ML20196J561
| ML20196J561 | |
| Person / Time | |
|---|---|
| Issue date: | 06/27/1988 |
| From: | NRC ADVISORY COMMITTEE ON NUCLEAR WASTE (ACNW) |
| To: | |
| References | |
| NACNUCLE-T-0001, NUDOCS 8807060409 | |
| Download: ML20196J561 (105) | |
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NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON NUCLEAR WASTE In the Matter of:
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ADVISORY COMMITTEE
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ON NUCLEAR WASTE
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AFTERNOON SESSION
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DATE:
June 27, 1988 i
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UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON NUCLEAR NASTE In the Matter Of:
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FIRST MEETING OF
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ADVISORY COMMITTEE
)
ON NUCLEAR WASTE
)
AFTERNOON SESSION kCMOff$bb Do Xot Ren)ovefrom AG ce l
PAGES:
95 through 180 PLACE:
Washington, D.C.
DATE:
June 27, 1988
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l HERITAGE REPORTING CORPORATION oskw n.pwvm l
1220 L Street, N.W., Suite 644 Washington, D.C. 20005 (202) 628-4888 1}70 09 880627 E:
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1 PUBLIC NOTICE BY THE (G
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2 UNITED STATES NUCLEAR REGULATORY COMMISSION'S 3
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 4
5 6
7 The contents of this stenographic transcript of the 8-proceedings of the United States Nuclear Regulatory 9
Commission's Advisory Committee on Reactor Safeguards (ACRS),
10 as reported herein, is an uncorrected record of the discussions 11 recorded at the meeting held on the above date.
12 No member of the ACRS Staff and no participant at 13 this meeting accepts any responsibility for errors or 14 inaccuracies of statement or data contained in this transcript.
15 16 17 18 19 20 21 22 23 24 25
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Heritage Reporting Corporation (202) 628-4888
I UNITED STATES NUCLEAR REGULATORY COMMISSION
( )1 2
ADVISORY COMMITTEE ON NUCLEAR WASTE 3
4 In the Matter of:
)
)
5 FIRST MEETING OF
)
ADVISORY COMMITTEE
)
6 ON NUCLEAR ~ WASTE
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7 AFTERNOON SESSION 8
9
- Monday, June 27, 1988 10 Room 1046 11 1717 H Street Washington, D.C.
12 r3 13 The above-entitled matter came on for hearing, U
14 l pursuant to notica, at 1:00 p.m..
BEFOR2 DR. DADE W. MOELLER f
Professor of Engineering in Environmental 16 Health Associate Dean for Continuing Education 37 School of Public Health Harvard University gg Boston, Massachusetts f
I' ACNW MEMBERS PRESENT:
I 0
DR. WILLIAM KERR Professor of Nuclear Engineering 21 Director, Office of Energy Research University of Michigan-22 Ann Arbor, Michigan 23 DR. PAUL G. SHEWMON Professor, Metallurgical Engineering Department
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Ohio State University l
Columbus, Ohio 3
Heritoge Reporting Corporation mm
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ACNW MEMBERS PRESENT (CONTINUED )
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2 DR. MARTIN STEINDLER Director, Chemical Technology Division 3
Argonne~ National Laboratory Argonne, Illinois 4
i DR. CLIFFORD G. SMITH ACNW COGNIZANT STAFF MEMBER:
6 Dr. Sidney J.S. Perry 7
ACRS 8
Raymond F.
Fraley, Executive Director 9
H.
Stanley Schofer, Technical Secretary 10 NRC 11 Dr. Thompson 12 NRC Staff
/-
13 John Lenahan i
Seth Copeland 14 ACNW Consultants 15 Melvin Carter 16 Donald Orth 17 Presenters:
ja Keith Cline Ed Regnier Leslie Jardine 19 William Millo 20 21 22 23 24
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PROCEEDINGS
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2 NT. MOELLER:
The meeting will come to order.
3
-The next topic that we're going to discuss will be 4
NRR regulatory activities on spent fuel storage and rod 5
consolidation, as well as low-level radioactive waste 6
treatment processes ~and the licensing thereof.
7 Our spokesman is John W.
Craig, chief of the 8
SPLD/ DEST branch.
Could you tell us what that is?
9 MR. CRAIG:
Good afternoon.
10 Is the mike on here?
Is it on?
11 MR. MOELLER:
You have to put the mike around your 12 neck.
13 MR. CRAIG:
My name's John Craig, and I'm a chief 14 of the plant systems branch in the division of engineering
()
15
.and systems technology in the office of nuclear reactor 16 regulation.
And I'd like to introduce my staff who work in 17 the area of fuel storage, rod consolidation, and low-level j
18 radioactive wastes.
19 The acting section leader for the plant systems 20 section is Dr. Charles Nichols, and the principal engineer 21 who reviews solid radioactive waste systems is Mr. Jay Lee.
22 And the engineer who's been reviewing fuel rod consolidation 23 is Amira Gill.
24 The objective of the briefing this afternoon is to 25 describe the NRR licensing activities with respect to solid Heritage Reporting Corporation (202) 628-4888 I
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radioactive waste management and spent fuel s* rage and
-( )
2 consolidation.
And before we go any further, if it's 3
acceptable, I'd like to rearrange the agenda a little bit so 4
that we can start with solid waste management systems with 5
Jay Lee.
6 MR. MOELLER:
To help me as you begin, I notice it 7
says on the agenda that you're going to discusa low-level 8
radioactive wastes-rod consolidation.
9 MR. LEE:
No, we meant just low-level radioactive 10 waste for myself.
11 MR. MOELLER:
Okay, and rod consolidation is not 12 low-level --
13 MR. LEE:
Right.
14 MR. MOELLER:
-- waste, I mean, I didn't think it P)
(_/
15 was.
Okay, fine.
16 MR. LEE:
I think Amira will address that part 17 later.
18 Good afternoon.
I'm Jay Lee, and I'm plant 19 assistance branch in NRR.
20 Today I will be presenting solid radioactive waste 21 management system.
Sometime it's abbreviated as LLW, but 22 really solid radioactive waste management system.
23 Sorry that the slide is not that clear, but this 24 particular slide present the interfaces we have, NRR and 25 NMSS.
What NRR is reviewing, respect to the licensing Heritage Reporting Corporation O
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activities, and what the NMSS does.
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2 NRR, we review the solid waste management system 3
in accordance with SRP, Section 11.4.
SRP stand for, of 4
course, standard review plan.
5 We review the design basis of the solid rad waste 6
system, design criteria, process capacity, equipment quality 7
group classification, radiation protection, fire protection, 8
quality assurance, radiation monitoring, radioactive 9
effluent control and monitoring, and process control system, 10 and also onsite inspection activities.
11 So you can see all uhe -- our main area of review 12 is process safety associated with operation of radioactive 13 waste management system.
14 Now, NMSS reviews waste form stability.
This is
()
15 basically 10 CFR 61 stuff.
Generic waste streams and 16 solidification media to assure that the final solidified 17 product meet the requirements specified in 10 CFR 61, and 18 also clarified in NRC or NMSS brx.ch technical position on 19 waste form, which was issued May 1983.
20 So '.his clarifies that between NRR and NMSS, as 21 we put up this slide first so during my presentation it will 22 be clear that what NRR, what we are doing as far as the 23 licensing activity's concerned.
24 MR, SMITH:
Smith, Clifford.
25 MR. LEE:
Yes.
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MR. SMITH:
Are we saying, then, that in reviewing
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2 a facility that you would review it first, and then you send 3
it over to NMSS?
4 MR. LEE:
NMSS --
5 MR. SMITH:
Or do you take into account NMSS's 6
requirements when you review it?
I'm a little confused 7
here.
8 MR. LEE:
Okay.
For the individual licensing 9
application from,our licensee, we review.
l'0 MR. SMITH:
That's -- okay.
11 MR. LEE:
Okay?
12 MR. SMITH:
All right.
13 MR. LEE:
Now, for the topical report, in generic 14 review, NHSS reviews all this vaste form stability and so 15
.forth --
16 MR. SMITH':
Okay.
