consistent with the guidanc'e and philosophy found in the ANS-8 standards. Examples are:
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| "(B) For new processes submit the results of the A&i and any revisions as part of t ise.
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| applicationfor amendment of the license under 70.31. " -
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| The time delay inherent in this process would result in enormous costs at no practical risk reduction. Currently the DOE does not review and approve criticality safety evaluations before the contractor can implement operations unless an Unreviewed Safety Question (USQ)is found. For criticality safety, this is usually avoided by having a reasonably conservative criticality accident described in the Safety Analysis Report.
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| "70.64 Baseline design criteria....
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| "(a) Licensees shall maintain.....unless.....not relied onfor safety.
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| "(l) Appropriate records ofthese items must be maintained.....throughout the life of thefacility.
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| "(2) . . . "
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| What does this mean? Would it include cans and process equipment such as 5-liter dissolution pots 4-liter Erlenmayer flasks,7-liter filter boats, etc. that are not fixed in place on a glovebox floor and that truly do provide meaningful criticality protection? What if there are several barriers to reaching the critical state, a combination of vessel geometry ;
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| and administrative controls such that none are dominant as is the case in many DOE l' operations?
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| I
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| . ESH-6-98-ADM-05 2 12/02/98- ,
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| p
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| " Appendix C to part 70 - Reportable Safen' Events "11(4 hours)....a deviationfrom safe operating conditions.....has thepotential, as identified in the ISA..... "
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| o This implies first that all gradations of upsets are identified in the IS A. This is impossible.
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| Fissile mass is controlled for nearly all processes, but the limit can be exceeded in an infinitely variable amount.. For example. consider an operator who is loading a melt crucible and ex~cceds the limit by 1,10. 50, or 150%. etc. due to a simple human addition error. These incidents are clearly of varying likelihoods and significance. Should they all have been discussed in the ISA? Again. impossible. While overmass in general has the potential ofleading to a criticality accident, should a 1% overmass be reported in 4 hours, 24 hours. 30 days. or not at all outside of the company? Within the DOE there is the flexibility to use a graded approach such that the process upset can bejudged to be of such little significance locally and of such little learning value globally that it is recorded and tracked internally only. The consequences of not using this common sense approach have been painfully and expensively documented within the DOE!
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| " Standard Review Plan, Chapter 5 "S.4.6 ISA Results "The nuclear criticality aspects of the applicant's ISA are acceptable ifthefedlowing criteria are met:
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| "I. The applicant conducts and maintains an ISA that identifies specific control ~
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| parameters.. .. "
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| Should this requirement be interpreted to mean that controls for every operation or process are identified in the ISA? If so, either the ISA would be continually out of date or
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| . the DOE contractors nationwide would be shut down. Due to hundreds ofindependent operations. processes and limits in larger facilities are changing weekly if not daily in some cases.
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| "7. a. At least one of the two controlledparameters... "
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| Thisimplies that there are only two controlled parameters, a very rare situation, and
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| - implies a misunderstanding of the double-contingency principle.
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| "5.4.5.2 NCS Limits "5.a controlledparameters: When using evperimental data, the applicant applies
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| ' industry-accepted safetyfactors..... 45%... 75%....etc. "
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| These .' industry-accepted' safety factors were never adopted by ANS-8. nor are they in any refereed publication. .in fact I have no idea where they are documented except possibly in NRC guidance for licensees. The DOE has no such formal. specific limits since there is no indication that they would reduce accident frequency; they would clearly have a tremendous cost impact on many DOE sites.
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| ESH-6-98 ADM-05 3 12/02/98 b
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| u
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| ,b,
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| .g
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| ;e "5.51+0ceduresfor Review "5.5.2 Safety Evaluation "14. The reviewer will determine that...... maintains a NCS review ofthe 1%l......that includes a review ofidentifiedpotential accident sequences that result in an inadvertent nuclear criticality. "
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| This does not state a representative worst-case' criticality scenario and thus it implies that this will be maintained for every operation in the ISA. This is contrary to the safety analysis guidance for DOE facilities and would be prohibitively expensive.
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| Closing Comment Most of the issues and concerns raised in this letter revolve around the reasonableness of.
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| the application of the words. This will likely be highely reviewer-dependent and the cost and safety impact cannot be known at this time. In a few cases, for example, ' industry-accepted safety factors,' the section should be deleted unless a stronger basis can be provided.
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| I otter my senices to the Commission in further reviewing and providing comments on the criticality safety aspects of these documents.
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| Sincerely.
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| Thomas P. McLaughlin, Group Leader Nuclear Criticality Safety Group Phone: (505)667-7628; FAX: (505)665-4970 e-mail: tyn a lanl em-cc: Shirley A. Jackson, Chairman, NRC Great J. Dieus, Commissioner, NRC ./
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| Nils J. Diaz, Commissioner, NRC Edward McGafligan, Jr., Commissioner, NRC Jeffery S. Merrifield, Commissioner, NRC Dr. William D. Travers, Executive Director for Operaitons, NRC ESH4-98-ADM 05 4 12/02/98 I
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| }.
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| Los Alamos National Laboratory Nuclear Criticality Safin Group (ESH-6)
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| P.O. Box 1663. Mail Stop F691 Date: 2 December 1998 Los Alamos. New Mexico 87545 Sy mbol- ES H-6 ADM-05 (5n5) 667-4789 t FAX: (505) 665-4970 Dr. Carl A. Paperiello, Director Office of Nuclear Material Safety and Safeguards U S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
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| ==Dear Dr. Paperiello:==
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| SUB, LECT: Draft Proposed 10CFR70 and Associated SPR I have recently become aware of the subject document and, as an employee of one of the National Laboratories, the following statement therein caught my eye. " Applicants operating existingfacilities that c<mid become neu IV subject to the ('omminion's authority, such as IXJEfacilities, u onld be expected to " I have read drafts of the NRC's review of the REDC facility at ORNL, and the criticality safety implications on Los Alamos portend to be enormous in cost and could be detrimental to worker safety if scarce criticality statTresources are required to respond to additional documentation requirements and prevented from spending time on the process floor.
| |
| Permit me to introduce myself so that it will be apparent where my interests lie. I am the Group Leader of the Nuclear Criticality Safety Group at Los Alamos National Laboratory Nuclear Criticality Safety Group (ESH-6). Aner ten years spent performing critical experiments, reactor design, and theoretical reactor safety research, I migrated into criticality safety fulltime and have been a practitioner for the last twenty years. Thus my interests and comments are limited to nuclear criticality safety. In the comments that follow, I am speaking as the Group Leader of the Nuclear Criticality Safety Group at Los Alamos National Laboratory.
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| Since there was little time to review this dran and its supporting documents, such as the Standard Review Plan (SRP), prior to the December 3-4 meeting, these comments are brief. Nevertheless, my quick review has convinced me that, while the intent of the revision. as stated and expressed in general terms such as ' performance-based' and ' risk-informed / is completely reasonable, as the saying goes' "the devil is in the details " I will highlight my concerns with a few examples, and provide a more thorough, documented review at a later date should that be appropriate. Also, I am aware of comments provided by Dr. Cecil Parks, representing the Nuclear Criticality Safety Division of the American Nuclear Society, and will not cover issues and points previously addressed
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| " Risk informed and performance based regulation"(words found in the Description of Proposed Action)
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| Operated by the Unnmity of Cahfimua for the Department or Energy An Equal opportunity Employer
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| l 1
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| 't While these words imply a consistency in philosophy with the ANS-8 standards . they are followed by numerous uses of the term consequence criteria' and discussions and examples using ' quantified likelihoods.' Basing requirements for documentation, QA.
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| training, etc. on the worst-case (of a criticality accident) consequence alone is cenainly not risk-informed regulation. Further. the attempt to have PRA or any other form of quanti 0ed risk assessment become a major part of the safety basis of nuclear criticality safety at any facility would be inappropriate at best. The data on which to base failure rates simply do not exist. Considering the track record of criticality safety in the U.S. and worldwide, as evidenced by the accident rate, one can only conclude that the absolute performar.ce and its trend have been very positive compared to other worksite hazards.
| |
| 70.62 ISA Requirements As a direct result of a criticality accident being labeled a 'high-consequence' event. there are potentially severe implications in the rule on required actions and documentation compared to how the DOE regulates criticality safety, the latter being consistent with the guidance and philosophy found in the ANS-8 standards. Examples are:
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| "(B) For newprocesses submit the results of the ISA and any revisions as part of the applicationfor amenthnent ofthe license under 70.34. "
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| The time delay inherent in this process would result in enormous costs at no practical risk reduction. Currently the DOE does not review and approve criticality safety evaluations before the contractor can implement operations unless an Unreviewed Safety Question (USQ)is found. For criticality safety. this is usually avoided by having a reasonably conservative criticality accident described in the Safety Analysis Report.
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| "70.64 Baseline design criteria....
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| "(a) Licensees shall maintain.....unless.....not relied onfor safety.
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| "(1) Appropriate records of these items must be maintained.....throughout the h*fe of thefaciliy.
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| "(2)..."
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| What does this mean? Would it include cans and process equipment such as 5-liter
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| . dissolution pots. 4-liter Erlenmayer Dasks,7-liier filter boats, etc. that are not fixed in place on a glovebox floor and that truly do provide meaningful criticality protection? What if there are several barriers to reaching the critical state. a combination of vessel geometry and administrative controls such that none are dominant as is the case in many DOE operations?
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| ESH-6-98-ADM-05 2 12/02/98
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| p t
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| 's "Appendit C to part 70 - Reportable Safety Events "H(4 hours)....a deviationfrom safe operating conditions... .has the potential, as identified in the 1%I,.... "
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| This implies first that all gradations of upsets are identified in the ISA. This is impossible.
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| Fissile mass is controlled for nearly all processes, but the limit can be exceeded in an infinitely variable amount. For example, censider an operator who is loading a melt cmcible and exceeds the limit by 1,10,50, or 150%, etc. due to a simple human addition error. These incidents are clearly of varying likelihoods and significance. Should they all have been discussed in the ISA? Again, impossible. While overmass in general has the potential oficading to a criticality accident, should a 1% overmass be reported in 4 hours, 24 hours,30 days, or not at all outside of the company? Within the DOE there is the flexibility to use a graded approach such that the process upset can be judged to be of such little significance locally and of such little learning value globally that it is recorded and tracked internally only. The consequences of not using this common sense approach have been painfully and expensively documented within the DOE!
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| " Standard Reviese Plan, Chapter 5 "5.4.61%i Results "The nuclear criticality aspects of the applicant's 1%i are acceptable if thefallouring criteria are met:
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| "1. The applicant conducts and maintains an 1%i that ider,tifies specific control parameters. . .. "
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| Should this requirement be interpreted to mean that controls for every operation or process are identified in the ISA? If so, either the ISA would be continually out of date or the DOE contractors nationwide would be shut down. Due to hundreds ofindependent operations. processes and limits in larger facilities are changing weekly if not daily in some cases.
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| "7. a. At least one of the two controlledparameters... "
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| This implies that there are only two controlled parameters, a very rare situation. and implies a misunderstanding of the double-contingency principle.
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| "5.4.5.2 NCS Limits "5.a controlledparameters: When using evperimental data, the applicant applies industry-accepted safen' factors..... 45%... 75%....ete. "
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| These ' industry-accepted' safety factors were never adopted by ANS-8, nor are they in any refereed publication. In fact I have no idea where they are documented except possibly in NRC guidance for licensees. The DOE has no such formal, specific limits since there is no indication that they would reduce accident frequency; they would clearly have a tremendous cost impact on many DOE sites.
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| ESH-6 98-ADM-05 3 12/02/98
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| C
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| :p-a l
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| "S.5 Proceduresfor Reviese "5.5.2 Safety Evaluation "14. The reviescer will determine that...... maintains a NCS reviese of the ISA......that l includes a review ofidentifiedpotential accident sequences that result in an '
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| l inadvertent nuclear criticality. "
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| 'This does not state a ' representative worst-case' criticality scenario and thus it implies that this will be maintained for every operation in the ISA. This is contrary to the safety analysis guidance for DOE facilities and would be prohibitively expensive.
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| Closing Comment Most of the issues and concerns raised in this letter revolve around the reasonableness of the application of the words. This will likely be highely reviewer-dependent and the cost and safety impact cannot be known at this time. In a few cases, for example, ' industry-accepted safety factors.' the section should be deleted unless a stronger basis can be provided 1 offer my senices to the Commission in funher reviewing and providing comments on the criticality safety aspects of these documents.
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| Sincerely, Thomas P. McLaughlin, Group Leader Nuclear Criticality Safety Group Phone: (505)667-7628; FAX: (505)665-4970 e-maili mmiitante cc: Shirley A. Jackson, Chairman, NRC Great J. Dieus, Commissioner, NRC Nils J. Diaz, Commissioner, NRC Edward McGafTigan, Jr., Commissioner, NRC Jeffery S. Merrifield, Commissioner, NRC Dr. William D. Travers, Executive Director for Operaitons, NRC
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| ' ESH-6-98 ADM-05 4
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| '12/02/98
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| 1 a
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| December 17,1998 1 Dr. Carl A. Paperiello, Director Office of Nuclear Material Safety and Safeguards ;
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| i U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
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| ==REFERENCE:==
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| 10 CFR Part 70 Nuclear Criticality Safety
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| ==Dear Dr. Paperiello:==
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| I At the December 3-4,1998 NRC Public Meeting on Amendment to 10 CFR Part 70 the Nuclear Energy Institute (NEI)' expressed concern over Nuclear Regulatory Commission (NRC) proposals addressing Nuclear Criticality Safety (NCS) in the j proposed revisions to 10 CFR Part 70. This letter documents such concerns and !
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| recommends improvements to the proposed rule revisions in the area of NCS. j Comments and recommendations on the rule implementation mechanisms, as detailed in the draft NUREG-1520, Standard Review Plan, will be presented by NEl i and industiy representatives at the NRC Workshop on NCS scheduled for January 13-11,1999. This letter defers discussion of highly technical implementation issues to q that Workshop and addresses only revision of the Part 70 rule language.
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| NEl supports the NRC's efforts to make the Part 70 rule consistent with the ANSI /ANS-8 NCS standards. In this regard, some modification of the language of the proposed revisions is, however, required to focus on the risks, rather than the ' consequences' and ' quantified likelihood' of accident sequences that could lead to potential nuclear criticalities. While nuclear criticality accidents can be considered high consequence events, their risks differ. Consequently, the safety controls required to mitigate the risk of an accident sequence and the assurances Dr. Carl A. Paperiello Nuclear Regulatory Commission December 17,1998 Page 2 to be applied to such controls should be graded according to the severity of the potential risk. Such safety controls and assurances to ensure sub-criticality in fuel
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| ' NEl is the organization responsible for establishing unified nuclear industry policy on matters affecting the nuclear energy industry. including the regulatory aspects of generic operational and technical issucs. NEl's members include all utilities licensed to operate commercial nuclear power plants in the United States. nuclear plant designers. major architect / engineering firms. fuel fabrication facilitics. materials licensecs, and other organizations and individuals involved in the nuclear energy industry.
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| . l cycle operations cannot be arbitrarily assigned, but must be selected (and graded) based upon the results of an Integrated Safety Analysis (ISA).
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| I NEl believes that Part 70 revisions should impose a level of regulatory oversight over NCS similar to that afforded to other potential hazards, as NCS differs in no intrinsic way from other industrial safety concems. Experience gained from U.S. nuclear criticality accidents indicates that a danger to health is posed only when a nuclear worker is within a few meters of a critical excursion (and with little inteivening l shielding). Nuclear criticality accidents pose no danger to the public.
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| The risk of an inadvertent nuclear criticality accident can be miniinized through application of risk-informed, performance-based regulation. A Part 70 license should include license commitments to manage NCS in accordance with ANS-8 guidelines.
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| It should define the broad, operational bases for a facility, within which limits the licensee may safely operate without additional NRC approval (or license amendment) and without burdensorne reporting requirements. A licensee should have the latitude to focus its NCS resources on high-risk nuclear criticality accident sequence prevention and to address safety issues within a licensee's corrective action program.
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| NEllooks fonvard to continuing our dialogue with the NRC on the Part 70 rulemaking. We should be pleased to address any questions which you or your staff may have on the industry's concerns and positions.
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| Sincerely, Marvin S. Fertel Enclosure cc: The Honorable Shirley A. Jackson. Chairman, NRC The Honorable Greta J. Dieus, Commissioner, NRC l The Honorable Nils J. Diaz, Commissioner, NRC The Honorable Edward McGaffigan, Jr., Commissioner, NRC .
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| l The Honorable Jeffrey S. Merrifield, Commissioner, NRC Dr. William D. Travers, Executive Director for Operations, NRC
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| 1 1
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| l ENCLOSURE NUCLEAR ENERGY INSTITUTE (NEI)
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| RECOMMENDED IANGUAGE CHANGES TO PART 70 CONCERNING NUCLEAR CRITICALITY SAFETY I. Proposed Changes to the Draft Language (a) Risk-InformedRegulation l Proposed revisions to 10 CFR Part 70 addressing Nuclear Criticality Safety (NCS) are ambiguous and could potentially be misinterpreted as ' consequence based' rather than ' risk-based' regulation. Proposed revisions continue to address the consequences and likelihood of an accident sequence, whereas they should focus regulatory attention on its risk. The language of the proposed revisions should be checked to ensure consistency with the concept of risk.
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| i The fuel cycle industry acknowledges that a nuclear criticality accident is an operating hazard whose risk must be adequately managed. The operating history of fuel cycle facilities has shown that serious risks to worker health and safety from a nuclear criticality accident are confined to within several meters of the critical excursion. Most criticality accidents in the U.S.A. have not exposed workers to greater than 100 rem. Never has the health of a member of the public been impacted by a nuclear criticality accident in the U.S.A. Risks to workers from other operating hazards such as fires, equipment malfunctions or hazardous chemical releases often pose greater health and safety concerns. In i.ddition, some high consequence accident scenarios may be low risk due to their inherently low likelihood of occurrence (e.g. certain natural phenomena).
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| Depending on the structures, systems and controls implemented at a facility, the risk of a nuclear criticality accident - the parameter to be addressed by the NRC in risk informed, performance based regulation - may indeed be low.
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| NCS revisions to Part 70 should consider application of a risk-informed, i performance-based methodology to:
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| . evaluate the risk (i.e. consequences and likelihood) of potential nuclear criticality accidents whether initiated by external events, process deviations or internal events e establish appropriate risk-based (graded) levels of protection to prevent nuclear criticality accidents e establish appropriate risk-based (graded) levels of assurance for items relied on for safety to ensure their availability and reliability l
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| l
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| \
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| I i
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| (b) Double Contingency ,
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| Draft revisions to 10 CFR Part 70 define, and now require, double contingency !
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| as a practice for managing the risk of a nuclear criticality accident. The fuel j cycle industry has for many years adhered to American National Standard j I
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| ANSI /ANS-8.1 and incorporated the double contingency principle into its NCS programs. The NRC has consistently endorsed the ANS!/ANS-8.1 double contingency principle and now does so in the draft revisions to Part 70.
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| NElis concemed with the manner in which NUREG-1520, Standard Review Plan (SRP) for review of Part 70 license applications, requires implementation of the j double contingency principle. To determine whether there are at least two
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| 'unlikely', independent and concurrent process changes necessary before a I criticality might occur (i.e. double contingency protection), industry has relied on the expertise, experience and judgment of nuclear criticality experts on a deterministic basis. Risk-informed decisions are reached by thoroughly understanding the system characteristics and performance in a nuclear criticality safety evaluation process. Whenever possible, the NRC favors use of a quantitative measure to judge implementation of, or compliance with, a principle or methodology. For example,in the case of adherence to the double contingency principle, the SRP requires assignment of specific, quantitative numerical frequencies to each of the controls to determine that a nuclear criticality accident is ' highly unlikely.' The SRP's definition of ' highly unlikely' as 4 4 a frequency of 104 is arbitrary and forces differentiation of 10 and 10 between two 'unlikely' events in a criticality accident scenario. Measuring compliance to these arbitrary, quantitative values is burdensome and problematic for both licensees and the NRC. In fact, quantification of NRC's expression of the principle of double contingency contradicts guidance of the American National Standard, which upholds the basic definition of the double contingency principle as adequate and sufficient. NEl recommends that industiy's current practice of detailed evaluation of credible accident sequences by experienced nuclear criticality engineers continue. Adherence to the ANS-8 guidance should also be continued.
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| (c) Graded Level of Protection ofitems Relled On For Safety The risk of a nuclear criticality for a given accident sequence is established by the ISA. As a nuclear criticality accident is defined to be a high consequence of concern,its risk becomes solely a function of the accident's likelihood of occurrence. Section 70.60(c) requires that graded safety contiols (or items relied on for safety) be implemented to prevent nuclear criticality accidents.
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| Systems of controls applied to a high risk sequence should, for example, be inherently more robust than those applied to an intermediate risk sequence.
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| The choice of controls will depend on the risk (or likelihood) of the accident o
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| b
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| sequence and may include one or more passive engineered, active engineered or administrative controls. While individual controls may vaiy in their relative importance or robustness,in aggregate, they must render the risk of a nuclear criticality accident to an acceptably (low) level, s70.60(c) also requires the licensee to ensure that safety controls, or items relied on for safety, to prevent a n:tclear criticality accident are continuously available and reliable. in practice, specific safety controls will not be operational during maintenance and i:al:bration testing and will not be required when the process is not operational or when Special Nuclear Material (SNM) is not present. The wording of s70.60(c) should, therefore, be modified to address the risk of a nuclear criticality accident (rather than its consequences and likelihood) and to assure that items relied on for safety are "...available and reliable when required to perform their safety functions."
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| Section s70.60(c) incorrectly identifies only the likelihood of external events as an element of risk from a nuclear criticality accident, thereby excluding the likelihood of process deviations or other internal events as an element of the risk evaluation. In actuality, the likelihood of a process deviation or other internal event initiating an accident sequence leading to a potential nuclear criticality is probably far greater than that posed by an external event. The language of s70.60(c) should be clarified.
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| (d) Nuclear Criticality: Quality Assurance As noted in comment (a) above, the proposed revisions to Part 70 are likely to be misinterpreted as ' consequence-based' rather than ' risk-based' regulation.
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| An example of this is clearly evident in the draft SRP s5.4.4.1 " Quality Assurance for NCS." Once the appropriate level of protection is afforded to an accident scenario through the identification and application of specific controls, then appropriate assurance must be provided for these controls to ensure their availability and reliability. Draft SRP s5.4.4.1(1) incorrectly requires that all criticality safety controls be afforded the highest level of assurance, while 170.60(d)(3)(vi) and draft SRP E5.4.4.1(5) correctly require the assurance level be commensurate with the importance of the safety function.
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| Through the application of the double contingency principle, the importance of any single criticality safety control may be less than that of some other control that is the only barrier to an accident with a consequence of concern.
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| Therefore, the highest level of assurance would not necessarily be warranted for criticality controls in accident scenarios with double contingency protection.
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| In addition, the reliability of individual controls should be considered when determining the appropriate level of assurance for criticality safety controls. For 1
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| l
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| )
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| example, passive engineered controls in general require a lesser degree of surveillance than active engineered controls or administrative controls.
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| However, each control should be considered relative to the envirolunent and application to detennine the levei of assurance needed to ensure the reliability and availability of the control.
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| (e) Historical Nuclear Criticality Data Part 70 license applications for operating facilities are required by s70.65(c) to 4 include "...a description of operational events, within the past 10 years, that had a significant impact on the safety of the facility." Detailed incident reports of nuclear criticality deviations or violations, including any corrective safety measures that were implemented, are submitted to the NRC at the time of the j incident or retained in licensees' records. As the NRC has on file, or available to >
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| them, voluminous information on all operational events, including nuclear criticality safety deviations, NEl sees little justification in submitting this information at the time of license application or renewal. NEl recommends that s70.65(c) be deleted from the Part 70 revisions.
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| II. Concludins! Remarks Nuclear criticality is an important, but readily manageable, operational hazard at fuel cycle facilities. NEl recommends that the proposed revisions of 10 CFR 70 be clarified to reduce their ambiguity and the possibility of interpreting them to be
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| ' consequence-based' rather than ' risk-based' regulations. The rule should permit industry to continue implementation of the double contingency principle as it has done without imposition of a probabilistic methodology. Part 70 should be consistent with American National Standard 8 that upholds the basic definition of the double contingency principle as adequate and sufficient. In support of risk-informed, performance-based regulation, the rule should grant a license applicant the flexibility to implement graded controls (and assurances) based on the results of the ISA.
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| Rd I fhen,Part 'O Cntwahry Safety (Papendo Lanot %RC \ et 4)Im*
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| a
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| - De'cember 17,1998-Dr. Carl A. Paperiello, Director
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| - Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
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| ==REFERENCE:==
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| 10 CFR Part 70 Nuclear Criticality Safety -
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| ==Dear Dr. Paperiello:==
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| ' At the December 3-4,1998 NRCPublic Meeting on Amendment to 10 CFR Part 70 the Nuclear Energy Institute (NEI)' expressed concern over Nuclear Regulatory Commission (NRC) proposals addressing Nuclear Criticality Safety (NCS) in the proposed revisions to 10 CFR Part 70. This letter documents such concems and recommends improvements to the proposed rule revisions in the area of NCS.
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| Comments and recommendations on the rule implementation mechanisms, as detailed in the draft NUREG 1520, Standard Review Plan, will be presented by NEl and industry representatives at the NRC Workshop on NCS scheduled for Januaiy 13-l 1,1999. This letter defers discussion of highly. technical implementation issues to that Workshop and addresses only revision of the Part 70 rule language.
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| NEl supports the NRC's efforts to make the Part 70 rule consistent with the' ANSI /ANS-8 NCS standards. In this regard, some modification of the language of the proposed revisions is, however, required to focus on the risks, rather than the ' consequences' and ' quantified likelihood' of accident sequences that could lead to potential nuclear criticalities. While nuclear criticality accidents can be considered high consequence events, their risks differ. Consequently, the safety controls required to mitigate the risk of an accident sequence and the assurances Dr. Carl A. Paperiello Nuclear Regulatoly Commission December 17,1998 +
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| Page 2 to be applied to such controls should be graded according to the severity of the
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| .g potential risk. Such safety controls and assurances to ensure sub-criticality in fuel
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| ' NEl is the organization responsible ror establishing unined nuclear industry policy on matters affecting the nuclear energy industry including the regulatory aspects of generic operational and technical issues. NEI's J
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| . me!nbers include all utilities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect / engineering firms, fuel fabrication facilitics, materials licensecs, and other organizations and individuals involved in the nuclear energy industry.
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| 6 cycle operations cannot be arbitrarily assigned, but must be selected (and graded) based upon the results of an Integrated Safety Analysis (ISA).
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| NEl believes that Part 70 revisions should impose a level of regulatory oversight over NCS similar to that afforded to other potential hazards, as NCS differs in no intrinsic way from other industrial safety concerns. Experience gained from U.S. nuclear criticality accidents indicates that a danger to health is posed only when a nuclcar worker is within a few meters of a critical excursion (and with little intervening shielding). Nuclear criticality accidents pose no danger to the public.
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| The risk of an inadvertent nuclear criticality accident can be minimized through application of risk-informed, performance-based regulation. A Part 70 license should include license commitments to manage NCS in accordance with ANS-8 guidelines.
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| 11 should define the broad, operational bases for a facility, within which limits the i
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| licensee may safely operate without additional NRC approval (or license amendment) and without burdensome reporting requirements. A licensee should have the latitude to focus its NCS resources on high-risk nuclear enticality accident sequence prevention and to address safety issues within a licensee's corrective action program.
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| NEllooks fonvard to continuing our dialogue with the NRC on the Part 70 rulemaking. We should be pleased to address any questions which you or your staff may have on the industry's concerns and positions.
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| Sincerely, Marvin S. Fertel l Enclosure j cc: The Honorable Shirley A. Jackson, Chairman, NRC The Honorable Greta J. Dicus, Commissioner, NRC The Honorable Nils J. Diaz, Commissioner, NRC J The Honorable Edward McGaffigan, Jr., Commissioner, NRC l The Honorable Jeffrey S. Merrifield, Commissioner, NRC J Dr. Williarn D. Travers, Executive Director for Operations, NRC i
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| O ENCLOSURE NUCLEAR ENERGY INSTITUTE (NEI)
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| RECOMMENDED LANGUAGE CHANGES TO PART 70 CONCERNING NUCLEAR CRITICALITY SAFETY I. Proposed Changes to the Draft Language (a) Risk-Informed Regulation Proposed revisions to 10 CFR Part 70 addressing Nuclear Criticality Safety (NCS) are ambiguous and could potentially be misinterpreted as ' consequence-based' rather than ' risk based' regulation. Proposed revisions continue to address the consequences and likelihood of an accident sequence, whereas they should focus regulatory attention on its risk. The language of the proposed revisions should be checked to ensure consistency with the concept of risk.
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| The fuel cycle industry acknowledges that a nuclear criticality accident is an i operating hazaid whose risk must be adequately managed. The operating I history of fuel cycle facilities has shown that serious risks to worker health and safety from a nuclear criticality accident are confined to within several meters I of the critical excursion. Most criticality accidents in the U.S.A. have not I exposed workers to greater than 100 rem. Never has the health of a member of the public been impacted by a nuclear criticality accident in the U.S.A. Risks to >
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| workers from other operating hazards such as fires, equipment malfunctions or hazardous chemical releases often pose greater health and safety concerns. In addition, some high consequence accident scenarios may be low risk due to their inherently low likelihood of occurrence (e.g. certain natural phenomena).
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| Depending on the structures, systems and controls implemented at a facility, the risk of a nuclear criticality accident - the parameter to be addressed by the NRC in risk-informed, performance-based regulation - may indeed be low.
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| NCS revisions to Part 70 should consider application of a risk-informed, performance-based methodology to:
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| . evaluate the rish (i.e. consequences and likelihood) of potential nuclear criticality accidents whether 'nitiated by external events, process deviations or internal events e establish appropriate risk-based (graded) levels of protection to prevent nuclear criticality accidents
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| . . establish appropriate risk-based (graded) levels of assurance for items relied on for safety to ensure their availability and reliability i
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| 1 l
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| I' o
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| '(b) Double Contingency Draft revisions to 10 CFR Part 70 define, and now require, double contingency as a practice for managing the tisk of a nuclear criticality accident. The fuel cycle industry has for many years adhered to American National Standard ANSI /ANS-8.1 and incorporated the double contingency principle into its NCS programs. The NRC has consistently endorsed the ANSI /ANS-8.1 double contingency principle and now does so in the draft revisions to Part 70.
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| NElis conceme'd with the manner in which NUREG-1520, Standard Review Plan
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| - (SRP) for review of Part 70 license applications, requires implementation of the double contingency principle. To determine whether there are at least two j
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| 'unlikely', independent and concurrent process changes necessary before a criticality might occur (i.e. double contingency protection), industry has relied on the expertise, experience and judgment of nuclear criticality experts on a deterministic basis. Risk-informed decisions are reached by thoroughly understanding the system characteristics and performance in a nuclear criticality safety evaluation process. Whenever possible, the NRC favors use of a quantitative measure to judge implementation of, or compliance with, a principle or methodology. For example,in the case of adherence to the double contingency principle, the SRP requires assignment of specific, quantitative numerical frequencies to each of the controls to determine that a nuclear criticality accident is ' highly unlikely.' The SRP's definition of ' highly unlikely' as 4
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| a frequency of 104 is arbitrary and forces differentiation of 104 and 10 between two 'unlikely' events in a criticality accident scenario. Measuring compliance to these arbitrary, quantitative values is burdensome and problematic for both licensees and the NRC. In fact, quantification of NRC's expression of the principle of double contingency contradicts guidance of the American National Standard, which upholds the basic definition of the double contingency principle as adequate and sufficient. NEl recommends that industry's current practice of detailed evaluation of credible accident sequences by experienced nuclear criticality engineers continue. Adherence to the ANS-8 guidance should j also be continued.
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| (c) Graded Level of Protection ofitems Relled On For Safety 1 The risk of a nuclear criticality for a given accident sequence is established by ,
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| the ISA. As a nuclear criticality accident is defined to be a high consequence of l concern, its risk becomes solely a function of the accident's likelihood of i occurrence. Section 70.60(c) requires that graded safety controls (or items l I
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| relied on for safety) be implemented to prevent nuclear criticality accidents.
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| Systems of controls applied to a high risk sequence should, for example, be j inherently more robust than those applied to an intennediate risk sequence. l The choice of controls will depend on the risk (orlikelihood) of the accident l
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| )
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| sequence and may include one or more passive engineered, active engineered or administrative controls. While individual controls may vary in their relative
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| 'importance or robustness, in aggregate, they must render the risk of a nuclear criticality accident to an acceptably (low) level.
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| s70.60(c) also requires the licensee to ensure that safety controls, or items relied on for safety, to prevent a nuclear criticality accident are continuously available and reliable. In practice, specific safety controls will not be operational during maintenance and calibration testing and will not be required when the process is not operational or when Special Nuclear Material (SNM) is not present. The wording of s70.60(c) should, therefore, be modified to address the risk of a nuclear criticality accident (rather than its consequences and likelihood) and to assure that items relied on for safety are " ..available and reliable when required to perform their safety functions."
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| Section s70.60(c) incorrectly identifies only the likelihood of external events as an element of risk from a nuclear criticality accident, thereby excluding the likelihood of process deviations or other internal events as an element of the risk evaluation. In actuality, the likelihood of a process deviation or other internal event initiating an accident sequence leading to a potential nuclear criticality is probably far greater than that posed by an external event. The language of s70.60(c) should be clarified.
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| (d) Nuclear Criticality: Quality Assurance As noted in comment (a) above, the proposed revisions to Part 70 are likely to be misinterpreted as ' consequence-based' rather than ' risk-based' regulation.
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| An example of this is clearly evident in the draft SRP E5.4.4.1 " Quality Assurance for NCS." Once the appropriate level of protection is afforded to an accident scenario through the identification and application of specific controls, then appropriate assurance must be provided for these controls to ensure their availability and reliability. Draft SRP s5.4.4.1(1) incorrectly requires that all criticality safety controls be afforded the highest level of assurance, while s70.60(d)(3)(vi) and draft SRP s5.4.4.1(5) correctly require the assurance level be commensurate with the importance of the safety function.
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| Through the application of the double contingency principle, the importance of any single criticality safety control may be less than that of some other control that is the only barrier to an accident with a consequence of concern.
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| Therefore, the highest level of assurance would not necessarily be warranted for criticality controls in accident scenarios with double contingency protection.
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| In addition, the reliability of individual controls should be considered when determining the appropriate level of assurance for criticality safety controls. For
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| example, passive engineered controls in general require a lesser degree of surveillance than active engineered controls or administrative controls.
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| However, each control should be considered relative to the emironunent and application to determine the level of assurance needed to ensure the reliability and availability of the control.
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| (e) Historical Nuclear Criticality Data Part 70 license applications for operating facilities are required by s70.65(c) to include "...a description of operational events, within the past 10 years, that had a significant impact on the safety of the facility." Detailed incident teports of nuclear criticality deviations or violations, including any corrective safety measures that were implemented, are submitted to the NRC at the time of the incident or retained in licensees' records. As the NRC has on file, or available to them, voluminous information on all operational events, including nuclear criticality safety deviations, NEl sees little justification in submitting this information at the time of license application or renewal. NEl reconunends that s70.65(c) be deleted from the Part 70 revisions.
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| : 11. Concluding Remarks Nuclear criticality is an important, but readily manageable, operational hazard at fuel cycle facilities. NEl recommends that the proposed revisions of 10 CFR 70 be clarified to reduce their ambiguity and the possibility ofinterpreting them to be
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| ' consequence-based' rather than ' risk-based' regulations. The rule should permit industry to continue implementation of the double contingency principle as it has cone without imposition of a probabilistic methodology. Part 70 should be consistent with American National Standard 8 that upholds the basic definition of the double contingency principle as adequate and sufficient. In support of risk-informed, performance-based regulation, the rule should grant a license applicant the flexibility to implement graded controls (and assurances) based on the results of the ISA.
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| ker ! 'filoPan Ys Cntiubbty brety IPapen& L.ener MC Ver 4;in..
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| /F NEI h
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| .sc.::a ist:3 4 5- .
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| ""~'"''*"*'
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| November 25.1998 Dr. Carl A. Paperiello, Director Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington D.C. 20555 0001
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| ==REFERENCE:==
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| New Programmatic Criteria in NUREG-1520 (Standard Review Plan)
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| ==Dear Dr. Paperiello:==
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| At the September 29.1998 NRC NuclearInc't stry Workshop on Part 70 Regulation the Nuclear Energy Institute (NElc agreed to cite specific examples of new programmatic criteria which have been included in the proposed NUREG.
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| 1520. NEI believes these new criteria in the Standard Review Plan (SRP) are not required by the rule itself and are also not needed to create a risk informed performance-based regulatory program for Part 70 licensees. Ten areas of principal concern in the SRP have been identified:
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| (1) Quality Assurance Criteria (2) Training and Qualifications (3) Fire Safety (4) Decommissioning (5) Human Systems Interfaces (6) Organization and Administration (7) Emergency Management (8) Configuration Management (9) Maintenance (10) Nuclear Criticality Safety NEI is concerned that new presecriptive, programmatic criteria introduced in the SRP without any specific basis in 10 CFR Part 70 will become de facto regulatory requirements. Although we recognize the SRP is only intended to be a staff
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| ' NEl is the organization responsible for establishing uni 6ed nuclear industry policy on matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEl's members include all utilities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architectiengineering Orms, fuel fabrication facilities, materials licensees, and other organizations and individuals involved in the nuclear energ;y industry.
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| a i o
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| Dr. Carl A. Paperiello Nuclear Regulatory Commission November 25.1998 Page 2 guidance document to ensure consistency in license application reviews, the SRP acceptance criteria can over time become minimum standards (' lowest rung on the acceptance ladder'). The prescriptiveness of the draft SRP language is of particular concern. Though possibly not intended, it often appears to prejudge the need to implement new programs and practices before an Integrated Safety Analysis (ISA) establishes their need. In accordance with a risk-informed, performance based regulatory approach, the SRP should reflect the philosophy that the licensee will propose appropriate programmatic activities based upon the risk significance identified in the ISA, and that the reviewer should expect a sound justification for each proposal from the licensee.
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| NEI also believes that if the NRC were to morc clearly distinguish between what information is expected in a license application for a new fuel cycle operation versus that required for the renewal of an existing license, the guidance provided to the NRC reviewer in the SRP might be different and more in line with the current industry proposals.
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| The Enclosure to this letter discusses in detail each area of concern. NEI would welcome a meeting with you and your staff to discuss our concerns and to address the new programmatic elements in the draft SRP. We look forward to assisting the NRC in modifying the proposed revisions to 10 CFR 70 and to preparing an SRP that effectively supports a risk-informed, performance based approach to nuclear fuel cycle facility regulation.
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| Sincerely, A-- MM Marvin S. Fertel Enclosure cc: The Honorable Shirley A. Jackson Chairman, NRC The Honorable Greta J. Dieus, Commissioner, NRC The Honorable Nils J. Diaz, Commissioner, NRC The Honorable Edward McGaffigan, Jr., Commissioner, NRC The Honorable Jeffrey S. Merrifield, Commissioner, NRC Dr. William D. Travers, Executive Director for Operations, NRC i
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| O ENCLOSURE NUCLEAR ENERGY INSTITUTE (NEI)
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| IDENTIFICATION OF NEW PROGRAMMATIC CRITERIA IN NUREG-1520 (STANDARD REVIEW PLAN)
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| I. Introduction NEI has' expressed concern about the programmatic criteria contained in the earlier draft NUREG-1520 " Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility" (SRP) and now in the current draft SRP (June.1998). NEI's July 2,1996 testimony before the Nuclear Regulatory Commission (NRC) noted that the prior draft SRP contained "significant new programmatic criteria in areas such as quality assurance. maintenance, chemical safety. fire protection, training, and human fauors." NEI did not disagree with the need to establish programs tailored to risk. out rather with the prescriptiveness of the guidance provided in the SRP. In SECY-97-137, the NRC Staff appeared to have acknowledged our concern and indicated that:
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| "In response to licensees' concerns, staffis now proposing that, rather than require multiple safetyprograms . . .
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| licensees have the flexibility to determine, based on the ISA results, the specific elements of the safety program that would be needed."[SECY 97137. Attachment 1. p.3]
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| -NEI presumed from this comment that the revised draft SRP would eliminate references to new programmatic criteria required for licensee compliance with Part 70. NEI also anticipated that the NRC would permit licensees to determine, based on the results of their own Integrated Safety Analyses (ISA), whether any changes would be required in their existing programs, procedures and practices in order to provide reasonable assurance that the consequences of concern set forth in 70.60(B) of the rule would not be exceeded.
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| Continued inclusion of these new programmatic criteria in the latest draft SRP was one of the principal reasons that NEI's subsequent testimony before the Commission on August 25,1998 expressed concern that much of the progress achieved to date had been " illusory," and that the draft SRP was a "significant departure from how we understood our rulemaking petition was being dispositioned" [ August 25.1998 Commission meeting transcript. Testimony of Marvin Fertel).
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| Attachment to the Letter to Dr. Carl A. Paperiello New Programmatic Criteria in NUREG-1520 (SRP)
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| Page 1 4
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| /.
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| Furthermore. industry experience with NRC Staff application of some of these new criteria in individual licensing dockets - without the benefit of a formal rule change or even a final SRP --is one of the principal reasons NEI has consistently advocated inclusion of an immediately-effective backfit provision in the revised rule.
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| NEI agrees that licensees should be required by rule and/or license condition to modify their existing programs, procedures or practices, as necessary, to provide reasonable assurance that the consequences of concern of 70.60(b) will not be exceeded. However, the decision to expand or modify existing programs, procedures or practices or to adopt entirely new ones should be based upon the results of the ISA. It is this aspect of the SRP prejudgment of the need to establish new programs and practices -- with which NEl has the greatest difficulty.
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| In response to the concern expressed by NEI at the September 29.1998 XRC-Industry Workshop on Part in Remdatiw chat the draft SRP inappropriately creates de facto new substantive regulatory standards that go beyond the existing. as well as the proposed. requirements of Part 70. the NRC requested NEI to identify its areas of greatest concern. Ten areas where new.
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| proposed regulatory standards which neither are justified in the rulemaking package nor constitute a risk informed, performance-based regulatory approach have been identified and are discussed in the balance of this Enclosure. NEI does not argue with the need for a licensee to address each of these areas in a license.
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| Rather, it solely objects to addressing the implied requirements already embedded in the proposed SRP.
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| II. New Prostrammatic Criteria II.1 Quality Assurance Criteria (Draft SRP 11.3)
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| The draft SRP includes criteria for licensee quality assurance (QA) programs which are drawn directly from ANSI /ASME NQA 1-1994 " Quality Assurance Requirements for Nuclear Facility Applications" (NQA-1). SRP Acceptance Criteria specifically reference NQA-1 and Section 9.1 of NUREG-1200 " Standard Review Plan for the Review of a License Application for a Low-Level Radioactive Waste Disposal Facility," which addresses QA during design, construction and operation. According to the SRP, the intent oflisting the 18 NQA-1 criteria is to establish " reasonable assurance of the implemented QA principles in the design, Attachment to the Letter to Dr. Carl A. Paperiello New Programmatic Cnteria in NUREG-1520 (SRP)
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| Page 2 l
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| l
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| construction. operation. maintenance. modification. and decommisioning pha.-e-of a facility's life." [ Draft SRP $11.3.31. The SRP mandates that all 18 criteria are to be addressed for both high and intermediate risk accident sequences. although their application is to be graded according to risk [ Draft SRP $11.3 4 3].
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| These provisions create significant new regulatory expectations and criteria.
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| ' Imposition of NQA-1 as a requirement for compliance with 10 CFR Part 70 is a new programmatic requirement. Part 70 currently contains no explicit QA requirements, except for plutonium processing and fuel cycle facilities which must meet 10 CFR Part 50. Appendix B standards [10 CFR (( 70 22(f. 70.23(b)]. The draft regulation would require that items relied on for safety and safety controls meet
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| " quality standards . . commensurate with the importance of the safety functions performed." [ Draft 10 CFR f 70.60(dx3)(vi)). While the NRC's Branch Technical Position on Alanagement Controls / Quality Assurance for Fuel Cycle Facilities,54 Fed.
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| Reg.11590 (31 arch 21.1989) advises licensees to put in place various QA functions (including an effective management organization, plant safety committee.e. p acedural controls, surveillar. e, inspnction and audit programs. and training programs). it does not specify implementation of a full NQA-1 QA program 1 The SRP " prejudges" that a licensee's quality program. must conform to the NQA 1 criteria.
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| Imposition of NQA-1 on fuel facility licensees would necessitate radical changes in virtually all affected licensees' quality programs. The NRC's own draft Regulatory Analysis predicts that under the draft rule and SRP, five of seven affected Part 70 licensees would be given " essentially no credit for existing [QA]
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| measures, and that an entirely new program would have to be established."