17 MR. LEE:
-- for the 10 CFR 61.
18 MR. SMITH:
Thank you.
19 MR. LEE:
Yes.
20 MR. STEINDLER:
When you say review, what do you 21 mean by that?
l 22 MR. LEE:
Really safety review.
We'll come to 23 that couple later in this slide, the why we are review!.ng, 24 how we are reviewing, and the what we are reviewing.
So if 25 you can --
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MR. STEINDLER:
Do you also approve or disapprove?
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2 MR. LEE:
You mean --
3 MR. STEINDLER:
The applicant comes to you with 4
either a document or a flow diagram or something similar to 5
that; you then make a decision as to whether that's 6
acceptable or not acceptable?
7 MR. LEE:
Yes, we do, in safety evaluation report.
8 MR. STEINDLER:
How do you transmit that 9
information to NMSS?
Or do you?
10 MR. LEE:
Oh, we do not.
11 MR. STEINDLER:
You do not.
12 MR. LEE:
No.
13 MR. STEINDLER:
One other question.
14 MR. CRAIG:
Excuse me.
15 MR. LER:
Yes.
16 MR. CRAIG:
In fact we do transmit it to NMSS, and 17 they would typically get a copy of it, and the approval 18 would be contingent upon, for instance, NMSS's approval of 19 the final product.
Okay, that's 20 solidification media.
We approve the process, but that's 21 not an overall approval, necessarily, with submittal.
It 22 takes both offices to approval, review an approval, prior to 23 complete acceptance.
24 MR. STEINDLER:
Okay.
One other question.
25 What sort of talents do you have in your branch or Heritage Reporting Corporation (202) 628-4888 s
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office or whatever related to processes and the kind of 2
O reviews and approval that you are currently exercising?
3 MR. LEE:
Think I'll let you answer that question, 4
John.
5 MR. CRAIG:
Other than Mr. Lee, we have some 6
national labs that provide consulting service to us in 7
preparation of technical evaluations.
8 MR. STEINDLER:
Do you assign this to an 9
engineering department or to what kind of expertise are you 10 able to muster?
11 MR. CRAIG:
We have, in addition to Jay and to 12 Charlie, who are both very experienced and knowledgeable in 13 radioactive waste areas, and another engineer who is in the 14 branch, they're typically engineering people who would then 15 work with the national labs.
~
16 MR. MOELLER:
Go ahead.
17 MR. LEE:
Okay.
18 Now this is the NRR solid radio -- the radioactive 19 waste review.
This is overview of the NRR review process.
20 First, NRR licensing activity, we just saw the 21 previous slide on overview of NRR-NHSS activities and its 22 interfaces.
Then we'll discuss the basis for NRR review of 23 liquid, gaseous, and solid radioactive waste management l
24 systems.
25 First subitem is the statutes, which really states Heritage Reporting Corporation (202) 628-4888 n
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the basis or the purpose of our review.
Then we go into
()
2 rules and regulations applicable to our review.
This will 3
give how we are reviewing the system.
4 Then third subitem is NRR review areas, what we 5
are reviewing and licensing, and we'll discuss SRP Section 6
11.4, waste sources and waste management systems.
Then last 7
item we'll discuss briefly on the recent events on the solid 8
rad waste system.
Recent we meant subsequent to the 10 CFR 9
61 became effective and implemented in December 1983.
l 10 Okay.
Why we are reviewing, the purpose for the 11 review of radioactive waste management systems are, number f
12 one, to assess environmental impact associated with 13 operation of a radioactive waste management system pursuant 14 to the National Environmental Policy Act of 1969, which is
(
15
.really NEPA, and also we review, in order to assess as low 16 as reasonably achievable criterion -- we abbreviate as ALARA 17
-- since we are discharging radioactive materials in the 18 gaseous and liquid effluents from the power plant to the 19 environment, and also we are transporting solid rad waste 20 system -- excuse me, solid rad waste product to the burial 21 site.
22 These are pursuant to the energy reorganization of 23 1974.
24 MR. MOELLER:
We have a question from Dr. Shewmon.
25 HR. SHENHON:
Do you expect them to use ALARA in Heritage Reporting Corporation Om (202) 628-4888
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the design that they do?
<~s
(,)
2 MR. LEE:
Yes, we use the design objectives 3
specified, such as in Appendix I to 10 CFR 50, will give you 4
the numerical design objectives on the design of the system 5
as well as operation.
6 MR. SHENMON:
That's interesting, because the DOE 7
man asserted that ALARA was not used in design this morning, 8
or expected to be used, as I understood him.
That was part c
10 MR. MOELLCR:
I think his comment was for the 11 accident, in designing for an accident with teams.
12 HR. SHEWMON:
No, he said we don't have any severe 13 accidents that are imaginable, and so we should jttet stay 14 with this reasonable five rem; we shouldn't -- nobody should 15
. expect us to use ALARA and have to go below that five rem 16 for the release.
17 MR. NICHOLS:
Let me clarify Jay's points.
18 Charles Nichols, 19 I think Appendix I, design objectivas, apply to 20 the design of the liquid and to the -- of the gaseous waste 21 management systems.
What we have with the solid is what's l
22 left over from that process.
And there's no specific dose 23 objectives associated with the design of the solid waste f
24 management system.
25 In all these systems, however, accidents or l
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potential accidents are considered in the design,'such as
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2 rupture of a --
3 MR. SHEWMON:
My point is really irrelevant to 4
this afternoon.
I think it's relevant to what we were told 5
this morning.
Thanks.
Go ahead.
.6 MR. MOELLER:
I didn't quite follow -
go ahead, 7
Cliff.
Go ahead.
8 MR. SMITH:
I thought this morning that he was in 9
fact saying that they didn't -
they used ALARA, but that 10 ALARA was only used under the normal operating 11 circumstances, and you were designing the process to meet 12 that, but that this morning he was specifically talking 13 about an accident.
And so I guess I don't see the 14 confusion.
15 MR. MOELLER:
I was confused to hear when you 16 saying you review, and obviously the effluent treatment 17 systems branch, or whatever it's called, would review 18 Appendix I, but what does that have to do with solid waste 19 disposal?
20
. MR. LEE:
Solid rad waste system is not directly 21 associated with Appendix I, you are right.
But the 22 operation of a solid rad waste system will contribute 23 potentially gaseous effluent from that particular system in 24 operation as well as liquid.
25 MR. MOELLER:
So you look at it in terms of Heritage Reporting Corporation (202) 628-4888
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compliance.
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2 MR. LEE:
Yes.
That is right.
3 MR. MOELLER:
Thank you. I understand now.
l 4
MR. STEINDLER:
One other question.
Did I hear
(
5 you say that you also reviewed ths process of transporting 6
the solid product from the reactor to the burial ground?
7 MR. LEE:
No, we do not.
8 That is the NMS'S area.
9 MR, STEINDLER:
Okay.
10 MR. LEE:
We do not go into the transportation.
11 MR. STEINDLER:
You have no role to play in 12 transportation at all?
13 MR. LEE:
Correct.
You are right.
14 MR. NICHOLS:
In the NEPA, it is incumbent upon us 15
.to assess the total environmental impact of licensing and in 16 that we do look at every impact, including waste disposal.
17 The NRC does that.
18 MR. STEINDLER:
Well, now I am thoroughly lost.
19 Who looks at the question of transporting solid 20 waste in some final package from the reactor to the disposal 21 area?
22 MR. NICHOLS:
That is regulated by NMSS.
23 MR. STEINDLER:
Do you do a NEPA review at NRR?
24 MR. NICHOLS:
When we issue a license, a 25 construction permit or an operating license to a plant, we Heritage Reporting Corporation O-(202) 628-4888
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look at the total impact of that on the environment.
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2 MR. STEINDLER:
So NRR looks at the environmental 3
impact of that transportation activity?
4 MR. NICHOLS:
It is covered in the statement.
5 MR. MOELLER:
Okay.
6 Go ahead, Mr. Lee.
7 MR. LEE:
Thank you.
8 (Viewgraph presented) 9 These are the rules and the cegulations applicable 10 to the review of the radioactive warce management system.