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| [SECY.98185. Attachment 3. Regulatory Analysis (Draft). p.20]
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| The imposition of NQA 1 on Part 70 licensees, whether on a graded basis or otherwise, is a substantial change in practice and policy. It should not be injected as a new " expectation" either in an SRP, or through informal case-by-case licensing action, unless specifically included as a Part 70 rule requirement.
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| Reference to NUREG 1200 on " design" and " construction" activities creates QA criteria for design and construction of non-plutonium Part 70 facilities. This is a new programmatic requirement that is not embodied in the Part 70 rule and that is not consistent with licenses that have been issued. Under existing regulations, only plutonium processing and fuel cycle facilities have been subjected to design and construction requirements in the licensing process (10 CFR ff 70.22(0. 70.23(b)].
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| ' NQA 1 is cited as a reference in the Branch Technical Position (BTP).
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| Attachment to the Letter to Dr. Cari A. Paperiello New Programmatic Criteria in NUREG 1520 (SRP)
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| Page 3
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| e Licenses issued untier Part 70 are for the pre--mn anti use of-peeml nuewar material. rather than for the construction and operation of facilities. Accordingly.
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| most Part 70 licensees have designed and constructed their facilities without NRC review and approval of design criteria or activities. or construction practices.
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| NRC licensing reviews have been focused on the as designed as-built facility and ;
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| its ability to assure safe possession and use of radioactive material. The creation ;
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| of QA criteria for design and construction of Part 70 facilities is not a requirement l of the Part 70 rule. The SRP does not address how existing licensed facilities !
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| would have to comply with these new design and construction requirements. !
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| Quality Assurance programs and measures should be appropriate to the level of risk of a potential accident sequence or appropriate to the level of responsibility of I a staff position. Quality Assurance progran.s must be graded and implemented '
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| according to risk. For example, the number of NQA 1 criteria which an individual ;
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| program must address - even for high and intermediate risk events - can only be established following completion of the appropriate ISA.
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| II.2 Traininst and Qualification (Draft SRP 611.4)
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| The draft SRP establishes the elements of a " Systems Approach to Training" !
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| (SAT) program as the acceptance criterion for licensees' training and qualification programs. It references NL* REG-1220, Rev.1 " Training Reciew Criteria and Procedures" as the primary regulatory reference document (Draft SRP li 11.4.4. 11.4 3]. In addition. the SRP states that the Staffis to ensure that personnel have the knowledge and skills necessary "to design (and) construct" the facility (Draft SRP $11.4.1].
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| The SRP introduces two new programmatic requirements: adoption of SAT and requiring staff to be knowledgeable in the design and construction of the facility.
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| These concepts have only been applied as a licensing requirement for certain specific job categories at commercial nuclear power plants under 10 CFR 50.
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| There is no requirement in the Part 70 rule which requires such a comprehensive level of staff training as that mandated in the SRP.
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| Risk-informed, performance-based regulation grants a licensee the latitude to establish the content, detail and comprehensiveness ofits staff training and qualification program. The scope of the program will be established based upon the results of the ISA and specifically by the graded level of risk associated with each operator task and the required level of responsibility. If the results of the ISA indicate a need for enhanced training of certain equipment operators (i.e. a Attachment to the Letter to Dr. Carl A. Papenello New Programmatic Criteria in NUREG-1520 (SRP)
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| Page 4
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| ~ vulnerability" exposed by the ISA). due m the iteenreei reliance on actions by those operators to prevent excessive radiation exposures, the licensee will determine the most appropriate way to address the training needs (e.g. increase the frequency of the operators' training expand the content of the training. or impose new qualification requirements). Such actions may be adequate and effective in addressing the identified vulnerability in the context of the licensee's existine training program. A SAT program may not be warranted.
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| Imposition of SAT criteria for nuclear power plant operators is required by a specific regulation (10 CFR 50.120) which establishes SAT as a formal regulatory requirement for certain designated categories of personnel. Proposed Part 70 revisions set a new and higher standard for performance (SAT)in the absence of a Part 70 regulation (analogous to Part 50.120) and before the results of an ISA demonstrate the need for that level of performance.
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| Extreme care should be taken in referring to NUREG 1220. a regulatory guidance 4 document created for nuclear power plant hee sees. as the basic regulatory {
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| reference for Part 70 facilities. The SRP does not justify how operator knowledge and skills in " design" and " construction" activities at non-plutonium licensed fuel cycle facilities enhances health and safety. Adoption of such standards represents a significant departure from current licensing practice and the rulemaking I package does not discuss the implications of this change. Different training requirements may be appropriate for new fuel cycle facilities, particularly if a new process or technology is to be used where there is a dearth of operating, safety and performance history. The SRP should differentiate between the staff training and qualification requirements for new and existing fuel cycle facilities.
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| The Qualifications, Training and Human Performance Requirements detailed in the SRP are very prescriptive and cumbersome, are inconsistent with current industry practice and will result in only a marginal positive impact on the effectiveness of facility training programs. Such requirements should only be established by the licensee using the results of the ISA.
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| II.3 Fire Safety (Draft SRP 67.0)
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| The draft SRP includes in its acceptance criteria provisions governing the development of a formal and comprehensive " Fire Protection Program" (FPP),
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| performance of Fire Hazards Analyses (FHAs), and development of specific Pre-Fire Plans (PFP). It also establishes the "most current versions of. . .
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| nationally recognized codes and standards" as the basis for Staff reviews [ Draft SRP Attachment to the Letter to Dr. Carl A. Paperiello New Programmatic Critena in NUREG-1520 (SRP)
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| Page 5
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| l I
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| i nationally recocnue.1 code. .ind standard.- a- t he basis for Staff reviews [Dra6.qP ;
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| 6 :.u]. and lists as references. 58 National Fire Protection Association (NFPA) codes and several other standards. The requirement for an FPP, FHA and PFP constitutes a new set of programmatic requirements.
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| Licensees currently have in place controls to prevent. detect and mitigate fires.
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| Part 70 does not require a licensee to establish formal FPPs, to perform detailed l FHAs, or to put in place specific PFPs as a license compliance condition. Unless '
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| the risk of an accident sequence justifies, or a specific provision written into the Part 70 rule mandates this comprehensive level of fire safety. FPPs, FHAs and PFPs may not be warranted.
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| In NRC Generic Letter 95 01 WRC Staff Technical Position on Fire Protection for Fuel Cycle Facilities" (January 26.1995). the NRC directed licensees to describe how they intended to implement the guidance set forth in its Branch Technical Position (BTP) on Fire Protection for Fuel Cycle Facilities," 57 Fed. R_e_g. 35607 (Aug.10.1992' which calls for conduct of FE/.s and for the preparation of PFPs.
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| In response. NEI and a number oflicensees stated that the Generic Letter appeared to establish new requirements. NEI. in particular, stated:
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| 1 There is no extant regulation nor order that requires . .
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| Part 70 licensees to prepare pre-fire plans [or) to conduct
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| [ ire hazard analyses. [ Letter Fehx .N!. IGilar. Jr. to Robert F.
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| Burnett. February 23.1995}.
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| The listing of 58 NFPA codes and the statement that the "most current eersions" of those codes will be utilized as the basis for Staff reviews clearly creates new regulatory expectations that may be very costly to achieve and may require licensees to continually upgrade their facilities to meet newly developed industry codes without any commensurate reduction in risk. Requiring facilities to be constructed in accordance with NFPA codes that are current at the time of construction is not an unreasonable regulatory demand. The SRP prejudges that application of these codes is necessary without the benefit of any ISA results, and thus departs from a risk-informed. performance based approach to regulation.
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| II.4 Decommissioning (Draft SRP 610.0)
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| NEI's comments on this portion of the SRP presume that the chapter on decommissioning is intended to apply to the review of new, renewed, amended or Attachment to the Letter to Dr. Carl A Paperiello New Programrnatic Criteriain NUREG 1520(SRP) s Page 6 l l
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| resubmitted"2 license application.< for oneratine facilities, and not to facihtie.-
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| that have permanently shutdown and intend to decommission.
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| The stated purpose of this portion of the draft SRP is to ensure that an "avolicants' plans for decommissioning . . . provide reasonable assurance that the avolicant will be able to decommission the facility safely and in accordance with NRC requirements." [ Draft SRP S 10.1 (emphasis added)].
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| The SRP discusses Reg. Guide 3.G5 " Standard Format and Content of Decommissioning Plans for Licensees under 10 CFR Parts 30, 40, and 70," and various other guidance documents. These documents are used primarily, if not exclusively, to review detailed decommissioning plans that are not required to be submitted until after a licensee has ceased active operations and intends to decommission its facility. Such plans, filed close to the time that actual decommissioning activities are expected to occur, must provide detailed information or., among other things, proposed release criteria for land and facilities, survey methods and criteria, procedures for radiological protection during the decommissioning process, and plans for waste disposal. By contrast, at present, licensees at operating faci'lities must simply submit a cost estimate for decommissioning and provide financial assurance through a decommissioning funding plan, as part of a licensing submittal.
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| At the time oflicense application the SRP requires submission of a detailed I decommissioning plan and detailed procedures to minimize contamination to the
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| {
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| environment. This constitutes a new programmatic requirement.
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| Forecasting the methodologies or technologies to be used to decommission a facility 20 to 40 years in the future is an unreasonable requirement. Technologies j which might be used for decontamination of equipment and soils may not yet be i commercially available at the time oflicense application. The implication of this !
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| portion of the SRP is that the detailed review criteria applicable to decommissioning plans.will become regulatory review criteria for " applicants" for licenses, renewals, amendments or resubmittals. Application of those review criteria at this early stage is premature and unnecessary. ,
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| 1 NEl believes that this entire chapter should be removed from the SRP and placed in a Regulatory Guidance document Draft 10 CFR {70.62(a)(1) requires existing licensees to resubmit new applications within four years of the !
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| effective date of the rule, even if their licensees are not due to expire.
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| Attachrnent to the Letter to Dr. Carl A. Papenello New Programmatic Criteria in NUREG 1520 (SRP)
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| Page 7
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| 1 l
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| II.5 Human-System Interfaces (Draft SRP S11.6)
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| The draft SRP requires ["f]ormal evaluation of human system interfaces" and requires licensees to have a formal process for " design. evaluation.
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| implementation, maintenance, and modification of human system interfaces . . "
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| [ Draft SRP ll 11.6.3. 11.6.4.3]. This includes periodic human; system interface reviews, employment of human-system interface " specialists," development of human-system " standards" and creation of an " inventory" of such interfaces. This portion of the SRP is a new programmatic requirement. It creates an entirely new and complex set of criteria that will require licensees to establish detailed programs and procedures to formally analyze interfaces between personnel and systems. Additionally, it prejudges that control of human-system interfaces is I needed, regardless of the results of the ISA. The draft SRP departs from a risk-informed, performance based approach to regulation.
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| II.6 Organization and Administration (Draft SRP 62.0) 1 There are several aspects of this chapter of the SRP that create new expectations and go well beyond the rule and existing practice. First. this portion of the SRP provides for NRC Staff review of the applicant's organizational structure and policies governing " design" and " construction" activities and the qualifications of design and construction personnel [ Draft SRP S2.3]. Licenses issued under Part 70 are not for the construction and operation of facilities, but rather for the possession and use of special nuclear material. Therefore, specifying policies on design and construction in the SRP is unwarranted. This represents a substantial change in policy and practice.
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| Second, the SRP provides for NRC Staff review of the " experience" and
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| " availability" of personnel for decommissioning oflicensed facilities [ Draft SRP S2.4.3]. Again, review of such details associated with the actual decommissioning process at the licensing stage is premature. What contractors and personnel will be available in 20 to 40 years to oversee decommissioning cannot reasonably be expected to be known now. This constitutes an unnecessary new requirement.
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| Finally, the SRP calls for NRC review and approval ofinternallicensee mechanisms for reporting safety concerns and for their investigation, assessment ;
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| and resolution. The SRP anticipates that the NRC will make an affirmative i determination that the license applicant " promotes an open environment that supports safety and is absent of any chilling effect that discourages prompt and open reporting of safety concerns" [ Draft SRP $2.4.3]. This aspect of the SRP goes far Attachment to the Letter to Dr. Carl A. Papenello New Programmatic Criteria in NUREG 1520 (SRP)
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| Page 8 l
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| r I
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| {
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| beyond even NRC regulation of reactor licensees. Part 70 licensees clearlv have an obligation under 10 CFR 70.7 not to discriminate against employees for raising safety concerns or for otherwise engaging in protected activities.
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| Enforcement action may be taken by the NRC for violations. Alllicensees have 4 programs to report and dispose of safety concerns and to ensure an open environment for raising such concerns. However, neither reactor nor materials l
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| licensees have been required, as part of a license review or otherwise, to establish i such programs, nor has the NRC ever to NEI's knowledge imposed " lack of a chilling environment" as a licensing standard. In early 1998 the NRC considered, f i but withdrew, a proposal to impose additional requirements to ensure that I licensees maintain a " safety conscious work environment" (SCWE). The NRC '
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| concluded that its existing non-binding Policy Statement on " Freedom of 1
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| Employees in the Nuclear Industry to Raise Safety Concerns Without Fear of '
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| Retaliation" and its regulat. ions prohibiting discrimination (e.g.10 CFR 70.7) are sufficient [63 Eed. Brg 6235 (Feb. 6.1998)]. Under those policies and regulations, the NRC has established certain " expectations" for its licensees and has given itself the enforcement tools to respond if discriminati in, or the potential for discrimination occurs. Furthermore, the NRC's Executive Director for Operations )
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| recently proposed to the Commission that it " consider the possibility of discontinuing any agency efforts to independently assess SCWE" [SECY 98176
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| ~ Proposed Options for Assessing a Licensee's Safety Conscious Work Ennronment"(July 21,1998)]. The Commission ultimately decided to continue the existing Staff practice of using inspection and enforcement techniques to ensure a SCWE [ Staff Requirement.
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| Memorandum (September 1.1998)]. Imposing licensine standards for the maintenance of a SCWE goes well beyond existing practice and requirements and is inconsistent with the Commission's February and September policy determinations. ;
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| II.7 Emergency Management (Draft SRP 68.0) l This portion of the SRP calls for the licensee to establish an adequate emergency response training program, not only for onsite workers, but also for "offsite emergency response personnel" in order to ensure that such personnel have adequate " knowledge of the emergency plan, assigned duties, and effectively respond to an actual emergency." The licensee must provide the " topics and general content" of the training used for "offsite emergency response personnel"
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| [ Draft SRP $8 4.3.2.11]. Part 70 currently requires that licensees provide a "brief description of {among other things] any special instructions and orientation tours the licensee would offer to fire, police, medical and other emergency personnel,"
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| but it does not require formal training of such personnel [to crR STo.22(i>(3)(x)].
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| NRC's own analysis did not identify significant off-site risks. The language of the Attachment to the Letter to Dr. Carl A. Paperiello New Programmatic Criteria in NUREG-1520 (SRP)
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| Page 9 L_m_____.___.
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| draft SRP appears, therefore, to go beyond existmg requirements and suggests an emergency response training program that is more akin to those established for commercial nuclear power plants.
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| The design of an Emergency Management program must be based on the risks that could be posed to public health and safety and the environment by the facility and its operation. Until such risks are assessed in an ISA, the components and requirements of an emergency management plan can not be accurately defined.
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| The SRP must allow the licensee to establish appropriate emergency response measures and to determine the extent of training which should be provided to "offsite emergency response personnel."
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| II.8 Configuration Management (Draft SRP 611.1)
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| This portion of the SRP creates detailed new criteria for the establishment of a formal configuration management program. NEI has not yet assessed all of the implications of the new configuration management criteria but one of the most significant, and entirely new requirements. is the expectation that licensees will be required to " reconstitute" their " designs" [ Draft SRP l11.1.3(6) 11.1.5.26). In particular:
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| * licensees must reconstitute "the current design bases, supporting analyses, requirements, and documentation that support items important to safety";
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| e if documentation has not "hept pace"with the as-built plant configuration, licensees must " walk down systems, update drawings and specifications, perform new calculations and analyses, and otherwise rebuild the design bases."[ Draft SRP $511.1.4.3.11.1.5.2}.
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| The provisions constitute a new programmatic requirement. Provisions for design bases reconstruction go well beyond existing requirements and, in fact, substantially exceed the requirements applied to nuclear power plants. Part 70 licenses do not " license" the design of a facility and so there should be no requirement to perform a reconstitution. Unless dictated by the results of an ISA, reconstitution of fecility design documents may not in itselflikely result in any significant improvement to safety.
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| Operators of new and existing fuel cycle facilities should commit to a configuratior inanagement program in their licenses. However, the requirement Attachment to the Letter to Dr. Carl A. Paperiello New Programmatic Criteria in NUREG 1520 (SRP)
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| Page 10
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| v*
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| that current licensees now undertake a major design reconstitution after their facilities have been safely operated for many years, seems unnecessary.
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| Justification of how this action might improve facility safety is lacking.
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| II,9 Maintenance (Draft SRP 611.2_)
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| This portion of the SRP creates entirely new requirements patterned after commercial nuclear power plant requirements and guidance for maintenance programs. It appears to apply the concepts of preventive and corrective maintenance to " human performance" and activities, and references a wide range of guidance documents applicable to reactor maintenance programs. [ Draft SRP li 11.2. 4. 3, 1 1.2. -] . For example, corrective and preventive maintenance practices are to be applied to " items relied on for safety" -- which are defined in the draft regulation (10 CFR 70.4) to include " activities of personnel" (Draft SRP l11.2.4.3].
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| The discussion of preventive maintenance specifically discusses "requalification and retraining of personnel" [ Draft SRP $ 11.2.4.3]. This is a unique and to the best of our knowledgn. unprecedented extension 4 the concept of a nuclear facility maintenance program, it is not clear what additional requirements this would add over the proposed training program criteria in SRP 11.4.
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| i The requirement for a nuclear power plant maintenance program is required by a specific regulation (10 CFR 50.65). In the absence of a corresponding requirement in the Part 70 rule, the NRC should not attempt to impose a highly prescriptive maintenance program either through the SRP or as a license condition.
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| The draft SRP contains very extensive requirements related to maintenance. For example, the draft SRP appears to require preventive maintenance and post maintenance functional tests, regardless of whether such activities are needed to ensure the proper functioning ofitems relied on for safety as identified by the ISA. l As a result, the draft SRP departs from a risk informed, performance based approach to regulation. ,
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| II-10 Nuclear Criticality Safety (Draft SRP 65.0)
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| This portion of the SRP calls for adherence to the well established principle of
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| " double contingency protection" in order to provide for adequate nuclear criticality safety and includes a definition of" double contingency protection" that is inconsistent with American National Standard ANSI /ANS-8.1. The SRP goes well beyond accepted international and nuclear industry practice by assigning specific, l Attachment to the Letter to Dr. Carl A. Paperiello New Programmatic Criteria in NUREG.1520 (SRP)
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| Page 11
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| n quantitative, numerical frequencie.e to each of the two controlled parameters or controls as an acceptance criterion. presumably m order to determine that a particular nuclear criticality accident is " highly unlikely. '
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| In particular, the SRP indicates that one controlled parameter or control should have a frequency of failure no greater than 104 per year. and the other controlled parameter or control should have a frequency of failure no greater than 10 2 per year [ Draft SRP 55.4 6). Neither ANSI nor, to the best of our knowledge, the NRC has adopted this quantitative approach to criticality safety in the past. Instead, it has been a well accepted practice for the determination as to whether there are at least two "unlikely," independent and concurrent process changes necessary before a nuclear criticality might occur (i.e. double contingency protection) to be made on the basis of the expertise, experience and judgment of nuclear criticality safety experts on a deterministic basis.
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| Risk-informed. performance based regulation allows a licensee to determine the risk of potential nuclear criticalities in his 'acility. The rish of a nuclear criticality from each credible accident sequence will be assessed in the facility's ISA. Results of the ISA will guide the licensee in selecting and implementing nuclear criticality mitigative practices and measures that are appropriate to the level of the nsk.
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| Adoption of these new quantitative standards will add considerably to the cost and complexity of performing nuclear criticality safety analyses. . Furthermore, adherence to the traditional methods of applying the double contingency
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| - protection principle will still enable licensees to evaluate and determine whether a nuclear criticality is "unlikely" as required by the' draft rule. In industry's view, if adherence to the double contingency protection principle is confirmed, then it follows that a nuclear criticality event would be " highly unlikely."
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| III. Conclusions The June 1998 draft SRP contains references to a wide range of new regulatory expectations and standards that have not been justified in the rulemaking package and are not consistent with a risk informed, performance-based regulatory program. The rulemaking record is replete with explanations as to the purpose of the requirements to perform ISAs. to adopt consequences of concern, to identify items relied on for safety, and to assure that such items remain available and reliable. It does not, however, explain at all the bases for the determination that the wide range of new programmatic criteria in the draft SRP is necessary or Attachment to the Letter to Dr. Carl A. Papenello New Programmatic Cnteria in NUREG-1520 (SRP)
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| Page - 12
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| r-i I
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| appropriate, Furthermore by generically pre-cr:ning .-uch criteria. without the benefit ofISA results the draft SRP runs counter to the effort to establish a risk informed, performance based regulatory regime. As NEI stated in its July.
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| 1996 testimony before the Commission. it believes that fundamental changes are required to both the draft Part 70 rule and the draft SRP before either can be adopted as regulatory standards or guidance.
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| Ref I Fdes Part o SMP Propammaue Commenw10-10-944 l
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| Attachment to the Letter to Dr. Carl A. Papenello New Programmatic Criteria m NUREG-1520 (SRP)
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| Page 13
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| NEI W
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| s;ts: :a .,i::. si- ; .
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| November 25.1998 """'"*"*'
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| Dr. Carl A. Paperiello, Director Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington. D.C. 20555 0001
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| ==REFERENCE:==
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| New Prograrnmatic Criteria in NUREG-1520 (Standard Review Plan)
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| ==Dear Dr. Paperiello:==
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| At the September 29.1998 NRC Nuclear Inc'estry Workshop on Part 70 Regulation the Nuclear Energy Institute (NElc agreed to cite specific examples of new programmatic criteria which have been included in the proposed NUREG-1520. NEI believes these new criteria in the Standard Review Plan (SRP) are not required by the rule itself and are also not needed to create a risk informed performance based regulatory program for Part 70 licensees. Ten areas of principal concern in the SRP have been identified:
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| (1) Quality Assurance Criteria (2) Training and Qualifications (3) Fire Safety (4) Decommissioning (5) Human Systems Interfaces (6) Organization and Administration (7) Emergency Management (8) Configuration Management (9) Maintenance (10) Nuclear Criticality Safety NEI is concerned that new presecriptive, programmatic criteria introduced in the SRP without any specific basis in 10 CFR Part 70 will become de facto regulatory requirements. ' Although we recognize the SRP is only intended to be a staff
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| ' NEl is the organization responsible for establishing uni 6ed nuclear industry policy on matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEI's members include all utilities licensed to operate commercial nuclear power plants in the United States. nuclear plant designers, major architect, engineering Orms fuel fabrication facilities, materials licensees, and other organizations and individuals involved in the nuclear energy industry.
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| Dr. Carl A. Paperiello Yuclear Regulatory Commission Noumber 25,1998 Page 2 guidance document to ensure consistency in license application reviews, the SRP acceptance criteria can over time become minimum standards (' lowest rung on the acceptance ladder'). The prescriptiveness of the draft SRP language is of particular concern. Though possibly not intended, it often appears to prejudge the need to implement new programs and practices before an Integrated Safety Analysis (ISA) establishes their need. In accordance with a risk informed, performance-based regulatory approach, the SRP should reflect the philosophy that the licensee will propose appropriate programmatic activities based upon the risk significance identified in the ISA, and that the reviewer should expect a sound justification for each proposal from the licensee.
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| NEI also believes that if the NRC were to morc clearly distinguish between what information is expected in a license application for a new fuel cycle operation versus that required for the renewal of an existing license. the guidance provided to the NRC reviewer in the SRP might be different and more in line with the current industry proposals.
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| The Enclosure to this letter discusses in detail each area of concern. NEI would welcome a meeting with you and your staff to discuss our concerns and to address the new programmatic elements in the draft SRP. We look forward to assisting the NRC in modifying the proposed revisions to 10 CFR 70 and to preparing an SRP that effectively supports a risk informed, performance based approach to nuclear fuel cycle facility regulation. !
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| Sincerely, f^-~ NN Marvin S. Fertel Enclosure cc: The Honorable Shirley A. Jackson Chairman, NRC The Honorable Greta J. Dicus, Commissioner, NRC i The Honorable Nils J. Diaz, Commissioner, NRC I The Honorable Edward McGaffigan, Jr., Commissioner, NRC The Honorable Jeffrey S. Merrifield, Commissioner, NRC Dr. William D. Travers, Executive Director for Operations, NRC l
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| i
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| L. 1 ENCLOSURE NUCLEAR ENERGY INSTITUTE (NEI)
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| IDENTIFICATION OF NEW PROGRAMMATIC CRITERIA 3 IN NUREG-1520 (STANDARD REVIEW PLAN) !
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| I. Introduction
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| )
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| NEI has expressed concern about the programmatic criteria contained in the I i
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| earlier draft NUREG-1520 " Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility" (SRP) and now in the current draft SRP (June,1998). NEl's July 2,1996 testimony before the Nuclear Regulatory ,
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| Commission (NRC) noted that the prior draft SRP contained "significant new programmatic criteria in areas such as quality assurance, maintenance, chemical safety, fire protection, training, and human fauors." NEI did not disagree with the need to establish programs tailored to risk. out rather with the .
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| l prescriptiveness of the guidance provided in the SRP. In SECY-97-137, the NRC f
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| Staff appeared to have acknowledged our concern and indicated that: l "In response to licensees' concerns, staffis nowproposing that, rather than require multiple safetyprograms . .
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| licensees have the flexibility to determine, based on the ISA results, the specific elements of the safety program that would be needed."[SECV 97-137. Attachment 1, p.3]
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| NEI presumed from this comment that the revised draft SRP would eliminate references to new programmatic criteria required for licensee compliance with i Part 70. NEI also anticipated that the NRC would permit licensees to determine, based on the results of their own Integrated Safety Analyses (ISA), whether any ;
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| changes would be required in their existing programs, procedures and practices in
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| {
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| order to provide reasonable assurance that the consequences of concern set forth in S70.60(B) of the rule would not be exceeded. l Continued inclusion of these new programmatic criteria in the latest draft SRP was one of the principal reasons that NEI's subsequent testimony before the Commission on August 25,1998 expressed concern that much of the progress achieved to date had been " illusory," and that the draft SRP was a "significant departure from how we understood our rulemaking petition was being dispositioned" [ August 25.1998 Commission meeting transcript. Testimony of Mamn Fertel].
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| Attachment to the Letter to Dr. Carl A. Papenello New Programmatic Criteria in NUREG 1520 (SRP)
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| Page 1 j
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| l
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| ~O Furthermore. industry experience with NRC Staff application of some of the.se l new criteria in individuallicensing dockets - without the benefit of a formal rule I change or even a final SRP -- is one of the principal reasons NEI has consistently advocated inclusion of an immediately-effective backfit provision in the revised I rule.
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| 1 NEI agrees that licensees should be required by rule and/or license condition to l modify their existing programs, procedures or practices, as necessary, to provide I reascnable assurance that the consequences of concern of 70.60(b) will not be exceeded. However, the decision to expand or modify existing programs, procedures or practices or to adopt entirely new ones should be based upon the results of the ISA. It is this aspect of the SRP -prejudgment of the need to establish neu> programs and practices - with which NEl has the greatest difficulty.
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| 'In response to the concern expressed by NEI at the September 29,1998 XRC-Industry (Vorkshop on Part in Rennlatio > chat the draft SRP inappropriately creates de facto new substantive regulatory standards that go beyond the existing. as well as the proposed, requirements of Part 70, the NRC requested NEI to identify its areas of greatest concern. Ten areas where new, proposed regulatory standards which neither are justified in the rulemaking package nor constitute a risk-informed, performance based regulatory approach have been identified and are discussed in the balance of this Enclosure. NEI does not argue with the need for a licensee to address each of these areas in a license.
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| Rather, it solely objects to addressing the implied requirements already embedded in the proposed SRP.
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| II. New Programmatic Criteria II.1 Quality Assurance Criteria (Draft SRP {11.3)
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| The draft SRP includes criteria for licensee quality assurance (QA) programs which are drawn directly from ANSI /ASME NQA-1-1994 " Quality Assurance Requirements for Nuclear Facility Applications" (NQA-1). SRP Acceptance Criteria specifically reference NQA-1 and Section 9.1 of NUREG-1200 " Standard Review Plan for the Review of a License Application for a Low Level Radioactive Waste Disposal Facility," which addresses QA during design, construction and operation. According to the SRP, the intent oflisting the 18 NQA-1 criteria is to establish " reasonable assurance of the implemented QA principles in the design, Attachment to the Letter to Dr. Carl A. Paperiello New Programmatic Cnteria in NUREG-1520 (SRP)
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| Page 2
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| N construction. operation. maintenance. moditication, and decommissioning pha-e-of a facility's life." [ Draft SRP 511.3.31 The SRP mandates that a_ll18 criteria are to be addressed for both high and intermediate risk accident sequences. although their application is to be graded according to risk [ Draft SRP 511.3.4 3].
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| These provisions create significant new regulatory expectations and criteria.
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| Imposition of NQA-1 as a requirement for compliance with 10 CFR Part 70 is a new programmatic requirement. Part 70 currently contains no explicit QA '
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| requirements, except for plutonium processing and fuel cycle facilities which must meet 10 CFR Part 50, Appendix B standards [10 CFR l 70 22(f). 70.23(b)]. The draft regulation would require that items relied on for safety and safety controls meet
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| " quality standards . commensurate with the importance of the safety functions performed." [ Draft 10 CFR j 70.60(dx3)(vi)]. While the NRC's Branch Technical Position on Management Controls / Quality Assurance for Fuel Cycle Facilities. 54 Fed.
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| Rec.11590 OIarch 21,1989) advises licensees to put in place various QA functions (including an effective management organization, plant safety committees, p acedural controls, surveillar. e. inspoction and audit programs. and training programs), it does not specify implementation of a full NQA-1 QA program 1 The SRP " prejudges" that a licensee's quality program, must conform to the NQA-1 criteria.
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| Imposition of NQA-1 on fuel facility licensees would necessitate radical changes in virtually all affected licensees' quality programs. The NRC's own draft Regulatory Analysis predicts that under the draft rule and SRP, five of seven affected Part 70 licensees would be given " essentially no credit for existing [QA]
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| measures, and that an entirely new program would have to be established."
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| (SECY-98185. Attachment 3. Regulatory Analysis (Draft), p.20]
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| The imposition of NQA 1 on Part 70 licensees, whether on a graded basis or otherwise, is a substantial change in practice and policy. It should not be injected as a new " expectation" either in an SRP, or through informal case by-case licensing action, unless specifically included as a Part 70 rule requirement.
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| Reference to NUREG-1200 on " design" and " construction" activities creates QA criteria for design and construction of non-plutonium Part 70 facilities. This is a new programmatic requirement that is not embodied in the Part 70 rule and that is not consistent with licenses that have been issued. Under existing regulations, only plutonium processing and fuel cycle facilities have been subjected to design and construction requirements in the licensing process (10 CFR ss 70.22(f). 70.23(b>].
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| ' NQA-1 is cited as a reference in the Branch Technical Position (BTP).
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| w Attachment to the Letter to Dr. Carl A. Papenello New Programmatic Criteria in NUREG-1520 (SRP)
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| Page 3
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| ,l
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| )
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| Licenses issued under Part 70 are for the pr.-e--ton and use of-pecial nuclear material. rather than for the construction and operation of facilities. Accordingly.
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| most Part 70 licensees have designed and constructed their facilities without NRC review and approval of design criteria or activities, or construction practices.
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| NRC licensing reviews have been focused on the as. designed, as. built facility and its ability to assure safe possession and use of radioactive material. The creation of QA criteria for design and construction of Part 70 facilities is not a requirement of the Part 70 rule. The SRP does not address how existing licensed facilities would have to comply with these new design and construction requirements.
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| Quality Assurance programs and measures should be appropriate to the level of risk of a potential accident sequence or appropriate to the level of responsibility of a staff position. Quality Assurance progran.s must be graded and implemented according to risk. For example, the number of SQA 1 criteria which an individual program must address - even for high and intermediate risk events - can only be established following completion of the appropriate ISA.
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| II.2 Training and Qualification (Draft SRP 611.4)
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| The draft SRP establishes the elements of a " Systems Approach to Training" (SAT) program as the acceptance criterion for licensees' training and qualification programs. It references NUREG-1220, Rev.1 " Training Review Criteria and Procedures" as the primary regulatory reference document (Draft SRP li 11.4.4. 11.43]. In addition, the SRP states that the Staffis to ensure that personnel have the knowledge and skills necessary "to design (and] construct" the facility (Draft SRP $11.4.1}.
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| The SRP introduces two new programmatic requirements: adoption of SAT and requiring staff to be knowledgeable in the design and construction of the facility.
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| These concepts have only been applied as a licensing requirement for certain specific job categories at commercial nuclear power plants under 10 CFR 50.
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| There is no requirement in the Part 70 rule which requires such a comprehensive level of staff training as that mandated in the SRP.
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| Risk informed, performance based regulation grants a licensee the latitude to establish the content, detail and comprehensiveness ofits staff training and qualification program. The scope of the program will be established based upon the results of the ISA and specifically by the graded level of risk associated with each operator task and the required level of responsibility. If the results of the ISA indicate a need for enhanced training of certain equipment operators (i.e. a Attachment to the Letter to Dr. Carl A Paperiello New Programmatic Cnteria in NtJREG-1520 (SRP)
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| Page 4
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| i O .
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| " vulnerability" exposed by the ISAn tiue to the been ee . reliance on actwn, by those operators to prevent excessive radiation exposures. the licensee will determine the most appropriate way to address the training needs (e.g. increase the frequency of the operators' training. expand the content of the training, or impose new qualification requirements). Such actions may be adequate and effective in addressing the identified vulnerability in the context of the licensee's existing training program. A SAT program may not be warranted.
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| Imposition of SAT criteria for nuclear power plant operators is required by a specific regulation (10 CFR 50.120) which establishes SAT as a formal regulatory requirement for certain designated categories of personnel. Proposed Part 70 revisions set a new and higher standard for performance (SAT)in the absence of a Part 70 regulation (analogous to Part 50.120) and before the results of an ISA demonstrate the need for that level of performance.
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| Extreme care should be taken in referring to NUREG-1220. a regulatory guidance document created for nuclear power plant hcensees. as the basic regulatory reference for Part 70 facilities. The SRP does not justify how operator knowledge and skills in " design" and " construction" activities at non-plutonium licensed fuel cycle facilities enhances health and safety. Adoption of such standards represents a significant departure from current licensing practice and the rulemaking package does not discuss the implications of this change. Different training requirements may be appropriate for new fuel cycle facilities, p.irticularly if a new process or technology is to be used where there is a dearth of operating, safety and performance history. The SRP should differentiate between the staff training and qualification requirements for new and existing fuel cycle facilities.
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| The Qualifications. Traming and Human Performance Requirements detailed in the SRP are very prescriptive and cumbersome, are inconsistent with current industry practice and will result in only a marginal positive impact on the effectiveness of facility training programs. Such requirements should only be established by the licensee using the results of the ISA.
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| 11.3 Fire Safety (Draft SRP 67.0)
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| The draft SRP includes in its acceptance criteria provisions governing the development of a formal and comprehensive " Fire Protection Program"(FPP),
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| performance of Fire Hazards Analyses (FHAs), and development of specific Pre Fire Plans (PFP). It also establishes the "most current versions of.
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| nationally recognized codes and standards" as the basis for Staff reviews [ Draft SRP Attachment to the Letter to Dr Carl A. Paperiello New Programmatic Critena in NUREG-1520(SRP)
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| Page 5
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| natmnally recocmze.1 emie .ind standard- a- e ban- for Staff review.- [ Draft e s i 3] and lists as references. 58 National Fire Protection Association (NFPA) codes and several other standards. The requirement for an FPP, FHA and PFP constitutes a new set of procrammatic requirements.
| |
| Licensees currently have in place controls to prevent. detect and mitigate fires.
| |
| Part 70 does not require a licensee to establish formal FPPs, to perform detailed FHAs, or to put in place specific PFPs as a license compliance condition. Unless the risk of an accident sequence justifies, or a specific provision written into the Part 70 rule mandates this comprehensive level of fire safety. FPPs. FHAs and PFPs may not be warranted.
| |
| In NRC Generic Letter 95 01 'NRC Staff Technical Position on Fire Protection for Fuel Cycle Facilities" (January 26,1995), the NRC directed licensees to describe how they intended to implement the guidance set forth in its Branch Technical Position (BTP) on Fire Protection for Fuel Cycle Facilities." 57 Fed. Rec. 35607 (Aug.10.199T which calls for conduct of FE/.s and for the preparation of PFPs.
| |
| In response. NEl and a number oflicensees stated that the Generic Letter appeared to establish new requirements. NEl. in particular, stated:
| |
| There is no extant regulation nor order that requires . .
| |
| Part 70 licensees to prepare pre fire plans [or] to conduct
| |
| [tre hazard anolyses. [Ietter Febx .\1. Ibliar. Jr. to Robert F.
| |
| Burnett. Februarv 23.1995}.
| |
| The listing of 58 NFPA codes and the statement that the "most current versions" of those codes will be utilized as the basis for Staff reviews clearly creates new regulatory expectations that may be very costly to achieve and may require licensees to continually upgrade their facilities to meet newly developed industry codes without any commensurate reduction in risk. Requiring facilities to be constructed in accordance with NFPA codes that are current at the time of construction is not an unreasonable regulatory demand. The SRP prejudges that l' application of these codes is necessary without the benefit of any ISA results, and thus departs from a risk-informed, performance-based approach to regulation.
| |
| 11.4 Decommissionimr (Draft SRP 610.0) l NEI's comments on this portion of the SRP presume that the chapter on !
| |
| decommissioning is intended to apply to the review of new, renewed, amended or l Attachment to the Letter to Dr. Carl A. Paperiello New Programmatic Criteria in NUREG 1520(SRP)
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| Page 6
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| | |
| V 1 o L
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| i
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| " resubmitted"2 license applications for oneratine facilities, and not to facilitin that have permanently shutdown and intend to decommission.
| |
| The stated purpose of this portion of the draft SRP is to ensure that an
| |
| ' "anolicants' plans for decommissioning . . . provide reasonable assurance that the'avolicant will be able to decommission the facility safely and in accordance with NRC requirements." [ Draft SRP $ 10.1 (emphasis added)].
| |
| The SRP discusses Reg. Guide 3.65. " Standard Format and Content of Decommissioning Plans for Licensees under 10 CFR Parts 30, 40, and 70," and various other guidance documents. These documents are used primarily, if not exclusively, to review detailed decommissioning plans that are not required to be submitted until after a licensee has ceased active operations and intends to decommission its facility. Such plans, filed close to the time that actual decommissioning activities are expected to occur. must provide detailed information or., among other things, proposed release criteria for land and facilities, survey methods and criteria. procedures for radiological protection during the decommissioning process, and plans for waste disposal. By contrast, at present, licensees at operating facilities must simply submit a cost estimate for decommissioning and provide financial assurance through a decommissioning funding plan. as part of a licensing submittal.
| |
| At the time oflicense application the SRP requires submission of a detailed decommissioning plan and detailed procedures to minimize contamination to the environment. This constitutes a new programmatic requirement.
| |
| Forecasting the methodologies or technologies to be used to decommission a facility 20 to 40 years in the future is an unreasonable requirement. Technologies which might be used for decontamination of equipment and soils may not yet be commercially available at the time oflicense application. The implication of this portion of the SRP is that the detailed review criteria applicable to decommissioning plans.will become regulatory review criteria for " applicants" for licenses, renewals, amendments or resubmittals. Application of those review criteria at this early stage is premature and unnecessary.
| |
| NEI believes that this entire chapter should be removed from the SRP and placed in a Regulatory Guidance document -
| |
| : Draft 10 CFR s70.62(a)(1) requires existing licensees to resubmit new applications within four years of the effective date of the rule. even if their licensees are not due to expire.
| |
| Attachment to the Letter to Dr. Carl A. Paperiello New Programmatic Criteria in NUREG-1520 (SRP)
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| Page 7
| |
| | |
| r ,
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| O I,
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| II.5 Human-Svstem Interfaces (Draft SRP 411.6) j l
| |
| The draft SRP requires ["f]ormal evaluation of human-system interfaces" and requires licensees to have a formal process for " design, evaluation.
| |
| implementation, maintenance, and modification of human-system interfaces . ."
| |
| [ Draft SRP li 11.6.3. 11.6.4.3). This includes periodic human system interface reviews, j
| |
| employment of human-system interface " specialists," development of human system " standards" and creation of an " inventory" of such interfaces. This ,
| |
| portion of the SRP is a new programmatic requirement. It creates an entirely I new and complex set of criteria that will require licensees to establish detailed programs and procedures to formally analyze interfaces between personnel and systems. Additionally, it prejudges that control of human-system interfaces is needed, regardless of the results of the ISA. The draft SRP departs from a (
| |
| risk-informed, performance-based approach to regulation.
| |
| l II.6 Oreanization and Administration (_ Draft SRP 62.0)
| |
| There are several aspects of this chapter of the SRP that create new expectations and go well beyond the rule and existing practice. First, this portion of the SRP provides for NRC Staff review of the applicant's organizational structure and policies governing " design" and " construction" activities and the qualifications of j design and construction personnel [ Draft SRP 52 3]. Licenses issued under Part 70 are not for the construction and operation of facilities, but rather for the possession and use of special nuclear material. Therefore, specifying policies on design and construction in the SRP is unwarranted. This represents a substantial change in policy and practice.
| |
| Second, the SRP provides for NRC Staff review of the " experience" and
| |
| " availability" of personnel for decommissioning oflicensed facilities [ Draft SRP l 52.4.3}. Again, review of such details associated with the actual decommissioning process at the licensing stage is premature. What contractors and personnel will be available in 20 to 40 years to oversee decommissioning cannot reasonably be ,
| |
| expected to be known now. This constitutes an unnecessary new requirement.
| |
| I Finally, the SRP calls for NRC review and approval ofinternallicensee
| |
| {
| |
| mechanisms for reporting safety concerns and for their investigation, assessment and resolution. The SRP anticipates that the NRC will make an affirmative determination that the license applicant " promotes an open environment that supports safety and is absent of any chilling effect that discourages prompt and i open reporting of safety concerns" (Draft SRP 62.4.3]. This aspect of the SRP goes far
| |
| )
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| J l
| |
| Attachment to the L.ctter to Dr. Carl A. Paperiello New Programmatic Criteria in NUREG-1520 (SRP) I Page 8 i
| |
| | |
| I beyond even NRC regulation of reactor licensee 3 Part 70 licensees clearly have an obligation under 10 CFR 70.7 not to discriminate against employees for raising safety concerns or for otherwise engaging in protected activities.
| |
| Enforcement action may be taken by the NRC for violations. Alllicensees have programs to report and dispose of safety concerns and to ensure an open environment for raising such concerns. However, neither reactor nor materials licensees have been required, as part of a license review or otherwise, to establish such programs, nor has the NRC ever to NEI's knowledge imposed " lack of a chilling environment" as a licensing standard. In early 1998 the NRC considered, but withdrew, a proposal to impose additional requirements to ensure that licensees maintain a " safety-conscious work environment" (SCWE). The NRC concluded that its existing non. binding Policy Statement on " Freedom of Employees in the Nuclear Industry to Raise Safety Concerns Without Fear of Retaliation" and its regulations prohibiting discrimination (e.g.10 CFR 70.7) are sufficient [63 red EfE.6235(Feb 6.1998)]. Under those policies and regulations, the NRC has established certain " expectations" for its licensees and has given itself the enforcement tools to respond if discriminati in, or the potential for discrimination occurs. Furthermore, the NRC's Executive Director for Operations recently proposed to the Commission that it " consider the possibility of discontinuing any agency efforts to independently assess SCWE" [SECY.98176
| |
| " Proposed Options for Asses 3:ng a Licensee's Safety Conscious Work Ennronment''(July 21,1998)]. The Commission ultimately decided to continue the existing Staff practice of using inspection and enforcement techniques to ensure a SCWE [ Staff Requirements Memorandum (September 1.1998)}. Imposing licensing standards for the maintenance of a SCWE goes well beyond existing practice and requirements and is inconsistent with the Commission's February and September policy determinations. {'
| |
| II.7 Emercency Manacement (Draft SRP 68.0) 1 This portion of the SRP calls for the licensee to establish an adequate emergency i response training program, not only for onsite workers, but also for "offsite !
| |
| emergency response personnel"in order to ensure that such personnel have !
| |
| adequate " knowledge of the emergency plan, assigned duties, and effectively j respond to an actual emergency." The licensee must prcvide the " topics and !
| |
| general content" of the training used for "offsite emergency response personnel"
| |
| [ Draft SRP $8A.3.211]. Part 70 currently requires that licensees provide a "brief '
| |
| description of[among other things) any specialinstructions and orientation tours the licensee would offer to fire, police, medical and other emergency personnel."
| |
| but it does not require formal training of such personnel [10 CFR 570.22(a)(3)(x)).
| |
| NRC's own analysis did not identify significant off-site risks. The language of the Attachment to the Letter to Dr Carl A. Papenello New Programmatic Cnteria in NUREG-1520 (SRP)
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| Page 9
| |
| | |
| 3 draft SRP appears. therefore. to go beyond existmg requirements and suggest.c an emergency response training program that is more akin to those established for commercial nuclear power plants.
| |
| 1 The design of an Emergency Management program must be based on the risks that could be posed to public health and safety and the environment by the facility and its operation. Until such risks are assessed in an ISA, the components and i requirements of an emergency managernent plan can not be accurately defined.
| |
| The SRP must allow the licensee to establish appropriate emergency response measures and to determine the extent of training which should be provided to
| |
| ~
| |
| "offsite emergency response personnel."
| |
| II.8 Configuration Management (Draft SRP 611.1)
| |
| This portion of the SRP creates detailed new criteria for the establishment of a l formal configuration management program. NEI has not yet assessed all of the implications of the new configuration management criteria, but one of the most significant. and entirely new requirernents, is the expectation that licensees will be required to " reconstitute" their " designs" (Draft SRP $11.1.3(GL 11.1.5.26). In particular:
| |
| * licensees must reconstitute "the current design bases, supporting analyses, requirements, and documentation that support items important to safety";
| |
| e if documentation has not "kept pace"with the as-built plant configuration, licensees must " walk down systems, update drawings and specifications, perform new calculations and analyses, and otherwise rebuild the design bases." \ Draft SRP E511.1.4.3.11.1.5.2}.
| |
| - The provisions constitute a new programmatic requirement. Provisions for design bases reconstruction go well beyond existing requirements and, in fact, substantially exceed the requirements applied to nuclear power plants. Part '70 l licenses do not " license" the design of a facility and so there should be no requirement to perform a reconstitution. Unless dictated by the results of an ISA, reconstitution of facility design documents may not in itselflikely result in any significant improvement to safety.
| |
| Operators of new and existing fuel cycle facilities should commit to a configuration management program in their licenses. However, the requirement Attachment to the Letter to Dr. Carl A. Papenello New Programmatic Criteria in NUREG-1520 (SRP)
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| Page 10 I
| |
| | |
| O-A that current licensees now undertake a major design reconstitution after their facilities have been safely operated for many years. seems unnecessary.
| |
| Justification of how this action might improve facility safety is lacking.
| |
| II.9 Maintenance (Draft SRP 611.2)
| |
| This portion of the SRP creates entirely new requirements patterned after commercial nuclear power plant requirements and guidance for maintenance programs. It appears to apply the concepts of preventive and corrective maintenance to " human performance" and activities, and references a wide range of guidance documents applicable to reactor maintenance programs. [ Draft SRP ff 11.2.4.3. 11.2.7]. For example, corrective and preventive maintenance practices are to be applied to " items relied on for safety" - which are defined in the draft regulation (10 CFR 70.4) to include " activities of personnel" [ Draft SRP $11.2.4.3].
| |
| The discussion of preventive maintenance specifically discusses "requalification and retraining of personnel" [ Draft SRP i ll.2 4.3]. This is a unique and to the best of our knowledgn. unprecedented extension af the concept of a nuclear facility maintenance program. It is not clear what additional requirements this would add over the proposed training program criteria in SRP 11.4.
| |
| The requirement for a nuclear power plant maintenance program is required by a specific regulation (10 CFR 50.65). In the absence of a corresponding requirement in the Part 70 rule, the NRC should not attempt to impose a highly prescriptive maintenance program either through the SRP or as a license condition.
| |
| The draft SRP contains very extensive requirements related to maintenance. For example, the draft SRP appears to require preventive maintenance and post -
| |
| maintenance functional tests, regardless of whether such activities are needed to ensure the proper functioning ofitems relied on for safety as identified by the ISA.
| |
| As a result, the draft SRP departs from a risk-informed, performance based approach to regulation.
| |
| II-10 Nuclear Criticality Safety (Draft SRP 65.0)
| |
| This portion of the SRP calls for adherence to the well-established principle of
| |
| " double contingency protection" in order to provide for adequate nuclear criticality safety and includes a definition of" double contingency protection" that is !
| |
| inconsistent with American National Standard ANSI /ANS 8.1. The SRP goes well beyond accepted international and nuclear industry practice by assigning specific, ;
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| Attachment to the Letter to Dr. Carl A. Papenello New Programmatic Criteria in Nt* REG-1520 (SRP)
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| Page iI
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| l V 1
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| J i
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| i quantitative, numencal frequencie.< to each of the two controlled parameter.s or f controls as an acceptance criterion. presumably m order to determine that a i particular nuclear criticality accident is " highly unlikely. '
| |
| 1 In particular, the SRP indicates that one controlled parameter or control should )
| |
| have a frequency of failure no greater than 104 per year. and the other controlled parameter or control should have a frequency of failure no greater than 10 2 per year [ Draft SRP 95A6]. Neither ANSI nor. to the best of our knowledge, the NRC has adopted this quantitative approach to criticality safety in the past. Instead, it has been a well accepted practice for the determination as to whether there are at least two "unlikely," independent and concurrent process changes necessary before a nuclear criticality might occur (i.e. double contingency protection) to be i made on the basis of the expertise, experience and judgment of nuclear criticality safety experts on a deterministic basis.
| |
| J Risk-informed. performance-based regulation allows a licensee to determine the risk of potential nuclear criticalities in his 'acility. The risk of a nuclear criticality d from each credible accident sequence will be assessed in the facility's ISA. Results of the ISA will guide the licensee in selecting and implementing nuclear criticality mitigative practices and measures that are appropriate to the level of the risk.
| |
| Adoption of these new quantitative standards will add considerably to the cost and complexity of performing nuclear criticality safety analyses. Furthermore, adherence to the traditional methods of applying the double contingency protection principle will still enable licensees to evaluate and determine whether a nuclear criticality is "unlikely" as required by the draft rule. In industry's view, if adherence to the double contingency protection principle is confirmed, then it follows that a nuclear criticality event would be " highly unlikely." l l
| |
| III. Conclusions The June 1998 draft SRP contains references to a wide range of new regulatory expectations and standards that have not been justified in the rulemaking package and are not consistent with a risk-informed, performance-based regulatory program. The rulemaking record is replete with explanations as to the i purpose of the requirements to perform ISAs, to adopt consequences of concern, to l identify items relied on for safety, and to assure that such items remain available j and reliable. It does not, however, explain at all the bases for the determination '
| |
| that the wide range of new programmatic criteria in the draft SRP is necessary or i
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| Attachment to the Letter to Dr. Carl A. Paperiello New Programmatu: Criteria in NUREG 1520(SRP)
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| Page 12
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| a' )
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| l I
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| e {
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| appropriate. Furthermore by generically pre.-er:ning .-uch criteria. without the benefit ofISA results, the draft SRP runs counter to the effort to establish a risk informed, performance-based regulatory regime. As NEI stated in its July, 1996 testimony before the Commission. it believes that fundamental changes are required to both the draft Part 70 rule and the draft SRP before either can be adopted as regulatory standards or guidance.
| |
| Ref I Fdes Part*0 SHP Prnrrammaue CommenwM to 944 Attachment to the Letter to Dr. Carl A. Paperiello New Programmatic Criteria in NUREG-1520 (SRP)
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| Page 13
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| e 11/30/98 Tentative Aaenda for December 3 - 4.1998. Public Meetina on Amendment to 10 CFR Part 70 Meeting Objective: The purpose of the meeting is to consider industry suggestions for specific changes to the language in the SECY-98-185 draft amendment to 10 CFR Part 70, and the associated draft standard review plan (SRP). Topics to be addressed are: (1) next steps 1 in the revision of 10 CFR Part 70; (2) chemical safety requirements; (3) SRP issues; (4) criticality safety in relation to risk-informed regulations; (5) the content of the integrated safety analysis (ISA) summary; (6) the role of the preliminary ISA in the regulatory process; and (7) other issues identified.
| |
| | |
| ==Background:==
| |
| | |
| Both the NRC staff and the Nuclear Energy Institute (NEI) briefed the Commission on August i 25,1998, regarding SECY-98-185, " Proposed Rulemaking - Revised Requirements for the Domestic Licensing of Special Nuclear Material," dated July 30,1998. Although both NEl and the staff are in agreement that Part 70 should be amended to require the performance of an ISA, disagreements about the details of that proposed requirement were identified at the Commission meeting. At a subsequent public meeting with staff on September 29,1998, NRC staff and industry iopresentatives identified the following issues and agreed on the need for an additional meeting:
| |
| 1, Risk-informed approach / graded approach to safety, and use of consequence and likelihood criteria;
| |
| : 2. Acceptance criteria applicable to criticality controls;
| |
| : 3. Regulation of Chemical Hazards: Consistency with existing NRC/ Occupational Safety and Health Administration (OSHA) Memorandum of Understanding (MOU), EPA requirements, and the AEA and NEPA;
| |
| : 4. ISA Summary: Content, level of detail, and form of submittal to NRC;
| |
| : 5. Standard Review Plan: Appropriateness of standards and level of prescriptiveness of acceptance criteria to support a performance based rule;
| |
| : 6. Preliminary ISA: Purpose and applicability to major amendments of existing licenses and to new license applications;
| |
| | |
| 'n Meeting Agenda:
| |
| '1. Next Steps in the development of a revised 10 CFR Part 70 A. NRC staff presentation B. Discussion
| |
| : 2. Chemical Safety A. Industry-NEl presentations l
| |
| : 1. Suggested changes to NRC draft Part 70 (with focus on sections 70.60(a),
| |
| 70.60(b) and draft Part 70 Appendices A and B.)
| |
| : 2. Suggested changes to NRC draft SRP (including chemical process safety chapter.)
| |
| B. Industry-NEl / NRC discussion of suggested changes C. Comments by other attendees
| |
| : 3. Standard Review Plan issues A. Industry-NEl presentation on SRP issues B. _ Industry-NEl / NRC discussion of issues C. Comments by other attendees
| |
| : 4. Criticality Safety: Relationship to Risk-Informed, Performance Based Regulations A. Industry-NEl presentations
| |
| : 1. Comments on criticality safety and the recommended directions B. Industry-NEl / NRC discussion C. Comments by other attendees 2-
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| . l
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| | |
| i i .
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| : 5. ISA Summary A. Industry-NEl presentations-
| |
| : 1. Comments on the ISA summary and the recommended directions B. . Industry-NEl / NRC discussion I
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| C. Comments by other attendees i
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| : 6. Preliminary ISA and other issues A. Industry /NEl Presentations
| |
| : 1. Comments on Preliminary ISA and other topics B. Industry-NEl / NRC discussion j C. Comments by other attendees 1
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| L __
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| | |
| 6
| |
| 's 11/30/98 Tentative Aaenda for December 3 - 4.1998. Public Meetino on Amendment to 10 CFR Part 70 Meeting Objective: The purpose of the meeting is to consider industry suggestions for specific changes to the language in the SECY-98-185 draft amendment to 10 CFR Part 70, and the associated draft standard review plan (SRP). Topics to be addressed are: (1) next steps in the revision of 10 CFR Part 70; (2) chemical safety requirernents; (3) SRP issues; (4) criticality safety in relation to risk-informed regulations; (5) the content of the integrated safety analysis (ISA) summary; (6) the role of the preliminary ISA in the regulatory process; and (7) other issues identified.
| |
| | |
| ==Background:==
| |
| | |
| Both the NRC staff and the Nuclear Energy Institute (NEI) briefed the Commission on August 25,1998, regarding SECY-98-185,
| |
| * Proposed Rulemaking - Revised Requirements for the Domestic Licensing of Special Nuclear Material," dated July 30,1998. Although both NEl and the staff are in agreement that Part 70 should be amended to require the performance of an ISA, disagreements about the details of that proposed requirement were identified at the Commission meeting. At a subsequent public meeting with staff on September 29,1998, NRC staff and industry representatives identified the following issues and agreed on the need for an additional meeting:
| |
| : 1. Risk-informed approach / graded approach to safety, and use of consequence and likelihood criteria;
| |
| : 2. Acceptance criteria applicable to criticality controls;
| |
| : 3. Regulation of Chemical Hazards: Consistency with existing NRC/ Occupational Safety and Health Adrninistration (OSHA) Memorandum of Understanding (MOU), EPA requirements, and the AEA and NEPA;
| |
| : 4. ISA Summary: Content, level of detail, and form of submittal to NRC;
| |
| : 5. Standard Review Plan: Appropriateness of standards and level of prescriptiveness of acceptance criteria to support a performance based rule;
| |
| : 6. Preliminary ISA: Purpose and applicability to major amendments of existing licenses and to new license applications;
| |
| .I_
| |
| w.
| |
| | |
| 's Meeting Agenda:
| |
| : 1. Next Steps in the development of a revised 10 CFR Part 70 A. NRC staff presentation B. Discussion
| |
| : 2. Chemical Safety A. Industry-NEl presentations
| |
| : 1. Suggested changes to NRC draft Part 70 (with focus on sections 70.60(a),
| |
| 70.60(b) and draft Part 70 Appendices A and B.)
| |
| : 2. Suggested changes to NRC draft SRP (including chemical process safety chapter.) j B. Industry-NEl / NRC discussion of suggested changes C. Comments by other attendees
| |
| ' 3. Standard Review Plan issues 1
| |
| ~ A. Industry-NEl presentation on SRP issues B. Inc.'ustry-NEl / NRC discussion of issues
| |
| - C. Comments by other attendees 4.- Criticality Safety: Relationship to Risk-informed, Performance Based Regulations
| |
| 'A. Industry-NEl presentations
| |
| : 1. Comments on criticality safety and the recommended directions B. Industry-NEl / NRC discussion C.- Comments by other attendees T-2-
| |
| | |
| l i !
| |
| 5.- ISA Summary A. - Industry NEl presentations is 1. Comments on the ISA summary and the recommended directions B. Industry-NEl / NRC discussion C. Comments by other attendees
| |
| : 6. Preliminary ISA and other issues A.' Industry /NEl Presentations
| |
| : 1. Comments on Preliminary ISA and other topics B. Industry-NEl / NRC discussion
| |
| ' C. Comments by other attendees i
| |
| I l
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| l 3-
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| | |
| 4 l
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| NUCLEAR EN[RGY I N SilI U T E November 4,1998 Dr. Carl A. Paperiello, Director i Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission j Two White Flint Center l Washington, D.C. 20555-0001 j i
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| | |
| ==REFERENCE:==
| |
| 10 CFR Part 70 Regulation of Chemical Hazards >
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| j
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| | |
| ==Dear Dr. Paperiello:==
| |
| | |
| At the September 29th NRC-Nuclear Industry Workshop on Part 70 Regulation you acknowledged that the language in the draft Part 70 revisions addressing regulation of hazardous chemicals required clarification. You requested the Nuclear Energy Institute (NEI)1 to propose corrections to this draft language.
| |
| Attachment 1 to this letter presents the changes that NEI recommends be i incorporated to accurately reflect NRC's regulatory jurisdiction over hazardous !
| |
| chemicals. Attachment 2 provides background information and explanations for l each recommended change. Attachment 3 is a red lined version of SECY-98-185 l which incorporates NEI's recommended changes.
| |
| i NEI is the organization responsible for establishing unified nuclear industry policy on matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEI's members include all utilities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect / engineering firtns, fuel fabrication facilities, materials licensees, and other organizations and individuals involved in the nuclear energy industry.
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| I n /76 i Sfetti, NW $Ulff 400 W ASMrNGTON. OC 20006-370s PHONE 202 739 8000 Far 202 7s5 4019
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| D Dr. Carl A. Paperiello ;
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| Nuclear Regulatory Commission j
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| . November 4,1998 J Page 2 NEI is pleased to have had the opportunity to provide this input to the NRC towards clarifying the draft rule language. We look forward to continuing the dialogue on the Part 70 rulemaking and to addressing any questions which you or your staff may have on the industry's concerns and positions.
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| Sincerely, 4 --- 1 Marvin S. Fertel cc: Chairman Shirley Ann Jackson J Commissioner Edward McGaffigan, Jr.
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| Commissioner Nils J. Diaz Commissioner Jeffrey S. Merrifield Commissioner Greta Joy Dicus William D. Travers, Emeritus Director of Operations 2
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| a
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| ________o
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| 4 ATTACHMENT 1 NUCLEAR ENERGY INSTITUTE (NEI)
| |
| RECOMMENDED LANGUAGE CHANGES TO PART 70 FOR REGULATION OF CHEMICAL HAZARDS I. Deficiencies in Draft Languare Proposed revisions to 10 CFR Part 70 will provide NRC regulatory jurisdiction over all " chemical hazards resulting from the processing of licensed'' radioactive material. The breadth of this jurisdiction exceeds that described in SECY-98185 and in the 1988 NRC/ OSHA Memorandum of Understanding (MOU). Proposed language in Part 70 can be construed to extend NRC regulation to any chemical hazard at a licensed fuel fabrication facility. NEI's principal objection to the draft Fart 70 language is its failure to clearly separate the regulatory responsibilities of the NRC and OSHA as established in the MOU. As written, the draft rule will result in redundant, overlapping regulatory oversight that will not improve public or worker health and safety.
| |
| II. Proposed Language Modificatignjii The draft language can be corrected primarily through clarification of the term
| |
| " hazardous chemicals" in Part 70.60 and addition of a new definition for !
| |
| " hazardous chemicals produced from radioactive materials."
| |
| The MOU grants NRC the responsibility of protecting against " chemical risk produced by radioactive materials." Chemical risk results from hazards posed by either (i) the radioactive material itself, or (ii) compounds created by reaction of the radioactive materials with other substances. To clarify NRC jurisdiction over these two chemical hazards the following changes are recommended:
| |
| (i) the term " hazardous chemicals" should be replaced by " radioactive materials or hazardous chemicals produced from radioactive materials" This change would apply to $70.60(b)(1)(ii)(B),
| |
| 570.60(b)(1)(iii)(c), f 70.60(b)(2)(i)(B) and S70.60(b)(2)(ii)(B) of the draft rule and throughout SECY-98-185.
| |
| (ii) the majority of the chemicals listed in Appendices A (AEGLs) and B (ERPGs) are non-radioactive, are not used in SNM processing and are not capable of being produced from radioactive materials. The proposed Rule revisions could be simplified by retaining references to the AEGL and ERPG standards, but deleting the actual tables of exposure limits which will be continually updated and modified.
| |
| 1
| |
| | |
| - (iii) the definition of"Hazardons Chemicals" includes the phrase
| |
| " ..cause significant damage to property or..." The NRC should not attempt to exercise jurisdiction over damage to property because such damage is not related to public health and safety. This clause should be deleted. The definition should, therefore, read as follows:
| |
| "Hamdous Chemicals means substances that are toxic, explosive, flammable, corrosive or reactive to the extent that they can endanger life if not adequately controlled."
| |
| (iv) inclusion of a definition for "Nazardous Chemicals Produced from Radioactive Materials"is required. This definition will reinforce the clear distinction between chemicals whose hazards are to be regulated by the NRC or by OSHA. Chemical hazards which could produce radiological consequences of concern are already regulated by the NRC. The new definition should build upon the existing definition of Hazardous Chemicals and should read:
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| " Hazardous Chemicals Produced from Radioactive Ma.tenala means Hazardous Chemicals either having radioactive material (s) as precursor compound (s) or formed through interaction with radioactive materials. They do not include chemicals merely added to, or used in, or recycled from, the processing of special nuclear material (SNM)."
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| III. Concluding Remarks The foregoing suggested changes more accurately reflect the language and intent of the NRC/ OSHA MOU and more clearly demarcate the regulatory responsibilities of the NRC and OSHA with respect to chemical safety. Adoption of these suggested changes will provide clarity of the areas over which the NRC has authority to regulate the chemical hazards of radioactive materials without engaging in the regulation of purely chemical hazards.
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| 2
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| i
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| ~ .
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| ATTACHMENT 2 NUCLEAR ENERGY INSTITUTE (NEI)
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| BACKGROUND INFORMATION ON RECOMMENDED LANGUAGE CHANGES TO PART 70 FOR REGULATION OF CHEMICAL HAZARDS I. Introduction The U.S. Nuclear Regulatory Commission (NRC) issued SECY-98-185, " Proposed Rulemaking - Revised Requirements for the Domestic Licensing of Special Nuclear Material" on July 30,1998 to obtain Commission approval to publish a proposed rule amending 10 CFR Part 70. One proposed amendment addresses chemical safety standards. This amendment would extend NRC regulatory jurisdiction to all " chemical hazards resulting from the processing oflicensed" radioactive material, a much broader scope than was originally mandated in the rulemaking.
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| The chemical safety standard amendment (l70.60) should reflect the separation of regulatory jurisdiction of hazardous chemicals between the Occupational Health and Safety Administration (OSHA) and the NRC as detailed in the " Memorandum of Understanding Between the Nuclear Regulatory Commission and the Occupational Safety and Health Administration; Worker Protection at NRC- l Licensed Facilities," 53 Eqd. Re_g. 43950 (Oct. 31,1988) (NRC/ OSHA MOU). The draft rule, however, extends NRC regulatory oversight to eighty-eight chemicals, '
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| a majority of which are neither used in fuel cycle operations nor pose radiation hazards to facility workers or the public. The proposed $70.60 specifies concentrations of these chemicals, exposure to which constitutes a " consequence of concern," necessitates assessment in the licensee's Integrated Safety Analysis l (ISA) and requires design and implementation of adequate safety measures.
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| Language in the draft rule can be construed to appreciably broaden the scope of i NRC authority into areas reserved for OSHA regulatory oversight.
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| At the September 29th NRC-Nuclear Industry Workshop on Part 70 Regulation the NRC concurred that the chemical safety rule amendment should conform to the NRC/ OSHA MOU and that the NRC should only regulate those hazards falling within its jurisdiction. The NRC requested NEI to offer suggestions to clarify the language of the draft rule to ensure that the regulatory authority of the NRC and OSHA is clearly' demarcated.
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| This Attachment 2 summarizes NEI's understanding of the scope of the NRC's authority to regulate chemical hazards at fuel-cycle facilities e.d explains the basis for each change in the draft Rule language presented in Attachment 1.
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| 1 i
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| o
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| II. NRC Authority to Regulate Chemical Hazards The NRC/ OSHA MOU identifies "four kinds of hazards that may be associated with NRC-licensed nuclear facilities":
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| : 1. " Radiation risk produced by radioactive materials;"
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| : 2. " Chemical risk produced by radioactive materials;"
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| : 3. " Plant conditions which affect the safety of radioactive materials and thus present an increased radiation risk to workers. For example, these might produce a fire or an explosion, and thereby cause a release of radioactive materials or an unsafe reactor condition; and,"
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| : 4. " Plant conditions which result in an occupational risk, but do not affect the safety oflicensed radioactive materials. For example, there might be exposure to toxic non-radioactive materials and other industrial hazards in the workplace."
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| The NRC/ OSHA MOU states that the NRC shall have the responsibility for protecting against the first three hazards, while OSHA shall be responsible for protecting workers from the fourth hazard.
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| NRC's sole responsibility for protecting the health and safety of the public from the first hazard (radiation risk) is clear and unambiguous.
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| NRC's responsibility with respect to the second hazard is limited to a narrow class of chemical hazards. " Consequences of concern" which the NRC is responsible for regulating are chemical hazards which either: (1) result from the hazardous properties of the radioactive material itself, or (2) are created by the chemical reaction of the radioactive material and one or more other substances. For example, radioactive compounds UFs and UF4 exhibit toxic properties whose hazards are subject to NRC regulation. The NRC would also regulate generation of HF formed through interaction of UFs and moisture (humidity) in the conversion process to ensure that any exposures are kept below ERPG threshold concentrations. NRC regulatory oversight would not, however, extend to an HF recovery circuit once the HF off gas scrubber condensates leave the conversion plant and are confirmed to contain only residual concentrations of radionuclides.
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| Protection of workers from HF chemical hazards at that point in the HF plant would, instead, revert to OSHA jurisdiction. Acids, ion exchange eluants and 2
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| ~
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| solvent extraction organic chemicals used in UO2 scrap recovery would also be subject to NRC regulatory oversight only when actually used in uranium recovery processing and regeneration; regulation of chemical hazards from their bulk storage and handling (prior to use or after regeneration) would be an OSHA responsibility.
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| Determination of whether or not a particular chemical should be regulated by the NRC is often a process-specific issue. For example, nitric acid to be used in the UO2 scrap recovery process may be stored on site prior to use and would not be regulated by the NRC (so long as it could not affect the safety oflicensed material). However, once the acid is used in the dissolution process and combines with UO2, the radiological and chemical hazards of the mixture would be subject to NRC regulation. ' On the other hand, once the UO2 is stripped from the acid (via ion or solvent exchange) to leave the acid sufficiently free of radiological contamination to permit its handling as a non-radioactive material, the acid would only be subject to NRC regulation ifit was stored or used in a manner that could affect the safety oflicensed material. Off-gas scrubber condensates of gaseous and volatile radionuclides may be sabject to NRC oversight depending I upon their composition and upon the radiation hazard they pose. Licensees will need to evaluate their own processes and chemicalinventories to determine the relevant controls that should apply to a particular chemical at various stages in their manufacturing processes. In summary, NRC regulatory authority over chemical hazards extends only to those chemicals stored at a licensed fuel facility that may affect the safety of SNM.
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| NRC's responsibility with respect to the third hazard entails protecting workers against increased radiation risk caused by plant conditions affecting the safety of radioactive material. Radiation releases could originate directly from fires or explosions or indirectly from releases of hazardous, non-radioactive substances that might incapacitate an essential plant operator who would then be unable to respond to an emergency and prevent a release of radiation. In all cases the consequence of concern to the NRC is the increased radiation risk to the worker --
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| nat the occupational risk of the precursor fire, explosion or chemical release event.
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| It is this radiation risk that should be of concern to the NRC. The occupational risks associated with the precursor events are the responsibility of OSHA.
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| With respect to the fourth hazard, OSHA retains the responsibility for ensuring the occupational safety of workers including their protection from unacceptable
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| . exposures to toxic, non-radioactive chemicals and other industrial hazards. In this case the consequence of concern is the (non radiation) risk associated with a particular plant condition.
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| 3
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| L .
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| III. Prior NRC Guidance on Chemical Hazards NUREG-1601 (Chemical Process Safety at Fuel Cycle Facilities) provides guidance for licensees to address chemical safety issues. In accordance with the MOU, it acknowledges that the NRC's responsibility is assurance of the safety oflicensed material and that its oversight of the risk posed by hazardous chemicals is limited to their effect on licensed material and increased radiation risk to workers:
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| "Most NRC fuel cycle licensees possess materials that are chemically hazardous and/or pose some sort of non-radiological risk. Chemical and radiological risks have been known to compound one another, and in many cases, radioactive materials are also chemically hazardous. A chemical explosion in a fuel cycle facility could disperse radioactive material,just as the radiation environment could make it more difficult to respond to a hazardous chemical spill.
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| . . . The MOU between NRC and OSHA on chemical safety issues makes provision for the NRC to assume responsibility for the control of risks which may affect radioactive materials.
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| (NUREG.1601, i 2)
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| NUREG-1601 goes on to state that:
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| "Thus the NRC does not regulate chemicals per se; rather, the NRC verifies that the interactions of chemicals with NRC.
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| Licensed nuclear materials and/or with equipment which processes, transports, or stores these licensed materials have been considered in the design of the equipment and facilities and in the operating and maintenanceprocedures."
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| (NUREG 1601, $2).
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| NUREG 1601 instructs licensees to conduct hazard audits to identify:
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| " potential chemical hazards of radioactive materials and radiation hazards caused by chemicals . . . . "
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| (NUREG 1601,52.2.1).
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| Although it advises licensees to identify non-radioactive chemicals, it does so in order to ensure the safety oflicensed material:
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| ' Chemicals tuhich do not contain licensed materials should also be identified as potential chemical hazards because . . . release of such chemicals may affect theprocess bv releasina the 4
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| r I
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| licensed material or may affect the confinement of the licensed \
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| material in a favorable geometry". (NUREG 1601, $2.3.1.1 emphasis added). ,
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| With regard to the effect of chemical hazards on the environment, the NRC emphasizes that a licensee need only identify those:
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| "[c]hemicals which can cause a release oflicensed material to the environment above NRC-prescribed limits..." ,
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| (NUREG 1601,62.3.1.2). i
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| )
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| IV. NEI Recommended Channes to Part 70 l The chemical safety standard amendment in 570.60 should be rewritten to clarify ,
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| NRC's regulatory jurisdiction over chemical risks posed by: l (1) Special Nuclear Material (SNM) i
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| - (2) radioactive compounds (e.g. UF.)
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| (3) radioactive compounds produced from radioactive materials during the processing of SNM (e.g. HF)
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| NRC jurisdiction shall not extend to chemical risks originating from non-radioactive reagents stored at a fuel fabrication facility, either prior to their use or following their regeneration, and to non-radioactive by product chemicals produced in the fuel fabrication operation. j i
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| To clarify the separate regulatory jurisdictions of the NRC and OSHA over i chemical hazards, the following language changes are recommended:
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| (i) Definition of" Hazardous Chemicals" The MOU grants NRC the responsibility of protecting against " chemical risk produced by radioactive materials." Chemical risk results from hazards posed by (i) the radioactive material itself (e.g. UFs), (ii) compounds created by reaction of the radioactive materials with other substances (e.g. HF) or (iii) compounds contaminated by SNM or radioactive chemicals (e.g. HNOs, TCE, NaCO3). Chemicals produced by reaction with, or contaminated by, SNM '
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| (instances (ii) or (iii) above) are subject to NRC authority. Once, however, they are sufficiently free of radiological contamination to permit handing as a non-radioactive material they would no longer be subject to NRC oversight. All chemicals used in the facility, whether or not they are radioactive or hazardous, could fall under NRC jurisdication if they in any way impacted the i safety oflicensed material. ;
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| 5 i
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| L To clarify NRC jurisdiction over these two chemical hazards the following changes are required, the term " hazardous chemicals" should be replaced by
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| " radioactive materials or hazardous chemicals produced from radioactive materials" This change would apply to 970.60(b)(1)(H)(B),
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| f 70.60(b)(1)(iH)(c), S70.60(b)(2)(i)(B) and j70.60(b)(2)(ii)(B) and throughout SECY-98-185.
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| * $70.60(b)(1)(H)(B):
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| " Radioactive materials or hazardous chemicals produced from radioactive materials in concentrations exceeding AEGL-3 (Appendix A) or ERPG 3 (Appendix B) criteria; or"
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| * f70.60(b)(1)(iii)(C):
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| " Radioactive materials or hazardous chemicals produced from radioactive materials in concentrations exceeding AEGL-2 (Appendix )
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| A) or ERPG 2 (Appendix B) criteria; or"
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| $70.60(b)(2)(i)(B):
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| " Radioactive materials or hazardous chemicals produced from radioactive materials in concentrations between AEGL-2 (Appendix A) or ERPG-2 (Appendix B) criteria and AEGL-3 (Appendix A) or ERPG 3 (Appendix B) criteria; or"
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| $70.60(b)(2)(H)(B):
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| " Radioactive materials or hazardous chemicals produced from radioactive materials in concentrations between AEGL-1 (Appendix A) or ERPG-1 (Appendix B) criteria and AEGL-2 (Appendix A) or ERPG-2 (Appendix B) criteria; or" As a result of these proposed changes, a licensee would need to provide reasonable assurance that:
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| (i) concentrations of those radioactive materials listed in the AEGLs or ERPGs will not exceed the relevant consequences of concern, and that l (H) concentrations of hazardous chemicals listed in the AEGLs or ERPGs that may be produced through interactions with radioactive materials will not exceed the relevant consequences of concern.
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| (ii) Deletion ofAppendices A and B A majority of the chemicals listed in Appendices A (AEGLs) and B (ERPGs) are non radioactive, are not used in SNM processing and are not capable of being produced from radioactive materials. The proposed Rule revisions would be simplified by retaining references to the AEGL and ERPG standards, but deleting the actual tables of exposure limits which will be continually updated and modified. If the Appendices are not deleted, each should include the !
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| follo ving statement:
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| 6 a
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| e "The values listed should only be used as a consequence of concern if the chemical in question is radioactive, or is produced from radioactive material at a particular facility."
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| (iii) Reference to Property Damage The definition of " Hazardous Chemicals" includes the phrase "...cause significant damage to property or..." The NRC should not attempt to exercise jurisdiction over damage to property because such damage is not related to public health and safety. This clause should be deleted. The definition shall, therefore, read as follows:
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| " Hazardous Chemicals means substances that are toxic, explosive, flammable, corrosive or reactive to the extent that they can endanger life if not adequately controlled."
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| (iv) " Hazardous Chemicals Produced from Radioactive Materials" Definition of an additional term - Hazardous Chemicals Produced from Radioactive Materials - should be added to reinforce the clear distinction between chemicals whose hazards are to be regulated by the NRC and those to be solely regulated by OSHA. Chemical hazards which could produce radiological consequences of concern are already regulated by the NRC. The new definition should build upon the existing definition of Hazardous Chemicals and should read:
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| " Hazardous Chemicals Produced from Radioactive Materials means Hazardous Chemicals either having radioactive material (s) as precursor compound (s) or formed through interaction with radioactive materials. They do not include chemicals merely added to, or used in, or recycled from, the processing of special nuclear material (SNM)."
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| Ref:1:NAlesNPs,t 70NCbeam:al Comments ORAX13-10-M) 7
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| *C. 4-S
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| [7590-01-P]
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| NUCLEAR REGULATORY COMMISSION 10 CFR Part 70 RIN 3150 - AF22 Revised Requirements for the Domestic Licensing of Special Nuclear Material AGENCY: Nuclear Regulatory Commission.
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| t ACTION: Proposed rule.
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| | |
| ==SUMMARY==
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| : The Nuclear Regulatory Commission (NRC) is proposing to amend its safety regulations in the provisions goveming the domestic licensing of special nuclear material (SNM) for licensees cuthorized to possess a critical mass of SNM, that are engaged ir' one of the following activities:
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| cnriched uranium processing; uranium fuel fabrication; uranium enrichment; enriched uranium htxafluoride conversion; plutonium processing; mixed-oxide fuel fabrication; scrap recovery; or any l
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| other activity involving a critical mass of SNM that the Commission determines could significantly affect public health and safety.-- The proposed amendments would identify appropriate consequence criteria
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| . end the level of protection needed to prevent or mitigate accidents that exceed these criteria; require Effected licensees to perform an integrated safety analysis (ISA) to identify potential accidents at the facility and the items relied on for safety; require the implementation of measures to ensure that the it;ms relied on for safety are continuously available and reliable; require the inclusion of the safety bases, including the results of the ISA, in.the license application; and allow for licensees to make certain changes to their facilities without prior NRC approval.
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| DATES: The comment period expires (insert 75 days after publication in the Federal Register.)
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| Comments received after this date will be considered if it is practical to do so, but, the Commission is cble to ensure consideration only for comments received on or before this date.-
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| ADDRESSES: Submit comments to: The Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC,20555-0001, Attention: Rulemakings and Adjudications Staff.
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| Deliver comments to: 11555 Rockville Pike, Rockville, Maryland, between 7:30 a.m. and 4:15
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| - p.m. on Federal workdays.
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| ATTACHMENT 1 9
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| p-You may also provide comments via NRC's interactive rulemaking website through the NRC home page (http://www.nrc. gov)."From the home page, select "Rulemaking" from the tool bar. The int:ractive rulemaking website can then be accessed by selecting "New Rulemaking Website." This sit 3 provides the ability to upload comments as files (any format), if your web browser supports that function. For information about the interactive rulemaking website, contact Ms. Carol Gallagher, (301) 415-5905; e-mail cag@nrc. gov.
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| FOR FURTHER INFORMATION, CONTACT:- Richard I. Milstein, Office of Nuclear Material
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| ' S:fety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC,20555-0001, t:l: phone (301) 415-8149; e-mail rim @nrc. gov.
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| SUPPLEMENTARY INFORMATION:
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| : 1. Backgroun'd ll. Description of P cposed Action
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| ' I. Background A near-criticality incident at a low enriched fuel fabrication facility in May of 1991 prompted NRC to r:; view its safety regulations for licensees that possess and process large quantities of SNM. [See "Pr: posed Method for Regulating Major Materials Licensees" (U.S. Nuclear Regulatory Commission, 1992) for additional details on the review.] As a result of this review, the Commission and the staff r cognized the need for revision of its regulatory base for these licensees and, specifically, for those possessing a critical mass of SNM. Further, the NRC staff concluded that to increase confidence in the margin of safety at a facility possessing this type and amount of material, a licensee should perform an ISA. An ISA is a systematic ATTACHMENT 2
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| Onalysis that identifies:
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| (1) Plant and external hazards and their potential for initiating accident sequences; (2) The potential accident sequences, their likelihood, and consequences; and {
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| (3) The structures, systems, equipment, components, and activities of personnel relied on to prevent or mitigate potential accidents at a facility.
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| NRC held public meetings with the nuclear industry on this issue during May and November of 1995. Industry's position on the need for revision of NRC regulations in Part 70 wCs articulated to the Commission by the Nuclear Energy Institute (NEI) at a July 2,1996,
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| )
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| meeting, and in the subsequent filing of a Petition for Rulemaking (PRM-70-7) by NEl with NRC in September 1996. NRC published in the Federal Register a notice of receipt of the PRM and requested public comments on August 21,1996 (61 FR 60057). The PRM requested that NRC amend Part 70 to:
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| (1) Add a definition for a uranium processing a'id fuel fabrication plant; 3 (2) Require the performance of an ISA, or acceptable alternative, at uranium processing, fuel fabrication, and enrichment plants; and (3) include a requirement for backfit analysis, under certain circumstances, within Port 70.
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| In SECY-97-137, dated June 30,1997, the NRC staff proposed a resolution to the NEl l PRM and i'ecommended that the Commission direct the staff to proceed with rulemaking. The NRC staff's recommended approach to rulemaking included the basic elements of the PRM, with some modification. In brief, NRC staff proposed to revise Part 70 to include the following major elements:
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| (1) Performance of a formal ISA, which would form the basis for a licensee's safety program. This requirement would apply to all licensed facilities (eycept reactors and the 90seous diffusion plants regulated under 10 CFR Part 76) or activities, subject to NRC regulation, that are authorized to possess SNM in quantities sufficient to constitute a potential for nuclear criticality; (2) Establishment of criteria to identify the adverse consequences that licensees must protect against; (3) inclusion of the safety bases in a license application (i.e., the identification of the j ATTACHMENT 2 1
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| 9 5 e potential accidents, the items relied on for safety to prevent or mitigate these accidents, and ths measures needed to ensure the continuous availability and reliability of these items). (This is in contrast to the PRM's approach, where the ISA results would not be included in the license application);
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| (4) Ability of licensees, based on the results of an ISA, to make certain changes without NRC prior approval; and (5) Consideration by the Commission, after initial conduct and implementation of the ISA by the licensees, of a qualitative backfitting mechanism to enhance regulatory stability.
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| In a Staff Requirements Memorandum (SRM) dated August 22,1997, the Commission
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| ... approved the staff's proposal to revise Part 70" and directed the NRC staff to "... submit a drcft proposed rule...by July 31,1998."
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| : 11. Description of Proposed Action The Commission has decided to grant, in part, the NEl PRM by initiating this rulemaking. Further, the proposed rule adopts the petitioner's proposal in part and modifies -
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| tha petitioner's proposal as indicated in the following discussion.
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| The Commission is proposing to modify Part 70 to provide increased confidence in the
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| - mirgin of safety at certain facilities authorized to process a critical mass of SNM. The Commission believes that this objective can be best accomplished through a risk-informed and performance-based regulatory approach that includes:
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| (1) The identification of appropriate consequence criteria and the level of protection needed to prevent or mitigate accidents that exceed such criteria; (2) The performance of an ISA to identify potential accidents at the facility and the items r:: lied on for safety; (3) The implementation of measures to ensure that the items relied on for safety are continuously available and reliable; (4) The inclusion of the safety bases, including the ISA results, in the license epplication; and (5) The allowance for licensees to make certain changes to their facilities without prior NRC approval.
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| The Commission's approach agrees in principle with the NEl petition. However, in ATTACHMENT 2
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| l contrast to the petition's suggestion that the ISA requirement be limited to " . uranium processing, fuel fabrication, and uranium enrichment plant licensees," the Commission would require the performance of an ISA for a broad range of Part 70 licensees that are authorized to possess a critical mass of SNM. The Part 70 licensees that would be affected include licensees engaged in one of the following activities: enriched uranium processing; uranium fuel fabrication; uranium enrichment; enriched uranium hexafluoride conversion; plutonium processing; mixed-oxide fuel fabrication; scrap recovery; or any other activity involving a critical mass of SNM that the Commission determines could significantly affect public health f cnd safety. The proposed rule would not apply to regulatees authorized to possess SNM under 10 CFR Parts 50,60,72, I cnd'76.
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| Furthermore, the Commission is not currently proposing, as suggested in the NEl p:tition, to include a backfit provision in Part 70. Based on the discussions at a public meeting held on May 28 1998, the purpose of the proposed backfit provision is to ensure that NRC staff does not impose safety controls that are not necessary to satisfy the performance r:quirements of Part 70, unless a quantitative cost-benefit analysis justifies this action. The Commission believes that once the safety bases, including the results of the ISA, are j incorporated in the license application, and the NRC staff has gained sufficient experience I with implementation of the ISA requirements, a qualitative backfit mechanism could be considered. Without a baseline determination of risk, as provided by the initial ISA process, it ;
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| is not clear how a determination of incremental risk, as needed for a backfit analysis, would be cccomplished. Furthermore, although NEl believes that a quantitative backfit approach is currently feasible, it would appear that a quantitative determination of incremental risk would r: quire a Probabilistic Risk Assessment, to which the industry has been strongly opposed. l Given the differences of opinion on this subject, the Commission requests public comment on its intent to defer consideration of a qualitative backfit provision in Part 70.
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| The majority of the proposed modifications to Part 70 are found in a new subpart,
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| " Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Material," that consists of @@70.60 through 70.74. These proposed modifications to Part 70, discussed in detail below, are required to increase confidence in the margin of safety and are in general accordance with the approach approved by the Commission in its August 22,1997, SRM. However, the Commission has decided that the new requirements should not apply to all licensees authorized to possess a critical mass of ATTACHMENT 2
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| SNM. Instead, the Commission has identified a subset ~of these licensees that, based on the rGlatively high level of risk associated with operations at these facilities, should be subject to the new requirements.- This change would exclude certain facilities (e.g., those authorized only to store SNM or use SNM in sealed form for research and educational purposes) from the new requirements, because of the relatively low level of risk at these facilities. This issue is further addressed in the discussion of $70.62.
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| Section 70 4. " Definitions."
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| The following fourteen definitions would be added to this section to provide a clear
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| ~
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| understanding of the meaning of the new subpart H, " Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Material:" Acute exposure, Acute exposure guideline levels, Controlled site boundary, Critical mass of SNM, Deviation from safe operating conditions, Double contingency Emergency response planning guidelines, Hazardous chemicals, Hazardous chemicals produced from radioactive materials, integrated safety analysis, items relied on for safety, New process, Results of the ISA, Unacceptable vulnerabilities, and Worker.
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| ATTACHMENT 2
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| E l
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| I Section 70.15," Nuclear reactors."
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| l A new section would be added to subpart B,
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| * Exemptions," that exempts nuclear reactors licensed under Part 50 from the new subpart H, " Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Material."
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| Section 70.22. " Contents of applications."
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| Paragraph (f) would be removed. Paragraph (f) currently requires that, for plutonium processing and fuel fabrication facilities, certain additional safety-related information be submitted with cn application. The new subpart H, " Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Material," would contain requirements for the submittal of information called for in paragraph (f) and is sufficient to allow the Commission to make a determination of adequacy.
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| Section 70.23. "Reouirements for the approval of soolications."
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| Paragraphs (a)(8), and (b) would be removed. These paragraphs currently require that the l C;mmission, to approve an application, determine that the construction of a plutonium processing and l f:brication facility meet certain conditions. These conditions would be covered in the new subpart H,
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| " Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Material."
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| Section 70.60. " Safety performance reouirements."
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| These requirements would establish the purpose of the new requirements, identify the potential adverse consequences that need to be protected against, establish the level of protection that is needed to ensure that the consequences of concem do not occur, and identify the safety program ,
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| ;lements that allow licensees to demonstrate their ability to provide an adequate level of protection.
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| I Section 70.60(at " Purpose."
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| This paragraph would address the following questions: Why are the new requirements i needed? What hazards need to be considered? Who are the intended beneficiaries? In general, the j new requirements are intended to ensure that workers', the general public, and the environment are i l
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| A worker, in the context of this rulemaking, is defined as an individual whose assigned duties in the course of employment involve exposure to radiation and/or radioactive material from licensed and unlicensed sources of radiation (i.e., I an individual who is subject to an occupational dose as in 10 CFR 20.1003).
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| ATTACHMENT 2 ;
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| l 1
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| | |
| protected from radiological and certain chemical hazards associated with plant operations. All hrzards, including fire, chemical, electrical, industrial, etc., that can potentially affect radiological safety, must be considered and addressed by licensees. In addition, chemical haaards risks produced by radioactive materials th:t recu!!' rem the preter !ng f !!:en::d nut!!:r m:ter!:! must also be considered.