11 This is how we review the system '.n licensing.
12 First we deal first wich 10 CFR 20, particularly 13 Section 20.106 has to do with radioactivity in effluents to 14 unrestricted areas.
15 And also 10 CFR 20, Appendix B.
We have Table 1 16 and 2 for the concentrations in air and water above a 17 natural background.
18 These are all in isotopic concentrations specified 19 in tables.
20 Then 10 CFR 50.34A, this is design objectives for 21 equipment to control releases of radioactive materials in 22 offluents.
Again, this covers design objectives.
23 The 10 CFR 50.36A deals with technical 24 specifications on effluents from nuclear power reactors.
25 This goes into really the operational procedures, how we
(-}
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regulate the operational procedures in conjunction with 10
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2 CFR 50.34A, or the design objectives.
3 Then we have 10 CFR 50, Appendix A, General Design 4
Criteria 60 which deals with the control of releases of 5
radioactive material to the environment.
.6 Then also GDC-64, the monitoring requirements for 7
the radioactivity releases..
8 Now in order to achieve 10 CFR 50.34A and 10 CFR 9
50.36A, we came up with 10 CFR 50, Appendix I, which 10 provides numerical guides for design objectives and the 11 limiting condition for operations to meet criterion ALARA 12 which was stated in 10 CFR 50.34A and also 36A.
13 In the numerical guides we do give liquid and 14 gaseous for the specific dose on total body and specific 15
. organs and for the skin and so forth.
16 So these are all the design objectives and 17 numerical guides.
These are the rules and the regulations 18 we use for reviewing radioactive waste management systems.
l 19 MR. MOELLER:
Now 10 CFR 20 is under revision?
20 MR. LEE:
Right.
21 MR. MOELLER:
What impact will the revision have 22 upon you?
t 23 MR. LEE:
Thai I cannot answer, that question.
We l
24 have another branch, the protection branch, who really deals 25 in the dose area.
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MR. MOELLER:
Well, but all of these effluent
()
2 releases, the concentrations in those tables in Appendix B, 3
are all, the vast majority, are going to be changed.
And 4
Appendix B limits the does to the public to half a rem per 5
year?
6 MR. LEE:
Yes.
7 MR. MOELLER:
The.new 10 CFR 20 in essence limits 8
it to about 50 millirem a year.
9 Are you people tooling up?
What impact is this 10 going to have on all of these systems?
11 MP. 12. 5 :
Wel l, I suppose we have to change a tech 12 spec.
'1 MR. CRAIG:
We have not made an evaluation yet.
14 MA. MOELLER:
Okay.
13 Well, another thing you are using -- and you cited 16 here Part 50, Appendix I -- the revised 10 CFR 20 uses the 17 effsetive dose equivalent approach which obsoletes -- if 18 that is a verb -- Appendix I.
19 So what are you going to do there?
Whose 20 responsibility is it to redo those regulations to make them 21 compatible, one with the other?
22 MR. NICHOLS:
I do not know the correct answer to 23 that but I am sure the first responsibility would fall upon 24 the utility.
25 MR. MOELLER:
The utility?
To change Appendix I Heritage Reporting Corporation O
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and make it correct?
()
2 MR. NICHOLS:
To comply with the regulations.
3 MR. CRAIG:
I think a different way to state that 4
would be we will have to evaluate the changes in the rule 5
with respect to how it impacts the existing reviews that we 6
do with respect to the new criteria.
7 And one of the things that we will here later on 8
is that we are in the process of -- obviously the SRP is 9
somewhat outdated -- revising the SRP and that will be one 10 of the things, as we move further into revision of Part 20, 11 to evaluate that impact, also.
12 But we do not know specifically how this will 13 impact us right now.
14 MR. MOELLER:
Okay.
15 (Viewgraph presented) 16 MR. LEE:
Now we come to Overview of NRR Review 17 Areas.
18 The radioactive waste management system consists 19 of liquid, gaseous and solid radioactive waste management 20 system as well as source term and process and effluent 21 radiation instrumentation.
22 Today's discussion that I present is really 23 limited to the solid radioactive waste management system 24 delineated in the Standard Review Plan Section 11.4.
25 MR. STEINDLZR:
I'm sorry.
Hight I ask?
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MR. LEE:
Yes.
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2 MR. STEINDLER:
Does that imply that you will not 3
deal with the topic of conversation of liquid waste to solid 4
waste?
5 MR. LEE:
How do you mean by conversion?
The 6
mixing ratio for that?
7 MR. STEIN 0LER:
Well, whatever the processes are 8
that a particular applicant has in mind to convert a liquid 9
waste stream to one that is solid and can be disposed of, 10 constitutes a process or several processes.
11 Do I understand that you will not discuss those 12 processes in the course of your discussion today?
13 MR. L1'E I do not plan to discuss that. I think, 14 that really belongs to NMSS.
15 MR. NICHOLS:
Let me make that point.
16 Let me clarify that if the liquid waste management 17 system in a nuclear power plant would result in liquid 18 effluent and also liquid water.
19 As a result of that kind of management, there are, 20 quote, "solid wastes" generated.
They are not really --
21 (Knocks on table) 22
--solid. I mean they are resins or filtered 23 whatever you call it.
And those ure considered, from the 24 Standard Review Plan 11.4, as solid waste to be managed.
25 The solidification of those wastes is a step in
(~'
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the solid waste management scheme of things.
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2 MR. STEINDLER:
And who is it that is responsible 3
for the review of whether these resins and things that are 4
not quite --
5 (Knocks on wood) 6
--solid, become solid when they leave the site and 7
are not dripping water or whatever?
8 MR. NICHOLS:
As Jay is pointing out, we are 9
performing a review on the Standard Review Plan 11.4 that is 10 supposed to provide that assurance.
11 MR. STEINDLER:
So Jay is the responsible one to 12 do that?
13 MR. NICHOLS:
Yes.
14 MR. MOELLER:
And in the safety evaluation report 15
,for the Perry Nuclear Power Plant NUREG 0887 Supplement No.
16 8 it says, quote, "The NUS topical report on the 17 solidification system was submitted to NRC review and was 18 epproved for referencing in license applications on May 19 30th, 1985", unquote.
20 Who did that?
Who approved that NUS topical 21 report?
22 MR. LEE:
Our branch did it for the process only.
23 Now NUS submitted that topical report and we reviewed prior 24 to 10 CFR 61 became effective and/or implemented.
25 MR. MOELLER:
But NRR reviewed it and approved it?
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-1 MR. LEE:
Yes we did.
()
2 MR. MOELLER:
Okay.
Thank you.
3 MR. NICHOLS:
Let me make a point about the review
'4 plan 11.4 that we have been going by.
It goes back to --
5 Jay?
6 MR. LEE:
July 1981.
7 MR. NICHOLS:
July 1981.
So it predates the 8
promulgation of Part 61.
9 MR. LEE:
We reviewed the many process reports 10 prior to when 10 CFR 61 became implemented as well as after 11 it became implemented.
12 MR. CRAIG:
But let me make a point of 13 clarification if I may.
14 The review that we would conduct would be a review
()
15
,of a topical which describes a process that a vendor would 16 assert that if this process is followed as described, that 17 the end form of waste form would be of a certain -- would 18 have certain characteristics.
19 The waste form itself and the acceptability of the 20 waste form is reviewed and evaluated by NMSS.
So it is a 21 separate topical.
22 There are really two that are required before the 23 program is completely approved or reviewed.
24 MR. STEINDLER:
One reviews the process --
25 MR. CRAIG:
Yes.
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MR. STEINDLER:
And if they conclude the process
("
g )%
2 is fol. lowed, the product would be satisfactory?
3 MR. CRAIG:
Yes.
4 MR. STEINDLER:
That is not enough.
Somebody else 5
does something else and that is where I got lost.
What 6
happens after that?
7 MR. CRAIG:
I am not your best witness.
8 MR. KERR:
What is the second part?
9 MR. CRAIG:
I will ask John Greeves from NMSS to 10 take a shot.
11 MR. GREEVES: Basically it was the purpose on one 12 of those early slides to show the separation between NRR and 13 NMSS; NRR reviewing the process control procedures. They do 14 an SER.
They send it to us and we concur on it.
And 15
. basically that is evaluating and processing the power plant.
16 A separate topical report which we talk to you 17 folks about comes to NMSS on the solidified product.
So 18 those are the two separate topical reports.
19 We approve or disapprove a product topical report 20 in terms of meeting the Part 61 burial ground requirements.
21 NRR approves or disapproves a process topical report at the 22 reactor for things like worker safety and the other items l
23 that Jay mentioned earlier.
l 24 I hope that clears up the fact that there are two l
25 topical reports and which office does what.
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-1 MR. MOELLER:
Did that answer it, Dr. Kerr?
()
2 DR. KERR:
I will not hold up the proceedings any 3
further.
4 Go ahead.
5 MR. MOELLER:
What I heard said was that NMSS 6
reviews the product.
7 MR. SMITH:
As I hear them, you have got a process 8
that you are going to design and build to produce a product.
9 You are looking at the process itself.
Now, if I 10 can make it simple, what comes out the pipeline has a 11 certain form, composition, characteristic.
NMSS has to pass 12 on the suitability of the characteristics of that product?
13 MR. LEE:
Correct.
14 MR. MOELLER:
And what is left -
go ahead.
15 MR. MOTL:
Can I take a turn at explaining this?
16 Is that permitted?
17 hR. MOELLER:
Give us your name and go to a 18 microphone, and if you could help clarify, you are welcome.
19 MR. MOTL:
My name is Gerry Motl from LA 20 Technologies.
We are the new name of NUS and submitted that 21 topical report.
22 I think the key clarification is the original 23 topical report we referred to as the process report but 24 really it was the system.
The initial topical report 2S described what the physical piece of equipment looked like, Heritage Reporting Corporation O
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the size of the pumps, how the mixing took place.
()
2 It was basically an equipment topical report.
L 3
This is before 10 CFR 61 came about.
4 The second topical report that the people from 5
NHSS are talking about is the formula, the recipe.
The 6
initial topical did not talk about how much cement, how much 7
chemical, how much waste you put in.
That is the recipe.
8 The recipe was addressed in the second topical 9
report.
The piece of equipment that did the mixing was 10 addressed in the first topical report.
11 MR. MOELLER:
Does that help?
12 DR. KERR:
Well, I cannot believe that it is done 13 that way, but if it really is, I guess that helps.
14 (Laughter) 15'.
MR. SHEWMON:
We will get later to the question of 16 if the process does not turn out the product that it said it 17 was going to turn out in this review document, who is 18 responsible for sort of what the feedback loop is or who 19 finds out about it?
20 (Pause) 21 That is a question.
22 Is it on the agenda?
23 MR. MOELLER:
That certainly should be.
24 Keep it in mind and try to clarify it.
25 MR. LEE:
Okay.
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(Viewgraph presented)
()
2 MR. LEE:
All right.
Then these are the areas 3
that NRR reviews:
Expected and design volumes of waste 4
streams to be processed and handled.
5 Here we are discussing the amount of wet waste 6
coming into the solid rad waste system to be solidified or 7
dewatered.
8 And then equipment and system capacities to meet 9
such a demand.
Make sure we have a--
10 DR. KERR:
What is meant by "reviewing expected 11 volumes of waste streams"?
12 MR. LEE:
Well, we are asking the licensee, from 13 all the systems in the power plant, how much waste will come 14 into the solid rad waste system to be solidified.
15 DR. KERR:
Okay.
16 MR. LEE:
Including dry radioactive waste, spent 17 resin, evaporator concentrates and also the decon waste and 18 oily waste, all this different type of waste; what do they 19 expect the volumes to be generated and therefore the solid 20 rad waste system has to process.
21 DR. KERR:
And they tell you that and thon what do 22 you do?
23 MR. LEE:
Then we compare with our number that we 24 have in our review.
For example, we use --
25 DR. KERR:
Now wait a minute.
I thought your Heritage Reporting Corporation O
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1 review was the review of what they tell you.
You mean you
()
2 go through the plant and make a separate analysis?
3 MR. LES:
Yes.
We do.
4 We have a source term in our NUREG 16 and in NUREG 5
17 how we independently calculate.
6 DR. KERR:
Well, do you do that for their 7
particular plant or a generic plant?
8 MR. LEE:
We do individual plants.
And we do look l
9 at the licensee's number, whether it is a reasonable number 10 for our licensing revicw.
But we have our own numbers.
And 11 whichever is more conservative, we tend to use the more 12 conservative number.
13 DR. KERR:
Why do you want a conservative number,.
14 rather than a real one?
It seems to me if you are designing 15
. equipment you would want to know what is going on.
16 MR. LEE:
Right.
Well here we have two numbers.
17 One is expected and the other one is a design.
18 Design volumes would be more conservative, our 19 number.
Expected volume would be perhaps from the 20 licensees.
Sometime I noticed that during our review that 21 the licensee's numbers are more conservative and sometimes 22 our numbers are more conservative.
23 They are different somewhat.
It depends on the 24 design.
25 DR. KERR:
Does "conservative" mean bigger?
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MR. LEE:
Larger /
larger volume.
[~h 2
MR. MOELLER:
Well, in essence the first two w) 3 bullets you are simply making sure that their rad waste 4
capacity is adequate to meet the cemand?
5 MR. LEE:
Correct.
So there won't be any overflow 6
in the tank or you have wet waste that you do not know what 7
to do with, that your system is incapable of meeting such a 8
demand.
9 DR. KERR:
Has there ever been a situation which 10 after both the groups reviewed this that the tank did run 11 over anyway?
12 MR. LEE:
I do not recall such a case.
13 Okay.
The third one then, the radionuclida 14 concentrations and the distribution for shielding and
()
15
. occupational exposures.
What are their expected fission, 16 activation and other products that come out from the plant 17 operation.
18 And to sea whether we have a proper design and if 19 we have a proper occupational exposure estimation in 20 accordance with Reg Guide 8.8.
21 MR. MOELLER:
And does your review of their 22 capacity to handle the waste streams consider failed fuel or 23 a standard for failed fuel or what?
24 MR. LEE:
In the case of PWR, we had an in-plant 25 measurement program, we called it.
We went out to operating Heritage Reporting Corporation O
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reactors.
I think there is about eight or nine reactors.
()
2 We went out with the contractor and we actually measured the 3
activities.
4 And we used that as our licensing number.
And, by 5
the way, that if not the conservative number.
That is the 6
actual numbers we measured.
7 In the case of a BWR as well, we did go out to the 8
BWR and measured the activity resulting from actual 9
operation.
And we are using those numbers for failed fuel.
10 We used to use such as
.1 per cent fuel defect in 11 the case of BE5 but we no longer do that. In the revised 12 source term, we use actual operational data.
13 Also we evaluate the design of the system, the 14 structures and the components of design bases and the
()
15
, criteria; whether all these tanks and the pressure vessels, 16 heat exchangers, valves, whether they meet applicable ASME 17 codes for their testing, fabrication, design and inspection.
18 We have those criteria in the Reg Guide 1.143.
We l
19 have a table there that each tank is supposed to meet i
20 particular ASME codes.
21 Then process control program.
The process control 22 program is a control document or a manual containing the 23 operational procedures that a plant is supposed to follow to l
24 produce acceptable waste form.
25 And then last, on-site storage capacity to make sure l
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they have enough wet waste as well as solidified waste, to
()
2 have enough storage in their design.
3 MR. STEINDLER:
In the course of the review of the 4
process and the process equipment, do you address the 5
question of what happens to the quality of the product if 6
there are changes in the process?
7 For example, if your process control system 8
malfunctions?
Or if the feed changes or some other process 9
upset takes place?
10 Is this within your review?
11 MR. LEE:
Yes.
Then they have to revise their 12 process control programs accordingly.
And they submit to us 13 their changes at the later date.
14 MR. STEINDLER:
I'm sorry.
I did not make myself 15
. clear.
16 MR. LEE:
Okay.