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| The question of NRC's authority to regulate chemical hasards nsks produced by radioactive materials at its fuel cycle facilities was raised after an accident in 1986 at a Part 40 licensed facility, in which a cylinder of uranium hexafluoride ruptured and killed a worker. The cause of the worker's death w:s the inhalation of hydrogen fluoride gas, which was produced from the chemical reaction of uranium h2xafluoride and water (humidity in air). As a result of that incident, NRC and the Occupational Safety cnd Health Administration (OSHA) established a memorandum of understanding (MOU) (1988) that identified the respective responsibilities of both agencies for the regulation of chemical hazards at nuclear facilities. The MOU identified the following four areas of responsibility. The NRC has r;sponsibility for the first three areas, whereas OSHA has respansibility for the fourth area:
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| (1) Radiation risk produced by radioactive materials; (2) Chemical risk produced by radioactive materials; (3) Plant conditions which that affect the safety of radioactive materials and, thus present an increased radiation risk to workers; and (4) Plant conditions which that result in an occupational risk, but do not affect the safety of licensed radioactive materials.
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| The purpose of the " Safety Performance Requirements," as defined in $70.60(a), is consistent with the NRC/ OSHA MOU.
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| Section 70.60fbt "Conseauences of concem."
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| The NRC is responsible for ensuring that workers and the general public are protected from the hazards involved in the handling, processing, and storage of SNM. All hazards (including fire and chemical) that could result in radiological consequences are a subject of NRC concern. In addition, th_e _
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| a4 chemical hasards risks oroduced from radioactive matenals retu!!Mg from the pre :::!ng Of
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| !! ented SNY th:t r!d erd!y :"ed : : erer m: d crOfth: p !!: are also a matter of NRC concem. Thus, NRC regulations need to address both radiological hazards the chemical consequences of radioactive materials and hazardous chemicals oroduced from radioactive materials.
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| The following discussion provides information, on the consequences of human exposure to radiation cnd hasardews hazards ehemicals produced by radioactive materials, that is relevant to the choice of ATTACHMENT 2
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| tppropriate consequence criteria. The actual choice of these criteria is discussed in
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| $970.60(b)(1)(ii)(A) and (B); 70.60(b)(1)(iii)(A) and (C); and 70.60(b)(2)(i)(A) and (B).
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| Radiolooical Consecuences. In the past, the regulation of licensees authorized to possess SNM, under 10 CFR Parts 70 and 20, has concentrated on radiation protection for persons involved in nuclear activities conducted under normal operations. The proposed amendments to Part 70 would sxplicitly address the potential exposure of workers or members of the public to radiation as a result of cccidents. Because accidents are unanticipated events that usually occur over a relatively short period of time, a regulation that seeks to assure adequate protection of workers and members of the public -
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| must limit the risk of such accidents. This can be accomplished by identifying appropriate consequence criteria and by limiting the likelihood of occurrence of the identified consequences. In sIlecting the radiological consequence criteria for use in the proposed rule, the Commission has examined the radiological criteria and design basis accident scenarios used in existing NRC regubtions to ensure that the proposed consequence criteria era consistent with criteria used in other Commission rules.
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| Chemical Consecuences. The processing of SNM may involve the use or production of hazardous chemicals. ' For example, low enriched uranium fuel fabrication facilities convert uranium h;xafluoride to uranium oxide by reaction with water (hydrolysis) to form uranyl fluoride and hydrogen flu:: ride. Uranyl fluoride, in addition to being radioactive, is a toxic uranium compound that can cause d: mage to the kidney. Hydrogen fluoride is highly toxic and poses a hazard to both workers and the general public. T.:9---dru: 2. .Scr' , !- din;; - c:n!:, -"d :f, :nd cr'".ud: 29, := r! 0 The effort to limit exposure of workers and the general public to hasardeus chemicals hazards produced by radioactive materials is based on two concerns:
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| ccute exposures that could result from accidental releases, and chronic exposures (i.e., multiple and repeated exposures occurring over a long period of time - days, months, or years), resulting from r: leases during normal operations.
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| Chemical consequence criteria corresponding to anticipated adverse health effects to humans frcm acute exposures (i.e., a single exposure or multiple exposures occurring within a short time - 24 h:urs or less) have been developed, or are under development, by a number of organizations. Of particular interest, the National Advisory Committee for Acute Guideline Levels for Hazardous Substances is developing Acute Exposure Guideline Limits (AEGLs) that will eventually cover approximately 400 industrial chemicals and pesticides. The committee, which works under the ATTACHMENT 2
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| cuspices of the U.S. Environmental Protection Agency (EPA) and the National Academy of Sciences
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| -(NAS), has identified a priority list of approximately 85 chemicals. Consequence criteria for 12 of these
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| > have currently been developed and criteria for approximately 30 additional chemicals per year are expected.
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| Another set of chemical consequence criteria, the Emergency Response Phnning Guidelines (ERPGs), has been developed by the American industrial Hygiene Association (AlHA) to provide cstimates of concentration ranges where defined adverse health effects might be observed because of short exposures to hazardous chemicals. ERPG criteria are widely used by those involved in casessing or responding to the release of hazardous chemical including "... community emergency pl:nners and response specialists, air dispersion modelers, industrial process safety engineers, implementers of environmental regulations such as the Superfund Amendment and Reauthorization Act, industrial hygienists, and toxicologists, transportation safety engineers, fire protection specialists, I
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| and govemment agencies...." (DOE Risk Manaaement Quarte IV.1997). Despite their general.
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| acceptance, there are cur,ently only approximately 80 ERP 3 criteria available, and some chemicals of importance (e.g., nitric acid) are not covered.
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| Federal regulations and intemal U.S. Department of Energy (DOE) guidance require the use of ERPGs for emergency planning. Recognizing that ERPGs exist for a limited number of chemicals, DOE's Subcommittee on Consequence Assessment and Protective Actions developed Temporary Emergency Exposure Limits (TEELs) so that DOE facilities could perform complete hazard analysis end consequence assessments, even for chemicals lacking ERPGs TEELs are not equivalent to ERPGs, but are approximations to ERPGs. They exist only until an ERPG is developed for a chemical.
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| ' As of July 1997,400 TEELs had been developed according to a methodology published in the Amoncan Industrial Hymene Joumal (1995). That methodology is not based directly on toxicological studies of the chemicals involved, but on a derived relationship between attemative exposure-limit parameters and the existing ERPG criteria. The use of the methodology results in a significant underestimation of the TEEL-22 level (0.6 mg/m8 ) for soluble uranium and would be inconsistent with the criterion on soluble uranium intake (i.e., 30 mg) proposed in this rule.
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| A fourth set of chemical consequence criteria that was considered potentially applicable for ccute exposure to hazardous chemicals is the immediately Dangerous to Life and Health (IDLH) criteria established by the National Institute for Occupational Safety and Health (NIOSH). However, 8 TEEL-2 is defined as the maximum airbome concentration below which it is believed that nearly allindividuals could be cxposed for up to 1 hour without exponencing or developing ineversible or other health effects or symptoms which could impair an individuars ability to take protective action.
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| ATTACHMENT 2
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| eccording to NIOSH, the IDLH criteria are defined " . only for the purpose of respirator selection." In addition, unlike the previously mentioned sets of criteria, there is only one IDLH level that has been defined. This would not facilitate the definition of multiple consequence levels for workers and the public, as intended in the proposed rule.
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| For chronic exposures of workers to hazardous chemicals during normal and off-normal cperations, the permissible exposure limits (PELs) established by OSHA in 29 CFR 1910 are cpplicable. However, these limits are not relevant for. acute exposures to hazardous chemicals produced from radioactive materials.
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| Given the status of these various sets of consequence criteria, the Commission has chosen AEGLs and ERPGs, in that order, as criteria to be used for acute short-term exposure to radioactive materials that may themselves pose a chemical hazard (e a uranium hexafluoride in certain situations) or hazardous chemicals produced from radioactive materials. If a given chemical has an AEGL associated with it, that criterion should be used. If not, the ERPG criterion, if available, should i
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| be used. Append!r ^. Centain: t50 cv:!!:b!: .^EGL v:!uec, a nd Appendh 9 cent:!n: the Ovci!:b!
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| ERoO v:!ue: fer h:: rdeu che-!: ! . If both AEGLs and ERPGs are available for a particular ch:mical, only the AEGL values will be presented. Although the TEEls cover a wide range of additional hazardous chemicals, the Commission has decided not to require their use at this time, because the methodology used to derive these values is not based on the toxicology of the chemicals involved and may, at least in certain cases, underestimate the limits. However, the use of the TEELs may be justified on a case-by-case basis in the absence of other applicable standards.
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| As a result of further study, new AEGL or ERPG values are expected to be established by the issuing organizations (EPA for AEGLs; AlHA for ERPGs). The Commission does not propose to cng ge in full, formal rulemaking with respect to these future changes, but will incorporate them h4he red $cd appnd!:e: !- a t nform by issuing an immediately effective final rule. The Commission l belhves that these purely technical changes or additions do not require comment and are, in addition, subject to the categorical exclusion in 10 CFR 51.22(c)(2). l General Acoroach The consequences of concem, identified in $970.60(b)(1) and (b)(2), describe those consequences that licensees must protect against2 The level of protection to be provided is discussed 8
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| The proposed rule does not address chemical and radiological consequences to workers and members of the public s nsulting from routine operations. These consequences are covered in other regulations (i.e.,10 CFR Part 20 and 29 CFR Part 1910).
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| ATTACHMENT 2
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| in 70.60(c) and depends on the severity of the consequences. The goalis to ensure an acceptable I:; vel of risk by limiting the likelihood of occurrence of the identified consequences. The consequences !
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| identified in $70.60(b)(1) of the proposed rule are considered to be high consequences and include the occurrence of a nuclear criticality, and accidental exposure of a worker or member of the public to high 1:;vels of radiation, radioactive materials or hazardous chemicals produced from radioactive matenals. !
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| The consequences identified in 70.60(b)(2) are considered to be intermediate consequences and include accidental exposure of a worker or member of the public to moderate levels of radiation.
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| radioactive materials, or hazardous chemicals produced from radioactive materials, and significant I
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| r;; leases of radioactive material to the environment. The proposed consequence criteria that are applicable to a member of the public are more restrictive than those that are applicable to a worker.
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| Also, within each category (worker and public), NRC recognizes that the proposed radiological criteria cra more restrictive (in terms of acute health effects) than the chemical criteria for a given level of ssverity (high or intermediate) and that this is consistent with current regulatory practice.
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| In some cases, a qualitative description of the consequence is used (e.g., a nuclear criticality);
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| in other cases, a numerical criterion is used. For cases where numerical criteria have been used, NRC has based the criteria on values that have been developed previously by NRC or other government cg:ncies or professional societies. Table 1 illustrates the radiological and chemical consequence crit ria used in the proposed rule. It should be noted that only those AEGLs or ERPGs associated with radioactive materials (e o 30 mo/m for uranium hexaflounde (ERPG-3)) are included in these conseauence cntena TABLE 1 Radiological and Chemical Consequence Criteria Worker Public CONSEQUENCE Radiological Chemical Radiological Chemical High > 1 Sv (100 rem) > AEGL-3 > 0.25 Sv (25 rem) > AEGL-2 (ERPG-3) (ERPG-2)
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| Intermediate < 1 Sv (100 rem) < AEGL-3 (ERPG- < 0.25 Sv (25 rem) < AEGL-2 I
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| : 3) (ERPG-2)
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| > 0.25 Sv (25 > 0.05 SV (5 rem) rem) > AEGL-2 (ERPG- > AEGL-1
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| : 2) (ERPG-1)
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| Section 70.60(b)(1) This paragraph defines "high consequences."
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| ATTACHMENT 2 I
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| ' Certain events that could occur at licensees' facilities are considered high-consequence events.
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| They include the occurrence of an inadvertent nuclear criticality, the exposure of a worker or member of the public to levels of radiation at which clinically observable biological damage could occur, or concentrations of radioactive materials or hazardous chemicals produced from radioactive materials at which death or life threatening injury could occur due to their 'oxic. explosive.
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| flammable; corrosive, or reactive properties.
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| Section 70.60(b)(1)(i). This paragraph deals with a nuclear criticality.
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| The occurrence of an inadvertent nuclear criticality is considered to be a high-consequence sv:nt. Although detecting and mitigating the consequences of a nuclear criticality are important cbjectives (see 10 CFR 70.63), the prevention of a nuclear criticality is a primary ,
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| NRC objective.
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| Section 70.60(b)(1)(ii)(A). This paragraph deals with an acute exposure of a worker to a i
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| rrdiation dose of 1 Sv (100 rem) or greater total effective dose equivalent (TEDE).
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| An acute exposure of a worker to a radiation dose of 1 Sv (100 rem) or greater TEDE is c:nsidered to be a high-consequence event. According to the National Council on Radiation ;
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| Protection and Measurements (NCRP,1971), life saving actions - including the "... search for and
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| . r:moval of injured persons, or entry to prevent conditions that would probably injure numbers of ,
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| people" - should be undertaken only when the ".~.. planned dose to the whole body shall not exceed 100
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| . rcms. This is consistent with a later NCRP position (NCRP,1987) on emergency occupational exposures, that states "...when the exposure may approach or exceed 1 Gy (100 rad) of low-LET
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| [ linear energy transfer) radiation (or an equivalent high-LET exposure) to a large portion of the body, in a short time, the worker needs to understand not only the potential for acute effects but h' e or she should also have an appreciation of the substantialincrease in his or her lifetime risk of cancer." The use of the 1-Sv (100-rem) criterion is not intended to imply that 1 Sv (100 rem) constitutes an ccceptable criterion for an emergency dose to a worker. Rather, this dose value has been proposed in this section as a reference value, which should be used by licensees to determine the level of protection (i.e., items relied on for safety, and measures to assure their continuous availability and reli1bility) needed to ensure an acceptably low level of risk to workers.
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| Section 70.60(b)(1)(ii)(B). This paragraph deals with an acute exposure of a worker to radioactive materials or hazardous chemicals produced from radioactive materials in concentrations ATTACHMENT 2
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| ~
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| cxceeding AEGL-3 or ERPG-3 limits.
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| An acute exposure of a worker to radioactive materials or hazardous chemicals produced from radioactive materials at concentrations that could cause death or life-threatening injuries from causes other than radiation is considered a high-consequence event. Two existing criteria,' AEGL-3* and ERPG-3, can be used to define such concentration levels. AEGL-3 is defined as "The airbome concentration (expressed in ppm or mg/m') of a substance at or above which it is predicted that the g2neral population, including susceptible, but excluding hypersusceptible, individuals, could experience lifs-threatening effects or death." ERPG-3 is defined as "The maximum airbome concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour without experiencing or d:;veloping life-threatening health effects." If, for a particular chemical, the AEGL-3 value is available, it should be used. Otherwise, the ERPG-3 value should be used. If there is no AEGL or ERPG value cvrilable, then the applicant should adopt a criterion that is comparable in severity to those that have been established for other chemicals.
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| Section 70.60(b)(1)(iii)(A). This paragraph deals with an acute exposure of a member of the public to a radiation dose of 0.25 Sv (25 rem) or greater TEDE.
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| The exposure of a member of the public to a radiation dose of 0.25 Sv (25 rem) TEDE is considered a high-consequence event. This is based on the criterion established in 10 CFR 100.11 " Determination of exclusion area, low population zone, and population center distance," and 10 CFR 50.34, " Contents of applications; technical information," where a whole-body dose of 0.25 Sv (25 rem) is used to determine the dimensions of the exclusion area and low population zone required for siting nuclear power reactors J 1
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| Section 70.60(b)(1)(iii)(B). This paragraph deals with an intake of 30 mg or greater of uranium in a s:luble form by a member of the public.
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| The intake of 30 mg of soluble uranium by a member of the public is considered a high-c:nsequence event. This choice, which is based on a review of the available literature (Pacific
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| . Northwest Laboratories (PNL),1994], is consistent with the selection of 30 mg of uranium as a criterion that was discussed during the Part 76 rulemaking, " Certification of Gaseous Diffusion Plants." In p:rticular, the final rule that established Part 76 (59 FR 48944; September 23,1994) stated that "The Three levels of consequences are defined for each chemical (AEGL-1 AEGL-2, and AEGL-3) for four different exposure times: 30 minutes: 1 hour; 4 hours; and 8 hours. The AEGL value for a 1-hour exposure is chosen for consistency with the definition of ERPG.
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| ATTACHMENT 2 l . . . . . . . . . .. .. . . .
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| l
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| NRC will consider whether the potential consequences of a reasonable spectrum of postulated cccident scenarios exceed... uranium intakes of 30 milligrams.. ." The final rule also stated that "The Commission's intended use of chemical toxicity considerations in Part 76 is consistent with its practice elsewhere (e.g.,10 CFR 20.1201(e)), and prevents any potential regulatory gap in public protection (gainst toxic effects of soluble uranium."
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| Section 70.60(b)(1)(iii)(C). This paragraph deals with an acute exposure of a member of the public to radioactive materials or hazardous chemicals produced from radioactive materials in concentrations exceeding AEGL-2 or ERPG-2 criteria.
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| An acute exposure of a member of the public to radioactive matenals or hazardous chemicals produced from radioactive matenals at concentrations that could cause irreversible health effects is considered a high-consequence event. Two existing criteria, AEGL-2 and ERPG-2, can be used to d fine such concentration levels.
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| AEGL-2 is defined as "The airborne concentration (expressed in ppm or mg/m*) of a substance at or cbove which it is predicted that the general population, including susceptible, but excluding .
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| hypersusceptible, individuals, could experience irreversible or other serious, long-lasting effects or impaired ability to escape." ERPG-2 is defined as "The maximum airborne concentration below which it is believed that nearly allindividuals could be exposed for up to 1 hour without experiencing or developing irreversible or other health effects or symptoms that could impair an individual's ability to
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| {
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| trke protective action." If, for a particular chemical, the AEGL-2 value is available, it should be used. I Otherwise the ERPG-2 value should be used. If there is no AEGL or ERPG value availabic, then the cpplicant should adopt a criterion that is comparable in severity to those that have been established for ether chemicals.
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| Section 70.60(b)(2)(i)(A). This paragraph deals with an acute exposure of a worker to a radiation dose of between 0.25 Sv (25 rem) and 1 Sv (100 rem) TEDE.
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| The exposure of a worker to a radiation dose between 0.25 Sv (25 rem) and 1 Sv (100 rem)
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| TEDE is considered an intermediate-consequence event. The basis for this choice is the use of 0.25 Sv (25 rem) as an exposure criterion in existing NRC regulations. For example, in 10 CFR 20.2202,
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| " Notification of incidents," immediate notification is required of a licensee if an individual receives "... a total effective dose equivalent of 0.25 Sv (25 rem) or more." Also, in 10 CFR 20.1206, " Planned special exposures," a licensee may authorize an adult worker to receive a dose in excess of normal occupational exposure limits if a dose of this magnitude does not exceed 5 times the annual dose limits ATTACHMENT 2 1 1
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| l
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| 6
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| [i.e.,0.25 Sv (25 rem)) during an individual's lifetime. In addition, the EPA's Protective Action Guides (U.S. Environmental Protection Agency,1992) and NRC's regulatory guidance (Regulatory Guide 8.29 1996) identify 0.25-Sv (25-rem) as the whole-body dose limit to workers for life-saving actions and protection of large populations. NCRP has also stated that a TEDE of 0.25 Sv (25 rem) corresponds to th) once-in-a-lifetime accidental or emergency dose for workers. However, its use is not intended to imply that 0.25 Sv (25 rem) constitutes an acceptable criterion for an emergency dose to a worker.
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| Rtther, this dose value has been proposed in this section as a reference value, which should be used
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| )
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| by licensees to determine the level of protection (i.e., items relied on for safety, and measures to assure their continuous availability and reliability) needed to ensure an acceptably low level of risk to workers.
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| Section 70.60(b)(2)(i)(B). This paragraph deals with an acute exposure of a worker to radioa:tive materials or hazsrdous chemicals produced ' em raeactive materials in concentrations between AEGL-2 (ERPG 2) and AEGL-3 (ERPG-3) criteria.
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| An acute exposure of a worker to radioactive matenals or hazardous chemicals produced from radioactive materials at concentrations that could cause irreversible health effects (but below concentrations that could cause death or life-threatening effects) is considered an intermediate-consequence event. Two existing standards, AEGL-2 and ERPG-2, can be used to define the concentration level for irreversible health effects [see definitions in 970.60(b)(1)(iii)(C), above). Two cdditional standards, AEGL-3 and ERPG-3, can be used to define the concentration level for death or lif2-threatening effects [see definitions in 970.60(b)(1)(ii)(B), above] . If, for a particular chemical, the AEGL values are available, they should be used. Otherwise the ERPG values should be used. If there cro no AEGL or ERPG values available, then the applicant should adopt criteria that are comparable in severity to those that have been established for other chemicals.
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| Section 70.60(b)(2)(ii)(A). This paragraph deals with an acute exposure of a member of the public to a radiation dose between 0.05 Sv (5 rem) and 0.25 Sv (25 rem) TEDE.
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| The exposure of a member of the public to a radiation dose between 0.05 Sv (5 rem) and 0.25 Sv (25 rem) is considered an intermediate-consequence event. NRC has used a 0.05-Sv (5-rem) cxposure criterion in a number of its existing regulations. For example,10 CFR 72.106, " Controlled creo of an ISFSI or MRS," states that "Any individuallocated on or beyond the nearest boundary of the controlled area shall not receive a dose greater than 5 rem to the whole body or any organ from ATTACHMENT 2 I
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| )
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| I
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| L any design basis accident." In addition, in the regulation of geologic repository operations,10 CFR 60.136, states that ".~..for Category 2 design basis events, no individual located on or beyond any point on the boundary of the preclosure controlled area will receive...a total effective dose equivalent of 5 remi.." A TEDE of 0.05 SV (5 rem)is also the upper limit of EPA's Protective Action Guides of
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| - between 0.01 to 0.05 Sv (1 to 5 rem) for emergency evacuation of members of the public in the event of an accidental release that could result in inhalation, ingestion, or absorption of radioactive materials.
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| Section 70.60(b)(2)(ii)(B). This paragraph deals with an acute exposure of a member of the public to radioactive materials or hazardous chemicals produced from radioactive materials in concentrations between AEGL-1 (ERPG-1) and AEGL-2 (ERPG-2) criteria.
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| An acute exposure of a member of the public to radioactive materials or hazardous chemicals produced from radioactive matenals at concentrations that could cause notable discomfort (but below concentrations that could cause irreversible effects) is considered an intermediate-consequence cvant. Two existing standards, AEGL-1 and ERPG-1, :an be used to define the concentration level for not:ble discomfort. AEGL-1 is defined as "The airborne concentration (expressed in ppm or mg/m3) of a substance at or above which it is predicted that the general population, including susceptible, but
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| - cxcluding hypersusceptible, individuals, could experience notable discomfort." ERPG-1 is defined as "Th3 maximum airbome concentration below which it is believed that nearly all individuals could be Exposed for up to 1 hour without' experiencing other than mild transient adverse effects or perceiving a j cle:rly defined, objectionable odor." Two additional standards, AEGL-2 and ERPG-2, can be used to define the concentration level for irreversible health effects [see definitions in 970.60(b)(1)(iii)(C),
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| - cbove).: If, for a particular chemical, the AEGL values are available, they should be used. Otherwise j
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| - the ERPG values should be used. If there are no AEGL or ERPG values available, then the applicant i should adopt criteria that are comparable in severity to those that have been established for other chemicals.
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| Section 70.60(b)(2)(iii). This paragraph deals with a release of radioactive material to the environment.
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| The release of radioactive material to the environment outside the restricted area in concentrations that, if averaged over a period of 24 hours, exceed 5000 times the values specified in
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| . Tcble 2 of Appendix B to Part 20, is considered an intermediate-consequence event. In contrast to I the other consequences criteria that directly protect workers and members of the public, the intent of
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| ~ this criterion is to ensure protection of the environment from the occurrence of accidents at certain I ficilities authorized to process greater than critical mass quantities of SNM. This implements NRC's ATTACHMENT 2 l
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| 4
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| responsibility for protecting the environment in accordance with the Atomic Energy Act of 1954, et sea.,
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| and the National Environmental Policy Act of 1969, et sea The value established for the environmental consequence criterion is identical to the NRC Abnormal Occurrence (AO) criterion that addresses the discharge or dispersal of radioactive material from its intended place of confinement. (Section 208 of the Energy Reorganization Act of 1974, as cmended, requires that AOs be reported to Congress on an annual basis.) In particular, AO reporting criterion 1.B.1 requires the reporting of an event that involves "...the release of radioactive material to En unrestricted area in concentrations which, if averaged over a period of 24 hours, exceed 5000 times the values specified in Table 2 of Appendix B to 10 CFR Part 20, unless the licensee has demonstrated compliance with 10 CFR 20.1301 using 10 CFR 20.1302(b)(1) or 10 CFR 20.1302(b)(2)(ii)," [ December 19,1996; 61 FR 67072). The concentrations listed in Table 2 of Appendix B to Part 20 apply to rtdioactive materials in air and water effluents to unrestricted areas. NRC established these concentrations based on an implicit effective dose equivalent limit of 0.5 mSv/yr (50 mremlyr) for each medium, assuming an individual were continuously exposed to the listed concentrations present in an unrestricted area for a year.
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| If an individual were continuously exposed for i day to concentrations of radioactive material 5000 times greater than the values listed in Appendix B to Part 20, the projected dose would be about
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| ~ 6.8 mSv (680 mrem), or 5000 x 0.5 mSv/yr x 1 day x 1 yr/365 day. In addition, a release of radioactive mrterial, from a facility, resulting in these concentrations would be expected to cause some Environmental contamination in the area affected by the release. This contamination would pose a 1:nger-term hazard to the environment and members of the public until it was properly remediated.
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| Depending on the extent of environmental contamination caused by such a release, the contamination could require considerable licensee resources to remediate. For these reasons, NRC considered the existing AO reporting criterion for discharge or dispersal of radioactive material as an appropriate consequence criterion in this rulemaking.
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| Several existing fuel fabrication licensees have chosen to demonstrate compliance with the public dose limit in 10 CFR 20.1301, using 10 CFR 20.1302(b)(1). However, in these cases, routine operations at the facilities do not release effluents that come anywhere close to approaching the Table 2 values in Appendix B to Part 20. Indeed, routine discharge of heavy metals such as uranium in concentrations that substantially exceed the Table 2 values in water or air effluents would be expected to cause extensive environmental contamination that would be difficult and expensive to remediate.
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| This has been demonstrated by the extensive and expensive decommissioning actions that have been r2 quired at former fuel fabrication facilities in the United States (see NRC's " Site Decommissioning ATTACHMENT 2
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| M:nagement Plan,' NUREG-1444). In addition, SNM-processing licensees would not be expected to use the compliance method in 10 CFR 20.1302(b)(2)(ii) because this is primarily directed at extemal r:diation hazards, whereas the materials released from SNM processing facilities primarily represent internal radiation and chemical nsks hazardc. Consequently, there is no need to retain the caveat r:garding alternative means of demonstrating compliance with the public dose limit, as found in the AO r: porting criterion.
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| Section 70.60(ct This paragraph deals with the graded level of protection.
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| This section addresses the level of protection a licensee must provide to ensure an acceptable degree of risk at its facility. That protection must be sufficient to reduce the likelihood of potential tecidents to levels commensurate with their consequences. In determining the appropriate level of protection that the licensee must provide, consideration may be given to the inherent likelihood of the cecident. By inherent, we mean the likelihood of the accident, assuming no controls are in place.
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| Thus, an accident that is initiated by an unlikely external eveat may require less protection (provided by tha licensee) than an accident, with identical consequence, that is initiated by a more frequent event.
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| For example, suppose a serious fire, with high consequences, could be started as the result of a process deviation that is estimated to occur once per year. The level of protection needed to prevent or mitigate this accident would be greater than that needed to protect against a similar fire resulting from an unlikely extemal event, such as an earthquake that might occur once in 500 years. Thus, licensees may take credit for inherent " unlikeliness" of an accident in determining the level of protection th:t needs to be applied.
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| The goal of applying a graded level of protection is to reduce the likelihood or consequences of cccidents' to ensure an acceptable level of risk at the licensee's facility. For each of the high-cnnsequence events identified in the proposed 670.60(b), the Commission believes that the occurrence of such an event should be highly unlikely to occur during any given year of plant operation.
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| For each of the intermediate-consequence events identified in the proposed $70.60(b), the C:mmission believes that the occurrence of such an event should be unlikely to occur during any given y:ar of plant operation.
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| The Commission has decided not to include a quantitative definition of "unlikely" and " highly unlikely" in the proposed rule, because a single definition for each term may not be appropriate.
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| 5 For exposures of workers or members of the public to radioactive matenals or hazardous chemicals produced from radioactive materials during normal operations, adherence to the existing requirements of 10 CFR 20 and 29 CFR 1910 should be sufficient to protect the public health and safety.
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| ATTACHMENT 2 l
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| i
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| )
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| E Depending on the type of facility and its complexity, the number of potential accidents and their c:nsequences, which are identified in the ISA, could differ markedly. Thus, even if the permitted lik:lihood for each gygr11 were quantitatively defined, the integrated risk for a given facility would depend on the number of such events that could occur and the consequences of those events. For ex mple, some facilities may have few potential accidents in the "high-consequence
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| * range while cthers may have many potential accidents in this range. Therefore, to ensure that the overall facility risk is acceptable for different types of facilities, guidelines for interpreting "likely" and " highly unlikely" m y need to be adjusted accordingly. To accommodate the potential variation in these guidelines, the Ccmmission believes that the standard review plan is the appropriate document to address these tsrms. The " Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility,"
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| which is being made available with the proposed rule, provides guidelines that can be applied to existing fuel cycle facilities. These guidelines have been selected so as to be consistent with the safety performance goals in the NRC Strategic Plan (NUREG-1614, Vol.1). The Commission intends to publish standard review plans for different types of facilities licer. sed by NRC, as the need arises.
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| Appropriate guidelines for such facilities can be addressed in the standard review plans at that time.
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| Section 70.60(dI.' This paragraph deals with the safety program. ;
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| E The performance of an ISA, and the establishment of measures to ensure the continuous !
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| avcilability and reliability of items relied on for safety, are the means by which licensees are able to demonstrate their ability to provide an adequate level of protection at their facilities. The ISA is a syst;matic analysis to identify plant and external hazards and their potential for initiating accident sequences; the potential accident sequences and their consequences; and the site, structures, systems, equipment, components, and activities of personnel, relied on for safety. As used here, integrated means joint consideration of, and protection from, all relevant hazards, including r:di: logical, criticality, fire, and chemical. The structure of the safety program recognizes the critical rob that the ISA plays in identifying potential accidents and the items relied on for safety. However, it clso recognizes that the performance of the ISA, by itself, will not ensure adequate protection. Instead, cn cffective management system is needed to ensure that, when called on, the items relied on for safety are continuously in place and operating properly.
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| There are four major steps in performing an ISA:
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| (1) Identify all hazards at the facility, including both radiological and non-radiological hazards.
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| Haz:rdous materials, their location, and quantit'es, i should be identified, as well as all hazardous conditions, such as high temperature and high pressure. In addition, any interactions that could result ATTACHMENT 2
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| * r
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| - in the generation of hazardous materials or conditions should be identified.
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| (2) Analyze the hazards to identify how they might result in potential accidents. These accidents could be caused by process deviations or other events internal to the plant, or by credible sxttmal events, including natural phenomena such as floods, earthquakes, etc. To accomplish the task of identifying potential accidents, the licensee needs to ensure that detailed and accurate inf rmation about plant processes is maintained and made available to the personnel performing the ISA.
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| (3) Determine the consequences of each accident that has been identified. For an accident with consequences at a high or intermediate level, as defined in 10 CFR 70.60(b), the likelihood of such an accident must be shown to be commensurate with the consequences, as required in the prcposed 10 CFR 70.60(c). Protection against accidents with consequences below the intermediate lev 11 threshold is assumed to be provided by adherence to existing NRC, OSHA, and EPA regulations.
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| (4) Identify the items relied on for safety (i.e., those items that are relied on to prevent or to mitigate the accidents identified in the ISA). Such items are needed to reduce the likelihood or consequences of the accidents to acceptable levels. The identification of items relied on for safety is rsquired only for accidents with consequences at a high or intermediate level, as defined in the proposed 10 CFR 70.60(b).
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| Manaaement control. ~ Although the ISA plays a critical role in identifying potential accidents cnd the items relied on for safety, the performance of an ISA will not, by itself, ensure adequate prctection. Instead, according to the proposed 10 CFR 70.60(d), an effective management system is needed to ensure that, when called on, the items relied on for safety are continuously available and r;li:ble (i.e., in place and operating properly). Maintenance measures must be in place to ensure the continuous availability and reliability of all hardware relied on for safety. Training measures must be csttblished to ensure that all personnel relied on for safety are appropriately trained to perform their stfety functions. Human-system interfaces and safety-related procedures must be developed and implemented to enable personnel relied on for safety to effectively carry out their duties. Changes in the confguration of the facility need to be carefully controlled to ensure consistency among the facility design and operational requirements, the physical configuration, and the facility documentation in Eddition, quality assurance measures need to be established to ensure that the items relied on for 1 sifsty and the measures used to ensure their continuous availability and reliability are of sufficient quility. Periodic audits and assessments of licensee safety programs must be performed to ensure i thrt facility operations are conducted in compliance with NRC regulations and protect the worker and ATTACHMENT 2 l l
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| l
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| the public health and safety. When abnormal events occur, investigations of those events must be carried out to prevent their recurrence and to ensure that they do not lead to more serious c:nsequences. Finally, to demonstrate compliance with NRC regulations, records that document
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| . sifety program activities must be maintained for the life of the facility.
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| Section 70.62. This section deals with requirements for the performance of ISAs and the filing cf ISA results and license applications. These requirements address the question of who should perform ISAs, when they should be performed, and what ISA information should be provided to NRC.
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| The performance of an ISA would be required of all licensees authorized to possess a critical miss of SNM, that are engaged in one of the following activities: enriched uranium processing; urtnium fuel fabrication; uranium enrichment; enriched uranium hexafluoride conversion; plutonium processing; mixed-oxide fuel fabrication; scrap recovery; or any other activity that the Commission
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| .d:t;rmines could signifi:ently affect public health and safety. The Commission believes that possession and processing of SNM in amounts sufficient to constitute a potential for nuclear criticality l
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| is a reasonable criterion for requiring the performance of an ISA.- Licensees meeting this criterion are cirIndy subject to nuclear criticality monitoring and alarm requirements that ensure an adequate response to a nuclear criticality event after it occurs. The performance of an ISA provides the means f r licensees to ensure adequate measures are taken to prevent a nuclear criticality event (or other high-consequence event) before it occurs. By limiting the requirement for performance of an ISA to licensees engaged in specific activities that involve major chemical or mechanical processing of SNM, the Commission recognizes that these activities involve a higher degree of risk than the activities of licensees who are authorized to possess critical quantities of SNM, but do not perform any mechanical or chemical processing of critical or near-critical quantities of the SNM.
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| These types of facilities include sub-critical assemblies, where the critical mass of material is fixed in place in such a manner that an inadvertent nuclear criticality is not credible; research facilities that are authorized to possess a critical quantity of material, but do not process more than a small fraction of that material at any one laboratory; facilities that are authorized only to store the material; cnd facilities no longer operating, for which the material is dispersed throughout the facility as residue in walls, floors, or other fixed structures. However, potentially hazardous activities involving cleanup and decommissioning at non-operating facilities would be subject to the ISA requirement.
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| The proposed rule would require current Part 70 licensees, for whom the rule would be
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| . applicable to develop compliance plans and submit them to NRC within 6 months of the effective date cf the rule. ; Each compliance plan would identify the processes that would be subject to an ISA, the ATTACHMENT 2 1 , )
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| L
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| 1
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| {
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| ISA approach that would be implemented for each processi and the schedule for com >leting the enalysis of each process. Licensees would be expected to complete their ISAs within 4 years of the cffective date of the rule, correct any unaccefeble vulnerabilities identified, and suomit tu NRC tne i results for evaluation, approval, and incorporation in the license. Pending the correction of any unacceptable vulnerabilities, licensees would be expected to implement appropriate compensatory msasures to ensure adequate protection. The process description in the ISA submittal should contain information that demonstrates the licensee's compliance with the nuclear criticality monitoring and
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| . elarm requirements in 10 CFR 70.24.
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| Applicants operating existing facilities that could become newly subject to the Commission's authority, such as DOE facilities, would be expected to perform ISAs and submit tt'e results as part of th:ir applications for licenses. The ISA submittals should contain information that demonstrates the licensees
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| * compliance with the nuclear criticality monitoring and alarm requirements in 10 CFR 70.24.
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| Applicants for licenses to operate new facilities or new processes at existing facilities would be expected to design their facilities or processes to protect against the occurrence of the adverse consequences identified in the proposed 10 CFR 70.60(b). In addition, the initial designs are expected to comply with the nuclear criticality monitoring and alarm requirements in 10 CFR 70.24 and the baseline design criteria in the proposed 10 CFR 70.64.
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| Based on these initial designs, the applicants are expected to perform preliminary ISAs before construction of facilities. If the ISA results show deficiencies in the design, the design should be modified to assure that the items and measures planned to protect against identified accidents are cdequate. On the other hand, if the ISA results show that a given item at a given facility is not relied on )
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| for safety, or that it does not require full adherence to the baseline criteria, then the facility design may be modified accordingly. The applicant is expected to submit the results of the preliminary ISA, based
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| )
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| on the modified design of the facility, to NRC before construction. However, NRC approvalis not I necessary for the applicant to proceed with construction. The submittal should include the identification cf all cases where a deviation from the baseline criteria is proposed, along with a justification for that decision. .The submittal of the preliminary ISA for review by NRC provides an opportunity for cpplicants to get early feedback on the design of their facilities or processes. It is much more cost-sffective to correct problems identified at the design stage than after the facility has been constructed.
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| After construction, but before operation, applicants would be expected to update their ISAs, based on as-built conditions, taking into account the results of the preliminary ISAs, and submit the results to NRC for approval. Any inconsistencies between the results of the updated ISAs and the preliminary ISAs should be identified in the submittals.
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| ATTACHMENT 2 4
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| i Section 70.64. This section deals with baseline design criteria for new facilities or new processes at existing facilities.
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| A major feature of the proposed amendments to Part 70 is the requirement that licensees and cpplicants for a license perform an ISA. The ISA process is applied to existing designs to identify high risks that could warrant additional preventive or mitigative measures. For new facilities or new processes at existing facilities, the proposed rule calls for the performance of the ISA before construction, and the updating of the ISA before beginning operations. However, for new processes cnd facilities, the Commission recognizes that good engineering practice dictates that certain minimum rzquirements be applied as design and safety considerations for any new nuclear process or facility.
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| Therefore, the Commission has specified baseline design criteria in 970.64 that are similar to the general design criteria in Part 50 Appendix A; Part 72, Subpart F; and 10 CFR 60.131. The baseline j
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| design criteria identify 10 initial safety design considerations, including: quality standards and records; n tural phenomena hazards; fire protection; environmental and dynamic effects *; chemical protection; cmergency capability; utility services; inspection, testing, and maintenance; nuclear criticality control; cnd instrumentation and controls. The baseline design criteria do not provide relief from compliance with the safety performance requirements of 970.60. The baseline design criteria are generally an Ecceptable set of initial design safety considerations, which may not be sufficient to assure adequate
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| - safnty for all new processes and facilities. The ISA process is intended to identify additional safety f;atures that may be needed On the other hand, the Commission recognizes that there may be -
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| processes or facilities for which some of the baseline design criteria may not be necessary or cppropriate, based on the results of the updated ISA. For such processes and facilities, any design
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| . f:stures that are inconsistent with the baseline design criteria should be identified and justified.
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| Section 70.65. This section deals with the additional content of applications.
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| There is additional information that would need to be submitted to NRC as part of a license cpplication to demonstrate compliance with the additional requirements that would be established in the proposed new subpart. This information is necessary to determine whether the applicant has '
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| . pr:vided an adequate level of protection at the facility. In particular, additionalinformation would be needed to demonstrate how the applicant's safety program complies with 10 CFR 70.60(d). This Envwonmental and dyname effects are effects that could be caused by ambient conditons. For example, an item relied on for safety will need to function within its expected envronment (i.e., under normal operating conditions, expected acadent conditions, etc.). These condatens could include high temperatures, or a corroseve environment. It could also include dynarmc changes in surroundmg conditions caused by an acodont (e.g., the bursting of a high-pressure pipe).
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| ATTACHMENT 2 ;
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| information would include a description of the plant site and structures; the processes analyzed in the ISA; an appropriate summary of the results of the ISA, including the accident sequences, the consequences and likelihoods of such sequences; and the items relied on for safety; and the measures cstablished to ensure the continuous availability and reliability of such items. The plant and process descriptions are needed to fully understand the results of the ISA, including the rationale for choosing the items relied on for safety. The evaluation of the applicant's safety program is a critical element in determining whether the facility is safe and should be issued a license. Finally, the license application, for an operating facility, should include a description of operational events that have occurred during th] past 10 years that had a significant impact on the safety of the facility. These events should be cddressed in the applicant's ISA to ensure that the range of accident sequences considered in the ISA cncompasses actual events that have occurred at the facility.
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| The license application demonstrates how the applicant intends to meet the requirements of P;rt 70. The application provides information about the applicant's facility and processes and comm.tments that ensure the health and safety of workers, the pneral public and the environment. To cnsure confidence that these commitments will be adhered to, and will not be changed without NRC knowledge or approval, the following condition will be inserted in the license: " Authorized use: For use in accordance with the statements, representations, and conditions in the application dated .
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| cnd supplements dated The application may be revised in accordance with the provisions of 10 CFR 70.72." This condition is similar to the ones currently in use. However, it would apply to the cntire license application (not just a portion of the application, as was done previously), and would cllow changes to be made without prior NRC approval, in accordance with 10 CFR 70.72.
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| Section 70.66. This section deals with records.
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| {
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| NRC confidence in the margin of safety at its licensed facilities depends, in part, on the ability of licensees to maintain a set of current, accurate, and complete records available for NRC inspection.
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| These records serve two major purposes. First, they can supplement information that has been submitted as part of the license application. For example, applicants would be required to submit the r::sults of their ISAs to NRC for review. However, there may be substantial amounts of supporting m:terial, at the licensed facility, relevant to that submittal, that NRC may wish to review. Second, records are often needed to demonstrate licensee compliance with applicable regulations and license commitments. It is important, therefore, that an appropriate system of recordkeeping be implemented 13 cllow easy retrieval of required information.
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| I l
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| ATTACHMENT 2 1
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| o
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| ~
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| . Section 70.68. This section deals with additional requirements for the approval of license cpplications.
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| In addition to the requirements found in the existing rule (i.e.,10 CFR 70.23 ), the Commission must determine that the requirements in the proposed new subpart,10 CFR 70.60 through 70.66, will be satisfied.
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| Section 70.72. This section deals with changes to site, structures, systems, equipment, components, and activities of personnel.'
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| Past incidents at fuel cycle facilities have often resulted from changes not fully analyzed, not cuthorized by management, or not adequately understood by facility personnel. Therefore, effective control of changes to a facility's site, structures, systems, equipment, components, and activities of personnel is a key element in assuring confidence in the margin of safety at that facility. Any such change needs to be considered and evaluateo by the licensee t,efore the change is made. If the licensee evaluates the change, based on its ISA, and finds hat it, at most, increases the risk at the f:cility to a minimal extent, then the licensee may make the change and then notify NRC within 60 d:ys.' Otherwise, the licensee would need to request a license amendment and get NRC approval before making the change. In either case, the change should be controlled by the licensee's configuration management system, and appropriate modifications to the license application (including, if cpplicable, the results of the ISA) should be submitted to NRC, Aside from providing increased confidence in the margin of safety, maintaining the license so that it reflects the current configuration of the facility would facilitate a relatively simple, cost-effective license renewal process. The ability of licensees to make certain changes to their facility without prior NRC approval, as allowed in this proposed requirement, is analogous to existing requirements in 10 CFR 70.32.
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| Section 70.73. This section deals with the renewal of licenses.
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| Under the proposed amendments to Part 70, changes to site, structures, systems, equipment, components, and activities of personnel, made by a licensee, would be reflected in the license cpplication, which would be submitted to NRC and incorporated as a condition of the license. This process would establish a "living" license that would be maintained on a current basis. As a result, the
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| ~ license renewal process is expected to be a pro forma activity in which NRC, based on its current knowledge of licensee activities, as reflected in the "living license," would approve the renewal with minimal additional review of the licensee's safety program. This approval would be contingent on the licensee satisfying any requirements associated with the National Environmental Policy Act of 1969 as ATTACHMENT 2
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| [.
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| implemented in 10 CFR Part 51.
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| Section 70.74. This section deals with additional reporting requirements.