17 MR. STEINDLER:
There is likely to be, it seems to 18 me, a variability in any process that you run that may feed 19 back to the variability in the quality of the product.
20 Do you, in the course of investigating the nature 21 of the process, identify and examine the relationship 22 between variability of the process, how it is carried out or 23 what the feed is, in relation to the variability in the 24 product, for example tensile strength or whatever the 25 product has to be?
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MR. LEE:
You mean feedback from our licensees?
(
2 MR. STEINDLER:
No.
In the_ course of your initial 3
review.
4 MR. NICHOLS:
Let me try to clarify that.
5 MR. STEINDLER:
Okay.
6 MR. NICHOLS:
What we do in reviewing the system 7
is to look for limitations imposed on the operation of the 8
system imposed by the licensee in their description, such as 9
pH or maybe some very specific chemical parameters.
10 We don't -- we don't try to correlate those in any 11 way, at least in our purview of the system, with the 12 characteristics that are looked at by NMSS.
13 MR. STEINDLER:
Fine.
14 (Viewgraph presented) 15 MR. LEE:
That represents pretty much the 16 procedure we use at NRR for the licensing activities.
Now I 17 understand that you are particularly interested in the 18 recent event that occurred at Millstone and Quad Cities and 19 TMI-2.
20 So we are ready to discuss briefly what really 21 happened first of all at the Millstone site, j
22 Last July 1987, Millstone decontaminated the 23 reactor water circulation system and the reactor water 24 clean-up system.
This is not their first decontamination i
25 process but this is their --Northeast Utility's second n
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chemical decontamination they had, fh s_)
2 And also they solidified two liners with a spent 3
resin used for processing decontamination wastes using a 4
pulp and cement.
5 Now this spent resin is the strong acid cation and 6
weak base anion and they use that for the Lomi process for 7
the decontamination waste.
8 They solidify that resin with cement.
The waste 9
form is Class A waste in accordance with 10 CFR 61.55.
10 Based on NRC rules and regulations, they did not 11 have to solidify or they did not have to stabilize this 12 decon waste but, our licensee, they did it in order to ship 13 it to the burial site which. requires that anything higher,
14 than 1 per cent chelating agent and/or higher than one hl 15
. microcurie per cc of total activity, the burial site 16 requires to solidify.
17 Therefore, they did solidify in order to meet the 18 burial site requirements, not the NRC requirements.
19 And the cement matrix -- this is after they 20 solidified -- increased in volume by three to four per cent.
f 21 This is the way they measured it in the liner, before and 22 after solidification took place.
23 The liner itself had one to two per cent bulge; 24 one liner.
The other liner was okay.
25 The licensee notified NRC Region 1.
Then we had a Heritage Reporting Corporation O
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meeting. NRR and NMSS research staff met with the licensee ll) 2 and thuir contractor to discuss the incident.
Particularly 3
we were interested in how this process control program 4
worked; how did they implement it, the NRC-approved process 5
control program.
e 6
DR. KERR:
Did the bulging violate some rule or P
7 regulation of was it just a matter of curiosity?
8 MR. LEE:
They did not v*uolate any rules or 9
regulations.
In fact, they followed our approved process 10 control program very thoroughly without any exception to it.
11 MR. PARRY:
Jay?
12 MR. LEE:
Yes?
13 MR. PARRY:
Excuse me, Dr. Kerr.
There is no 14 limitation on the bulging of a liner per se.
There may have I
15 been a problem about getting it in and out of a cask and
[
16 that I am not sure about.
17 But the bulge in and of itself was not a violation 16 of any sort.
19 MR. SHEWMON:
The liner is plastic enough so that 20 it can take a several per cent change?
21 MR. PARRY:
It is 11 gauge steel on the sides.
22 The bottome are flat.
I think maybe reinforced perhaps a 23 quarter of an inch.
24 LN Technologies is here and they can describe 25 that.
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MR. SHEWMON:
Is it plain carbon steel?
()
2 MR. PARRY:
Yes, plain carbon steel.
3 MR. SHEWMON:
It can take a several per cent 4
deformation without bulging?
5 MR. PARRY:
Easily.
6 MR. LEE:
It's about 6.5 in diameter, six feet 7
high.
9 MR. MOELLER:
Go ahead.
9 MR. SMITH:
Well, maybe you mentioned it, but in 10 here you say, your next to the last bullet, the three groups 11
-- NRR, NMSS and RES -- met to discuss the incident.
12 MR. LEE:
Yes.
13 MR. SMITJ!
I want to know what "the incident" 14 was?
()
15 MR. LEE:
Pardon?
16 MR. SMITH:
What was "the incident"?
17 MR. LEE:
Oh, The incident was the bulging of one 18 particular liner.
19 MR. SMITH:
But I understand that is not a 20 problem.
21 MR. LEE:
That's right.
But we were interested 22 again with respect to how did the process control program l
23 work at the Millstone.
How did the process control program 24
-- whether they deviated, for example, from the approved l
25 process control program which might have caused this i
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particular bulge.
()
2 MR. PARRY:
I think it might have helped, Jay, if 3
you had started with the earlier events and showed the 4
sensitivity to bulging and distortion of liners, 5
There were two other events that Jay is going to 6
get to, to discuss that.
7 MR. LEE:
Also, particularly we are looking into 8
any generic implication of the Lomi process with the cement.
9 That was also our secondary interest.
10 MR. STEINDLER:
Let me ask one question.
11 Am I correct in assuming that the handling of low 12 level waste containing chelating agents is done by the 13 Commission on a case by case basis?
14 I mean are there some words to that effect
)
15
,somewhere in the regulations?
16 MR. LEE:
The review, you man?
17 MR. STEINDLER:
Yes.
The licensing of either the 18 process or the product.
19 MR. MOELLER:
Your third bullet says the waste 20 form is Class A.
21 MR. LEE:
Correct.
22 MR. MOELLER:
Now that means the least hazardous 23 class.
24 MR. LEE:
Right.
25 MR. MOELLER:
What are the implications of that?
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Can you have liquids in Class A waste?
()
2 MR. LEE:
I think I will let John answer that 3
question.
4 MR. GREEVES:
No, you cannot have liquids.
5 MR. LEE:
The procedure was complete and still 6
good and still applicable and why this liner bulged, I think 7
the licensee is still looking into it.
8 MR. SHEWMON:
Does this mean that the next time 9
one bulges it is no, never mind, and we don't care how much 10 they bulge because they are following all the regulations?
11 MR. LEE:
So long as the waste classification is 12 Class A.
13 MR. MOELLER:
Do you know that the growth in the 14 volume had terminated and it would not have later split the
(
15 container?
16 MR. LEE:
That particular liner is at site now for 17 some months.
And we were told it is not progressing any 18 worse than what it bulged the first time.
19 MR. MOELLER:
So I guess, Paul, the point here is 20 that in this particular case, because it was Class A waste, 21~
there was really no regulatory problem.
22 However, the experience ties into the experiences 23 Jack pointed out that he is now going to reveal, the Quad 24 Cities and TMI and so forth.
l 25 So in that light it is of interest.
t l
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MR. SHEWMON:
Okay.
f'l 2
MR. PARRY:
So Jay, you might want to describe the J
3 singular contour of the bulged liner as not just merely a 4
hemispherical or smooth contour bulge.
5 MR. LEE:
Right.
. hat is what happened.
6 MR. PARRY:
There was a circumferential bulge 7
going around the liner that was quite localized.
8 MR. LEE:
Right.
9 MR. PARRY:
Relatively speaking.
It wasn't that 10 the whole sides bulged out smoothly.
11 MR. LEE:
Yes.
12 MR. PARRY:
It went down, as I remember the 13 pictures, it went down; there was a bulge like reverse 14 Gibson Girl kind of shape, and then tapered down to the
()
15 bottom.
16 MR. MOELLER:
And so you are saying the waste 17 itself expanded unevenly?
18 MR. PARRY:
Apparently.
19 MR. MOELLER:
Apparently.
20 (Viewgraph presented) 21 MR. LEE:
The next event is the Quad Cities.
This I
22 is more recent than Millstone.
In May 1988, Quad Cities 23 decontaminated the reactor water circulation system.
l 24 Again, this is not their first decontamination 25 they had done, but I think their lith or 12th chemical l
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1 decontamination.
And this is believed to be happened that
.fD 2
first time, w/
3 They solidified only one liner with a spent resin.
4 Again, it is the same type of a spent resin that Millstone 5
used for processing decontamination waste and using the same 6
Portland cement 7
During this process, premature solidification 8
occurred.
And the licensee and the contractor, based on our 9
discussion at the meeting, we found out that the licensee 10 and its contractor did not follow the NRC-approved process 11 control program.
12 This is different from the Millstone case.
They 13 used instead what we call surrogate or simulated waste 14 sample instead of actual waste for making up the process
()
15
, control program sample.
16 At this time, the licensee is evaluating the 17 incident with help of their contractor to determine the 18 potential cause why this premature solidification did occur.
19 And the current status of the liner is that it has 20 been placed in a cask while future actions can be decided.
21 And the licensee will submit to NRC the complete 22 incident report and revise the process control program by 23 the end of July 1988.
24 Now this particular event is different from the 25 Millstone case in that they did deviate from the process Heritage Reporting Corporation O
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control program.
And also they had two different 2
.f,,\\
contractors to perform this work.
3 One contractor did the decontamination work and 4
the other contractor did the solidification process.
While 5
the Millstone case, only one contractor did both the 6
decontamination as well as the solidification.
7 And the only common point is that they did use the 8
same Lomi process.
They~ decontaminated the same system in 9
the BWR.
And they also used the same solidification media.
10 DR. KERR:
What was it that was decontaminated?
11 The prior system?
12 MR. LEE:
Pardon?
13 DR. KERR:
What was decontaminated?
14 MR. LEE:
They decontaminated the reactor water 15 recirculation system in the boiling water reactor.
16 DR. KERR:
Thank you.
17 MR. STEINDLER:
I'm a little confused still.
18 You say the contractor did not follow the NRC-19 approved process control --
20 MR. LEE:
Progratu.
21 MR. STEINDLER:
program.
And then you say 22 that the contractor tried the whole thing out on the basis 23 of a surrogate sample?
24 MR. LEE:
Yes.
25 MR. STEINDLER:
What is it that you folks approved Heritage Reporting Corporation (202) 628-4888
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when the applicant provided for you a description of the
.(O 2
process?
w/
3 MR. LEE:
We approved their process control 4
program to use actual waste that they were about to 5
solidify, not simulated or surrogate samples, which really 6
what they used was fresh resin saturated with Lomi 7
chemicals.
8 MR. STEINDLER:
Does that imply that the program 9
that you approved had never been done prior to submitting 10 the process description to you?
11 MR. LEE:
They do not submit, for example, this 12 decon procedure for prior approval.
Once we approve their 13 process control program, they will be using continuously.
14 And we understood that previously also they did not use
()
15
. actual waste samples but they used a simulated waste sample 16 instead.
17 MR. STEINDLER:
So the process control program 18 that you approve can then be used with any kind of incoming 19 waste to be processed?
20 MR. LEE:
Yes.
21 MR. STEINDLER:
Thank you.
22 MR. MOELLER:
That seems -
go ahead.
l 23 MR. SMITH:
I am a little confused, i
l 24 When you review this process, I presume at that 1
25 time when it is submitted to you by the vondor that the l
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vendor has checked this process out using what might be
()
2 considered actual waste or are you saying they don't.
That 3
you never know what is really going to happen until it 4
happens?
5 MR. LEE:
In the vendor's proposal or contractor's 6
topical report, they do not use actual radioactive waste, of 7
course.
8 They are based on generic waste streams, all types 9
of generic streams that they expect to have at the power 10 plant.
11 MR. SMITH:
So how do you know what you are going 12 to get?
13 MR. LEE:
So really the process control program 14 specifies that you take about three samples from the
/3
(_/
15
,three different samples; well, same sample but three 16 different containers.
17 Then what you do is you add a slightly different 18 amount of solidification media and whatever the 19 solidification requires to see whether they properly 20 solidify before it actually solidifies.
21 And if it does, they can proceed with that l
22 particular formulation.
23 MR. MOELLER:
The way I am hearing it, the process 24 can be approved to treat any type of waste but included 25 within the process is a trial and error series of tests with Heritage Reporting Corporation O
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small aliquots of waste to find out what is the best mix to I'T 2
use.
%)
3 It is like the old jar test.
4 MR. SMITH:
Yes, but it is assimilated.
5 MR. MOELLER:
Well, no.
I thought at that point 6
it was actual waste.
7 Am I wrong?
8 MR. LEE:
In the Quad Cities, they did not use --
9 MR. MOELLER:
That was a mistake there.
10 MR. LEE:
Right.
11 MR. MOELLER:
The fellow misunderstood or he asked 12 for the wrong thing.
13 MR. PARRY:
No.
Excuse me. That is apparently 14 normal procedure, r
(_)/
15 MR. SMITH:
See, that is what I was asking.
16 MR. LEE:
Up to this point, yes.
17 MR. MOELLER:
Well then what is the use in the 18 process of telling them they must take two or three samples 19 of the waste and find out what is the best mix?
What is the 20 use of those instructions if the waste samples they take are 21 not even the real waste?
22 MR. LEE:
That is what actually happened at the 23 Quad Cities and that is what we found out.
24 MR. MOELLER:
But I thought that was a mistake 25 there and that normally they did take actual waste.
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MR. LEE:
No.
In the case of Quad Cities and
()
2 Commonwealth Edison, we understand that they did repeat it 3
that way earlier.
4 In other words, the earlier solidification, they 5
also took simulated waste as well and NRC didn't know about 6
it.
7 MR. MOELLER:
But,your regulations state they 8
should test the real waste, is that correct?
9 MR. LEE:
Correct.
10 MR. MOELLER:
Okay.
Thank you.
11 (Viewgraph presented) 12 MR. LEE:
Okay.
Lastly then what happs.ned at TMI 13 a few years ago --
14 MR. MOELLER:
Excuse me. There was no way ag0in
()
15 that the vendor or the contractor to the utility who is 16 solidifying this waste, there is no way for them to look at 17 it and say hey, this doesn't look like the real stuff?
l 18 or didn't' they monitor it to see if it was 19 radioactive?
j 20 MR. LEE:
They knew.
They knew it was not.
21 MR. MOELLER:
Oh, they knew, too, and figured that 22 is okay.
23 MR. LEE:
Yes.
24 MR. MOELLER:
But what's the use of doing the 25 test?
Why even bother with them?
They are a waste of time, l
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I would think.
2 MR. LEE:
Well, that I think you have to ask to gm 3
the utilities or contractors.
4 MR. PARRY:
Jay, isn't there visual observation 5
into the liners and that is in fact how they have seen a 6
couple of these incidents occur?
7 Isn' t that correct, Jay?
8 MR. LEE:
Correct.
In the Quad Cities case, yes.
9 Okay.
TMI.
I believe the NMSS staff briefed the j
10 committee in more detail last March so we were not prepared 1
1 11 to repeat that again, other than we would like to point out 12 what happened since last March and the current studies of 13 five liners on site, what happened since you did receive a 14 briefing from NMSS.
{}
15 The liner 23, which degraded, was split open. The 16 contents were vacuumed to the HICs.
And also the liner 19 17 was also degraded and the contents were vacuumed to the 18 HICs.
19 The liner 20, 22 and 7, which did not solidify, 20 will be "sluced" (sic) to the HICs.
And these aren't 21 transported to the burial site yet.
And that is what the 22 current conditions are of these liners.
23 And determination of cause.
It is still under 24 investigation by the licensee and the contractor and 25 sometime in the near future we expect to receive their Heritage Reporting Corporation
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reports.
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MR. PARRY:
Excuse me.
I just wanted to clarify.
\\/-
3 Actually one of the five liners -- I think No. 21 -- was 4
solidified and shipped off fairly promptly, within two or 5
three months.
6 And it was only when they came to ship the other 7
next two off that they found the bulged liners, is that 8
correct, Jay?