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| The new requirements that would be incorporated in the proposed changes to Part 70 suggest a revised approach for reporting of events to NRC. This new approach, based on consideration of the consequences of concern established in 10 CFR 70.60(b)', is intended to replace and expand on the cpproach licensees have currently been using for reporting nuclear criticality events under Bulletin 91-
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| : 01. The new approach would cover all types of events, not just nuclear criticality events, and establish c timeframe for reporting that is scaled according to risk. The new reporting requirements are intended to supplement the requirements in the existing Part 70. A more detailed discussion of the new r:quirements is found in the discussion of Appendix A_ G to Part 70.
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| Append!r ^ "^ cute Exeer'.r: Gu!d:!!n: Lev !: "EGL '" Th!:-cppend!r cent:!n: the ^.EGL v !uer, for
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| * hour expectre:, th:t h:v: 5::n ect:M!:hed by EP^ There v:!uer ::: ref:renced !- 10 CF .
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| ~ 70.SO(b).
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| Appnd!: 9. "EPPG" Th!: 2ppnd!x cent:!n: the EPoG ve'uer th:t h: : b::n : t9!! hed by ^'".A
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| ^ :: v !ce: :r: :f:renced !- 10 CFo 70.S0(b).
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| Appendix A_C.
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| * Reportable Events" To effectively fulfill its responsibilities, NRC needs to be aware of conditions that could result in cn imminent danger to the worker or to public health and safety, in the event of an accident, NRC must be able to respond accurately to requests for information by the public and the media. in addition, to the extent possible, NRC needs to be able to provide appropriate assistance to licensees in their efforts to address potential emergencies. Once safe conditions have been restored after an cv:nt, NRC has an interest in disseminating information on the event to the nuclear industry and other int; rested parties, to reduce the likelihood that the event will occur in the future. Finally, NRC must track the performance of individual licensees and the industry as a whole to fulfill its statutory mandate t3 protect the health and safety of the worker and the public.
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| NRC intends to take a graded approach for reporting licensee events, as illustrated in Table 2.
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| According to this approach, licensees would report events based on whether actual consequences have occurred or whether a potential for such consequences exists. The most serious events, and those that must be reported within the shortest timeframe (1 hour) are high-consequence events that have actually occurred. Intermediate-consequence events that have actually occurred should be ATTACHMENT 2
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| m p .
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| l r: ported within 4 hours.
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| Events that could potentially lead to a consequence of concem should also be reported.
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| External conditions, such as a hurricane, tornado, or flood, that could pose a threat to safety at a f cility, should be reported within 4 hours. Deviations from safe operating conditions should be rrported within a time period that depends on the severity of the potential consequence and whether or not the licensee is able to correct the deviation within the specified period. A deviation from safe cperating conditions means that a parameter that is controlled to ensure adequate protection is outside its established safety limits, or that an item relied on for safety is no longer operational or has been degraded so that it cannot perform its intended function. The reporting requirements for deviations from safe operating conditions are intended to be generally consistent with the reporting scheme 6stablished under Bulletin 91-01. For example, if a nuclear criticality control identified in the ISA is no longer operational, or degraded so that it cannot perform its intended function, that situation should be rcported to NRC. If the control cannot be reestablished within 4 hours of discovery, the report should i be made before expiration of the 4-hour time period, if the control has been reestablished within 4 h urs of discovery, the report should be made within 24 hours. The term " reestablish"is intended to msan that the control identified in the ISA is made operative. Therefore, if a control fails and an ad-hoc control, not identified in the ISA, is established within 4 hours of discovery, a report to NRC would still h:ve to be made before expiration of the 4-hour time period.
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| Another category of potential events that should be reported is one that involves the existence of cn unsafe condition that is not identified in the ISA. This condition could be caused by a deviation from established safe operating conditions, or by an unanticipated and unanalyzed set of circumstances. The timefrarne for reporting this type of event would depend on how long it takes the licensee to remove the unsafe condition, and restore normal operations. If the licensee were unable to restore normal operating conditions within 4 hours, the report would need to be made before expiration of the 4-hour period. If the licensee were able to remove the unsafe condition and restore normal operations within 4 hours, the report would need to be made within 24 hours.
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| in l l
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| | |
| REFERENCES Graig, D.K., pt RL,, " Alternative Guideline Limits for Chemicals Without Environmental Response Planning Guidelines," American Industrial Hygiene Association Journal,1995.
| |
| Fisher, D.R., Hui, T.E., Yurconic, M., and Johnson, J.R., " Uranium Hexafluoride Public Risk,"
| |
| Pacific Northwest National Laboratory, PNL-10065, Richland, WA, August 1994.
| |
| National Council on Radiation Protection and Measurements (NCRP), " Basic Radiation Protection Criteria," NCRP Report No. 39, Washington, DC,1971.
| |
| National Council on Radiation Protection and Measurements (NCRP), " Recommendations on Limits for Exposure to lonizing Radiation," NCRP Report No. 91, Washington, DC,1987.
| |
| U.S. Nuclear Regulatory Commission, " Proposed Methods for Regulating Major Materials Licensees," NUREG-1324, Washington, DC, February 1992.
| |
| U.S. Nuclear Regulatory Commission / Occupational Safety and Health Administration (OSHA),
| |
| " Memorandum of Understanding Between NRC and OSHA; Worker Protection at NRC-Licensed Facilities" (53 FR 43950; October 31,1988).
| |
| U.S. Nuclear Regulatory Commission, " Certification of Gaseous Diffusion Plants" (59 FR 48944; September 23,1994).
| |
| U.S. Nuclear Regulatory Commission. " Abnormal Occurrence Reports: Implementation of Section 208 of Energy Reorganization Act of 1974" (61 FR 67072; December 19,1996).
| |
| U.S. Nuclear Regulatory Commission, " Site Decommissioning Management Plan," NUREG-1444, Washington, DC, October 1993.
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| U.S. Nuclear Regulatory Commission, " Strategic Plan, Fiscal Year 1997 - Fiscal Year 2002,"
| |
| NUREG-1614, Washington, DC, September 1997.
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| ATTACHMENT 2
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| U.S. Environmental Protection Agency, " Manual of Protective Action Guides and Protective Actions for Nuclear incidents, EPA-400-R-92-001, May 1992.
| |
| U.S. Nuclear Regulatory Commission, " Instruction Concerning Risks from Occupational Radiation Exposure," Regulatory Guide 8.29, Rev.1, February 1996.
| |
| Theide, L., " Emergency information Where it's Needed," DOE Risk Management Quarterly, Vol 5, No 2, Richland, WA, May 1997.
| |
| These documents are available for inspection and copying for a fee at the NRC Public Dccument Room,2120 L Street, N.W. (Lower Level), Wash;ngton DC 20555-0001.
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| Copies of NUREG-1324, NUREG-1614, and NUREC-1444 may also be purchased from the Superintendent of Documents, U.S. Govemment Printing Office, P.O. Box 37082, Washington DC 20402-9328. Copies are also available from the National Technical Information Service,5285 Port Royal Road, Springfield VA 22161.
| |
| Regulatory Guide 8.29 may be purchased from the Govemment Printing Office (GPO) at the current GPO price. Information on current GPO prices may be obtained by contacting the Superintendent of Documents, U.S. Govemment Printing Office, P.O. Box 37082, Washington DC 20402-9328. Issued guides may also be purchased from the National Technical Information Service on a standing order basis. Details on this service may be obtained by writing NTIS,5285 Port Royal Road, Springfield, VA 22161.
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| Copies of the following draft regulatory guidance documents are available by request from the NRC Public Document Room: " Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility" (Draft NUREG-1520); " Integrated Safety Analysis Guidance Document" (Draft NUREG-1513); and " Example Elements of an ISA Submittal- Process Descriptions and Accident Analysis Summary."
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| Finding of No Significant EnvironmentalImpact: Availability The Commission has determined, under the National Environmental Policy Act of 1969, as cmended, and the Commission's regulations in subpart A of 10 CFR Part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment, and ATTACHMENT 2
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| therefore an environmental impact statement is not required.
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| The proposed amendments to Part 70 are intended to provide increased confidence in the margin of safety at certain facilities that possess a critical mass of SNM. To accomplish this objective, the amendments: (1) identify appropriate consequence criteria and the level of protection needed to prevent or mitigate accidents that exceed such criteria; (2) require affected licensees to perform an ISA to identify potential accidents at the facility and the items relied on for safety; (3) require the implementation of measures to ensure that the items relied on for safety are continuously available and
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| - reliable; and (4) require the inclusion of the safety bases, including the results of the ISA, in the license application. The language, in the proposed rule, that defines an environmental consequence of concem, is relevant to the question of environmental impact. Licensees would be required to provide an adequate level of protection against a "... release of radioactive material to the environment outside the restricted area in concentrations that, if averaged over 24 hours, exceed 5000 times the values specified in Table 2 of Appendix B to 10 CFR Part 20." implementation of the new amendments, including the requiremer.t to protect against etents that coulc' damage the environment, is expected to result in a significant improvement in licensees' (and NRC's) understanding of the risks at their facilities and their ability to ensure that those risks are acceptable. For existing licensees, any deficiencies identified in the ISA would need to be promptly addressed. For new licensees, operations would not begin unless licensees demonstrated an adequate level of protection against potential accidents identified in the ISA. As a result, the safety and environmental impact of the new amendments is positive. There wili be less adverse impact on the environment from operations carried out in .
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| accordance with the proposed rule than if those operations were carried out in accordance with the existing Part 70 regulation.
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| The determination of this environmental assessment is that there will be no significant offsite impact to the public from this act' ion. However, the general public should note that NRC welcomes public participation. NRC has also committed to complying with Executive Order (EO) 12898, " Federal Actions to Address Environmental Justice in Minority Populations and Low-income Populations," dated February 11,1994, in all its actions. Therefore, NRC has also determined that there are no disproportionate, high, and adverse impacts on minority and low-income populations. In the letter and spirit of EO 12898, NRC is requesting public comment on any environmentaljustice considerations or questions that the public thinks may be related to this proposed rule, but somehow were not addressed. Comments on any aspect of the Environmental Assessment, including environmental justice, may be submitted to NRC, as indicated under the ADDRESSES heading.
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| NRC has sent a copy of the environmental assessment and this proposed rule to all State ATTACHMENT 2
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| Liaison Officers and requested their comments on the Environmental Assessment. The Environmental Assessment is available for inspection at the NRC Public Document Room,2120 L Street NW. (Lower Level), Washington, D.C. Single copies of the environmental assessment are available from Richard 1.
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| Milstein, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC,20555-0001, telephone (301) 415-8149; e-mail: rim @nrc. gov.
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| Paperwork Reduction Act Statement This proposed rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501, et sea.). This rule has been submitted to the Office of Management and Budget (OMB) for review and approval of the papenwork requirements.
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| The public reporting burden for this information collection is estimated to average 70 hours per response. and the recordkeeping burden is estimated to average 500 hours per licensee, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the information collection. The U.S. Nuclear Regulatory Commission is seeking public comment on the potential impact of the information collections contained in the proposed rule and on the following issues:
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| 1, is the proposed information collection necessary for the proper performance of NRC's function? Will the information have practical utility?
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| : 2. Is the burden estimate accurate?
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| : 3. Is there a way to enhance the quality, utility, and clarity of the information to be collected?
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| : 4. How can the burden of the information collection be minimized, including the use of Cutomated collection techniques?
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| Send comments on any aspect of this proposed information collection, including suggestions for reducing the burden, to the Records Management Branch (T-6-F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet electronic mail at bjs1@nrc. gov; and to the Desk Officer, Office of information and Regulatory Affairs, NEOB-10202 (3150-000g), Office of Management and Budget, Washington, DC 20503.
| |
| Comments to OMB on the information collections or on the above issues should be submitted by (insert 30 days after publication in the Federal Register). Comments received after this date will be considered if it is practical to do so, but assurance of consideration cannot be given to comments received after this date.
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| ATTACHMENT 2
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| O Public Protection Notification if an information collection does not display a currently valid OMB control number, NRC may not conduct nor sponsor, and a person is not required to respond to the information collection.
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| Regulatory Analysis The Commission has prepared a draft regulatory analysis on this proposed regulation. The analysis examines the costs and benefits of the attematives considered by the Commission. The draft analysis is available for inspection in the NRC Public Document Room,2120 L Street N.W. (Lower Level), Washington, D.C. Single copies of the analysis may be obtained from Barry T. Mendelsohn, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC, telephone (301) 415- 7262, e-mail: btm1@nrc. gov. I The Commission requests public comment on the draft regulatory analysis. Comments on the draft analysis may be submitted to NRC as indicated under :he ADDRESSES heading.
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| [
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| l i
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| Regulatory Flexibility Certification As required by the Regulatory Flexibility Act, as amended,5 U.S.C. 605(b), the Commission
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| {
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| certifies that this proposed rule, if adopted, would not have a significant economic impact on a l
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| substantial number of small entities. This proposed rule would affect major nuclear fuel fabrication facilities that are authorized to possess a critical mass of SNM. These licensees do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act, nor the size standards published by NRC (10 CFR 2.810).
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| Backfit Analysis NRC has determined that the backfit rule does not apply to this proposed rule; therefore, a backfit analysis is not required for this proposed rule because these amendments do not involve any provisions that would impose backfits as defined in 10 CFR Chapter 1.
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| List of Subjects in 10 CFR Part 70 Criminal penalties, Hazardous materials transportation, Material control and accounting, Nuclear materials, Packaging and containers, Radiation protection, Reporting and recordkeeping requirements, Scientific equipment, Security measures, Special nuclear material.
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| For the reasons set out in the preamble and under the authority of the Atomic Energy Act of f ATTACHMENT 2 1
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| I 4
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| 1 I
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| 1954, as amended; the Energy Reorganization Act of 1974, as amended. and 5 U.S.C. 553, NRC is proposing to adopt the following amendments to Part 70 Part 70 -- DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL
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| : 1. The authority citation for Part 70 continues to read as follows:
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| AUTHORITY: Secs. 51, 53,161,182,183, 68 Stat. 929, 930, 948, 953, 954, as amended, sec.
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| 234, 83 Stat. 444, as amended (42 U.S.C. 2071, 2073, 2201, 2232, 2233, 2282, 2297f); secs. 201, as amended,202,204,206,88 Stat.1242, as amended, 1244,1245,1246 (42 U.S.C. 5841, 5842, 5845, 5846). Sec.193,104 Stat. 2835, as amended by Pub. L. 104-134,110 Stat.1321,1321-349 (42 U.S.C. 2243).
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| Sections 70.1(c) and 70.20a(b) also issued under secs. 135,141, Pub. L. 97-425,96 Stat.
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| 2232,2241 (42 U.S.C.10155,10161). Section 70.7 also issued under Pub. L. 95-601, sec.10,92 Stat. 2951 (42 U.S.C. 5851). Section 70.21(g) also issued ander sec.122, 68 Stat. 939 (42 U.S.C.
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| 2152). Section 70.31 also issued under sec. 57d, Pub. L 93-377, 88 Stat. 475 (42 U.S.C. 2077).
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| Sections 70.36 and 70.44 also issued under sec.184, 68 Stat. 954, as amended (42 U.S.C. 2234).
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| Section 70.61 also issued under secs. 186,187,68 Stat. 955 (42 U.S.C. 2236,2237). Section 70.62 also issued under sec.108, 68 Stat. 939, as amended (42 U.S.C. 2138).
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| : 2. The undesignated center heMing " GENERAL PROVISIONS" is redesignated as "Subpart A
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| - General Provisions."
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| : 3. In 10 CFR 70.4, the definitions of Acute exposure, Acute exposure guideline levels (AEGLs),
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| Controlled site boundary, Critical mass of SNM, Deviation from safe operating conditions, Double contingency, Emergency response planning guidelines (ERPGs), Hazardous chemicals, Hazardous chemicals produced from radioactive materials. Integrated safety analysis (ISA), items relied on for safety, New process, Results of the ISA, Unacceptable vulnerabilities, and Worker are added, in alphabetical order, as follows:
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| $ 70.4 Definitions.
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| Acute exposure means a single exposure or multiple exposures occurring within a short time (24 hours or less).
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| ATTACHMENT 2
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| | |
| Acute exposure auideline levels (AEGLs) mean chemical concentration levels, established by the National Advisory Committee for Acute Guideline Levels for Hazardous Substances, that, for a defined exposure, would result in anticipated adverse health effects to humans. The following three levels have been established:
| |
| (1) AEGL-1 means the airborne concentration (expressed in ppm or mg/m') of a substance at or above which it is predicted that the general population, including susceptible but excluding hypersusceptible individuals, could experience notable discomfort.
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| (2) AEGL-2 means the airbome concentration (expressed in ppm or mg/m') of a substance at or above which it is predicted that the general population, including susceptible but excluding hypersusceptible individuals, could experience irreversible or other serious, long-lasting effects or impaired ability to escape.
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| (3) AEGL-3 means the airbome concentration (expressed in ppm or mg/m*) of a substance at or above which it is predicted that the general population, iricluding susceptible but excluding hypersusceptible individuals, could experience life-threatening effects or death.
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| Controlled site boundary means the physical barrier surrounding the facility that is used by the licensee to control access. It may or may not coincide with the property boundary.
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| Critical mass of SNM means special nuclear materialin a quantity exceeding 700 grams of contained uranium-235; 520 grams of uranium-233; 450 grams of plutonium; 1500 grams of contained uranium-235, if no uranium enriched to more than 4 percent by weight of uranium-235 is present; 450 grams of any combination thereof; or one-half such quantities if massive moderators or reflectors made of graphite, heavy water, or beryllium may be present.
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| Qgyiation from f.afe operatina conditions means that a parameter that is controlled to ensure adequate protection is outside its established safety limits, or that an item relied on for safety has been lost or has been degraded so that it cannot perform its intended function.
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| Double continoency means a process design that incorporates sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a nuclear criticality accident is possible.
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| . . . . . l l
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| ATTACHMENT 2 I
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| | |
| 1 Emeraency response otannina auidelines (ERPGs) mean chemical concentration levels, established by the American Industrial Hygiene Association, that, for a defined exposure, would result in anticipated adverse health effects on humans. The following three levels have been established:
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| (1) ERPG-1 means the maximum airbome concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour without experiencing other than mild transient adverse effects or perceiving a clearly defined, objectionable odor.
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| (2) ERPG-2 means the maximum airbome concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour without experiencing or developing irreversible or other health effects or symptoms which could impair an individual's ability to take protective action.
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| (3) ERPG-3 means the maximum airbome concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour without experiencing or developing life-threatening health effects.
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| Hazardous chemicals means substances that are toxic, explosive, flammable, corrosive or reactive to the extent that they can cer 9;9f!=nt deme; te pr:My cr endanger life if not adequately controlled.
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| Hazardous chemicals produced from radioactive materials means Hazardous chemicals either havina radioactive material (s) as precursor compound (s) or formed throuah interaction with radioactive materials Thev do not include chemicals merely added to used in. or recycled from. the processina of special nuclear material Intearated safety analysis (ISA) means a systematic analysis to identify plant and extemal hazards and their potential for initiating accident sequences, the potential accident sequences, their likelihood and consequences, and the site, structures, systems, equipment, components, and activities of personnel that are relied on for safety. As used here, integrated means joint consideration of, and protection from, all relevant hazards, including radiological, nuclear criticality, fire, and chemical.
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| Items relied on for safety means structures, systems, equipment, components, and activities of personnel that are relied on to prevent or to mitigate potential accidents at a facility.
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| New process means, for a particular licensee, a change in the basic method for processing special nuclear material, where the new method is not currently specifically authorized by the' NRC license.
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| A1TACHMENT 2
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| | |
| Results of the ISA means the information obtained as a result of performing an ISA. It includes the identification of: (1) the radiological and non-radiological hazards at the facility; (2) the accident sequences that could result from such hazards; (3) the consequence and likelihood of occurrence of each accident sequence; and (4) the items relied on for safety.
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| Unacceptable vulnerabilities mean deficiencies in the items relied on for safety or the measures used to assure the continuous availability and reliability of such items that need to be corrected to ensure an adequate level of protection as defined in 10 CFR 70.60(c). j Worker means an individual whose assigned duties in the course of employment involve exposure to radiation and/or radioactive material from licensed and unlicensed sources of radiation (i.e., an individual who is subject to an occupational dose as in 20 CFR 20.1003).
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| : 4. The undesignated center heading " EXEMPTIONS" is redesignated as "Subpart 8 -
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| Exemptions."
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| $$ 70.13a and 70.14 [ Redesignated)
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| : 5. Sections 70.13a and 70.14 are redesignated as @ 70.14 and 70.17, respectively.
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| : 6. Section 70.15 is added to read as follows:
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| 6 70.15 Nuclear reactors.
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| The regulations in Subpart H do not apply to nuclear reactors licensed under 10 CFR Part 50.
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| : 7. The undesignated center heading " GENERAL LICENSES" is redesignated as "Subpart C -
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| General Licenses."
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| : 8. The undesignated center heading " LICENSE APPLICATIONS"is redesignated as "Subpart D - Ucense Applications."
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| 9 70.22 [ amended) i 1
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| : 9. In 10 CFR 70.22, paragraph (f) is removed and paragraphs (g) through (n) are redesignated
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| {
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| es (f) through (m).
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| ATTACHMENT 2
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| 1 t
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| 1 1
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| l 70.23 [ amended]
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| i
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| : 10. In 10 CFR 70.23, paragraph (a)(8) is removed, paragraph (b)is removed and reserved, and paragraphs (a)(9) through (a)(12) are redesignated as (a)(8) through (a)(11), respectively.
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| : 11. The undesignated center heading " LICENSES" is redesignated as "Subpart E - Licenses."
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| : 12. The undesignated center heading " ACQUISITION, USE AND TRANSFER OF SPECIAL NUCLEAR MATERIAL, CREDITORS' RIGHTS," is redesignated as "Subpart F - Acquisition, Use, And Transfer Of Special Nuclear Material, Creditors' Rights." )
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| : 13. The undesignated center heading "SPECIAL NUCLEAR MATERIAL CONTROL RECORDS, REPORTS AND INSPECTIONS" is redesignated as "Subpart G - Special Nuclear j Material Control Records, Reports, And Inspections."
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| : 14. The undesignated center heading " MODIFICATION AND REVOCATION OF LICENSES"is rGdesignated as "Subpart 1 - Modification and Revocation of Licenses."
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| @$ 70.61 and 70.62 [ redesignated]
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| : 15. Sections 70.61 and 70.62 are redesignated as 9 70.81 and 70.82, respectively.
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| : 16. The undesignated center heading " ENFORCEMENT' is redesignated as "Subpart J -
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| Enforcement."
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| $$ 70.71 and 70.72 [ redesignated]
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| : 17. Sections 70.71 and 70.72 are redesignated as $970.91 and 70.92, respectively. ;
| |
| : 18. In Part 70, a new "SUBPART H" ($$ 70.60 - 70.74) is added to read as follows:
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| Subpart H - Additional Requirements for Certain Licensees Authorized To Possess a Critical Mass of Special Nuclear Material Sec.
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| 70.60 Safety performance requirements.
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| 70.62 Requirements for the performance of ISAs and the filing of ISA results and license applications. ,
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| ATTACHMENT 2
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| 70.64' Baseline design criteria for new facilities or new processes at existing facilities.
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| 70.65 Additional content of applications.
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| 70.66 Records.'
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| 70.68 . Additional requirements for approval of license application.
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| 70.72 Changes to facility structures, systems, equipment, components, and activities of personnel.
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| 70.73 Renewal oflicenses. 4 70.74 ' Additional reporting requirements.
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| ]
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| 670.60 Safety performance reauirements.
| |
| (a) Purpose. Each licensee engaged in enriched uranium processing, uranium fuel fabrication, uranium enrichment, enriched uranium hexafluoride conversion, plutonium processing, mixed-oxide fuel fabrication, scrap recovery, or any other activity that the Commission determines could significantly affect public health and safety, shall provide protection to its workers, the genera!
| |
| public, and the environtrent against radiological (including n.yclear criticality), chemical, and fire hazards that could result in the adverse consequences identified in paragraph (b) of this section.
| |
| Consideration must be given to radiological consequences from all causes (including those resulting from fires and hazardous chemicals), and those chemical and environmental consequences produced by radioactive materials 'h:t u!d :::et ' rem t6: precerr!n;; ef :p !:! nut!: r meter!:!.
| |
| (b) Conseauences of concern. Each licensee shall protect against the occurrence of the following high and intermediate adverse consequences that could result from accidents involving the handling, storage, or processing of licensed special nuclear material:
| |
| (1) Hiah consecuences.
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| (i) A nuclear criticality; (ii) Acute exposure of a worker to -
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| (A) A radiation dose of 1 Sv (100 rem) or greater total effective dose equivalent; or (B) Radioactive materials or Mhazardous chemicals produced from radioactive materials in concentrations exceeding AEGL-3 (.^;; nd!: ^.) or ERPG-3 (.^;; d!r B) criteria; or (iii) Acute exposure of a member of the public outside the controlled site boundary to:
| |
| (A) A radiation dose of 0.25 Sv (25 rem) or greater total effective dose equivalent; (B) An intake of 30 mg or greater of uranium in a soluble form; or (C) Radioactive materials or Mhazardous chemicals produced from radioactive materials in concentrations exceeding AEGL-2 (.^;; db ^) or ERPG-2 (.^;; nd!: S) criteria.
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| (2) Intermediate consecuences.
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| ATTACHMENT 2
| |
| | |
| l (i) Acute exposure of a worker to --
| |
| (A) A radiation dose between 0.25 Sv (25 rem) and 1 Sv (100 rem) total effective dose equivalent; or (B) Rs x.petive materials or Whazardous chemicals produced from radioactive materials in concentrations between AEGL-2 (.^ppendir ^.) or ERPG-2 (Appendix 9) criteria and AEGL-3 (Append!r ^.) or ERPG-3 (Ap; nd!r 9) criteria; or (ii) Acute exposure of a member of the public outside the controlled site boundary to -
| |
| (A) A radiation dose between 0.05 Sv (5 rem) and 0.25 Sv (25 rem) total effective dose equivalent; or (B) Radioactive materials or Whazardous chemicals produced from radioactive materials in concentrations between AEGL-1 (Ar~end!r ^) or ERPG-1 (Append!x 9) criteria and AEGL-2 )
| |
| (Appnd!r ^.) or ERPG-2 (Append!r 9) criteria; or (iii) Release of radioactive material to the envi onment outside the restricted area in concentrations that, if averaged over a period of 24 hours, exceed 5000 times the values specified in Table 2 of Appendix B to 10 CFR Part 20.
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| J (c) Graded level of orotection. Each licensee shall provide a level of protection that is !
| |
| commensurate with the severity of the consequences resulting from credible accidents and the likelihood of any external events (e.g., natural phenomena) assumed to initiate or propagate such accidents. This graded level must apply to the items relied on for safety, identified in paragraph (d)(2)(iv) of this section, and to the measures used to assure their continuous availability and reliability, identified in paragraph (d)(3) of this section. The application of a graded level of protection must assure that -
| |
| (1) The occurrence of any of the high consequences identified in paragraph (b)(1) of this section is highly unlikely; and (2) The occurrence of any of the intermediate consequences identified in paragraph (b)(2) of this section, is unlikely. 1 (d) Safety oroaram. Each licensee shall establish and maintain a safety program that provides reasonable assurance that the accident consequences identified in paragraph (b) of this section are Odequately protected against in accordance with paragraph (c).
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| (1) Each licensee shall compile and maintain a set of process safety information to enable the '
| |
| performance of an integrated safety analysis (ISA). This process safety information must include information pertaining to the hazards of the materials used or produced in the process, information pertaining to the technology of the process, and information pertaining to the equipment in the process.
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| ATTACHMENT 2
| |
| | |
| (2) Each licensee shall perform an ISA to identify -
| |
| : (i) All radiological and non-radiological hazards (e.g., chemical, fire, electrical, and mechanical);-
| |
| (ii) Potential accident sequences caused by process deviations or other events internal to the plant (e.g., fires, explosions, or chemical releases) and credible extemal events, including natural phenomena (e.g., hurricanes, floods, tomadoes, earthquakes, tsunami, and seiches), fires, explosions, or chemical releases occurring offsite; (iii) The consequence and likelihood of occurrence of each accident sequence identified pursuant to paragraph (d)(2)(ii) of this section; and (iv) Items relied on for safety (i.e., structures, systems, equipment, components, and activities of personnel), that are relied on to prevent or mitigate those accidents identified under paragraph (d)(2)(ii) of this section, that exceed the consequences of concem stated in paragraph (b) of this
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| - se: tion.
| |
| (3) To ensure the continuous availability and reliability of items relied on for safety identified under p'aragraph (d)(2)(iv) of this section, each licensee shall demonstrate that -
| |
| (i) Structures, systems, equipment, and components relied on for safety are designed, constructed, inspected, calibrated, tested, and maintained, as necessary, to ensure the continuous ability to perform their safety functions to satisfy paragraph (c) of this section. Items subject to this
| |
| )
| |
| requirement include but are not limited to: principal structures of the plant; passive barriers relied on for safety (e.g., piping, glove boxes, containers, tanks, columns, vessels); active systems, equipment, and components relied on for safety; sampling and measurement systems used to convey information about the safety of plant operations; instrumentation and control systems used to monitor and control the behavior of systems relied on for safety; and utility service systems relied on for safety.
| |
| - (ii) Personnel are trained, tested, and retested, as necessary, to ensure that they understand, recognize the importance of, and are qualified to perform their safety duties to satisfy paragraph (c) of this section; (iii) Procedures relied on for safety are developed, reviewed, approved, and distributed to ensure that personnel are able to perform their safety duties to satisfy paragraph (c) of this section.
| |
| (iv) Human-system interfaces are designed and implemented to ensure that personnel relied on for safety are able to perform their safety duties to satisfy paragraph (c) of this section.
| |
| (v) Configuration changes to' site, structures, process, systems, equipment, components, computer programs, personnel, procedures, and documentation are managed so that such
| |
| ' modifications are reviewed, documented, communicated, and implemented in a systematic, controlled ATTACHMENT 2 2
| |
| | |
| manner to satisfy paragraph (c) of this section.
| |
| (vi) All items relied on for safety identified under paragraph (d)(2)(iv) of this section and measures' established under paragraphs (d)(3)(i) through (d)(3)(v) of this section must meet quality standards that are commensurate with the importance of the safety functions performed. Management shall establish appropriate quality assurance policies and procedures to ensure that all items relied on for safety perform their safety functions and are continuously available and reliable.
| |
| (4) Each licensee shall conduct audits and assessments of its safety program to ensure that an adequate level of protection is maintained at the facility.
| |
| (5) Each licensee shall investigate abnormal events and take corrective action to min'imize the recurrence of these events.
| |
| (6) Each licensee shall establish records that will demonstrate that the requirements of paragraphs (d)(1), (d)(2), (d)(3), (d)(4), and (d)(5) of this section have been met. Each licensee shall maintain these records for the lifetime of the plant.
| |
| 670.62 Reauirements for the performance of ISAs and the filina of ISA results and license aoolications.
| |
| (a) Each applicant for a license under this subpart and each current licensee subject to this
| |
| . subpart shall perform an ISA as described in $70.60(d)(2).
| |
| (1) Each current licensee shall-(i) Wdhin 6 months of the effective date of this rule, submit, for NRC approval, a compliance plan that describes the ISA approach that will be used, the processes that will be analyzed, and the schedule for completing the analysis of each process; and (ii) Wdhin 4 years of the effective date of this rule, perform an ISA in accordance with the compliance' plan submitted under paragraph (a)(1)(i) of this section, correct any unacceptable vulnerabilities identified in the ISA, and submit the results of the ISA as part of the license application contents identified in $70.65 to NRC, for approval. Pending the correction of any unacceptable vulnerabilities identified in the ISA, the licensee shall implement appropriate compensatory measures ,
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| to ensure adequate protection. The process description in the ISA submittal must include information _
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| i that demonstrates the licensee's compliance with the design requirements for nuclear criticality monitoring and alarms in $70.24.
| |
| (2) Each applicant operating a facility that is newly subject to the Commission's authority shall j perform an ISA, correct any unacceptable vulnerabilities identified in the ISA, and submit the results of ;
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| the ISA as part of the license application contents identified in $970.22 and 70.65 to NRC, for approval. l The process description in the ISA submittal must include information that demonstrates the applicant's l ATTACHMENT 2 l
| |
| | |
| m 6-compliance with the d' esign requirements for nuclear criticality monitoring and alarms in 70.24.
| |
| (3) Each applicant for a license to operate a new facility or a new process at an existing facility shall-(i) Initially design the facility or process to protect against the occurrence of the adverse consequences identified in $70.60(b), meet the nuclear criticality monitoring and alarm requirements of
| |
| $70.24, and meet the baseline design criteria in $70.64; (ii) Perform a preliminary ISA and submit the results to NRC before construction of the facility or process. The results of the preliminary ISA must demonstrate an adequate level of protection, as .i defined in $70.60(c), against occurrence of the adverse consequences in $70.60(b). The preliminary ISA submittal shall include facility and process description and design information that demonstrates the applicant's incorporation of the nuclear criticality monitoring and alarm requirements in $70.24, and
| |
| ~ the baseline design criteria in $70.64. Any proposed relaxation in the application of the baseline design criteria, pursuant to $70.64(a), must be identified and justified in the preliminary ISA submittal; and (iii) Before beginning operations, update the preliminary ISA and correct any unacceptable vulnerabilities identified in the ISA. The updated ISA must be based on as-built conditions and must take into account the results of the preliminary ISA. Any inconsistencies between the results of the updated ISA and the preliminary ISA must be identified.
| |
| (A) For new facilities submit the results of the ISA, as part of the license application contents identified in $$70.22 and 70.65, to NRC for approval.
| |
| (B) For new processes submit the results of the ISA and any revisions of the approved license application as part of an application for amendment of the license under $70.34.
| |
| (b) If the decommissioning of a facility involves potentially hazardous activities such as chemical treatment of wastes, each licensee shall perform an ISA of the decommissioning process, correct any unacceptable vulnerabilities identified in the ISA, and submit the results to NRC for approval before beginning such decommissioning activities.
| |
| 670.64 Baseline desian criteria for new facilities or new orocesses at existina facilities.
| |
| - (a) Applicants shall address the following baseline design criteria in the design of new facilities or design of new processes at existing facilities, before performing the preliminary ISA, in accordance with 570.62(a)(3)(ii). Applicants shall address these baseline design criteria in establishing minimum requirements for all items in their process design and description, which is provided in the application for a license or license amendment. Licensees shall maintain the application of these criteria unless ATTACHMENT 2
| |
| | |
| r the preliminary ISA, submitted before construction, pursuant to $70.62(b)(3)(iii), demonstrates that a given item is not relied on for safety or does not require adherence to the specified criteria.
| |
| (1) Quality standards and records. The design must be established and implemented in cccordance with a quality assurance program, to provide adequate assurance that items relied on for safety will satisfactorily perform their safety functions. Appropriate records of these items must be maintained by or under the control of the licensee throughout the life of the facility.
| |
| (2) Natural phenomena hazards. The design must provide for adequate protection against natural phenomena with consideration of the most severe documented historical events for the site.
| |
| (3) Fire protection. The design must provide for adequate protection against fires and explosions:
| |
| (4) Environmental and dynamic effects. The design must provide for adequate protection from cnvironmental conditions and dynamic effects associated with normal operations, maintenance, testing, and postulated accidents that could lead to loss of safety functions.
| |
| (5) Chemical crotection. The design must provide for adequate protection against chemical hazards of radioactive materials and of hazardous chemicals produced from radioactive materials r9tM te the eter:;:,5:nt;, d pr::::9;; Of!! rred nuter rtrir!.
| |
| (6) Emeroency capability. The design must provide for emergency capability to maintain control of:
| |
| (i) Licensed material; (ii) Evacuation of personnel; and (iii) Onsite emergency facilities and services that facilitate the use of available offsite services.
| |
| (7) Utility services. The design must provide for continued operation of essential utility services, including reliable and timely emergency power to items relied on for safety.
| |
| (8) Inspection. testina. and maintenance. The design of items relied on for safety must provide for periodic inspection, testing, and maintenance, to ensure their continued function and readiness.
| |
| (g) Nuclear criticality control. The design must provide for nuclear criticality controlincluding adherence to the double-contingency principle.
| |
| (10) Instrumentation and controls. The design must provide for inclusion of instrumentation
| |
| : cnd control systems to monitor and control the behavior of items relied on for safety.
| |
| (b) Facility and system design and plant layout must be based on defense-in-depth practices.
| |
| Features must be incorporated that enhance safety by reducing challenges to items relied on for safety. Where practicable, passive systems and features must be selected over active systems and features, to increase overall system reliability.
| |
| ATTACHMENT 2
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| | |
| 670.65 Additional content of aoolications.
| |
| In addition to the contents required by 70.22, each application for a license to possess a critical mass of special nuclear material for use in the activities described in 70.60(a), must contain -
| |
| (a) A description of the applicant's site, structures, and the processes analyzed in the ISA; (b) A description of the applicant's safety program established under $70.60(d), including the results of the ISA and the measures established to ensure the continuous availability and reliability of items relied on for safety; and (c) For currently operating facilities, a description of operational events, within the past 10 years, that had a significant impact on the safety of the facility.
| |
| 670.66 Records.
| |
| The applicant or licensee shall establish and maintain onsite, readily available for Commission inspection, a system of legible, current, accurate, complete, and easily retrievable records to document application-related and license-related information required by applicable parts of this chapter, Commission action, license condition, and commitments by the applicant or licensee. Records must be retained for the period specified by the applicable parts of this chapter, Commission action, license condition, and commitments made by applicant or licensee. If a retention period is not otherwise specified, these records must be retained until the Commission terminates the license or determines that they are no longer required.
| |
| 670.68 - Additional reauirements for aooroval of license aoolication An application for a license to possess a critical mass of SNM w!:1 be approved if the Commission determines that the applicant has complied with the requirements of $70.'23 and $$70.60 through 70.66.
| |
| b 670.72 Chances to site. structures. systems. eouioment. components. and activities of personnel.
| |
| (a) Except for a new process, subject to the requirements of $70.62(a)(3), any change to mte, structures, systems, equipment, components, and activities of personnel must be evaluated by the licensee before the change, to determine whether the change increases the likelihood or j consequences of an accident at the facility. The evaluation must be based on the licensee's ISA l results, developed in accordance with $70.60(d)(2), and other safety program information, developed in accordance with $70.60(d)(3), which are part of the license application contents identified in 970.65.
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| ATTACHMENT 2
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| | |
| (b) A licensee may make a change to site, structures, systems, equipment, components, and activities of personnel, without prior Commission approval,if the change--
| |
| (1) Results in, at most, a minimal increase in the likelihood or consequences of an accident previously evaluated in the ISA; (2) Would not create the potential for an accident different from any previously evaluated in the ISA; and (3) is not inconsistent with NRC requirements and license conditions.
| |
| (c) For any change authorized under paragraph (b) of this section, the licensee shall submit revised pages to the license application, including any changes in the results of the ISA, to NRC within 60 days ofinitiation of the change.
| |
| (d) For any change that is not authorized under paragraph (b) of this section, the licensee shall file an application for an amendment of its license, as specified in $70.34, that authorizes the change.
| |
| As part of the applicatiol for the amendment, the 1.censee shall perform an ISA of the change and submit any revisions of the ISA and the license application t) NRC for approval. The licensee shall also provide, as required by Part 51 of this chapter, any necessary revisions to its environmental report.
| |
| (e) The licensee shall maintain records of changes to its facility carried out under paragraph (a) of this section. These records must include a written evaluation that provides the bases for the determination that the changes do not require prior Commission approval under paragraph (b) of this section. These records must be maintained until termination of the license.
| |
| 670.73 Renewal of licenses.
| |
| Applications for renewal of a license must be filed in accordance with SQ 2.109,70.21,70.22, 70.33,70.38, and 70.65. Information provided in applications, including the results of the ISA, must be current, complete, and accurate in all material respects. Information contained in previous applications, statements, or reports filed with the Commission under the license may be incorporated by reference, provided that these references are clear and specific.
| |
| 670.74 Additional reportino reauirements.
| |
| (a) Reports to NRC Operations Center.
| |
| (1) Each licensee shall report to the NRC Operations Center the events described in paragraphs I, il, and 111 of Appendix A_C to Part 70.
| |
| (2) Reports must be made by a knowledgeable licensee representative and by any method that ATTACHMENT 2
| |
| | |
| O will ensure compliance with the required time period (1,4, or 24 hours) for reporting.
| |
| (3) The information provided must include a description of the event and other related information as described in paragraph V of Appendix A C to Part 70.
| |
| (4) Followup information to the reports must be provided until all information required to be reported in paragraph (a)(3) of this section is complete.
| |
| (5) Duplicate reports to the Commission are not required for events when the reports are made in compliance with other parts of this chapter, provided that the reports comply with the requirements of this section conceming addressees, information content, and timeliness of filing.
| |
| (6) Each licensee shall provide reasonable assurance that reliable communication with the NRC Operations Center is available during each event.
| |
| - (b) Written reports.
| |
| (1) Each licensee shall provide a written report to NRC, of the events described in paragraph IV of Appendix A_C to Pari 70, within 30 days of discovery. The written report must contain the information described in paragraph VI of Appendix 1C to Part 70.
| |
| (2) Each licensee who makes a report required by paragraph (a) of this section shall submit a written foi,lowup report within 30 days of the initial report. The written report shall contain the information as described in paragraph VI of Appendix A_C to Part 70.