9 MR. LEE:
Right.
10 MR. STEINDLER:
Is it the general thought that the 11 contents of liners 20, 22 and 7 that were sluiced to HICs --
12 MR. LEE:
Will be sluiced to HICs.
13 MR. STEINDLER:
-- will eventually solidify 14 themselves or are they going to be processed?
()
15 MR. LEE:
So long as they have it in the HICs, 16 they can transport it to the burial site without solidifying 17 it.
18 MR. STEINDLER:
Are these Class A wastes?
19 MR. LEE:
No.
Class B.
But they were put into 20 the HICs.
21 MR. STEINDLER:
There is no requirement to 22 transport and bury only solid waste? I assume these things 23 are semi-solid at best.
24 MR. LEE:
The NRC provision allows you to use HICs 25 as an alternative to the solidifying or stabilizing.
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MR. PARRY:
I am sorry to intrude again, Jay, but
(}
2 3
MR. TOKAY.
Excuse me, but can I interject here?
4 The high integrity container -- by the way, my 5
name is Mike Tokay, for the benefit of the Reporter.
You 6
have to have less than one per cent liquid in a high 7
integrity container so we wouldn't allow transport of 8
liquids in any kind of container.
9 It would have to be solidified before it was 10 shipped.
11 MR. LEE:
But couldn't they just lose it, dewater 12 it?
13 MR. TOKAY:
They could dewater it, but it cannot 14 contain more than one per cent by volume.
I) 15 MR. LEE:
Right.
Right.
They could have 16 dewatered it.
17 MR. SREWMON:
Could you tell me what that word 18 "sluced" (sic) is?
19 MR. LEE:
The transporting of the spent resin with 20 the water.
In other words --
21 MR. SHEWMON:
It is not spelled correctly then.
l 22 MR. LEE:
Okay.
l 23 MR. SHEWMON:
There is an "i" in it.
t 24 MR. SHEWMON:
But you mean you will take an awful 25 lot of water and wash this stuff out and then send all of Heritage Reporting Corporation C,)s (202) 628-4888 l
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that water in a HIC or do you solidify it after that?
2 MR. LEE:
We return sluiced water back to the
{}
3 plant.
In other words, you have a water pressure.
With the 4
water, it flushes the spent resin from the spent resin tank.
5 Put it in a HIC or any other container.
6 Then you dewater it and excess water will return 7
to either spent resin tank or the licensee's liquid 8
radioactive waste management system.
9 MR. SHEWMON:
That is what you mean by sluicing?
10 MR. LEE:
Right.
11 MR. PARRY:
Jay, again excuse me for intruding, 12 but this occurred in 1985.
In 1985, Barnwell would not 13 accept any waste from TMI-2, period.
14 And the waste had to then go to Hanford.
Hanford
()
15 would only accept solidified waste.
So that was why the 16 waste from TMI-2 was being solidified.
17 Now, the regulations or the directive by the 18 Governor or whomever has changed and they are now able to 19 take the material from TMI-2 that has been sluiced out of 20 the previous containers into HICs; has been dewatered to 21 take it below the one per cent liquid limitation; and then 22 can be transported directly and disposed of directly at 23 Barnwell.
24 I believe that is essentially the situation.
25 MR. LEE:
Right.
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MR. MOELLER:
Well, if you took a resin, a tank or
(
2' a HIC full of resin, and had a valve at the bottom and just
%-)
3 drained all the water out of it you could --
4 MR. LEE:
Yes.
You have a filtration system, 5
MR. MOELLER:
Right. Would it be less than one per 6
cent water?
7 MR. LEE:
Yes.
Right.
8 MR. MOELLER:
It will?
9 MR. LEE:
Yes.
10 MR. MOELLER:
Okay.
11 MR. PARRY:
Free water.
12 MR. MOELLER:
Okay.
Free water.
All right.
13 The meeting will resume, and I gather our next 14 topic, then, is spent fuel rod storage and consolidation.
()
15 MS. GILL:
Right.
16 MR. MOELLER:
And Amira Gill.
17 MS. GILL:
My name is Amira Gill.
I work in plant 18 systems branch in division of engineering and system 19 technology in NRR.
The original reviewer is Terao 20 Varjorr.nca, and he's not here today, so bear with me while I 21 go through the presentation.
22 There is a mistake in the slide.
The first line l
23 should be spent fuel storage design basis is one and one
(
24 third full core.
The SRP Section 9.1.2, the standard review l
L 25 plan, Section 9.1.2, requires that each utility have enough Heritage Reporting Corporation
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storage for one full core and one refueling for one plant.
2
'If it's a dual plants, then we require one full core and two
()
3 discharge for -- from refueling.
4 That would take a utility for five years.
In 5
1960s, when these plants started, the idea was we're going 6
to have reprocessing and we don't need more than five years 7
of spent fuel discharge.
However, the philosophy changed 8
since then, and the 1977 federal policy came with no 9
reprocessing.
And the 1982 nuclear waste policy act was 10 that each utility is responsible for its discharged fuel.
11 As a result, all the utilities -- well, most 12 utilities are stuck with more generated spent fuel than they 13 have storage for.
So the alternatives that these -- well, 14 19-- in the 1982 nuclear waste policy act, the utilities
()
15 were responsible for their own discharge.
So the 16 alternatives they had was rerack with high-density fuel, 17 reduction in the spent fuel generation by extending the fuel 18 burnup, or finding independent spent fuel storage 19 installations.
For example, we have Oconee and Surry who 20 went to dry sites, dry fuel sites, as NMSS going to talk 21 about that in details later.
22 And the last option they had was rod 23 consolidation.
24 MR. MOELLER:
Now in some of the material provided 1
f 25 to us, I read, and I read it several times and I couldn't l
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believe it, it said that rod consolidation really did not
()
2 increase your storage capacity.
Now was that -- did I 3
misinterpret it?
4 MS. GILL:
Well, it's -- it did and it did not.
5 It did in the sense that you' re consolidating two spent fuel 6
assemblies to take the -- in one canister that would take 7
one fuel assembly space, and it did not in the sense that 8
you had for example space for 1000 -- that's just an example 9
-- 1000 spent fuel assemblies, and now you have still space 10 for only 1000 reconsolidated storage canister.
11 So in that sense you didn't, and in the sense that 12 you consolidated two, you did.
That's maybe where the 13 confusion is.
14 MR. MOELLER:
But --
()
15 MR. CRAIG:
Roughly you get about a two-to-one.
16 You can store twice as much fuel, and Amira's -- has a slide 17 a little bit later.
The high-density racks, if you will, 18 the rectangular spaces are designed such that they would 19 hold a spent fuel assembly.
When you go through the 20 consolidation process, you basically push the rods into a 21 box that will still fit in that same hole, if you will.
The 22 difference is you' re putting the spent fuel rods from two 23 assemblies, whereas if you have it as the assembly it will 24 only hold one.
25 MR. MOELLER:
So you're doubling up for capacity.
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MR. CRAIG:
Right.
()
2 MS. GILL:
Yes.
3 I'm going to jump just to show what I mean here.
4 This is a consolidated storage canister, which 5
hold two spent fuel asremblies.
So we had two spent fuel 6
assemblies put in this particular canister.
7 MR. CRAIG:
Just as another note, the Department 8
of Energy maintains that NUREG 9
provided you a positive background 10 that has the status of spent fuel storage at each reactor in 11 the country.
Did we provide that, James?
12 MR. PARRY:
Yes, we --
13 MR. CRAIG:
Okay, good.
14 MR. PARRY:
provided that.
()
15 MR. MOELLER:
Now, jumping ahead a speck, in 16 assessing the safety of the consolidation process, you 17 stated, at least what we were provided, project 18 descriptions, spent fuel consolidation demonstration, it 19 stated that 10 CFR 100 and NUREG-0800 applied.
Now what is 20 NUREG-0800?
21 MR. CRAIG:
The standard review plan --
22 MR. MOELLER:
Oh, that's the standard review plan.