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| ATTACHMENT 2
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| . . . .O n . ,. ..k e.n. n.... e n . . .M e. ... I,j ATTACHMENT 2
| |
| | |
| r o
| |
| j 12 24. Appendix 6 G to Part 70 is added to read as follows:
| |
| Appendix A C to Part 70 - Reportable Safety Events As required by 10 CFR 70.74, licensees who are authorized to possess a critical mass of special nuclear material shall report the following safety events (see table A-1 of this appendix):
| |
| : 1. Events to be reported within 1 hour of discovery, followed by a written report within 30 days.
| |
| (a) An accident from the processing of licensed material that resulted in any of the following consequences:
| |
| (1) A nuclear criticality.
| |
| (2) Acute exposure of a worker to -
| |
| (i) A radiation dose of 1 Sv (100 rem) or greater total effective dose equivalent, or I
| |
| (ii) Radioactive materials or Whazardous chemicals produced from radioactive materials in concentrations exceeding AEGL-3 (App nd!r ^) or ERPG-3 (Append!r 9) criteria.
| |
| (3) Acute exposure of a member of the public outside the controlled site boundary to -
| |
| (i) A radiation dose of 0.25 Sv (25 rem) or greater total effective dose equivalent, (ii) An intake of 30 mg or greater of uranium in a soluble form, or (iii) Radioactive materials or Whazardous chemicals produced from radioactive materials in concentrations exceeding AEGL-2 (Append!x ^.) or ERPG-2 (Append!x 9) criteria.
| |
| II. Events to be reported within 4 hours of discovery, followed by a written report within 30 days.
| |
| (a) An accident from the processing of licensed material that resulted in any of the following consequences:
| |
| (1) Acute exposure of a worker to -
| |
| (i) A radiation dose between 0.25 Sv (25 rem) and 1 Sv (100 rem) total effective dose equivalent, or (ii) Radioactive materials or Whazardous chemicals produced from radioactive materials in concentrations between AEGL-2 (Append!r ^) or ERPG-2 (Append!r 9) criteria and AEGL-3 (.^;pnd!r A) or ERPG-3 (.^;pnd!x 9) criteria.
| |
| (2) Acute exposure of a member of the public outside the controlled site boundary to -
| |
| I (i) A radiation dose between 0.05 Sv (5 rem) and 0.25 Sv (25 rem) total effective dose equivalent, or (ii) Radioactive materials or Whazardous chemicals produced from radioactive materials in ATTACHMENT 2 1
| |
| 1
| |
| | |
| concentrations between AEGL-1 (Append!r ^) or ERPG-1 (Appen& S) criteria and AEGL-2 (Append!r A) or ERPG-2 (.^ ?per9 S) criteria.
| |
| (3) Release of radioactive material to the environment outside the restricted area in concentrations that, if averaged over a period of 24 hours, exceed 5000 times the values specified in Table 2 of Appendix B to 10 CFR Part 20.
| |
| (b) A deviation from safe operating conditions that has not been corrected within 4 hours and has the potential, as identified in the ISA, for causing an accident with one or more of the consequences specified in paragraph l(a) of this appendix.
| |
| (c) An external condition that poses a threat to the performance of items that are relied on for safety (e.g., site, structures, systems, equipment, components, or activities of personnel). These conditions would include natural phenomena (e.g., hurricanes, floods, tornados, earthquakes), fires, or chemical releases.
| |
| (d) A potentially unsafe condition that has not been corrected within 4 hours and that has not been identified or analyzed in the integrated safety analysis (ISA).
| |
| 111. Events to be reported within 24 hours of discovery, followed by a written report within 30 days.
| |
| (a) A deviation from safe operating conditions that was corrected within 4 hours and had the potential, as identified in the ISA, for causing an accident with one or more of the consequences specified in paragraph l(a) of this appendix.
| |
| (b) A deviation from safe operating conditions that has not been corrected within 24 hours and has the potential, as identified in the ISA', for causing an accident with one or more of the consequences specified in paragraph II(a) of this appendix. ,
| |
| (c) A potentially unsafe condition that was corrected within 4 hours and was not identified or analyzed in the ISA. l IV. Events to be reported in writing, to NRC, within 30 days of discovery.
| |
| (a) A deviation from safe operating conditions that was ccrrected within 24 hours and had the potential, as identified in the ISA, for causing an accident with one or more of the consequences specified in paragraph ll(a) of this appendix.
| |
| V. Licensee reports to the NRC Operations Center, as required by 10 CFR 70.74(a), shall include, to the extent that the information is applicable and available at the time the report is made, the following:
| |
| (a) Caller's name and position title.
| |
| (b) Date, time, and location of the event.
| |
| ATTACHMENT 2
| |
| | |
| 4 (c) Description of the event, including --
| |
| (1) Sequence of occurrences leading to the event, including degradation or failure ofitems j relied on for safety. l (2) Radiological or chemical hazards involved including isotopes, quantities, and chemical and .
| |
| physical form of any material released.
| |
| (3) Actual or potential health and safety consequences to the workers, the public, and the environment, including relevant chemical and radiation data for actual personnel exposures (e.g., level of radiation exposure, concentration of chemicals, and duration of exposure).
| |
| (4) ltems that are relied on to prevent or to mitigate the health and safety consequences, and whether the ability of those items to function has been affected by the event.
| |
| - (5) For events involving deviations from safe operating conditions, the process parameters that are deviant, the normal operating and safety limits on these parameters, and the current values of these parameters.
| |
| (d) Extemal conditions affecting the event.
| |
| (e) Additional actions taken by the licensee in response to the event.
| |
| (f) Status of the event (e.g., whether the event is on-going or was terminated).
| |
| (g) Current and planned site status, including any declared emergency class.
| |
| (h) Notifications related to the event that were made or are planned to any local, State, or other Federal agencies.
| |
| (i) Issue of a press release by the licensee related to the event that was made or is planned.
| |
| VI. Licensee written reports required by 10 CFR 70.74(b) shall consist of a completed NRC Form 366 and shall be forwarded to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001. Each written report must include the following information:
| |
| (1)_ Complete applicable information required by paragraph V of this appendix.
| |
| (2) Whether the event was identified in the ISA.
| |
| (3) Cause of the event, including all factors that contributed to the event.
| |
| (4) Corrective actions taken to prevent occurrence of similar or identical events in the future. l l
| |
| l ATTACHMENT 2 l
| |
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| | |
| Dated at Rockvil!e, Maryland, this day of .1998.
| |
| For the Nuclear Regulatory Commission.
| |
| John C. Hoyle, Secretary of the Commission.
| |
| .s i ATTACHMENT 2
| |
| | |
| 'L
| |
| +
| |
| DISCUSSION OF PROPOSED RULE'S MAJOR ELEMENTS Consequences of concem. An important element in the proposed rule is the identification of specific -
| |
| consequences against which licensees must provide adequate protection [10 CFR 70.60(b))._ These consequences, which are applicable to workers and members of the public, are categorized according titheir level of severity (high and intermediate). Because accidents at fuel cycle facilities could result
| |
| ' [ in human exposure to both radiological and certain chemical hazards, the proposed rule has adopted criteria that address both types of consequences., This approach satisfies the U.S. Nuclear Regulatory Commission's (NRC's) primary responsibility for radiation protection, in addition to its responsibility to j prgtect workers and the public from the chemical risk produced from radioactive materials hasards r"5;;':- 5:;xxx5;; f Mx:rd nur'::- rtrt!.
| |
| Grcdedlevelof Protection. To ensure an acceptable level of risk at facilities that possess a critical miss of special nuclear material, the proposed rule [10 CFR 70.60(c)] calls for licensees to provide a grcded level of protection against potential accidents. That level of protection must be sufficient to r: duce the likelihood of such accidents to levels commensurate with their consequences. _ Thus, cccording to the proposed rule, the occurrence of any high-consequence event should be " highly unlikely," while the occurrence of any intermediate-consequence event should be "unlikely." Although the rule does not define the terms " highly unlikely" and "unlikely," the draft Standard Review Plan
| |
| . provides criteria forjudging the likelihood of potential accidents. This guidance is based on a combination of qualitative and quantitative indicators, but does not require a probabilistic risk cssessment.
| |
| lnt: grated Safety Analysis (lSA). According to the proposed rule [10 CFR 70.60(d)], licensees must demonstrate, based on the performance of an ISA, their ability to provide an adequate level of protection against potential accidents. An ISA is a systematic analysis to identify plant and extemal hazards and their potential for initiating accident sequences; the potential accident sequences and their liktlihood and consequences; and the items (i.e., site, structures, systems, equipment, components, cnd activities of personnel) that are relied on for safety.
| |
| M:ssures to ensure continuous availability and reliability. Although the ISA plays a crttical role in identifying potential accidents and the items relied on for safety, the performance of an ISA will not, by itself, ensure adequate protection instead, as required by the proposed rule [10 CFR 70.60(d)], an
| |
| . effective management system is needed to cnsure that, when called upon, the items relied on for s"fety are in place and operating properly. Maintenance measures must be in place to ensure the continuous availability and reliability of all hardware relied on for safety. Training measures must be established to ensure that all personnel whose actions are relied on for safety are appropriately trained to perform their safety functions. Human-system interfaces and safety-related procedures must be developed and implemented to enable personnel relied on for safety to effectively carry out their duties.
| |
| Changes in the configuration of the facility need to be carefully controlled to ensure consistency among the facility design and operational requirements, the physical configuration, and the facility documentation. In addition, quality assurance measures need to be established to ensure that the itsms relied on for safety, and the measures used to ensure their continuous availability and reliability,
| |
| - tre of sufficient quality. Periodic audits and assessments of licensee safety programs must be performed to ensure that facility operations are conducted in compliance with NRC regulations and protect the public health and safety. When operational events occur, investigations of those events must be carried out to prevent their recurrence and to ensure that they do not lead to more serious consequences. Finally, to demonstrate compliance with NRC regulations, records that document safety program activities must be maintained for the life of the facility.
| |
| | |
| 6 Inclusion of safety bases in the application and changes to the safety bases. The performance of the ISA to identify the items relied on for safety and the measures established to ensure the continuous availability and reliability of such items are important elements in increasing confidence in the margin of safety. Nevertheless, without formal commitments to implement these items and measures, and to keep NRC informed of any changes in such commitments, the safety bases could become uncertain ov:r time. Thus, the proposed rule calls for the incorporation of licensee commitments to these items cnd measures in the license application. In addition, all changes in such commitments shall be submitted to NRC as part of a revised license application, including any changes in the ISA results (10 CFR 70.72). The rule does, however, allow for certain changes to be made, based on the results of tha ISA, without prior NRC approval, as long as such changes result in, at most, a minimal increase in the risk of accidents at the facility.
| |
| l ATTACHMENT 1 l 3
| |
| | |
| 8
| |
| '1F '
| |
| NUCl!AR [NERGY IN5111UTE l
| |
| I November 4,1998 Dr. Carl A. Paperiello, Director Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Two White Flint Center Washington, D.C. 20555-0001
| |
| | |
| ==REFERENCE:==
| |
| 10 CFR Part 70 Regulation of Chemical Hazards
| |
| | |
| ==Dear Dr. Paperiello:==
| |
| | |
| 1 At the September 29th NRC-NuclearIndustry Workshop on Part 70 Regulation you acknowledged that the language in the draft Part 70 revisions addressing regulation of hazardous chemicals required clarification. You requested the Nuclear Energy Institute (NEI)1 to propose corrections to this draft language.
| |
| Attachment 1 to this letter presents the changes that NEI recommends be incorporated to accurately reDect NRC's regulatory jurisdiction over hazardous chemicals. Attachment 2 provides background information and explanations for each recommended change. Attachment 3 is a red lined version of SECY-98-185 which incorporates NEI's recommended changes.
| |
| 1 i
| |
| ' NEI is the organization responsible for establishing unified nuclear industry policy on matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEI's members include all utilities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect / engineering fums, fuel fabrication facilities, materials licensees, and other organizations and indiv; duals involved in the nuclear energy industry.
| |
| i 1
| |
| l l
| |
| l l
| |
| 1776 4 57titi. NW $UITE 400 W ASHINGTON. DC 20006-3708 PHONE 202 739 8000 FAI 202 785 4019 4
| |
| | |
| - Dr. Carl A. Paperiello Nuclear Regulatory Commission November 4,1998 -
| |
| Page 2
| |
| . NEI is pleased to have had the opportunity to provide this input to the NRC towards clarifying the draft rule language. We look forward to continuing the dialogue on the Part 70 rulemaking and to addressing any questions which you or your staff may have on the industry's concerns and positions.
| |
| Sincerely, Marvin S.- Fertel cc: Chairman Shirley Ann Jackson Commissioner Edward McGaffigan, Jr.
| |
| Commissioner Nils J. Diaz '
| |
| Commissioner Jeffrey S. Merrifield Commissioner Greta Joy Dieus William D. Travers, Emeritus Director of Operations 2
| |
| e
| |
| | |
| ,e i l
| |
| , 1 ATTACHMENT 1 l
| |
| NUCLEAR ENERGY INSTITUTE (NEI)
| |
| RECOMMENDED LANGUAGE CHANGES TO PART 70 '
| |
| FOR REGULATION OF CHEMICAL HAZARDS l 4
| |
| -L Deficiencies in Draft Languare Proposed revisions to 10 CFR Part 70 will provide NRC regulatory jurisdiction over all " chemical hazards resulting from the processing oflicensed" radioactive material. The breadth of this jurisdiction exceeds that described in SECY-98-185 and in the 1988 NRC/ OSHA Memorandum of Understanding (MOU). Proposed i language in Part 70 can be construed to extend NRC regulation to any chemical hazard at a licensed fuel fabrication facility. NEI's principal objection to the draft Part 70 language is its failure to clearly separate the regulatory responsibilities of !
| |
| the NRC and OSHA as established in the MOU. As written, the draft rule will result in redundant, overlapping regulatory oversight that will not improve public or worker health and safety.
| |
| II. Pronosed Lanruare Modifications The draft language can be corrected primarily through clarification of the term l
| |
| " hazardous chemicals"in Part 70.60 and addition of a new definition for l
| |
| " hazardous chemicals produced from radioactive materials."
| |
| The MOU grants NRC the responsibility of protecting against " chemical risk produced by radioactive materials." Chemical risk results from hazards posed by either (i) the radioactive material itself, or (H) compo mds created by reaction of the radioactive materials with other substances. To clarify NRC jurisdiction over :
| |
| these two chemical hazards the following changes are recommended:
| |
| (i) the term " hazardous chemicals" should be replaced by " radioactive materials or hazardous chemicals produced from radioactive materials" This change would apply to $70.60(b)(1)(H)(B),
| |
| $70.60(b)(1)(iii)(c), $70.60(b)(2)(i)(B) and $70.60(b)(2)(H)(B) of the draft rule and throughout SECY 98-185.
| |
| (H) the majority of the chemicals listed in Appendices A (AEGLs) and B (ERPGs) are non-radioactive, are not used in SNM processing and are not capable of being produced from radioactive materials. The ;
| |
| proposed Rule revisions could be simplified by retaining references to the AEGL and ERPG standards, but deleting the actual tables of exposure limits which will be continually updated and modified.
| |
| I !
| |
| u
| |
| | |
| - (iii) the definition of" Hazardous Chemicals" includes the phrase
| |
| " ..cause significant damage to property or..." The NRC should not attempt to exercise jurisdiction over damage to property because such damage is not related to public health and safety. This clause should be deleted. The definition should, therefore, read as follows:
| |
| " Hazardous Chemicals means substances that are toxic, explosive, flammable, corrosive or reactive to the extent that they can endanger life if not adequately controlled."
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| (iv) inclusion of a definition for " Hazardous Chemicals Produced from Radioactive Materials"is required. This definition will reinforce the clear distinction between chemicals whose hazards are to be regulated by the NRC or by OSHA. Chemical hazards which could produce radiological consequences of concern are already regulated by the NRC. The new definition should build upon the existing definition of Hazardous Chemicals and should read:
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| " Hazardous Chemicals Produced from Radioactive Materials means Hazardous Chemicals either having radioactive material (s) as precursor compound (s) or formed through interaction with radioactive materials. They do not -
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| include chemicals merely added to, or used in, or recycled ;
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| from, the processing of special nuclear material (SNM)."
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| III. Concluding Remarks The foregoing suggested changes more accurately reflect the language and intent of the NRC/ OSHA MOU and more clearly demarcate the regulatory- (
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| responsibilities of the NRC and OSHA with respect to chemical safety. Adoption of these suggested changes will provide clarity of the areas over which the NRC has authority to regulate the chemical hazards of radioactive materials without engaging in the regulation of purely chemical hazards.
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| 2
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| .0 ,
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| ATTACHMENT 2 NUCLEAR ENERGY INSTITUTE (NEI)
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| BACKGROUND INFORMATION ON RECOMMENDED LANGUAGE CHANGES TO PART 70 FOR REGULATION OF CHEMICAL HAZARDS I. Introduction The U.S. Nuclear Regulatory Commission (NRC) issued SECY-98-185, " Proposed Rulemaking - Revised Requirements for the Domestic Licensing of Special Nuclear Material" on July 30,1998 to obtain Commission approval to publish a proposed rule amending 10 CFR Part 70. One proposed amendment addresses chemical safety standards. This amendment would extend NRC regulatory jurisdiction to all" chemical hazards resulting from the processing oflicensed" radioactive material, a much broader scope than was originally mandated in the rulemaking.
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| The chemical safety standard amendment (S70.60) should reflect the separation of regulatory jurisdiction of hazardous chemicals between the Occupational Health and Safety Administration (OSHA) and the NRC as detailed in the " Memorandum of Understanding Between the Nuclear Regulatory Commission and the Occupational Safety and Health Administration; Worker Protection at NRC-Licensed Facilities," 53 F33. Beg. 43950 (Oct. 31,1988) (NRC/ OSHA MOU). The draft rule, however, extends NRC regulatory oversight to eighty-eight chemicals, a majority of which are neither used in fuel cycle operations nor pose radiation hazards to facility workers or the public. The proposed $70.60 specifies concentrations of these chemicals, exposure to which constitutes a " consequence of concern," necessitates assessment in the licensee's Integrated Safety Analysis (ISA) and requires design and implementation of adequate safety measures.
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| Language in the draft rule can be construed to appreciably broaden the scope of NRC authority into areas reserved for OSHA regulatory oversight.
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| At the September 29th NRC Nuclear Industry Workshop on Part 70 Regulation the NRC concurred that the chemical safety rule amendment should conform to the NRC/ OSHA MOU and that the NRC should only regulate those hazards falling within its jurisdiction. The NRC requested NEI to offer suggestions to clanfy the language of the draft rule to ensure that the regulatory authority of the NRC and OSHA is clearly demarcated.
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| This Attachment 2 summarizes NEI's understanding of the scope of the NRC's authority to regulate chemical hazards at fuel-cycle facilities and explains the basis for each change in the draft Rule language presented in Attachment 1.
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| I a
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| II. NRC Authority to Regulate Chemical Hazards The NRC/ OSHA MOU identifies "four kinds of hazards that may be associated with NRC-licensed nuclear facilities":
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| : 1. " Radiation risk produced by radioactive materials;"
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| : 2. " Chemical risk produced by radioactive materials;"
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| : 3. " Plant conditions which affect the safety of radioactive I materials and thus present an increased radiation risk to workers. For example, these might produce a fire or an explosion, and thereby cause a release of radioactive materials or an unsafe reactor condition; and,"
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| l
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| : 4. " Plant conditions which result in an occupational risk, but do not affect the safety oflicensed radioactive materials. For example, there might be exposure to toxic non radioactive 1 materials and other industrial hazards in the workplace."
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| The NRC/ OSHA MOU states that the NRC shall have the responsibility for protecting against the first three hazards, while OSHA shall be responsible for protecting workers from the fourth hazard.
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| 1 NRC's sole responsibility for protecting the health and safety of the public from l the first hazard (radiation risk) is clear and unambiguous. J NRC's responsibility with respect to the second hazard is limited to a narrow class of chemical hazards. " Consequences of concern" which the NRC is responsible for regulating are chemical hazards which either: (1) result from the hazardous properties of the radioactive material itself, or (2) are created by the chemical reaction of the radioactive material and one or more other substances. For example, radioactive compounds UFs and UF4 exhibit toxic properties whose hazards are subject to NRC regulation. The NRC would also regulate generation of HF formed through interaction of UFs and moisture (humidity) in the conversion process to ensure that any exposures are kept below ERPG threshold concentrations. NRC regulatory oversight would not, however, extend to an HF recovery circuit once the HF off-gas scrubber condensates leave the conversion plant and are confirmed to contain only residual concentrations of radionuclides.
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| Protection of workers from HF chemical hazards at that point in the HF plant would, instead, revert to OSHA jurisdiction. Acids, ion exchange eluants and 2
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| | |
| solvent extraction organic chemicals used in UO2 scrap recovery would also be subject to NRC regulatory oversight only when actually used in uranium recovery processing and regeneration; regulation of chemical hazards from their bulk storage and handling (prior to use or after regeneration) would be an OSHA responsibility.
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| Determination of whether or not a particular chemical should be regulated by the NRC is often a process-specific issue. For example, nitric acid to be used in the UO2 scrap recovery process may be stored on site prior to use and would not be regulated by the NRC (so long as it could not affect the safety oflicensed material). However, once the acid is used in the dissolution process and combines with UO2, the radiological and chemical hazards of the mixture would be subject to NRC regulation. On the other hand, once the UO2 is stripped from the acid (via ion or solvent exchange) to leave the acid sufficiently free of radiological contamination to permit its handling as a non-radioactive material, the acid would only be subject to NRC regulation ifit was stored or used in a manner that could affect the safety oflicensed material. Off-gas scrubber condensates of gaseous and volatile radionuclides may be sabject to NRC oversight depending upon their composition and upon the radiation hazard they pose. Licensees will need to evaluate their own processes and chemicalinventories to determine the relevant controls that should apply to a particular chemical at various stages in their manufacturing processes. In summary, NRC regulatory authority over chemical hazards extends only to those chemicals stored at a licensed fuel facility that may affect the safety of SNM.
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| NRC's responsibility with respect to the third hazard entails protecting workers against increased radiation risk caused by plant conditions affecting the safety of radioactive material. Radiation releases could originate directly from fires or '
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| explosions or indirectly from releases of hazardous, non radioactive substances that might incapacitate an essential plant operator who would then be unable to j respond to an emergency and prevent a release of radiation. In all cases the '
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| consequence of concern to the NRC is the increased radiation risk to the worker -
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| not the occupational risk of the precursor fire, explosion or chemical release event. ;
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| It is this radiation risk that should be of concern to the NRC. The occupational risks associated with the precursor events are the responsibility of OSHA. ;
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| With respect to the fourth hazard, OSHA retains the responsibility for ensuring the occupational safety of workers including their protection from unacceptable exposures to toxic, non radioactive chemicals and other industrial hazards. In this case the consequence of concern is the (non-radiation) risk associated with a particular plant condition.
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| 3
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| III. Prior NRC Guidance on Chemical Hazards NUREG-1601 (Chemical Process Safety at Fuel Cycle Facilities) provides guidance for licensees to address chemical safety issues. In accordance with the MOU, it acknowledges that the NRC's responsibility is assurance of the safety oflicensed material and that its oversight of the risk posed by hazardous chemicals is limited I to their effect on licensed material and increased radiation risk to workers:
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| "Most NRC fuel cycle licensees possess materials that are chemically hazardous andfor pose some sort of non-radiological risk. Chemical and radiological risks have been known to compound one another, and in many cases, radioactive materials are also chemically hazardous. A chemical explosion in a fuel cycle facility could disperse radioactive material,just as the radiation environment could make it more difficult to respond to a hazardous chemical spill.
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| . . . The MOU between NRC and OSHA on chemical safety issues makes provision for the NRC to assume responsibility for the control of risks which may affect radioactive materials.
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| (NUREG 1601, i 2)
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| NUREG-1601 goes on to state that:
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| Thus the NRC does not regulate chemicals per se; rather, the NRC verifies that the interactions of chemicals with NRC-licensed nuclear materials and/or with equipment which processes, transports, or stores these licensed materials have been considered in the design of the equipment and facilities and in the operating and maintenanceprocedures."
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| (NUREG.1601, 62).
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| NUREG-1601 instructs licensees to conduct hazard audits to identify:
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| ' potential chemical hazards of radioactive materials and radiation hazards caused by chemicals . . . . "
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| (NUREG.1601, 62.2.1).
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| Although it advises licensees to identify non-radioactive chemicals, it does so in order to ensure the safety oflicensed material:
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| ' Chemicals which do not contain licensed materials should also be identified as potential chemical hazards because . . . release of such chemicals may affect the process by releasing the 4
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| licensed material or may affect the confinement of the licensed material in a favorable geometry". (NUREG-1601,52.3.1.1 emphasis added).
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| With regard to the effect of chemical hazards on the environment, the NRC emphasizes that a licensee need only identify those:
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| "[c]hemicals which can cause a release oflicensed material to the environment above NRC-prescribed limits..."
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| (NUREG-1601,12.3.1.2).
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| IV. NEI Recommended Changes to Part 70 The chemical safety standard amendment in 70.60 should be rewritten to clarify NRC's regulatory jurisdiction over chemical risks posed by:
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| (1) Special Nuclear Material (SNM)
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| (2) radioactive compounds (e.g. UF6)
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| (3) radioactive compounds produced frow Mioactive materials during the processing of SNM (e.g. HF)
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| NRC jurisdiction shall not extend to chemical risks originating from non-radioactive reagents stored at a fuel fabrication facility, either prior to their use or following their regeneration, and to non-radioactive by-product chemicals produced in the fuel fabrication operation.
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| To clarify the separate regulatory jurisdictions of the NRC and OSHA over chemical hazards, the following language changes are recommended:
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| (i) Definition of" Hazardous Chemicals" The MOU grants NRC the responsibility of protecting against " chemical risk produced by radioactive materials." Chemical risk results from hazards posed by (i) the radioactive material itself (e.g. UFe), (ii) compounds created by reaction of the radioactive materials with other substances (e.g. HF) or (iii) compounds contaminated by SNM or radioactive chemicals (e.g. HNO3, TCE, NaCO3). Chemicals produced by reaction with, or contaminated by, SNM (instances (ii) or (iii) above) are subject to NRC authority. Once, however, they are sufficiently free of radiological contamination to permit handing as a non-radioactive material they would no longer be subject to NRC oversight. All chemicals used in the facility, whether or not they are radioactive or hazardous, could fall under NRC jurisdication if they in any way impacted the safety oflicensed material.
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| 5
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| 4 -
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| To clarify NRC jurisdiction over these two chemical hazards the following changes are required, the term " hazardous chemicals" should be replaced by
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| " radioactive materials or hazardous chemicals produced from radioactive materials" This change would apply to $70.60(b)(1)(U)(B),
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| f 70.60(b)(1)(iii)(c), S70.60(b)(2)(i)(B) and f 70.60(b)(2)(ii)(B) and throughout SECY-98-185.
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| . S70.60(b)(1)(ii)(B):
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| " Radioactive materials or hazardous chemicals oroduced from radioactive materials in concentrations exceeding AEGL 3 (Appendix A) or ERPG-3 (Appendix B) criteria; or"
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| . [70.60(b)(1)(iii)(C):
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| " Radioactive materials or hazardous chemicals oroduced from radioactive materials in concentrations erceeding AEGL-2 (Appendix A) or ERPG 2 (Appendix B) criteria; or" e i70.60(b)(2)(i)(B):
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| " Radioactive materials or hazardous chemicals uroduced from radioactive materials in concentrations between AEGL-2 (Appendix A) or ERPG-2 (Appendix B) criteria and AEGL-3 (Appendix A) or ERPG-3 (Appendix B) criteria; or"
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| . f70.60(b)(2)(ii)(B):
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| "Badioactive materials or hazardous chemicals uroduced from radioactive materials in concentrations between AEGL-1 (Appendix A) or ERPG4 (Appendix B) criteria and AEGL-2 (Appendix A) or ERPG 2 (Appendix B) criteria; or" As a result of these proposed changes, a licensee would need to provide reasonable assurance that:
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| (i) concentrations of those radioactive materials listed in the AEGLs or ERPGs will not exceed the relevant consequences of concern, and that (ii) concentrations of hazardous chemicals listed in the AEGLs or ERPGs that may be produced through interactions with radioactive materials will not exceed the relevant consequences of concern.
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| (ii) Deletion ofAppendices A and B A majority of the chemicals listed in Appendices A (AEGLs) and B (ERPGs) are non radioactive, are not used in SNM processing and are not capable of being prcduced from radioactive materials. The proposed Rule revisions would be simplified by retaining references to the AEGL and ERPG standards, but deleting the actual tables of exposure limits which will be continually updated and modified. If the Appendices are not deleted, each should include the following statement:
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| 6
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| 1
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| \
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| I 1
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| 'The values listed should only be used as a consequence of concern if the chemical in question is radioactive, or is produced from radioactive material at a particular facility."
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| (iii) Reference to Property Damage The dennition of " Hazardous Chemicals" includes the phrase "...cause significant damage to property or..." The NRC should not attempt to exercise jurisdiction over damage to property because such damage is not related to public health and safety. This clause should be deleted. The definition shall, therefore, read as follows:
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| . " Hazardous Chemicals means substances that are toxic, I
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| explosive, flammable, corrosive or reactive to the extent that they can endanger life if not adequately controlled."
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| (iv) " Hazardous Chemicals Produced from Radioactive Materials" Definition of an additional term - Hazardous Chemicals Produced from Radioactive Materials - should be added to reinforce the clear distinction between chemicals whose hazards are to be regulated by the NRC and those to i be solely regulated by OSHA. Chemical hazards which could produce .
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| radiological consequences of concern are already regulated by the NRC. The new definition should build upon the existing definition of Hazardous Chemicals and should read:
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| " Hazardous Chemicals Produced from Radioactive Materiala means Hazardous Chemicals either having radioactive material (s) as precursor compound (s) or formed ;
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| through interaction with radioactive materials. They do not J include chemicals merely added to, or used in, or recycled j from, the processing of special nuclear material (SNM)." i i
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| Raf. I:\Sims\ Pert 70\h-=1 Comments OdLBX13-10-98) d 7
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| h
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| ~
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| : m. ,. .. .. .. .
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| *Ie J
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| ~
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| 1 I
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| [7590-01-P] ;
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| NUCLEAR REGULATORY COMMISSION l 10 CFR Part 70
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| - RIN 3150 - AF22 i Revised Requirements for the Domestic Licensing of Special Nuclear Material i i
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| l AGENCY: Nuclear Regulatory Commission.
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| ACTION: Proposed rule. ;
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| | |
| ==SUMMARY==
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| : The Nuclear Regulatory Commission (NRC) is proposing to amend its safety regulations ;
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| in the provisions goveming the domestic licensing of special nuclear material (SNM) for licensees i tuthorized to possess a critical mass of SNM, that are engaged ir' one of the following activities: ;
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| cnriched uranium processing; uranium fuel fabrication; uranium enrichment; enriched uranium l h;xafluoride conversion; plutonium processing; mixed-oxide fuel fabrication; scrap recovery; or any i other activity involving a critical mass of SNM that the Commission determines could significantly affect public health and safety. The proposed amendments would identify appropriate consequence criteria !
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| End the level of protection needed to prevent or mitigate accidents that exceed these criteria; require i cffected licensees to perform an integrated safety analysis (ISA) to identify potential accidents at the facility and the items relied on for safety; require the implementation of measures to ensure that the y it:ms relied on for safety are continuously available and reliable; require the inclusion of the safety bases, including the results of the ISA, in.the license application; and allow for licensees to make certain changes to their facilities without prior NRC approval.
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| DATES: The comment period expires (insert 75 days after publication in the Federal Register.)
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| Comments received after this date will be considered if it is practical to do so, but, the Commission is cble to ensure consideration only for comments received on or before this date.
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| - ADDRESSES: Submit comments to: The Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC,20555-0001, Attention: Rulemakings and Adjudications Staff.
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| Deliver comments to: 11555 Rockville Pike, Rockville, Maryland, between 7:30 a.m. and 4:15 p.m. on Federal workdays. l ATTACHMENT 1 l i
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| l 3-
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| w You may also provide comments via NRC's interactive rulemaking website through the NRC home page (http://www.nrc. gov). *From the home page, select "Rulemaking" from the tool bar. The
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| - int ractive rulemaking website can then be accessed by selecting "New Rulemaking Website." This site provides the ability to upload comments as files (any format), if your web browser supports that
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| - function.. For information about the interactive rulemaking website, contact Ms. Carol Gallagher, (301) 415-5905; e-mail cag@nrc. gov.
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| FOR FURTHER INFORMATION, CONTACT: Richard 1. Milstein, Office of Nuclear Material
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| - Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC, 20555-0001, tzlIphone (301) 415-8149; e-mail rim @nrc. gov.
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| SUPPLEMENTARY INFORMATION:
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| : 1. Background ll. ~ Description of P cposed Action
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| | |
| ===1. Background===
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| A near-criticality incident at a low enriched fuel fabrication facility in May of 1991 prompted NRC ta review its safety regulations for licensees that possess and process large quantities of SNM. [See
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| ' Proposed Method for Regulating Major Materiale Licensees" (U S Nuclear Regulatory Commission 1992) for additional details on the review.) As a result of this review, the Commission and the staff recognized the need for revision of its regulatory base for these licensees and, specifically, for those possessing a critical mass of SNM. Further, the NRC staff concluded that to increase confidence in the marDi n of safety at a facility possessing this type and amount of material, a licensee should perform an ISA. An ISA is a systematic I
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| I ATTACHMENT 2
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| 4 e' analysis that identifies:
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| (1) Plant and external hazards and their potential for initiating accident sequences; (2) The potential accident sequences, their likelihood, and consequences; and (3) The structures, systems, equipment, components, and activities of personnel relied on to prevent or mitigate potential accidents at a facility.
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| NRC held public meetings with the nuclear industry on this issue during May and l
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| November of 1995. Industry's position on the need for revision of NRC regulations in Part 70 '
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| was articulated to the Commission by the Nuclear Energy Institute (NEI) at a July 2,1996, m:eting, and in the subsequent filing of a Petition for Rulemaking (PRM-70-7) by NEl with ,
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| NRC in September 1996. NRC published in the Federal Register a notice of receipt of the I PRM and requested public comments on August 21,1996 (61 FR 60057). The PRM l r: quested that NRC amend Part 70 to:
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| (1) Add a definition for a uranium processing and fuel fabrication plant; (2) Require the performance of an ISA, or acceptable alternative, at uranium processing, fuel fabrication, and enrichment plants; and (3) include a requirement for backfit analysis, under certain circumstances, within Pcrt 70.
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| In SECY-97-137, dated June 30,1997, the NRC staff proposed a resolution to the NEl PRM and recommended that the Commission direct the staff to proceed with rulemaking. The <
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| NRC staffs recommended approach to rulemaking included the basic elements of the PRM, with some modification. In brief, NRC staff proposed to revise Part 70 to include the following m;jor elements:
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| (1) Performance of a formal ISA, which would form the basis for a licensee's safety program. This requirement would apply to all licensed facilities (except reactors and the gaseous diffusion plants regulated under 10 CFR Part 76) or activities, subject to NRC rcgulation, that are authorized to possess SNM in quantities sufficient to constitute a potential !
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| for nuclear criticality; (2) Establishment of criteria to identify the adverse consequences that licensees must protect against; i (3) inclusion of the safety bases in a license application (i.e., the identification of the ATTACHMENT 2 3 l
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| 1
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| . l
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| 1 1
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| I l
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| potential accidents, the items relied on for safety to prevent or mitigate these accidents, and th3 measures needed to ensure the continuous availability and reliability of these items). (This is in contrast to the PRM's approach, where the ISA results would not be included in the <
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| lic:nse application);
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| (4) Ability of licensees, based on the results of an ISA, to make certain changes I without NRC prior approval; and (5) Consideration by the Commission, after initial conduct and implementation of the l ISA by the licensees, of a qualitative backfitting mechanism to enhance regulatory stability, in a Staff Requirements Memorandum (SRM) dated August 22,1997, the Commission
| |
| ... approved the staff's proposal to revise Part 70" and directed the NRC staff to "... submit a draft proposed rule...by July 31,1998."
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| : 11. Description of Proposed Action i The Commission has decided to grant, in part, the NEl PRM by initiating this rul: making. Further, the proposed rule adopts the petitioners proposalin part and modifies the petitioner's proposal as indicated in the following discussion.
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| The Commission is proposing to modify Part 70 to provide increased confidence in the m rgin of safety at certain facilities authorized to process a critical mass of SNM. The Commission believes that this objective can be best accomplished through a risk-informed and performance-based regulatory approach that includes:
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| {
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| (1) The identification of appropriate consequence criteria and the level of protection n:cded to prevent or mitigate accidents that exceed such criteria; (2) The performance of an ISA to identify potential accidents at the facility and the items r: lied on for safety; (3) The implementation of measures to ensure that the items relied on for safety are continuously available and reliable; (4) The inclusion of the safety bases, including the ISA results, in the license cpplication; and (5) The allowance for licensees to make certain changes to their facilities without prior NRC approval.
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| The Commission's approach agrees in principle with the NEl petition. However, in ATTACP. MENT 2 J
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| i i
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| contrast to the petition'.s suggestion that the ISA requirement be limited to ". . uranium processing, fuel fabrication, and uranium enrichment plant licensees," the Commission would r quire the performance of an ISA for a broad range of Part 70 licensees that are authorized to possess a critical mass of SNM. The Part 70 licensees that would be affected include licensees engaged in one of the following activities: enriched uranium processing; uranium fuel fabrication; uranium enrichment; enriched uranium hexafluoride conversion; plutonium processing; mixed-oxide fuel fabrication; scrap recovery; or any other activity involving a critical mass of SNM that the Commission detemines could significantly affect public health and safety. The proposed rule would not apply to regulatees authorized to possess SNM under 10 CFR Pads 50,60,72, cnd 76.
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| Furthermore, the Commission is not currently proposing, as suggested in the NEl pstition, to include a backfit provision in Part 70. Based on the discussions at a public meeting held on May 28,1998, the purpose of the proposed backfit provision is to ensure that NRC staff does not impose safety controls that are not necessary to satisfy the performance requirements of Part 70, unless a quantitative cost-benefit analysis justifies this action. The Commission believes that once the safety bases, including the results of the ISA, are incorporated in the license application, and the NRC staff has gained sufficient experience with implementation of the ISA requirements, a qualitative backfit mechanism could be considered. Without a baseline determination of risk, as provided by the initial ISA process, it is not clear how a determination of incremental risk, as needed for a backfit analysis, would be accomplished. Furthermore, although NEl believes that a quantitative backfit approach is currently feasible, it would appear that a quantitative determination of incremental risk would require a Probabilistic Risk Assessment, to which the industry has been strongly opposed.
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| Given the differences of opinion on this subject, the Commission requests public comment on its intent to defer consideration of a qualitative backfit provision in Part 70.
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| The majority of the proposed modifications to Part 70 are found in a new subpart,
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| " Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Matpal," that consists of $$70.60 through 70.74. These proposed
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| - modifications to Part 70, discussed in detail below, are required to increase confidence in the margin of safety and are in general accordance with the approach approved by the Commission in its August 22,1997, SRM. However, the Commission has decided that the new requirements should not apply to all licensees authorized to possess a critical mass of ATTACHMENT 2
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| 4 SNid. Instead, the Commission has identified a subset of these licensees that, based on the r;latively high level of risk associated with operotions at these facilities, should be subject to tha new requirements. This change would excluile certain facilities (e.g., those authorized only to store SNM or use SNM in sealed form for research and educational purposes) from the new requirements, because of the relatively low sevel of risk at these facilities. This issue is further addressed in the discussion of 970 62.
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| Section 70.4. " Definitions."
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| The following fourteen definitions would be added to this section to provide a clear understanding of the meaning of the new subpart H, " Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Material:" Acute exposure, Acute expos"re guideline levels, Controlled site boundary, Critical mass of SNM, Deviation from safe operating conditions, Double contingency, Emergency response planning guidelines, Hazardous chemicals, Hazardous chemicals produced from radioactive materials, Integrated safety analysis, items relied on for safety, New process, Results of the ISA, Unacceptable vulnerabilities, and Worker.
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| ATTACHMENT 2
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| Section 70.15, " Nuclear reactors."
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| ' A new section would be added to subpart B, " Exemptions," that exempts nuclear reactors licensed under Part 50 from the new subpart H, " Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Material."
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| ' Section 70.22. " Contents of aoolications."
| |
| Paragraph (f) would be removed. Paragraph (f) currently requires that, for plutonium processing and fuel fabrication facilities, certain additional safety-related information be submitted with an application. The new subpart H, " Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Material," would contain requirements for the submittal of
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| 'information called for in paragraph (f) and is sufficient to allow the Commission to make a d; termination of adequacy.
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| Section 70.23. "Reauirements for the aooroval of apolications.'
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| Paragraphs (a)(8), and (b) would be removed. These paragraphs currently require that the Commission, to approve an application, determine that the construction of a plutonium processing and fabrication facility meet certain conditions. These conditions would be covered in the new subpart H,
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| " Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Material."
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| Section 70.60. " Safety performance reauirements."
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| These requirements would establish the purpose of the new requirements, identify the potential adverse consequences that need to be protected against, establish the level of protection that is needed to ensure that the consequences of concern do not occur, and identify the safety program tiements that allow licensees to demonstrate their ability to provide an adequate level of protection.
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| Section 70.60(at " Purpose."
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| This paragraph would address the following questions: Why are the new requirements needed? What hazards need to be considered? Who are the intended beneficiaries? In general, the new requirements are intended to ensure that workers', the general public, and the environment are 8-A worker, in the context of this rulemaking, is defined as an individual whose assigned duties in the course of
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| = 6 mployment involve exposure to radiauon and/or radioactive material from licensed and unlicensed sources of radiation (i.e.,
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| an individual who is subject to an occupatonal dose as in 10 CFR 20.1003).
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| ATTACHMENT 2
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| | |
| protected from radiological and certain chemical hazards associated with plant operations. All hazards, including fire, chemical, electrical, industrial, etc., that can potentially affect radiological safety, must be considered and addressed by licensees. In addition, chemical h:::rd: risks produced by radioactive materials th:t rece!!' rem th: prece: !ng of !!=n::d ne !: r m:teri:! must also be considered.
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| The question of NRC's authority to regulate chemical h:: rde risks produced by radioactive materials at its fuel cycle facilities was raised after an accident in 1986 at a Part 40 licensed facility, in which a cylinder of uranium hexafluoride ruptured and killed a worker. The cause of the worker's death w!s the inhalation of hydrogen fluoride gas, which was produced from the chemical reaction of uranium h:xafluoride and water (humidity in air). As a result of that incident, NRC and the Occupational Safety cnd Health Administration (OSHA) established a memorandum of understanding (MOU) (1988) that id;ntified the respective responsibilities of both agencies for the regulation of chemical hazards at nuclear facilities. The MOU identified the following four areas of responsibility. The NRC has r;sponsibility for the first three areas, whereas OSHA has respansibility for the fourth area:
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| (1) Radiation risk produced by radioactive materials; (2) Chemical risk produced by radioactive materials; (3) Plant conditions which that affect the safety of radioactive materials and. thus. present an increased radiation nsk to workers; and (4) Plant conditions which that result in an occupational risk, but do not affect the safety of licensed radioactive materials.
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| The purpose of the " Safety Performance Requirements," as defined in $70.60(a), is consistent with the NRC/ OSHA MOU.
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| Section 70.60(b). "Consecuences of concem."
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| The NRC is responsible for ensuring that workers and the general public are protected from the h:zards involved in the handling, processing, and storage of SNM. All hazards (including fire and chemical) that could result in radiological consequences are a subject of NRC concem. In addition, the au chemical hasards risks produced from radioactive materials rece!!Mg frem the preter !ng Of
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| !!=r d SM'? thet Or!d dired!y :" d ::: d:r er m -6 r Of tS: pub!!: are also a matter of NRC concem. Thus, NRC regulations need to address both radiological hazards; the chemical consequences of radioactive materials and hazardous chemicals produced from radioactive materials.
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| The following discussion provides information, on the consequences of human exposure to radiation cnd haaerdous hazards ehemisals produced by radioactive materials, that is relevant to the choice of ATTACHMENT 2
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| appropriate consequence criteria. The actual choice of these criteria is discussed in
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| $$70.60(b)(1)(ii)(A) and (B); 70.60(b)(1)(iii)(A) and (C); and 70.60(b)(2)(i)(A) and (B).
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| Radioloaical Conseauences. In the past, the regulation of licensees authorized to possess SNM, under 10 CFR Parts 70 and 20, has concentrated on radiation protection for persons involved in nuclear activities conducted under normal operations. The proposed amendments to Part 70 would explicitly address the potential exposure of workers or members of the public to radiation as a result of tecidents. Because accidents are unanticipated events that usually occur over a relatively short period of time, a regulation that seeks to assure adequate protection of workers and members of the public must limit the risk of such accidents.- This can be accomplished by identifying appropriate consequence criteria and by limiting the likelihood of occurrence of the identified consequences. In l sIlecting the radiological consequence criteria for use in the proposed rule, the Commission has eximined the radiological criteria and design basis accident scenarios used in existing NRC !
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| rrgulations to ensure that the proposed consequence cr;teria era consistent with criteria used in other Commission rules. 1 Chemical Conseauences. The processing of SNM may involve the use or production of hazardous chemicals. For example, low enriched uranium fuel fabrication facilities convert uranium hexafluoride to uranium oxide by reaction with water (hydrolysis) to form uranyl fluoride and hydrogen fluoride. Uranyl fluoride, in addition to being radioactive, is a toxic uranium compound that can cause d: mage to the kidney. - Hydrogen fluoride is highly toxic and poses a hazard to both workers and the ,
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| g2neral public. T; '- -f:ur 2 c'::':, !9 fM;; n n!:, "d: 09, :nd rr!$ed: : !d, ::: 9:
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| The effort to limit exposure of workers and the general public to hananlows chemicals hazards produced by radioactive materials is based on two concems:
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| ccute exposures that could result from accidental releases, and chronic exposures (i.e., multiple and repeated exposures occurring over a long period of time - days, months, or years), resulting from ral:ases during normal operations.
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| Chemical consequence criteria corresponding to anticipated adverse health effects to humans from acute exposures (i.e., a single exposure or multiple exposures occurring within a short time - 24 hours or less) have been developed, or are under development, by a number of organizations. Of particular interest, the National Advisory Committee for Acute Guideline Levels for Hazardous Substances is developing Acute Exposure Guideline Limits (AEGLs) that will eventually cover tpproximately 400 industrial chemicals and pesticides. The committee, which works under the ATTACHMENT 2
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| r I
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| cuspices of the U.S. Environmental Protection Agency (EPA) and the National Academy of Sciences
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| - (NAS), has identified a priority list of approximately 85 chemicals. Consequence criteria for 12 of these h:ve currently been developed and criteria for. approximately 30 additional chemicals per year are sxpected.
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| Another set of chemical consequence criteria, the Emergency Response Planning Guidelines (ERPGs), has been developed by the American Industrial Hygiene Association (AlHA) to provide Estimates of concentration ranges where defined adverse health effects might be observed because of short exposures to hazardous chemicals. ' ERPG criteria are widely used by those involved in cssessing or responding to the release of hazardous chemical including "... community emergency pl:nners and response specialists, air dispersion modelers, industrial process safety engineers, implementers of environmental regulations such as the Superfund Amendment and Reauthorization
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| . Act, industrial hygienists, and toxicologists, transportation safety engineers, fire protection specialists, cnd govemment agencies...." (DOE Risk Manaaement Quarte IV.1997). Despite their general Ecceptance, there are currently only approximately 80 ERP 3 criteria available, and some chemicals of importance (e.g., nitric acid) are not covered.
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| Federal regulations and intemal U.S. Department of Energy (DOE) guidance require the use of
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| ' ERPGs for emergency planning. Recognizing that ERPGs exist for a limited number of chemicals, DOE's Subcommittee on Consequence Assessment and Protective Actions developed Temporary Emergency Exposure Limits (TEELs) so that DOE facilities could perform complete hazard analysis and consequence assessments, even for chemicals lacking ERPGs. TEELs are not equivalent to ERPGs, but are approximations to ERPGs. They exist only until an ERPG is developed for a chemical.
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| As of July 1997,400 TEELs had been developed according to a methodology published in the American Industrial Hvaiene Joumal (1995). That methodology is not based directly on toxicological studies of the chemicals involved, but on a derived relationship between attemative exposure-limit parameters and the existing ERPG criteria. The use of the methodology results in a significant underestimation of the TEEL-22 level (0.6 mg/m') for soluble uranium and would be inconsistent with the criterion on soluble uranium intake (i.e.,30 mg) proposed in this rule.
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| A fourth set of chemical consequence criteria that was considered potentially applicable for ccute exposure to hazardous chemicals is the immediately Dangerous to Life and Health (IDLH) criteria established by the National Institute for Occupational Safety and Health (NIOSH). However, 8
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| TEEL-2 is defined as the maximum airbome concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour without expenencing or developing irreversible or other health effects or symptoms which could impair an individuars ability to take protective schon ATTACHMENT 2 l
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| l 1
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| p.
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| according to NIOSH, the IDLH criteria are defined ". . only for the purpose of respirator selection." In addition, unlike the previously mentioned sets of criteria, there is only one IDLH level that has been d; fined. This would not facilitate the definition of multiple c:nsequence levels for workers and the public, as intended in the proposed rule.
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| For chronic exposures of workers to hazardous chemicals during normal and off-normal operations, the permissible exposure limits (PELs) established by OSHA in 29 CFR 1910 are applicable. However, these limits are not relevant for. acute exposures to hazardous chemicals produced from radioactive materials.
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| Given the status of these various sets of consequence criteria, the Commission has chosen AEGLs and ERPGs, in that order, as criteria to be used for acute short-term exposure to radioactive materials that may themselves pose a chemical hazard (e o uranium hexafluoride in certain situations) or hazardous chemicals produced from radioactive materials. If a given chemical has an AEGL . associated with it, that criterion should be used. If not, the ERPG criterion, if available, should be used. A?;rnfr ^. centr! .: t'.: cer"r''r .^5GL rr'_er, 0 7d .^;; rt 9 centr!n: t'.: Ovr"r''r if both AEGLs and ERPGs are available for a particular chsmical, only the AEGL values will be presented. Although the TEELs cover a wide range of cdditional hazardous chemicals, the Commission has decided not to require their use at this time, because the methodology used to derive these values is not based on the toxicology of the chemicals involved and may, at least in certain cases, underestimate the limits. However, the use of the TEELs miy be justified on a case-by-case basis in the absence of other applicable standards.
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| As a result of further study, new AEGL or ERPG values are expected to be established by the hauing organizations (EPA for AEGLs; AlHA for ERPGs). The Commission does not propose to sng ge in full, formal rulemaking with respect to these future changes, but will incorporate them inthe
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| ::fE:d :;;:nf!::: 5 *n ! t.r. by issuing an immediately effective final rule. The Commission beli;ves that these purely technical changes or additions do not require comment and are, in addition, subject to the categorical exclusion in 10 CFR 51.22(c)(2).
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| General Anoroach The consequences of concem, identified in $$70.60(b)(1) and (b)(2), describe those consequences that licensees must protect against'. The level of protection to be provided is discussed 8
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| The proposed rule does not address chemical and radiological consequences to workers and members of the public resulting from routine operations. These consequences' are covered in other regulations (i.e.,10 CFR Part 20 and 29 CFR .
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| Pat 1910).
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| ATTACHMENT 2 )
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| 1 I
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| in $70.60(c) and depends on the severity of the consequences. The goalis to ensure an acceptable I;v;l of risk by limiting the likelihood of occurrence of the identified consequences. The consequences id:ntified in 70.60(b)(1) of the proposed rule are considered to be high consequences and include the occurrence of a nr ear criticality, and accidental exposure of a worker or member of the public to high 1;v;ls of radiation, radioactive materials or hazardous chemicals produced from radioactive materials.
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| Tha consequences identified in 970.60(b)(2) are considered to be intermediate consequences and include accidental exposure of a worker or member of the public to moderate levels of radiation, radioactive materials. or hazardous chemicals produced from radioactive materials, and significant r; leases of radioactive material to the environment. The proposed consequence criteria that are applicable to a member of the public are more restrictive than those that are applicable to a worker. )
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| Also, within each category (worker and public), NRC recognizes that the propo: 3d radiological criteria I cra more restrictive (in terms of acute health effects) than the chemical criteria for a given level of stverity (high or intermediate) and that this is consistent with current r:gulatory practice.
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| In some cases, a qualitative description of the conseque.nce is used (e.g., a nuclear criticality);
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| )
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| in other cases, a numerical criterion is used. For cases where numerical criteria have been used, NRC has based the criteria on values that have been developed previously by NRC or other government cg:ncies or professional societies. Table 1 illustrates the radiological and chemical consequence crit:ria used in the proposed rule, it should be noted that only those AEGLs or ERPGs associated with radioactive materials (e o 30 malm for uranium hexaflounde (ERPG-3)) are included in these consecuence criteria.
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| TABLE 1 Radiological and Chemical Consequence Criteria Worker Public CONSEQUENCE Radiological Chemical Radiological Chemical High > 1 Sv (100 rem) > AEGL-3 > 0.25 Sv (25 rem) > AEGL-2 (ERPG-3) (ERPG-2)
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| Intermediate < 1 Sv (100 rem) < AEGL-3 (ERPG- < 0.25 Sv (25 rem) < AEGL-2
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| : 3) (ERPG-2)
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| > 0.25 Sv (25 > 0.05 Sv (5 rem) i rem) > AEGL-2 (ERPG- > AEGL-1 l
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| : 2) (ERPG-1)
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| Section 70.60(b)(1) This paragraph defines "high consequences."
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| ATTACHMENT 2
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| Certain events that could occur at licensees' facilities are considered high-consequence events.
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| They include the occurrence of an inadvertent nuclear criticality, the exposure of a worker or member of the public to level's of radiation at which clinically observable biological damage could cccur, or concentrations of radioactive materials or hazardous chemicals produced from radioactive materials at v hich death or life threatening injury could occur due to their toxic. explosive.
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| flammable. corrosive, or reactive properties.
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| Section 70.60(b)(1)(i). This paragraph deals with a nuclear criticality.
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| The occurrence of an inadvertent nuclear criticality is considered to be a high-consequence cv nt. Although detecting and mitigating the consequences of a nuclear criticality are important obj:ctives (see 10 CFR 70.63), the prevention of a nuclear criticality is a primary NRC objective.
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| Section 70.60(b)(1)(ii)(A). This paragraph deals with an acbe exposure of a worker to a radiation dose of 1 Sv (100 rem) or greater total effective dose equivalent (TEDE).
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| An acute exposure of a worker to a rads + ion dose of 1 Sv (100 rem) or greater TEDE is considered to be a high-consequence event. According to the National Council on Radiation Protection and Measurements (NCRP,1971), life saving actions -including the " .. search for and rsmoval of injured persons, or entry to prevent conditions that would probably injure numbers of people" - should be undertaken only when the "... planned dose to the whole body shall not exceed 100 rems." This is consistent with a later NCRP position (NCRP,1987) on emergency occupational exposures, that states "...when the exposure may approach or exceed 1 Gy (100 rad) of low-LET
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| [ linear energy transfer) radiation (or an equivalent high-LET exposure) to a large portion of the body, in a short time, the worker needs to understand not only the potential for acute effects but h'e or she should also have an appreciation of the substantialincrease in his or her lifetime risk of cancer." The use of the 1-Sv (100-rem) criterion is not intended to imply that 1 Sv (100 rem) constitutes an tcceptable criterion for an emergenc/ dose to a worker. Rather, this dose value has been proposed in this section as a reference va!ue, which should be used by licensees to determine the level of protection (i.e., items relied on for safety, and measures to assure their continuous availability and rsliability) needed to ensure an acceptably low level of risk to workers.
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| Section 70.60(b)(1)(ii)(B). This paragraph deals with an acute exposure of a worker to radioactive materials or hazardous chemicals produced from radioactive materials in concentrations ATTACHMENT 2
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| exceeding AEGL-3 or ERPG-3 limits.
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| An acute exposure of a worker to, radioactive materials or hazardous chemicals produced from radioactive materials at concentrations that could cause death or life-threatening injuries from causes other than radiation is considered a high-consequence event. Two existing criteria, AEGL-3' and ERPG-3, can be used to define such concentration levels. AEGL-3 is defined as "The airbome concentration (expressed in ppm or mg/m*) of a substance at or above which it is predicted that the general population, including susceptible, but excluding hypersusceptible, individuals, could experience lif3-threatening effects or death." ERPG-3 is defined as "The maximum airborne concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour without experiencing or -
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| d;veloping life-threatening health effects." If, for a particular chemical, the AEGL-3 value is available, it should be used.: Otherwise, the ERPG-3 value should be used. If there is no AEGL or ERPG value cvrilable, then the applicant should adopt a criterion that is comparable in severity to those that have been established for other chemicals.
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| Section 70.60(b)(1)(iii)(A). This paragraph deals with an acute exposure of a member of the public to a radiation dose of 0.25 Sv (25 rem) or greater TEDE.
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| The exposure of a member of the public to a radiation dose of 0.25 Sv (25 rem) TEDE is considered a high-consequence event. This is based on the criterion established in 10 CFR 100.11, " Determination of exclusion area, low population zone, and population center distrnce," and 10 CFR 50.34, " Contents of applications; technical information," where a whole-body dose of 0.25 Sv (25 rem)is used to determine the dimensions of the exclusion area and low population zone required for siting nuclear power reactors.
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| t Section 70.60(b)(1)(iii)(B). This paragraph deals with an intake of 30 mg or greater of uranium in a soluble form by a member of the public.
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| The intake of 30 mg of soluble uranium by a member of the public is considered a high-consequence event. This choice, which is based on a review of the available literature [ Pacific Northwest Laboratories (PNL),1994], is consistent with the selection of 30 mg of uranium as a criterion that was discussed during the Part 76 rulemaking, " Certification of Gaseous Diffusion Plants." in
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| _ particular, the final rule that established Part 76 (59 FR 48944; September 23,1994) stated that "The Three levels of consequences are defined for each chemical (AEGL-1, AEGL-2, and AEGL-3) for four different exposure times: 30 minutes; 1 hour; 4 hours; and 8 hours. The AEGL value for a 1-hour exposure is chosen for consistency with the
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| ' definition of ERPG.
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| ATTACHMENT 2
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| NRC will consider whether the potential consequences of a reasonable spectrum of postulated cccident scenarios exceed... uranium intakes of 30 milligrams... " The final rule also stated that "The Commission's intended use of chemical toxicity considerations in Part 76 is consistent with its practice elsewhere (e.g.,10 CFR 20.1201(e)), and prevents any potential regulatory gap in public protection cgainst toxic effects of soluble uranium."
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| Section 70.60(b)(1)(iii)(C). This paragraph deals with an acute exposure of a member of the public to radioactive materials or hazardous chemicals produced from radioactive materials in concentrations exceeding AEGL-2 or ERPG-2 criteria.
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| An acute exposure of a member of the public to radioactive materials or hazardous chemicals produced from radioactive matenals at concentrations that could cause irreversible health effects is considered a high-consequence event. Two existing criteria, AEGL-2 and ERPG-2, can be used to d; fine such concentration levels.
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| AEGL-2 is defined as "The airbome concentration (expressed in ppm or mg/m 8) of a substance at or cbove which it is predicted that the general population, including susceptible, but excluding hypersusceptible, individuals, could experience irreversible or other serious, long-lasting effects or impaired ability to escape." ERPG-2 is defined as "The maximum airborne concentration below which it is believed that nearly allindividuals could be exposed for up to 1 hour without experiencing or dnveloping irreversible or other health effects or symptoms that could impair an individuars ability to tike protective action." If, for a particular chemical, the AEGL-2 value is available, it should be used.
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| Otherwise the ERPG-2 value should be used. If there is no AEGL or ERPG value available, then the cpplicant should adopt a criterion that is comparable in severity to those that have been established for other chemicals.
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| I Section 70.60(b)(2)(i)(A). This paragraph deals with an acute exposure of a worker to a rrdiation dose of between 0.25 Sv (25 rem) and 1 Sv (100 rem) TEDE.
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| The exposure of a worker to a radiation dose between 0.25 Sv (25 rem) and 1 Sv (100 rem)
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| TEDE is considered an intermediate-consequence event. The basis for this choice is the use of 0.25 Sv (25 rem) as an exposure criterion in existing NRC regulations. For example, in 10 CFR 20.2202,
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| ' Notification of incidents," immediate notification is required of a licensee if an individual receives " . a t:tal effective dose equivalent of 0.25 Sv (25 rem) or more." Also, in 10 CFR 20.1206, " Planned special exposures," a licensee may authorize an adult worker to receive a dose in excess of normal occupational exposure limits if a dose of this magnitude does not exceed 5 times the annual dose limits ATTACHMENT 2 1
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| l l
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| j
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| [i.e.,0.25 Sv (25 rem)] during an individual's lifetime. In addition, the EPA's Protective Action Guides (U.G. Environmental Protection Agency,1992) and NRC's regulatory guidance (Regulatory Guide 8.29, 1996) identify 0.25-Sv (25-rem) as the whole-body dose limit to workers for life-saving actions and protection of large populations. NCRP has also stated that a TEDE of 0.25 Sv (25 rem) corresponds to th3 once-in-a-lifetime accidental or emergency dose for workers. However, its use is not intended to imply that 0.25 SV (25 rem) constitutes an acceptable criterion for an emergency dose to a worker.
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| R*ther, this dose value has been proposed in this section as a reference value, which should be used by licensees to determine the level of protection (i.e., items relied on for safety, and measures to Essure their continuous availability and reliability) needed to ensure an acceptably low level of risk to I
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| workers.
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| Section 70.60(b)(2)(i)(B1 This paragraph deals with an acute exposure of a worker to radioa:tive materials or hazrrdous chemicals oroduced ' cm ratoactive materials in concentrations between AEGL-2 (ERPG-2) and AEGL-3 (ERPG-3) criteria.
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| An acute exposure of a worker to radioactive matenals or hazardous chemicals oroducea from radioactive materials at concentrations that could cause irreversible health effects (but below concentrations that could cause death or life-threatening effects) is considered an intermediate-consequence event. Two existing standards, AEGL-2 and ERPG-2, can be used to define the concentration level for irreversible health effects (see definitions in $70.60(b)(1)(iii)(C), above). Two cdditional standards, AEGL-3 and ERPG-3, can be used to define the concentration level for death or lifs-threatening effects (see definitions in 970.60(b)(1)(ii)(B), above] . If, for a particular chemical, the AEGL values are available, they should be used. Otherwise the ERPG values should be used. If there tra no AEGL or ERPG values available, then the applicant should adopt criteria that are comparable in s3 verity to those that have been established for other chemicals.
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| Section 70.60(b)(2)(ii)(A). This paragraph deals with an acute exposure of a member of the public to a radiation dose between 0.05 Sv (5 rem) and 0.25 Sv (25 rem) TEDE.
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| The exposure of a member of the public to a radiation dose between 0.05 Sv (5 rem) and 0.25 Sv (25 rem) is considered an intermediate-consequence event. NRC has used a 0.05-Sv (5-rem) cxposure criterion in a number of its existing regulations. For example,10 CFR 72.106, " Controlled cr;a of an ISFSI or MRS " states that "Any individual located on or beyond the nearest boundary of tha controlled area shall not receive a dose greater than 5 rem to the whole body or any organ from I
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| ATTACHMENT 2 l
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| l
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| )
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| l
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| ]
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| any design basis accident." In addition, in the regulation of geologic repository operations,10 CFR l 60.136, states that " ..for Category 2 design basis events, no individual located on or beyond any point on the boundary of the preclosure controlled area will receive...a total effective dose equivalent of 5 rcm...." A TEDE of 0.05 Sv (5 rem)is also the upper limit of EPA's Protective Action Guides of between 0.01 to 0.05 Sv (1 to 5 rem) for emergency evacuation of members of the public in the event of an accidental release that could result in inhalation, ingestion, or absorption of radioactive materials.
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| {
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| Section 70.60(b)(2)(ii)(B). This paragraph deals with an acute exposure of a member of the
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| {
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| public to radioactive materials or hazardous chemicals produced from radioactive materials in concentrations between AEGL-1 (ERPG-1) and AEGL-2 (ERPG-2) criteria.
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| An acute exposure of a member of the public to radioactive matenals or hazardous chemicals oroduced from radioactive matenals at concentrations that could cause notable discomfort (but below concentrations that could cause irreversible effects) is considered an intermediate-consequence cv;nt. Two existing standards, AEGL-1 and ERPG-1, :an be used to define the concentration level for '
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| notable discomfort. AEGL-1 is defined as "The airborne concentration (expressed in ppm or mg/m3) of a substance at or above which it is predicted that the general population, including susceptible, but excluding hypersusceptible, individuals, could experience notable discomfort." ERPG-1 is defined as
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| 'Th] maximum airborne concentration below which it is believed that nearly allindividuals could be cxposed for up to 1 hour without experiencirig other than mild transient adverse effects or perceiving a cle rly defined, objectionable odor." Two additional standards, AEGL-2 and ERPG-2, can be used to define the concentratioa level for irreversible health effects [see definitions in 970.60(b)(1)(iii)(C),
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| tbove). If, for a particular chemical, the AEGL values are available, they should be used. Otherwise the ERPG values should be used. If there are no AEGL or ERPG values available, then the applicant should adopt criteria that are comparable in severity to those that have been established for other chemicals.
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| Section 70.60(b)(2)(iii). This paragraph deals with a release of radioactive material to the environment.
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| The release of radioactive material to the environment outside the restricted area in !
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| concentrations that, if averaged over a period of 24 hours, exceed 5000 times the values specified in !
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| Tcble 2 of Appendix B to Part 20, is considered an intermediate-consequence event. In contrast to the other consequences crit 3ria that directly protect workers and members of the public, the intent of this criterion is to ensure protection of the environment from the occurrence of accidents at certain f:cilities authorized to process greater than critical mass quantities of SNM. This implements NRC's ATTACHMENT 2
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| rssponsibility for protecting the environment in accordance with the Atomic Energy Act of 1954, et seq ,
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| and the National Environmental Policy Act of 1969, et sea.
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| The value established for the environmental consequence criterion is identical to the NRC Abnormal Occurrence (AO) criterion that addresses the discharge or dispersal of radioactive material from its intended place of confinement. (Section 208 of the Energy Reorganization Act of 1974, as Emended, requires that AOs be reported to Congress on an annual basis.) In particular, AO reporting criterion 1.B.1 requires the reporting of an event that involves "...the release of radioactive material to en unrestricted area in concentrations which, if averaged over a period of 24 hours, exceed 5000 times tha values specified in Table 2 of Appendix B to 10 CFR Part 20, unless the licensee has demonstrated compliance with 10 CFR 20.1301 using 10 CFR 20.1302(b)(1) or 10 CFR 20.1302(b)(2)(ii)," (December l
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| 19,1996; 61 FR 67072). The concentrations listed in Table 2 of Appendix B to Part 20 apply to l
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| rcdioactive materials in air and water effluents to unrestricted areas. NRC established these i concentrations based on an implicit effective dose equivalent limit of 0.5 mSv/yr (50 mrem /yr) for each f medium, assuming an individual were continuously exposed to the listed concentrations present in an unrestricted area for a year.
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| If an individual were continuously exposed for 1 day to concentrations of radioactive material l 5000 times greater than the values listed in Appendix B to Part 20, the projected dose would be about 6.8 mSv (680 mrem), or 5000 x 0.5 mSv/yr x 1 day x 1 yr/365 day. In addition, a release of radioactive mLterial, from a facility, resulting in these concentrations would be expected to cause some environmental contamination in the area affected by the release. This contamination would pose a longer-term hazard to the environment and members of the public until it was properly remediated.
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| Depending on the extent of environmental contamination caused by such a release, the contamination could require considerable licensee resources to remediate. For these reasons, NRC considered the existing AO reporting criterion for discharge or dispersal of radioactive material as an appropriate consequence criterion in this rulemaking.
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| Several existing fuel fabrication licensees have chosen to demonstrate compliance with the public dose limit in 10 CFR 20.1301, using 10 CFR 20.1302(b)(1). However, in these cases, routine
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| . operations at the facilities do not release effluents that come anywhere close to approaching the Table 2 vilues in Appendix B to Part 20. Indeed, routine discharge of heavy metals such as uranium in concentrations that substantially exceed the Table 2 values in water or air effluents would be expected to cause extensive environmental contamination that would be difficult and expensive to remediate.
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| This has been demonstrated by the extensive and expensive decommissioning actions that have been r: quired at former fuel fabrication facilities in the United States (see NRC's " Site Decommissioning ATTACHMENT 2
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| Mrnagement Plan," NUREG 1444). In addition, SNM-processing licensees would not be expected to use the compliance method in 10 CFR 20.1302(b)(2)(ii) because this is primarily directed at extemal radiation hazards, whereas the materials released from SNM processing facilities primarily represent internal radiation and chemical risks h:: rt. Consequently, there is no need to retain the caveat r:garding alternative means of demonstrating compliance with the public dose limit, as found in the AO rzporting criterion.
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| Section 70.60(c). This paragraph deals with the graded level of protection.
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| This section addresses the level of protection a licensee must provide to ensure an acceptable d; gree of risk at its facility. That protection must be sufficient to reduce the likelihood of potential recidents to levels commensurate with their consequences. In determining the appropriate level of protection that the licensee must provide, consideration may be given to the inherent likelihood of the accident. By inherent, we mean the likelihood of the accident, assuming no controls are in place.
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| Thus, an accident that is initiated by an unlikely external eve.it may require less protection (provided by tha licensee) than an accident, with identical consequence, that is initiated by a more frequent event.
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| i For example, suppose a serious fire, with high consequences, could be started as the result of a j process deviation that is estimated to occur once per year. The level of protection needed to prevent or mitigate this accident would be greater than that needed to protect against a similar fire resulting from an unlikely external event, such as an earthquake that might occur once in 500 years. Thus, licensees may take credit for inherent " unlikeliness" of an accident in determining the level of protection that needs to be applied.
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| The goal of applying a graded level of protection is to reduce the likelihood or consequences of eccidents' to ensure an acceptable level of risk at the licensee's facility. For each of the high-consequence events identified in the proposed 670.60(b), the Commission believes that the occurrence of such an event should be highly unlikely to occur during any given year of plant operation.
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| For each of the intermediate-consequence events identified in the proposed 570.60(b), the Commission believes that the occurrence of such an event should be unlikely to occur during any given yzr of plant operation.
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| The Commission has decided not to include a quantitative definition of "unlikely" and
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| * highly unlikely" in the proposed rule, because a single definition for each term may not be appropriate.
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| 5 For exposures of werkers or members of the public to radioactive materials or hazardous chemicals produced from radioactive materials during normal operations, adherence to the existing requirements of 10 CFR 20 and 29 CFR 1910 should be sufficient to protect the public health and safety.
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| ATTACHMENT 2
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| Depending on the type of facility and its complexity, the number of potential accidents and their consequences, which are identified in the ISA, could differ markedly. Thus, even if the permitted likIlihood for each event were quantitatively defined, the integrated risk for a given facility would depend on the number of such events that could occur and the consequences of those events. For cx;mple, some facilities may have few potential accidents in the "high-consequence" range while others may have many potential accidents in this range. Therefore, to ensure that the overall facility risk is acceptable for different types of facilities, guidelines for interpreting *likely" and " highly unlikely" may need to be adjusted accordingly. To accommodate the potential variation in these guidelines, the Commission believes that the standard review plan is the appropriate document to address these trrms. The " Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility,"
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| which is being made available with the proposed rule, provides guidelines that can be applied to existing fuel cycle facilities. These guidelines have been selected so as to be consistent with the safety performance goals in the NRC Strategic Plan (NUREG-1614, Vol.1). The Commission intends to publish standard review plans for different types of facilities licer. sed by NRC. as the need arises.
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| Appropriate guidelines for such facilities'can be addressed in the standard review plans at that time.
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| Section 70.60(d). This paragraph deals with the safety program.
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| ISA. The performance of an ISA, and the establishment of measures to ensure the continuous avillability and reliability of items relied on for safety, are the means by which licensees are able to demonstrate their ability to provide an adequate level of protection at their facilities. The ISA is a syst;matic analysis to identify plant and extemal hazards and their potential for initiating accident .
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| sequences; the potential accident sequences and their consequences; and the site, structures, systIms, equipment, components, and activities of personnel, relied on for safety. As used here, integrated means joint consideration of, and protection from, all relevant hazards, including radiological, criticality, fire, and chemical. The structure of the safety program recognizes the critical role that the ISA plays in identifying potential accidents and the items relied on for safety. However, it tiss recognizes that the performance of the ISA, by itself, will not ensure adequate protc non. Instead, an effective management system is needed to ensure that, when called on, the items rel. d on for safity are continuously in place and operating properly.
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| There are four major steps in performing an ISA:
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| (1) Identify all hazards at the facility, including both radiological and non-radiological hazards.
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| Haz rdous materials, their location, and quantities, should be identified, as well as all hazardous
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| _. conditions, such as high temperature and high pressure. In addition, any interactions that could result ATTACHMENT 2
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| in the generation of hazardous materials or conditions should be identified.
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| (2) Analyze the hazards to identify how they might result in potential accidents. These accidents could be caused by process deviations or other events intemal to the plant, or by credible extsrnal events, including natural phenomena such as Soods, earthquakes, etc. To accomplish the trsk of identifying potential accidents, the licensee needs to ensure that detailed and accurate information about plant processes is maintained and made available to the personnel performing the ISA.
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| (3) Determine the consequences of each accident that has been identified. For an accident with consequences at a high or intermediate level, as defined in 10 CFR 70.60(b), the likelihood of
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| .such an accident must be shown to be commensurate with the consequences, as required in the proposed 10 CFR 70.60(c). Protection against accidents with consequences below the intermediate livIl threshold is assumed to be provided by adherence to existing NRC, OSHA, and EPA regulations.
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| (4) Identify the items relied on for safety (i.e., those items that are relied on to prevent or to mitigate the accidents identified in the ISA). Such items are needed to reduce the likelihood or consequences of the accidents to acceptable levels. The identification of items relied on for safety is required only for accidents with cons'equences at a high or intermediate level, as defined in the proposed 10 CFR '70.60(b).
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| Manaoement control. Although the ISA plays a critical role in identifying potential accidents and the items relied on for safety, the performance of an ISA will not, by itself, ensure adequate protection. Instead, according to the proposed 10 CFR 70.60(d), an effective management system is .
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| needed to ensure that, when called on, the items relied on for safety are continuously available and reli!ble (i.e., in place and operating properly). Maintenance measures must be in place to ensure the continuous availability and reliability of all hardware relied on for safety. Training measures must be cstiblished to ensure that all personnel relied on for safety are appropriately trained to perform their ccfety functions. Human-system interfaces and safety-related procedures must be developed and implemented to enable personnel relied on for safety to effective!y carry out their duties. Changes in L the configuration of the facility need to be carefully controlled to ensure consistency among the facility design and operational requirements, the physical configuration, and the facility documentation. In cddition, quality assurance measures need to be established to ensure that the items relied on for
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| ~ saf;ty and the measures used to ensure their continuous availability and reliability are of sufficient
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| = quility. Periodic audits and assessments of licensee safety programs must be performed to ensure thit facility operations are conducted in compliance with NRC regulations and protect the worker and ATTACHMENT 2
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| tha public health and safety. When abnormal events occur, investigations of those events must be
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| . carried out to prevent their recurrence and to ensure that they do not lead to more serious consequences. Finally, to demonstrate compliance with NRC regulations, records that document safety program activities must be maintained for the life of the facility.
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| Section 70.62. ~ This section deals with requirements for the performance of ISAs and the filing cf ISA results and license applications. These requirements address the question of who should perform ISAs, when they should be performed, and what ISA information should be provided to NRC.
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| The performance of an ISA would be required of all licensees authorized to possess a critical miss of SNM, that are engaged in one of the following activities: enriched uranium processing; uranium fuel fabrication; uranium enrichment; enriched uranium hexafluoride conversion; plutonium processing; mixed-oxide fuel fabrication; scrap recovery; or any other activity that the Commission dit rmines could signifi:ently affect public health and safety. The Commission believes that possession and processing of SNM in amourits sufficient to constitute a potential for nuclear criticality is a reasonable criterion for requiring the performance of an ISA. Licensees meeting this criterion are tirrady subject to nuclear criticality monitoring and alarm requirements that ensure an adequate response to a nuclear criticality event after it occurs. The performance of an ISA provides the means for licensees to ensure adequate measures are taken to prevent a nuclear criticality event (or other high-consequence event) before it occurs. By limiting the requirement for performance of an ISA to licensees engaged in specific activities that involve major chemical or mechanical processing of SNM, the Commission recognizes that these activities involve a higher degree of risk than the activities of licensees who are authorized to possess critical quantities of SNM, but do not perform any mechanical
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| . cr chemical processing of critical or near-critical quantities of the SNM.
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| These types of facilities include sub-critical assemblies, where the critical mass of material is
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| ' fixed in place in such a manner that an inadvertent nuclear criticality is not credible; research facilities that are authorized to possess a critical quantity of material, but do not process more than a small
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| . fraction of that material at any one laboratory; facilities that are authorized only to store the material; cnd facilities no longer operating, for which the material is dispersed throughout the facility as residue
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| . in walls, floors, or other fixed structures. However, potentially hazan:lous activities involving cleanup cnd decommissioning at non-operating facilities would be subject to the ISA requirement.
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| The proposed rule would require current Part 70 licensees, for whom the rule would be epplicable to develop compliance plans and submit them to NRC within 6 months of the effective date cf the rule. Each compliance plan would identify the processes that would be subject to an ISA, the ATTACHMENT 2
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| ISA approach that would be implemented for each process, and the schedule for completing the
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| . analysis of each process. Licensees would be expected to complete their ISAs within 4 years of the effective date of the rule, correct any unacceptable vulnerabilities identified, and submit to NRC the rssults for evaluation, approval, and incorporation in the license. Pending the correction of any unacceptable vulnerabilities, licensees would be expected to implement appropriate compensatory m:asures to ensure adequate protection. The process description in the ISA submittal should contain information that demonstrates the licensee's compliance with the nuclear criticality monitoring and alirm requirements in 10 CFR 70.24.
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| Applicants operating existing facilities that could become newly subject to the Commission's authority, such as DOE facilities, would be expected to perform ISAs and submit the results as part of th:ir applications for licenses. The ISA submittats should contain information that demonstrates the licensees' compliance with the nuclear criticality monitoring and alarm requirements in 10 CFR 70.24.
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| Applicants for licenses to operate new facilities or new processes at existing facilities would be expected to design their facilities or processes to protect against the occurrence of the adverse consequences identified in the proposed 10 CFR 70.60(b). In addition, the initial designs are expected to comply with the nuclear criticality monitoring and alarm requirements in 10 CFR 70.24 and the baseline design criteria in the proposed 10 CFR 70.64.
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| Based on these initial designs, the applicants are expected to perform preliminary ISAs before construction of facilities. If the ISA results show deficiencies in the design, the design should be modified to assure that the items and measures planned to protect against identified accidents are 4dequate. On the other hand, if the ISA results show that a given item at a given facility is not relied on i for safe ty, or that it does not require full adherence to the baseline criteria, then the facility design may be modified accordingly. The applicant is expected to submit the results of the preliminary ISA, based i on the modified design of the facility, to NRC before construction. However, NRC approval is not n:cessary for the applicant to proceed with construction. The submittal should include the identification j of all cases where a deviation from the baseline criteria is proposed, along with a justification for that decision. The submittal of the preliminary ISA for review by NRC provides an opportunity for cpplicants to get early feedback on the design of their facilities or processes. It is much more cost-cffective to correct problems identified at the design stage than after the facility has been constructed.
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| After construction, but before operation, applicants would be expected to update their ISAs, based on as-built conditions, taking into account the results of the preliminary ISAs, and submit the
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| ~ results to NRC for approval. Any inconsistencies between the results of the updated ISAs and the pr;liminary ISAs should be identified in the submittals.
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| ATTACHMENT 2
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| I Section 70.64. This section deals with baseline design criteria for new facilities or new
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| . processes at existing facilities.
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| A major feature of the proposed amendments to Part 70 is the requirement that licensees and cpplicants for a license perform an ISA. The ISA process is applied to existing designs to identify high
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| ' risks that could warrant additional preventive or mitigative measures. For new facilities or new processes at existing facilities, the proposed rule calls for the performance of the ISA before construction, and the updating of the ISA before beginning operations. However, for new processes and facilities, the Commission recognizes that good engineering practice dictates that certain minimum rcquirements be applied as design and safety considerations for any new nuclear process or facility.
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| Thsrefore, the Commission has specified baseline design criteria in g70.64 that are similar to the general design criteria in Part 50 Appendix A; Part 72, Subpart F; and 10 CFR 60.131. The baseline .
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| d2 sign criteria identify 10 initial safety design considerations, including: quality standards and records; natural' phenomena hazards; fire protection; environmental and dynamic effects *; chemical prote'ction; cmsrgency capability; utility services; inspection, testing, and maintenance; nuclear criticality control; cnd instrumentation and controls. The baseline design criteria do not provide relief from compliance with the safety performance requirements of $70.60. The baseline design criteria are generally an gcceptable set of initial design safety considerations, which may not be sufficient to assure adequate enfaty for all new processes and facilities. The ISA process is intended to identify additional safety futures that may be needed. On the other hand, the Commission recognizes that there may be processes or facilities for which some of the baseline design criteria may not be necessary or cppropriate, based on the results of the updated ISA. For such processes and facilities, any design futures that are inconsistent with the baseline design criteria should be identified and justified. l Section 70.65. This section deals with the additional content of applications.
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| There is additional information that would need to be submitted to NRC as part of a license
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| : spplication to demonstrate compliance with the additional requirements that would be established in !
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| ' the proposed new subpart. This information is necessary to determine whether the applicant has provided an adequate level of protection at the facility. In particular, additional information would be i needed to demonstrate how the applicant's safety program complies with 10 CFR 70.60(d). This Envronmental and dynamic effects are effects that could be caused by ambient conditions. For example, an item relied
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| - on for safety will need to funcbon within its expected environment (i.e., under normal operating conditions, expected accident
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| ' conditions, etc.). These conditions could include high temperatures, or a corrosive environment. It could also include dynamic changes in surrounding conditions caused by an acodent (e.g., the bursting of a high-pressure pipe). i ATTACHMENT 2
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| information would include a description of the plant site and structures: the processes analyzed in the ISA; an appropriate summary of the results of the ISA, including the accident sequences, the consequences and likelihoods of such sequences; and the items relied on for safety; and the measures Established to ensure the continuous availability and reliability of such items. The plant and process I descriptions are needed to fully understand the results of the ISA, including the rationale for choosing ths items relied on for safety. The evaluation of the applicant's safety program is a critical element in datermining whether the facility is safe and should be issued a license. Finally, the license application, for an operating facility, should include a description of operational events that have occurred during
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| , tha past 10 years that had a significant impact on the safety of the facility. These events should be gddressed in the applicant's ISA to ensure that the range of accident sequences considered in the ISA (ncompasses actual events that have occurred at the facility.
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| The license application demonstrates how the applicant intends to meet the requirements of Part 70. The application provides information about the applicant's facility and processes and comm.tments that ensure the health and safety of workers, the general public and the environment. To cnsure confidence that these commitments will be adhered to, and will not be changed without NRC knowledge or approval, the following condition will be inserted in the license: " Authorized use: For use in cecordance with the statements, representations, and conditions in the application dated .
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| and supplements dated The application may be revised in accordance with the provisions of 10 CFR 70.72." This condition is similar to the ones currently in use. However, it would apply to the )
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| Entire license application (not just a portion of the application, as was done previously), and would j cllow changes to be made without prior NRC approval, in accordance with 10 CFR 70.72. ,
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| J Section 70.66. This section deals with records.
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| NRC confidence in the margin of safety at its licensed facilities depends, in part, on the ability of licensees to mintain a set of current, accurate, and complete records available for NRC inspection.
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| These records serve two major purposes. First, they can supplement information that has been submitted as part of the license application. For example, applicants would be required to submit the results of their ISAs to NRC for review. However, there may be substantial amounts of supporting !
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| m;terial, at the licensed facility, relevant to that submittal, that NRC may wish to review. Second, records are often needed to demonstrate licensee compliance with applicable regulations and license commitments. It is important, therefore, that an appropriate system of recordkeeping be implemented to allow easy retrieval of required information.
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| ATTACHMENT 2
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| Section 70.68. This section deals with additional requirements for the approval of license
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| )
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| applications.
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| In addition to the requirements found in the existing rule (i.e.,10 CFR 70.23 ), the Commission must determine that the requirements in the proposed new subpart,10 CFR 70.60 through 70.66, will be satisfied.
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| Section 70.72. This section deals with changes to site, structures, systems, equipment, components, and activities of personnel.
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| Past incidents at fuel cycle facilities have often resulted from changes not fully analyzed, not authorized by management, or not adequately understood by facility personnel. Therefore, effective control of changes to a facility's site, structures, systems, equipment, components, and activities of personnelis a key element in assuring confidence in the margin of safety at that facility. Any such change needs to be considered and evaluated by the licensee t,efore the change is made. If the licensee evaluates the change, based on its ISA, and finds that it, at most, increases the risk at the facility to a minimal extent, then the licensee may make the change and then notify NRC within 60 drys. Otherwise, the licensee would need to request a license amendment and get NRC approval before making the change. In either case, the change should be controlled by the licensee's configuration management system, and appropriate modifications to the license application (including, if applicable, the results of the ISA) should be submitted to NRC. Aside from providing increased confidence in the margin of safety, maintaining the license so that it reflects the current configuration of the facility would facilitate a relatively simple, cost-effective license renewal process. The ability of licensees to make certain changes to their facility without prior NRC approval, as allowed in this proposed requirement, is analogous to existing requirements in 10 CFR 70.32.
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| Section 70.73. This section deals with the renewal of licenses.
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| Under the proposed amendments to Part 70, changes to site, structures, systems, equipment, components, and activities of personnel, made by a licensee, would be reflected in the license epplication, which would be submitted to NRC and incorporated as a condition of the license. This process would establish a "living" license that would be maintained on a current basis. As a result, the license renewal process is expected to be a pro forma activity in which NRC, based on its current knowledge of licensee activities, as reflected in the "living license," would approve the renewal with l
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| minimal additional review of the licensee's safety program. This approval would be contingent on the j licensee satisfying any requirements associated with the National Environmental Policy Act of 1969 as )
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| ATTACHMENT 2
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| implemented in 10 CFR Part 51. I l
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| l Section 70.74. This section deals with additional reporting requirements.
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| The new requirements that would be incorporated in the proposed changes to Part 70 suggest a revised approach for reporting of events to NRC. This new approach, based on consideration of the consequences of concern established in 10 CFR 70.60(b), is intended to replace and expand on the approach licensees have currently been using for reporting nuclear criticality events under Bulletin 91-
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| : 01. The new approach would cover all types of events, not just nuclear criticality events, and establish e timeframe for reporting that is scaled according to risk. The new reporting requirements are intended to supplement the requirements in the existing Part 70. A more detailed discussion of the new requirements is found in the discussion of Appendix _A G to Part 70.
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| Appendix ^ "^ cute Exce:t te Guide!he Leve!: (AEGLc" TS!:-cppand!x cent:!n: the AEGL v !uet, fer
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| * 50er expecuret, th:t h:v: been :t:b!!:hed by EP^ TSO: v:!ue: Or ref ren00d !" 10 CFR MM%
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| Append!x E. "EPoG" 75!: 2pp^nd!x contain: the EPoG v !ue: th:t have been ect b!!:hed by ,^'"^
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| The:: v !ue: cre referen d !" 10 CFo 70.SO(b).
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| Appendix A G. " Reportable Events" To effectively fulfill its responsibilities, NRC needs to be aware of conditions that could result in en imminent danger to the worker or to public health and safety. In the event of an accident, NRC must be able to respond accurately to requests for information by the public and the media. In addition, to the extent possible, NRC needs to be able to provide appropriate assistance to licensees in their efforts to address potential emergencies. Once safe conditions have been restored after an cv;nt, NRC has an interest in disseminating information on the event to the nuclear industry and other int: rested parties, to reduce the likelihood that the event will occur in the future. Finally, NRC must track the performance of individual licensees and the industry as a whole to fulfill its statutory mandate ta protect the health and safety of the worker and the public.
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| NRC intends to take a graded approach for reporting licensee events, as illustrated in Table 2.
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| According to this approach, licensees would report events based on whether actual consequences h:ve occurred or whether a potential for such consequences exists. The most serious events, and those that must be reported within the shortest timeframe (1 hour) are high-consequence events that have actually occurred. Intermediate-consequence events that have actually occurred should be ATTACHMENT 2
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| rcported within 4 hours.
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| Events that could potentially lead to a consequence of concem should also be reported.
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| External conditions, such as a hurricane, tornado, or flood, that could pose a threat to safety at a ficility, should be reported within 4 hours. Deviations from safe operating conditions should be' raported within a time period that depends on the severity of the potential consequence and whether or not the licensee is able to correct the deviation within the specified period. A deviation from safe operating conditions means that a parameter that is controlled to ensure adequate protection is outside its established safety limits, or that an item relied on for safety is no longer operational or has been degraded so that it cannot perform its intended function. The reporting requirements for deviations L from safe operating conditions are intended to be generally consistent with the reporting scheme established under Bulletin 91-01. For example, if a nuclear criticality control identified in the ISA is no longer operational, or degraded so that it cannot perform its intended function, that situation should be i r ported to NRC. If the control cannot be reestablished within 4 hours of discovery, the report should be made before expiration of the 4-hour time period. If the con'rol has been reestablished within 4 hours of discovery, the report should be made within 24 hours. The term " reestablish"is intended to man that the control identified in the ISA is made operative. Therefore, if a control fails and an ad-hoc control, not identified in the ISA, is established within 4 hours of discovery, a report to NRC would still have to be made before expiration of the 4-hour time period.
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| Another category of potential events that should be reported is one that involves the existence of cn unsafe condition that is not identified in the ISA. This condition could be caused by a deviation from established safe operating conditions, or by an unanticipated and unanalyzed set of circumstances. The timefrarne for reporting this type of event would depend on how long it takes the licensee to remove the unsafe condition, and restore normal operations, if the licensee were unable to r; store normal operating conditions within 4 hours, the report would need to be made before expiration of the 4-hour period. If the licensee were able to remove the unsafe condition and restore normal operations within 4 hours, the report would need to be made within 24 hours.
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| REFERENCES Graig, D.K., et al.., " Alternative Guideline Limits for Chemicals Without Environmental Response Planning Guidelines," American Industrial Hygiene Association Joumal,1995.
| |
| I Fisher, D.R., Hui, T.E., Yurconic, M., and Johnson, J.R., " Uranium Hexafluoride Public Risk,"
| |
| Pacific Northwest National Laboratory, PNL-10065, Richland, WA, August 1994.
| |
| National Council on Radiation Protection and Measurements (NCRP), " Basic Radiation j Protection Criteria," NCRP Report No. 39, Washington, DC,1971.
| |
| National Council on Radiation Protection and Measurements (NCRP), " Recommendations on Limits for Exposure to lonizing Radiation," NCRP Report No. 91, Washington, DC,1987.
| |
| U.S. Nuclear Regulatory Commission, " Proposed Methods for Regulating Major Materials Licensees," NUREG-1324, Washington, DC, February 1992.
| |
| l U.S. Nuclear Regulatory Commission / Occupational Safety and Health Administration (OSHA),
| |
| " Memorandum of Understanding Between NRC and OSHA; Worker Protection at NRC-Licensed Facilities" (53 FR 43950; October 31,1988).
| |
| U.S. Nuclear Regulatory Commission, " Certification of Gaseous Diffusion Plants" (59 FR 48944; September 23,1994).
| |
| U.S. Nuclear Regulatory Commission, " Abnormal Occurrence Reports: Implementation of Section 208 of Energy Reorganization Act of 1974" (61 FR 67072; December 19,1996).
| |
| U.S. Nuclear Regulatory Commission, " Site Decommissioning Management Plan," NUREG-1444, Washington, DC, October 1993.
| |
| U.S. Nuclear Regulatory Commission, " Strategic Plan, Fiscal Year 1997 - Fiscal Year 2002,"
| |
| NUREG-1614, Washington, DC, September 1997.
| |
| ATTACHMENT 2
| |
| | |
| U.S. Environmental Protection Agency," Manual of Protective Action Guides and Protective Actions for Nuclear incidents, EPA-400-R-92-001, May 1992.
| |
| ' U.S. Nuclear Regulatory Commission, " Instruction Concerning Risks from Occupational Radiation Exposure," Regulatory Guide 8.29, Rev.1, February 1996.
| |
| Theide, L., " Emergency information Where It's Needed," DOE Risk Management Quarterly, Vol 5, No 2, Richland, WA, May 1997.
| |
| i These documents are available for inspection and copying for a fee at the NRC Public Dccument Room,2120 L Street, N.W. (Lower Level), Wash;ngton DC 20555-0001.
| |
| Copies of NUREG-1324, NUREG-1614, and NUREC-1444 may also be purchased from the Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington DC 20402-9328. Copies are also available from the National Technical Information Service, 5285 Port Royal Road, Springfield VA 22161.
| |
| Regulatory Guide 8.29 may be purchased from the Government Printing Office (GPO) at the current GPO price. Information on current GPO prices may be obtained by contacting the Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washingtoa DC 20402-9328. Issued guides may also be purchased from the National Technical Information Service on a standing order basis. Details on this service may be obtained by writing NTIS,5285 Port Royal Road, Springfield, VA 22161.
| |
| Copies of the following draft regulatory guidance documents are available by request from the NRC Public Document Room: " Standard Review P!an for the Review of a License Application for a Fuel Cycle Facility" (Draft NUREG-1520); " Integrated Safety Analysis Guidance Document" (Draft NUREG-1513); and " Example Elements of an ISA Submittal- Process Descriptions and Accident Analysis Summary."
| |
| Finding of No Significant EnvironmentalImpact: Availability The Commission has determined, under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in subpart A of 10 CFR Part 51, tnat this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment, and 1
| |
| ATTACHMENT 2
| |
| | |
| L ,-
| |
| therefore an environmental impact statement is not required.
| |
| The proposed amendments to Part 70 are intended to provide increased confidence in the margin of safety at certain facilities that possess a critical mass of SNM. To accomplish this objective, the amendments: (1) identify appropriate consequence criteria and the level of protection needed to prevent or mitigate accidents that exceed such criteria; (2) require affected licensees to perform an ISA
| |
| . to identify potential accidents at the facility and the items relied on for safety; (3) require the implementation of measures to ensure that the items relied on for safety are continuously available and reliable; and (4) require the inclusion of the safety bases, including the results of the ISA, in the license application. The language, in the proposed rule, that defines an environmental consequence of concem, is relevant to the question of environmental impact. Licensees would be required to provide an adequate level of protection against a " .. release of radioactive material to the environment outside
| |
| - the restricted area in concentrations that, if averaged over 24 hours, exceed 5000 times the values specified in Table 2 of Appendix B to 10 CFR Part 20." implementation of the new amendments, including the requiremer.t to protect against etents that coulc' damage the environment, is expected to result in a significant improvement in licensees' (and NRC's) understanding of the risks at their facilities and their ability to ensure that those risks are acceptable. For existing licensees, any deficiencies identified in the ISA would need to be promptly addressed. For new licensees, operations would not begin unless licensees demonstrated an adequate level of protection against potential accidents identified in the ISA. . As a result, the safety and environmental impact of the new amendments is positive. There will be less adverse impact on the environment from operations carried out in <
| |
| accordance with the proposed rule than if those operations were carried out in accordance with the existing Part 70 regulation.
| |
| The determination of this environmental assessment is that there will be no significant offsite impact to the public from this action. However, the general public should note that NRC welcomes public participation. NRC has also committed to complying with Executive Order (EO) 12898, " Federal Actions to Address Environmental Justice in Minority Populations and Low-income Populations," dated February 11,1994, in all its actions. Therefore, NRC has also determined that there are no
| |
| - disproportionate, high, and adverse impacts on minority and low-income populations. In the letter and spirit of EO _12898, NRC is requesting public comment on any environmental justice considerations or questions that the public thinks may be related to this proposed rule, but somehow were not addressed. Comments on any aspect of the Environmental Assessment, including environmental justice, may be submitted to NRC, as indicated under the ADDRESSES heading.
| |
| NRC has sent a copy of the environmental assessment and this proposed rule to all State ATTACHMENT 2 I
| |
| | |
| ?
| |
| Liaison Officers and requested their comments on the Environmental Assessment. The Environmental Assessment is available for inspection at the NRC Public Document Room,2120 L Street NW. (Lower Level), Washington, D.C. Single copies of the environmental assessment are available from Richard I.
| |
| Milstein, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC, 20555-0001, telephone (301) 415-8149; e-mail: rim @nrc. gov.
| |
| Paperwork Reduction Act Statement This proposed rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501, et sea.). This rule has been submitted to the Office of Management and Budget (OMB) for review and approval of the paperwork requirements.
| |
| The public reporting burden for this information collection is estimated to average 70 hours per response. and the recordkeeping burden is estimated to average 500 hours per licensee, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the information collection. The U.S. Nuclear Regulatory Commission is seeking public comment on the potential impact of the information collections contained in the proposed rule and on the following issues:
| |
| : 1. Is the proposed information collection necessary for the proper performance of NRC's function? Will the information have practical utility?
| |
| : 2. Is the burden estimate accurate? l
| |
| : 3. Is there a way to enhance the quality, utility, and clarity of the information to be collected?
| |
| : 4. How can the burden of the information collection be minimized, including the use of automated collection techniques?
| |
| Send comments on any aspect of this proposed information collection, including suggestions for reducing the burden, to the Records Management Branch (T-6-F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Intemet electronic mail at bjs1@nrc. gov; and to the Desk Officer, Office of information and Regulatory Affairs, NEOB-10202 (3150-0009), Office of Management and Budget, Washington, DC 20503.
| |
| Comments to OMB on the information collections or on the above issues should be submitted by (insert 30 days after publication in the Federal Register). Comments received after this date will be considered if it is practical to do so, but assurance of consideration cannot be given to comments received after this date.
| |
| ATTACHMENT 2
| |
| | |
| s Public Protection Notification if an information collection does not display a currently valid OMB control number, NRC may not conduct nor sponsor, and a person is not required to respond to the information collection.
| |
| Regulatory Analysis
| |
| {
| |
| The Commission has prepared a draft regulatory analysis on this proposed regulation. The analysis examines the costs and benefits of the alternatives considered by the Commission. The draft '
| |
| analysis is available for inspection in the NRC Public Document Room,2120 L Street N.W. (Lower Level), Washington, D.C. Single copies of the analysis may be obtained from Barry T. Mendelsohn, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC, telephone (301) 415- 7262, e-mail: btm1@nrc. gov.
| |
| The Commission requests public comment on the draft regulatory analysis. Comments on the draft analysis may be submitted to NRC as indicated under :he ADDRESSES heading.
| |
| Regulatory Flexibility Certification As required by the Regulatory Flexibility Act, as amended,5 U.S.C. 605(b), the Commission certifies that this proposed rule, if adopted, would not have a significant economic impact on a substantial number of small entities. This proposed rule would affect major nuclear fuel fabrication
| |
| - facilities that are authorized to possess a critical mass of SNM. These licensees do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act, nor the size standards published by NRC (10 CFR 2.810).
| |
| Backfit Analysis NRC has determined that the backfit rule does not apply to this proposed rule; therefore, a
| |
| )
| |
| backfit analysis is not required for this proposed rule because these amendments do not involve any ;
| |
| provisions that would impose backfits as defined in 10 CFR Chapter 1. ,
| |
| 1 List of Subjects in 10 CFR Part 70 Criminal penalties, Hazardous materials transportation, Material control and accounting, Nuclear materials, Packaging and containers, Radiation protection, Reporting and recordkeeping requirements, Scientific equipment, Security measures, Special nuclear material.
| |
| For the reasons set out in the preamble and under the authority of the Atomic Energy Act of ATTACHMENT 2
| |
| | |
| ,l 1954, as amended; the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, NRC is proposing to adopt the following amendments to Part 70 Part 70 - DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL
| |
| : 1. The authority citation for Part 70 continues to read as follows:
| |
| AUTHORITY: Secs. 51,53,161,182,183,68 Stat. 929,930,948,953,954, as amended, sec.
| |
| 234, 83 Stat. 444, as amended (42 U.S.C. 2071, 2073, 2201, 2232, 2233, 2282, 2297f); secs. 201, as amended, 202, 204, 206, 88 Stat.1242, as amended, 1244,1245,1246 (42 U.S.C. 5841,5842,5845, )
| |
| 5846). Sec.193,104 Stat. 2835, as amended by Pub. L. 104-134,110 Stat.1321,1321-349 (42 U.S.C. 2243).
| |
| Sections 70.1(c) and 70.20a(b) also issued under secs. 135,141, Pub. L. 97-425, 96 Stat.
| |
| 2232,2241 (42 U.S.C.10155,10161). Section 70.7 also issued under Pub. L. 95-601, sec.10, 92 Stat. 2951 (42 U.S.C. 5851). Section 70.21(g) also issued Jnder seC.122,68 Stat. 939 (42 U.S.C.
| |
| 2152). Section 70.31 also issued under sec. 57d, Pub. L. 93-377,88 Stat. 475 (42 U.S.C. 2077).
| |
| Sections 70.36 and 70.44 also issued under sec.184,68 Stat. 954, as amended (42 U.S.C. 2234).
| |
| Section 70.61 also issued under secs. 186,187, 68 Stat. 955 (42 U.S.C. 2236, 2237). Section 70.62 also issued under sec.108,68 Stat. 939, as amended (42 U.S.C. 2138).
| |
| : 2. The undesignated center heading " GENERAL PROVISIONS" is redesignated as "Subpart A
| |
| - General Provisions."
| |
| : 3. in 10 CFR 70.4, the definitions of Acute exposure, Acute exposure guideline levels (AEGLs),
| |
| Controlled site boundary, Critical mass of SNM, Deviation from safe operating conditions, Double contingency, Emergency response planning guidelines (ERPGs), Hazardous chemicals, Hazardous chemicals produced from radioactive materials, Integrated safety analysis (ISA), items relied on for safety, New process, Results of the ISA, Unacceptable vulnerabilities, and Worker are added, in alphabetical order, as follows:
| |
| 9 70.4 Definitions.
| |
| Acute exoosure means a single exposure or multiple exposures occurring within a short time (24 hours or less).
| |
| ATTACHMENT 2
| |
| | |
| E-h Acute exoosure auideline levels (AEGLs) mean chemical concentration levels, established by the National Advisory Committee for Acute Guideline Levels for Hazardous Substances, that, for a defined exposure, would result in anticipated adverse health effects to humans. The following three levels have been established:
| |
| (1) AEGL-1 means the airbome concentration (expressed in ppm or mg/m*) of a substance at or above which it is predicted that the general population, including susceptible but excluding hypersusceptible individuals, could experience notable discomfort.
| |
| (2) AEGL-2 means the airbome concentration (expressed in ppm or mg/m*) of a substance at or above which it is predicted that the general population,' including susceptible but excluding hypersusceptible individuals, could experience irreversible or other serious, long-lasting effects or impaired ability to escape.
| |
| (3) AEGL-3 means the airbome concentration (expressed in ppm or mg/m*) of a substance at or above which it is predicted that the general population, including susceptible but excluding hypersusceptible individuals, could exparience life-threatening effects or death.
| |
| Controlled site boundary means the physical barrier surrounding the facility that is used by the licensee to control access. It may or may not coincide with the property boundary.
| |
| Critical mass of SNM means special nuclear material in a quantity exceeding 700 grams of contained uranium-235; 520 grams of uranium-233; 450 grams of plutonium; 1500 grams of contained uranium-235, if no uranium enriched to more than 4 percent by weight of uranium-235 is present; 450 grams of any combination thereof; or one-half such quantities if massive moderators or reflectors made i of graphite, heavy water, or beryllium may be present.
| |
| Deviation from safe operatina conditions means that a parameter that is controlled to ensure adequate protection is outside its established safety limits, or that an item relied on for safety has been lost or has been degraded so that it cannot perform its intended function.
| |
| Double continoency means a process design that incorporates sufficient factors of safety to
| |
| _ requ a east two unlikely, independent, and concurrent changes in process conditions before a i re tl nuclear criticality accident is possible.
| |
| ATTACHMENT 2
| |
| | |
| Emeroency response olannina auidelines (ERPGs) mean chemical concentration levels, established by the American Industrial Hygiene Association, that, for a defined exposure, would result in anticipated adverse health effects on humans. The following three levels have been established: l (1) ERPG-1 means the maximum airbome concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour without experiencing other than mild transient adverse
| |
| )
| |
| effects or perceiving a clearly defined, objectionable odor.
| |
| (2) ERPG-2 means the maximum airbome concentration below which it is believed that nearly
| |
| {
| |
| all individuals could be exposed for up to 1 hour without experiencing or developing irreversible or other health effects or symptoms which could impair an individual's ability to take protective action.
| |
| (3) ERPG-3 means the maximum airbome concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour without experiencing or developing life-threatening 1 health effects.
| |
| . . . . . i Hazardous chemicals means substances that are toxic, explosive, flammable, corrosive or reactive to the extent that they can : rer !;r5: rt deme; t prepedy er endanger life if not adequately controlled.
| |
| Hazardous chemicals produced from radioactive materials means Hazardous chemicals either havina radioactive material (s) as precursor compound (s) or formed throuah interaction with radioactive materials They do not include chemicals merelv added to. used in or recycled from. the processina of special nuclear material inteorated safety analysis (ISA) means a systematic analysis to identify plant and extemal hazards and their potential for initiating accident sequences, the potential accident sequences, their likelihood and consequences, and the site, structures, systems, equipment, components, and activities of personnel that are relied on for safety. As used here, integrated means joint consideration of, and protection from, all relevant hazards, including radiological, nuclear criticality, fire, and chemical.
| |
| Items relied on for safety means structures, systems, equipment, components, and activities of personnel that are relied on to prevent or to mitigate potential accidents at a facility.
| |
| New process means, for a particular licensee, a change in the basic method for processing special nuclear material, where the new method is not currently specifically authorized by the NRC license.
| |
| ATTACHMENT 2
| |
| _a
| |
| | |
| l
| |
| \
| |
| l Results of the ISA means the information obtained as a result of performing an ISA. It includes the identification of: (1) the radiological and non-radiological hazards at the facility; (2) the accident sequences that could result from such hazards; (3) the consequence and likelihood of occurrence of each accident sequence; and (4) the items relied on for safety.
| |
| Unacceptable vulnerabilities mean deficiencies in the items relied on for safety or the measures used to assure the continuous availability and reliability of such items that need to be corrected to ensure an adequate level of protection as defined in 10 CFR 70.60(c).
| |
| Worker means an individual whose assigned duties in the course of employment involve exposure to radiation and/or radioactive material from licensed and unlicensed sources of radiation (i.e., an individual who is subject to an occupational dose as in 20 CFR 20.1003).
| |
| : 4. The undesignated center heading " EXEMPTIONS" is redesignated as "Subpart B -
| |
| Exemptions."
| |
| 9% 70.13a and 70.14 [ Redesignated]
| |
| : 5. Sections 70.13a and 70.14 are redesignated as $$ 70.14 and 70.17, respectively.
| |
| : 6. Section 70.15 is added to read as follows:
| |
| 6 70.15 Nuclear reactors.
| |
| The regulations in Subpart H do not apply to nuclear reactors licensed under 10 CFR Part 50.
| |
| l
| |
| : 7. The undesignated center heading " GENERAL LICENSES" is redesignated as "Subpart C -
| |
| General Licenses."
| |
| : 8. The undesignated center heading " LICENSE APPLICATIONS" is redesignated as "Subpart D - License Applications." i 9 70.22 [ amended] l
| |
| : 9. In 10 CFR 70.22, paragraph (f) is removed and paragraphs (g) through (n) are redesignated cs (f) through (m).
| |
| ATTACHMENT 2
| |
| | |
| 9 70.23 [ amended]
| |
| ' 10. In 10 CFR 70.23, paragraph (a)(8) is removed, paragraph (b) is removed and reserved, cnd paragraphs (a)(9) through (a)(12) are redesignated as (a)(8) through (a)(11), respectively.
| |
| : 11. The undesignated center heading " LICENSES" is redesignated as "Subpart E - Licenses."
| |
| : 12. The undesignated center heading " ACQUISITION, USE AND TRANSFER OF SPECIAL NUCLEAR MATERIAL,- CREDITORS' RIGHTS," is redesignated as "Subpart F - Acquisition, Use, And Transfer Of Special Nuclear Material, Creditors' Rights."
| |
| : 13. The undesignated center heading "SPECIAL NUCLEAR MATERIAL CONTROL RECORDS, REPORTS AND INSPECTIONS" is redesignated as "Subpart G - Special Nuclear Material Control Records, Reports, And Inspections."
| |
| : 14. The undesignated center heading "MODIFICAT'ON AND REVOCATION OF LICENSES"is redesignated as "Subpart I - Modification and Revocation of Licenses."
| |
| 99 70.61 and 70.62 [ redesignated)
| |
| : 15. Sections 70.61 and 70.62 are redesignated as $970.81 and 70.82, respectively.
| |
| : 16. The undesignated center heading " ENFORCEMENT is redesignated as "Subpart J -
| |
| Enforcement." ~
| |
| j 99 70.71 and 70.72 (redesignated]
| |
| : 17. Sections 70.71 and 70.72 are redesignated as $970.91 and 70.92, respectively.
| |
| : 18. In Part 70, a new "SUBPART H" (99 70.60 - 70.74) is added to read as follows:
| |
| Subpart H - Additional Requirements for Certain Licensees Authorized To Possess a Critical Mass of Special Nuclear Material Sec.
| |
| 70.60 . Safety pe'r formance requirements.
| |
| 70.62 Requirements for the performance of ISAs and the filing of ISA results and license applications.
| |
| ATTACHMENT 2
| |
| | |
| p 70.64 Baseline design criteria for new facilities or new processes at existing facilities.
| |
| 70.65 Additional content of applications.
| |
| '70.66 Records.
| |
| 70.68 ' Additional requirements for approval of license application.
| |
| 70.72 Changes to fLeility structures, systmas, equipment, components, and activities of personnel.
| |
| - 70.73 Renewal of licenses.
| |
| 70.74 Additional reporting requirements.
| |
| 670 60 Safety performance reauirements.
| |
| l l (a) Purpose. Each licensee engaged in enriched uranium processing, uranium fuc!
| |
| fabrication, uranium enrichment, enriched uranium hexafluoride conversion, plutonium processing, mixed-oxide fuel fabrication, scrap recovery, or any other activity that the Commission determines could significantly affect public health and safety, shall provide protection to its workers, the genera:
| |
| public, and the environtrent against radiological (including nyclear criticality), chemical, and fire hazards that could result in the adverse consequences identified in paragraph (b) of this section.
| |
| ' Concideration must be given to radiological consequences from all causes (including those resulting from fires and hazardous chemicals), and those chemical and environmental consequences produced
| |
| - by rsdioactive materials th:t ree!d rere" 40- tM: prerrrr! .;; ef e--"r' er'er- rtr-L. !.
| |
| (b) Conseauences of concem. Each licensee shall protect against the occurrence of the following high and intermediate adverse consequences that could result from accidents involving the ,
| |
| handling, storage, or processing of licensed special nuclear material:
| |
| (1) Hiah conseauences.
| |
| (i) A nuclear criticality; (ii) Acute exposure of a worker to -
| |
| (A) A radiation dose of 1 SV (100 rem) or greater total effective dose equivalent; or (B) Radioactive materiab or Mhazardous chemicals produced from radioactive materials in concentrations exceeding AEGL-3 (^;;r-9 ^) or ERPG-3 (.^;;:nfM S) criteria; or
| |
| - (iii) Acute exposure of a member of the public outside the controlled site boundary to:
| |
| (A) A radiation dose of 0.25 Sv (25 rem) or greater total effective dose equivalent; (B) An intake of 30 mg or greater of uranium in a soluble form; or (C) Radioactive materials or Whazardous chemicals produced from radioactive materials in concentrations exceeding AEGL-2 (.^;;:nf::: A) or ERPG-2 (^;;:nd::: 9) criteria.
| |
| g2) Intermediate consecuences.
| |
| ATTACHMENT 2 l
| |
| 1
| |
| !- l
| |
| | |
| p (i) Acute exposure of a worker to -
| |
| (A) A radiation dose between 0.25 Sv (25 rem) and 1 Sv (100 rem) total effective dose equivalent; or (B) Radioactive materials or Whazardous chemicals produced from radioactive materials in concentrations between AEGL-2 (.^; pen 9 ^) or ERPG-2 (.^; pent 9) criteria and AEGL-3
| |
| (.^;per9 ^.) or ERPG-3 (.^; pent 9) criteria; or (ii) Acute exposure of a nu nber of the public outside the controlled site boundary to -
| |
| (A) A radiation dose between 0.05 Sv (5 rem) and 0.25 Sv (25 rem) total effective dose equivalent; or (B) Radioactive materials or Whazardous chemicals produced from radioactive materials in concentrations between AEGL-1 (.^;per9 ^) or ERPG-1 (.^; pent 9) criteria and AEGL-2
| |
| (.^;per9.^) or ERPG-2 (.^;per9 9) criteria; or (iii) Release of radioactive material to tha envi onment outside the restricted area in concentrations that, if averaged over a period of 24 hours, exceed 5000 times the values specified in Table 2 of Appendix B to 10 CFR Part 20.
| |
| (c) Graded level of protection. Each licensee shall provide a level of protection that is commensurate with the severity of the consequences resulting from credible accidents and the likelihood of any extemal events (e.g., natural phenomena) assumed to initiate or propagate such accidents. This graded level must apply to the items relied on for safety, identified in paragraph (d)(2)(iv) of this section, and to the measures used to assure their continuous availability and reliability, identified in paragraph (d)(3) of this section, The application of a graded level of protection must assure that -
| |
| (1) The occurrence of any of the high consequences identified in paragraph (b)(1) of this section is highly unlikely; and (2) The occurrence of any of the intermediate consequences identified in paragraph (b)(2) of this section,is enlikely.
| |
| (d) Safety orooram. Each licensee shall establish and maintain a safety program that provides rGasonable assurance that the acc' dent consequences identified in paragraph (b) of this section are adequately protected against in accordance with paragraph (c).
| |
| (1) Each licensee shall compile and maintain a set of process safety information to enable the performance of an integrated safety analysis (ISA). This process safety information must include information pertaining to the hazards of the materials used or produced in the process, information pertaining to the technology of the process, and information pertaining to the equipment in the process.
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| ATTACHMENT 2
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| P. . ,
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| . (2) Each licensee shall perform an ISA to identify -
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| (i) All radiological and non-radiological hazards (e.g., chemical, fire, electrical, and mechanical);
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| (ii) Potential accident sequences caused by process deviations or other events internal to the plant (e.g., fires, explosions, or chemical releases) and credible external events, including natural ;
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| phenomena (e.g., hurricanes, floods, tomadoes, earthquakes, tsunami, and seiches), fires, explosions, or chemical releases occurring offsite; (iii) The consequence and likelihood of occurrence of each accident sequence identified I'
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| pursuant to paragraph (d)(2)(ii) of this section; and (iv) ltems relied on for safety (i.e., structures, systems, equipment, components, and activities of personnel), that are relied on to prevent or mitigate those accidents identified under paragraph (d)(2)(ii) of this section, that exceed the consequences of concem stated in paragrapn (b) of this 1
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| se: tion.
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| (3) To ensure the continuous availability and reliability of items relied on for safety identified ;
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| under giaragraph (d)(2)(iv) of this section, each licensee shall demonstrate that -
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| (i) S.ructures, systems, equipment, and components relied on for safety are designed, ;
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| constructed, inspected, calibrated, tested, and maintained, as necessary, to ensure the continuous ability to perform their safety functions to satisfy paragraph (c) of this section. Items subject to this j requirement include but are not limited to: principal structures of the plant; passive barriers relied on for safety (e.g., piping, glove boxes, containers, tanks, columns, vessels); active systems, equipment, and components relied on for safety; sampling and measurement systems used to convey information cbout the safety of plant operations; instrumentation and control systems used to monitor and control )
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| 4 the behavior of systems relied on for safety; and utility service systems relied on for safety.
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| (ii) Personnel are trained, tested, and retested, as necessary, to ensure that they understand, recognize the importance of, and are qualified to perform their safety duties to satisfy paragraph (c) of this section; (iii) Procedures relied on for safety are developed, reviewed, approved, and distributed to ensure that personnel are able to perform their safety duties to satisfy paragraph (c) of this section.
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| (iv) Human-system interfaces are designed and implemented to ensure that personnel relied l on for safety are able to perform their safety duties to satisfy paragraph (c) of this section. l (v) Configuration changes to site, structures, process, systems, equipment, components, computer programs, personnel, procedures, and documentation are managrA so that such i modifications are reviewed, documented, communicated, and implemented in a systematic, controlled !
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| l ATTACHMENT 2
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| {
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| i 1
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| manner to satisfy paragraph (c) of this section.
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| . (vi) All items relied on for safety identified under paragraph (d)(2)(iv) cf this section and measures established under paragraphs (d)(3)(i) through (d)(3)(v) of this section must meet quality standards that are commensurate with the importance of the safety functions performed. Management shall establish appropriate quality assurance policies and procedures to ensure that all items relied on for safety perform their safety functions and are continuously available and reliable.
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| (4) Each licensee shall conduct audits and assessments of its safety program to ensure that an adequate level of protection is maintained at the facility.
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| (5) Each licensee shall investigate abnormal events and take corrective action to minimize the recurrence of these events.
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| (6) Each licensee shall establish records that will demonstrate that the requirements of paragraphs (d)(1), (d)(2), (d)(3), (d)(4), and (d)(5) of this section have been met. Each licensee shall maintain these records for the lifetime of the plant.
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| 670.62 Reauirements for the oerformance of ISAs and the filina of ISA results and license aoolications.
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| (a) Each applicant for a license under this subpart and each current licensee subject to this subpart shall perform an ISA as described in 970.60(d)(2).
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| (1) Each current licensee shall-(i) Within 6 months of the effective date of this rule, submit, for NRC approval, a compliance-plan that describes the ISA approach that will be used, the processes that will be analyzed, and the schedule for completing the analysis of each process; and (ii) Wdhin 4 years of the effective date of this rule, perform an ISA in accordance with the compliance plan submitted under paragraph (a)(1)(i) of this section, correct any unacceptable vulnerabilities identified in the ISA, and submit the results of the ISA as part of the license application contents identified in 970.65 to NRC, for approval. Pending the correction of any unacceptable vulnerabilities identified in the ISA, the licensee shall implement appropriate compensatory measures to ensure adequate protection. The process description in the ISA submittal must include information that demonstrates the licensee's compliance with the design requirements for nuclear criticality monitoring and alarms in 570.24.
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| (2) Each applicant operating a facility that is newly subject to the Commission's authority shall perform an ISA, correct any unacceptable vulnerabilities identified in the ISA, and submit the results of the ISA as part of the license application contents identified in $$70.22 and 70.65 to NRC, for approval.
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| . The process description in the ISA submittal must include information that demonstrates the applicant's ATTACHMENT 2
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| )
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| Lm.- . - . . . - .. - __ -. ..
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| .__..--__.-.1
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| E compliance with the design requirements for nuclear criticality monitoring and alarms in @70.24.
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| (3) Each applicant for a license to operate a new facility or a new process at an existing facility shall -
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| (i) Initially design the facility or process to protect against the occurrence of the adverse consequences identified in 70.60(b), meet the nuclear criticality monitoring and alarm requirements of 970.24, and meet the baseline design criteria in 70.64; (ii) Perform a preliminary ISA and submit the results to NRC before construction of the facility or process. The results of the preliminary ISA must demonstrate an adequate level of protection, as defined in 970.60(c), against occurrence of the adverse consequences in @70.60(b). The preliminary ISA submittal shall include facility and process description and design information that demonstrates the applicant's incorporation of the nuclear criticality monitoring and alarm requirements in @70.24, and the baseline design criteria in 970.64. Any proposed relaxation in the application of the baseline design criteria, pursuant to @70.64(a), must be icientified and justified in the preliminary ISA submittal; and (iii) Before beginning operations, update the preliminary ISA and correct any unacceptable vulnerabilities identified in the ISA. The updated ISA must be baced on as-built conditions and must take into account the results of the preliminary ISA. Any inconsistencies between the results of the updated ISA and the preliminary ISA must be identified.
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| (A) For new facilities submit the results of the ISA, as part of the license application contents identified in 6 70.22 and 70.65, to NRC for approval.
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| (B) For new processes submit the results of the ISA and any revisions of the approved license Cpplication as part of an application for amendment of the license under 70.34.
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| (b) If the decommissioning of a facility involves potentially hazardous activities such as chemical treatment of wastes, each licensee shall perform an ISA of the decommissioning process, correct any unacceptable vulnerabilities identified in the ISA, and submit the results to NRC for approval before beginning such decommissioning activities.
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| 670.64 Baseline desian criteria for new facilities or new processes at existino facilities.
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| (a) Applicants shall address the following baseline design criteria in the design of new facilities or design of new processes at existing facilities, before performing the preliminary ISA, in accordance with $70.62(a)(3)(ii). Applicants shall address these baseline design criteria in establishing minimum requirements for all items in their process design and description, which is provided in the application for a license or license amendment. Licensees shall maintain the application of these criteria unless ATTACHMENT 2
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| c W
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| the preliminary ISA, submitted before construction, pursuant to 570.62(b)(3)(iii), demonstrates that a
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| : given item is not relied on for safety or does not require adherence to the specified criteria.
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| (1) Quality standards and records. The design must be established and implemented in i Eccordance with a quality assurance program, to provide adequate assurance that items relied on for
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| ~ tafety will satisfactorily perform their safety functions. Appropriate records of these items must be maintained'by or under the control of the licensee throughout the life of the facility.
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| (2) Natural ohenomena hazards. The design must provide for adequate protection against natural phenomena with consideration of the most severe documented historical events for the site.
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| (3) Fire protection. The design must provide for adequate protection against fires and explosions.
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| (4) Environmental and dynamic effects. The design must provide for adequate protection from Environmuntal conditions and dynamic effects associated with normal operations, maintenance, testing, and postulated accidents that could lead to loss of safety functions.
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| (5) Chemical crotection. The design must provide for adequate protection against chemical hazards pf radioactive materials and of hazardous chemicals produced from radioactive materials tr':trd !: th: : terr;r, 5:rd!!r;, Ord preer:6; Of!!:rrrrd nur!rrr mrtd:!.
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| (6) Emeroency capabi!ity. The design must provide for emergency capability to maintain control of; (i) Licensed material; l (ii) Evacuation of personnel; and (iii) Onsite emergency facilities and services that facilitate the use of available offsite services.
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| (7) Utility services. The design must provide for continued operation of essential utility services, including reliable and timely emergency power to items relied on for safety.
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| (8) Insoection. testina. and maintenance. The design af items relied on for safety must provide for periodic inspection, testing, and maintenance, to ensure their continued function and readiness.
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| (g) Nuclear criticality control. The design must provide for nuclear criticality controlincluding adherence to the double-contingency principle.
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| (10) Instrumentation and controls. The design must provide for inclusion of instrumcotation End control systems to monitor and control the behavior of items relied on for safety.
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| (b) Facility and system design and plant layout must be based on defense-in-depth practices.
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| Features must be incorporated that enhance safety by reducing challenges to items relied on for safety. Where practicable, passive systems and features must be selected over active systems and features, to increase overall system reliability.
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| ATTACHMENT 2 l
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| l
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| f~
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| 670.65 Additional content of aoolications, in addition to the contents required by $70.22, each application for a license to possess a critical mass of special nuclear material for use in the activities describcd in 970.60(a), must contain -
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| (a) A description of the applicant's site, structures, and the processes analyzed in the ISA; (b) A description of the applicant's safety program established under $70.60(d), including the results of the ISA and the measures established to ensure the continuous availability cnd reliability of items relied on for safety; and (c) For currently operating facilities, a description of operational events, within the past 10
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| . years, that had a significant impact on the safety of the facility.
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| 670.66 Records.
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| The applicant or licensee shall establish and maintain onsite, readily available for Commission
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| ' inspection,'a system of legible, current, accurate, complete, and easily retrievable records to' document application-related and license-related information required by applicable parts of this chapter, Commission action, license condition, and commitments by the applicant or licensee. Records must be retained for the period specified by the applicable parts of this chapter, Commission action, license condition, and commitments made by applicant or licensee. If a retention period is not otherwise !
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| specified, these records must be retained until the Commission terminates the license or determines that they are no longer required.
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| 670.68 Additional reauirements for aor teval of license aoolication.
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| An application for a license to possess a critical mass of SNM will be approved if the l Commission determines that the applicant has complied with the requirements of 970.23 and $$70.60 through 70.66.
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| 670.72 Chances to site. stsctures. systems. eauipment. components. and activities of personnel.
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| (a); Except for a new process, subject to the requirements of $70.62(a)(3), any change to site, structures, systems, equipment, components, and activities of personnel must be evaluated by the licensee before the change, to determine whether the change increases the Skelihood or i consequences of an accdont at the facility. The evaluation must be based on the licensee's ISA results, developed in accordance with 670.60(d)(2), and other safety program information, developed !
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| in accordance with $70.60(d)(3), which J part of the license application contents identified in 970.65.
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| ATTACHMENT 2
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| 4 9 (b) A licensee may make a change to site, structures, systems, equipment, ' components, and activities of personnel, without prior Commission approval,if the change--
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| (1) Results in, at most, a minimal increase in the likelibcod or consequences of an accident previously evaluated in the ISA; (2) Would not create the potential for an accident different from any previously evaluated in the ISA; and (3) Is not inconsistent with NRC requirements and license conditions.
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| (c) For any change authorized under paragraph (b) of this section, the licensee shall submit revised pages to the license application, including any changes h the results of the ISA, to NRC within 60 days ofinitiation of the change.
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| (d) For any change that is not authorized under paragraph (b) of this section, the licensee shall file an application for an amendment of its license, as specified in $70.34, that authorizes the change.
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| As part of the applicatio.1 for the amendment, the 1.censee shall perform an ISA of the change and submit any revisions of the ISA and the license application t) NRC for approval. The liceasee shall also provide, as required by Part 51 of this chapter, any necessary revisions to its environmental report.
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| (e) The licensee shall maintain records of changes to its facility carried out under paragraph (a) of this section. These records must include a written evaluation that provides the bases for the determination that the changes ao not require prior Commission approval under paragraph (b) of this section. These records must be maintained until termination of the license.
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| 670.73 Renewal of licenses.
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| Applications for renewal of a license must be filed in accordance with $$ 2.109,70.21,70.22, 70.33,70.38, and 70.65. Information provided in applications, including the results of the ISA, must be current, complete, and accurate in all material respects. Information contained in previous applications, statements, or reports filed with the Commission under the license may be incorporated by reference, provided that these references are clear and specific.
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| 670.74 Additional reportina reauirements.
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| (a) Reports to NRC Operations Centa_r.
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| (1) Each licensee shall report to the NRC Operations Center the events described in paragraphs I, il, and lli of Appendix AC to Part 70.
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| (2) Reports must be made by a knowledgeable licensee representative and by any method that ATTACHMENT 2
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| will ensure compliance with the required time period (1,4 or 24 hours) for reporting.
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| (3) The information provided must include a description of the event and other related information as described in paragraph V of Appendix A G to Part 70.
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| (4) Followup information to the reports must be provided until all information required to be reported in paragraph (a)(3) of this scetion is complete. 1 (5) Duplicate reports to the Commission are not required for events when the reports are made in compliance with other parts of this chapter, provided that the reports comply with the requirements of l
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| this section conceming addressees, information content, and timeliness of filing.
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| (6) Each licensee shall provide reasonable assurance that reliable communication with the NRC Operations Center is available during each event.
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| (b) Written reports. j (1) Each licensee shall provide a written report to NRC, of the events described in paragraph IV l of Appendix A_G to Pan 70, within 30 days of discovery. The written report must contain the informati on described in paragraph VI of Appendix iC to Part 70.
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| (2) Each licensee who makes a report required by paragraph (a) of this section shall submit a written fol,lowup report within 30 days of the initial report. The written report shall contain the I information as described in paragraph VI of Appendix A_G to Part 70.
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| ATTACHMENT 2 i
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| l ATTACHMENT 2 1
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| 1
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| \
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| l l
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| E o
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| l 12 34. Appendix A G to Part 70 is added to read as follows:
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| Appendix A C to Part 70 - Reportable Safety Events As required by 10 CFR 70.74, licensees who are authorized to possess a critical mass of special nuclear material shall report the following safety events (see table A-1 of this appendix):
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| : 1. Events to be reported within 1 hour of discovery, followed by a written report within 30 days.
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| (a) An accident from the processing of licensed material that resulted in any of the following i consequences:
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| (1) A nuclear criticality.
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| (2) Acute expcettre of a worker to --
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| (i) A radiation dose of 1 Sv (100 rem) or greater total effective dose equivalent, or I
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| (ii) Radioactive materials or Hhazardous chemicals produced from radioactive materials in concentrations exceeding AEGL-3 (^.ppnd!x ^.) or ERPG-3 (^; pend!x 9) criteria.
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| (3) Acute exposure of a member of the public outside the controlled site boundary to -
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| (i) A radiation dose of 0.25 Sv (25 rem) or greater total effective dose equivalent, (ii) An intake of 30 mg or greater of uranium in a soluble form, or (iii) Radioactive materials or Hhazardous chemicals produced from radioactive materials in concentretions exceeding AEGL-2 (Append!x A) or ERPG-2 (Appnd!: 9) criteria.
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| : 11. Events to be reported within 4 hours of discovery, followed by a written repcrt within 30 days.
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| (a) An accident from the processing of licensed material that resulted in ar.y of the following consequences:
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| (1) Acute exposure of a worker to -
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| (i) A radiation dose between 0.25 Sv (25 rem) and 1 Sv (100 rem) total effective dose equivalent, or (ii) Radioactive materials or Hhazardous chemicals produced from radioactive materials in concentrations between AEGL-2 (Appnd!r ^.) or ERPG-2 (Appndix 9) criteria and AEGL-3 (.^ppnd!x A) or ERPG-3 (.^;pnd!x 9) criteria.
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| (2) Acute., exposure of a member of the public outside the controlled site boundary to -
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| (i) A radiation dose between 0.05 Sv (5 rem) and 0.25 Sv (25 rem) total effective dose equivalent, or (ii) Radioactive materials or Mhazardous chemicals p_r,_oduced from radioactive materials in ATTACHMENT 2
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| concentrations between AEGL-1 (.^dpendr ^.) or ERPG-1 (^ppent 9) criteria and AEGL-2 (Appendr A) or ERPG-2 (Appent 9) criteria.
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| (3) Release of radioactive material to the environment outside the restricted area in concentrations that, if averaged over a period of 24 hours, exceed 5000 times the values specified in Table 2 of Appendix B to 10 CFR Part 20.
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| (b) A deviation from safe operating coriditions that has not been corrected within 4 hours and has the potential, as identified in the ISA, for causing an accident with one or more of the consequences specifed in paragraph l(a) of this appendix.
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| (c) An extemal condition that poses a threat to the performance of items that are relied on for
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| - safety (e.g., site, structures, systems, equipment, components, or activities of personnel). These conditions would include natural phenomena (e.g., hurricanes, floods, tomados, earthquakes), fires, or chemical releases.
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| (d) A potentially unsafe condition that has not been corrected within 4 hours and that has not been identified or analyzed in the integrated safety analysis (ISA).
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| Ill. Events to be reported within 24 hours of discovery, followed by a written report within 30 days.
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| (a) A deviation from safe operating conditions that was corrected within 4 hours and had the potential, as identified in the ISA, for causing an accident with one or more of the consequences specified in paragraph l(a) of this appendix.
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| (b) A deviation from safe operating conditions that has not been corrected within 24 hours and has the potential, as identified in the ISA, for causing an accident with one or more of the consequences specified in paragraph II(a) of this appendix.
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| (c) A potentially unsafe condition that was corrected within 4 hours and was net identifed or analyzed in the ISA.
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| IV. Events to be reported in writing, to NRC, within 30 days of discovery.
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| (a) A deviation from safe operating condition,,, that was corrected with?n 24 hours and had the potential, as identified in the ISA, for causing an accident with one or more of the consequences specified in paragraph II(a) of this appendix.
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| V. Licensee reports to the NRC Operations Center, as required by 10 CFR 70.74(a), shall include, to the extent that the information is applicable and available at the time the report is made, the following:
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| (a) Caller's name and position title.
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| (b) Date, time, and location of the event. !
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| i ATTACHMENT 2 i
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| +
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| .e (c) Description of the event, including -
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| (1) Sequence of occurrences leading to the event, including degradation or failure of items l
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| relied on for safety. i (2) Radiological or chemical hazards involved including isotopes, quantities, and chemical and physical form of any material released.
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| 3 (3) Actual or potential health and safety consequences to the workers, the public, and the j II . environment, including relevant chemical and radiation data for actual personnel exposures (e.g., level of radiation exposure, concentration of chemicals, and duration of exposure).
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| (4) ltems that are relied on to prevent or to mitigate the health and safety consequences, and whether the ability of those items to function has been affected by the event.
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| l (5) For events involving deviations from safe operating condiiicns, the process parameters that l are deviant, the normal operating and safety limits on these parameters, and the current values of these parameters. !
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| (d) Extemal conditions affecting the event. l (e) Additional actions taken by the licensee in response to the event.
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| (f) Status of the event (e.g., whether the event is on-going or was terminated). l (g) Current and planned site status, including any declared emergency class.
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| (h) Notifications related to the event that were made or are planned to any local, State, or other Federal agencies. l 1
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| (i) lasue of a press release by the licensee related to the event that was made or is planned. i V!. Licensee written reports required by 10 CFR 70.74(b) shall consist of a completed NRC i Form 366 and shall be forwarded to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001 Each written report must include the following information:
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| (1) Complete applicable information required by partpraph V of this appendix.
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| - (2) Whether the event was identified in the ISA.
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| (3) Cause of the event, including all factors that contributed to the event.
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| (4) Corrective actions taken to prevent occurrence of similar or identical events in the future.
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| ATTACHMENT 2
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| ,l Dated at Rockville, Maryland, this ~ day of 1998.
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| For the Nuclear Regulatory Commission.
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| John C. Hoyle, ;
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| I Secretary of the Commission.
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| t I
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| ATTACHMENT 2
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| t t ,
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| DISCUSSION OF PROPOSED RULE'S MAJOR ELEMENTS Consequences of concem. An important element in the proposed rule is the identification of specific consequences against which licensees must provide adequate protection [10 CFR 70.60(b)]. These consequences, which are applicable to workers and members of the public, are categorized according to their level of severity (high and intermediate). Because accidents at fuel cycle facilities could result l in human exposure to both radiological and certain chemical hazards, the proposed rule has adopted criteria that address both types of consequences. This approach satisfies the U.S. Nuclear Regulatory Commission's (NRC's) primary responsibility for radiation protection, in addition to its responsibility to l ~ protect workers and the public from the chemical risk produced from radioactive materials hasards reru!!5; fre- 5: prerech; cf !!cred nur' err meter!:!.
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| Graded Levelof Protection. To ensure an acceptable level of risk at facilities that possess a critical miss of special nuclear material, the proposed rule [10 CFR 70.60(c)] calls for licensees to provide a graded level of protection against potential accidents. That level of protection must be sufficient to reduce the likelihood of such accidents to levels commensurate with their consedi uences. Thus, cecording to the proposed rule, the occurrence of any high-consequence event should be " highly unlikely," while the occurrence of any intermediate-consequence event should be "unlikely." Although tha rule does not define the terms " highly unlikely" and "unlikely," the draft Standard Review Plan provides criteria for judging the likelihood of potential accidents. This guidance is based on a combination of qualitative and quantitative indicators, but does not require a probabilistic risk casessment.
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| Intrgrated Safety Analysis (lSA). According to the proposed rule [10 CFR 70.60(d)], licensees must ,
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| - damonstrate, based on the performance of an ISA, their ability to provide an adequate level of '
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| protection against potential accidents. An ISA is a systematic analysis to identify plant and extemal hazards and their potential for initiating accident sequences; the potential accident sequences and their
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| . lik:lihood and consequences; and the items (i.e., site, structures, systems, equipment, components, cnd activities of personnel) that are relied on for safety.
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| M:asures to ensure continuous availability and reliability. Although the ISA plays a critical role in l idsntifying potential accidents and the items relied on for safety, the performance of an ISA will not, by '
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| itself, ensure adequate protection. Instead, as required by the proposed rule [10 CFR 70.60(d)], an eff2ctive management system is needed to ensure that, when called upon, the items relied on for s;fety are in place and operating properly. Maintenance measures niust be in place to ensure the continuous availability and reliability of all hardware relied on for safety. Training measures must be
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| .cstablished to ensure that all personnel whose actions are relied on for safety are appropriately trained ts perform their safety functions. Human-system interfaces and safety-related procedures must be d;veloped and implemented to enable personnel relied on for safety to effectively carry out their duties.
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| Changes in the configuration of the facility need to be carefully controlled to ensure consistency among tha facility design and operational requirements, the physical configuration, and the facility documentation. In addition, quality assurance measures need to be established to ensure that the itrms relied on for safety, and the measures used to ensure their continuous availability and reliability, cre of sufficient quality. Periodic audits and assessments of licensee safety programs must be performed to ensure that facility operations are conducted in compliance with NRC regulations and protect the public health and safety. When operational events occur, investigations of those events must be carried out to prevent their recurrence and to ensure that they do not lead to more serious consequences. Finally, to demonstrate compliance with NRC regulations, records that document Jsafety program activities must be maintained for the life of the facility.
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| | |
| c Inclusion of safety bases in the application and changes to the safety bases. The performance of the ISA to identify the items relied on for safety and the measures established to ensure the continuous '
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| availability and reliability of such items are important elements in increasing confidence in the margin of safety. Nevertheless, without formal commitments to implement these items and measures, and to ,
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| kt p NRC informed of any changes in such commitments, the safety bases could become uncertain l ov r time. Thus, the proposed rule calls for the incorporation of licensee commitments to these items j End measures in the license application. In addition, all changes in such commitments shall be submitted to NRC as part of a revised license application, including any changes in the ISA results (10 !
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| I CFR 70.72). The rule does, however, allow for certain changes to be made, based on the results of th2 ISA, without prior NRC approval, as long as such changes result in, at most, a minimalincrease in tha risk of accidents at the facility.
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| ATTACHMENT 1 3}}
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