23 And now the 10 CFR 100, I believe the important -- and I'm 24 tying in to a discussion we had this morning -- the 25 important fact there is that although that lets you have 25 Heritage Reporting Corporation
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reta whole body to the nearest offsite person, you -- that's
(~'
2 a 25 rem based upon some assumed source term or some release
\\
3 from this fuel.
Does it assume the same as for a nuclear 4
power plant reactor accident, in other words, all the gap 5
activity and half of this, and so forth?
6 MR. CRAIG:
The accident of -- that's -- the 7
postulated accident is a heavy load drop of one spent fuel 8
assembly and the postulation of the release of all the gap 9
activity --
10 MB. MOELLER:
Okay.
11 MR. CRAIG:
-- in a single assembly.
12 MR. MOELLER:
Then the standard release, okay.
13 Thank you.
14 So that should not be misinterpreted in terms of -
()
15
- you can't directly compare that to the five rem --
16 MR. CRAIG:
No.
17 MR. MOELLER:
-- for and so forth.
18 Now one other thing.
I realize I'm getting off to 19 other things, but there were a number of items that worried 20 me about this.
It says in Page 5 of 19 in Part 2 on the 21 safety evaluation of the spent fuel consolidation 22 demonstration, again material that was provided to us, it 23 said for these accident conditions, however, and they're 24 assuming an accident to evaluste, you know, whether it's 25 being properly designed for, it says the double contingency Heritage Aaporting Corporation
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principle of ANSI 8.17-1984 is applied.
This states that
(}
2' one is not required to assume two unlikely independent 3
concurrent events to ensure protection against a criticality 4
accident.
5 Is that well accepted that we never have to assume 6
that two unlikely independent events will occur 7
concurrently?
8 MS. GILL:
I really don't know the answer to that 9
one.
10 MR. CRAIG:
It's my understanding that the --
11 MR. MOELLER:
What is the answer?
12 MR. CRAIG:
-- that that was directed toward the 13 assumed single failure of a system upon which you rely and 14 concurrent with an initiating event, and it's the same kind
()
15 of philosophy that we --
16 MR. MOELLER:
No, excuse me, I think -- I don't 17 think so, and maybe I'm wrong.
Let me read you the next 18 sentence.
19 It says thus, for accident conditions, the 20 presence of soluble boron in the storage pool can be assumed 21 as a realistic initial condition, since not assuming its 22 presence would be a second unlikely event.
23 It doesn't mean that two things can fail at once; 24 it just means people -- it's impossible for people to make 25 two mistakes concurrently.
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MS. GILL:
So they're being more conservative by
()
2 assuming you have boron already in the pool.
3 MR. MOELLER:
No, they're being less conservative, 4
because if the boron is in the pool, you wouldn't have 5
criticality.
6 Yes, you could have it, we -- I don't know.
7 MR. ORTH:
Dade, I read that about four times.
8 MR. MOELLER:
I did, too.
9 MR. ORTH:
I finally got to the conclusion that 10 the first independent event was the actual spill and taking 11 apart of two loads of fuel, which shouldn't happen the way 12 they're dealing with it.
The second unlikely event was the 13 boron in the pool being left out.
14 Now, so if it is a standard condition that boron
()
15 is always in the pool, then somehow removing the boron is an 16 unlikely event.
And I think that it, to me, it sounded like 17 two -- that boron was being used twice, its presence and 18 then its being left out in that double contingency.
But I 19 think that that is one contingency.
The other one was the 20 original initiating event that had been mentioned, then be 21 somehow spilling a couple of fuel pieces uncontrolled.
22 I think it was just poorly worded.
23 MR. MOELLER:
Well, are you -- do you accept it, 24 then, I gather?
25 MR. ORTH:
Well, it -- my little notes to myself Heritage Reporting Corporation s,/
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were I would have to be sure what they've done to ensure 2
that boron's there.
(^)T j
n 3
MR. MOBLLER:
Right.
4 MR. ORTH:
So it's a question of neither accepting 5
or not accepting until we find out is boron every there, 6
because boron wasn't involved in any of the other analyses _
7 that we used in that section, which is why I was going to 8
ask that question.
9 MR. STEINDLER:
In an open plant, say York 10 Canyons, what do you assume when you go through a 11 criticality analysis?
Single failure or double failure?
12 MR. ORTH:
Double.
13 MR. MOELLER:
Sure, and --
14 MR. STEINDLER:
That's basically, I think, Dade's
()
15 point.
16 MR. MOELLER:
-- and what we're doing here, we're 17 putting more and more fuel in a smaller and smaller space.
18 And like it says, if you should have four percent enrichment 19
-- I think I'm reading it correctly -- that the K-effective 20 would be 1.15.
Well, that's a heck of a K-effective.
21 MS. GILL:
Well, you' re not doing it at the 22 beginning of the life of the fuel.
23 MR. MOELLER:
Right, so I realize the -- it's 24 depleted.
25 MS. GILL:
Yes, you've depleted the fuel to a Heritage Reporting Corporation (202) 628-4888 g
145 1.
certain extent, and they do a criticality analysis of these 2
O spent fuel assemblies to see which two they should put 3
together to get a K-effective less than.95.
4 MR. MOELLER:
Well, now if we wanted a detailed 5
discussion of this sometime, whom would we call upon?
6 MS. GILL:
Maybe reactor systems branch.
7 MR. CRAIG:
Reactor systems branch for a 8
criticality analysis.
9 MR. MOELLER:
Okay, thank you.
10 Go ahead, excuse me.
11 MS. GILL:
Now, the process that's presented in 12 this picture is the one that was done in the Prairie Island 13 plant.
They have consolidated 36 fuel assemblies, and put 14 them in storage canister, which takes, you know, each 15 storage canister take two spent fuel assemblies.
16 And the way the process is handled is you bring 17 the spent fuel assembly, which is shown here, and you bring 18 it to where you're going to do the fuel handling steps area, 19 and then you bring what you call the transition canister, 20 which is brought right under it, and then the fuel assembly 21 guide plate on top and on the bottom are cut loose, and then 22 you put the transition canister under this spent fuel
(
23 assembly and you push the rods so that they go from a l
24 rectangular shape to a triangular shape.
5 In other words, you consolidate them such that i
l Heritage Reporting Corporation (202) 628-4888 l
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they' re going down on a 30-degree angle to be consolidated,
()
2 ready to be transferred into the consolidated canister as 3
half the size of what they originally were.
4 After this is done, you transfer the transition 5
canister into the area where the storage canister that takes 6
two fuel assemblies, you transfer the transition canister to 7
where the storage canister is, and that's where you push the 8
what you call the frame up, such that the storage canister 9
and the transition canister frame aro moving up and the 10 storage caniste.i moves up to where the spent fuel in the 11 transition canister is, and you keep moving it up until all 12 the rods are transferred from the transition canister to the 13 spent fuel canister.
14 When this is done, you get like half of this --
()
15 this is the consolidated storage canister, and this 16 represent where two spent fuel assemblies are consolidated 17 into this one storage canister.
18 This shows the storage canister and how the 19 thermohydraulics and how the flow can cools the spent fuel, 20 and that there -- they have open holes in the bottom and the 21 top such that the flow can -- the water can flows upward and 22 give cooling to the spent fuel assemblies.
23 Now, right now in the standard review plan we have 24 the criticality analysis is supposed to be presented in 25 Section 9.1.2, and we don't really have anything that i
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handles consolidation per se.
And Prairie Islano is going
()
2 to come up with a proposal to consolidate the rest of their 3
spent fuel because they don't have any cpace for it.
And 4
that will require extensive review from the staff, and 5
eventually we'll have to come up with new standard review 6
plan 9.1.2, 9.1.3, 9.1.4, and 9.1.5.
7 The licensee has to go through several things to 8
make sure that, you know, we don't -
you don't get 9
criticality, and you have enough coolant in the cooling in 10 the spent fuel pool.
And a few of the things that are 11 prerequisites to do a consolidation like that is you have to 12 cool the spent fuel at least two years after it's been 13 discharged from the reactor, and the fuel has to have at 14 least the minimum burnup required, and they have to go
(
15
.through the calculation for criticality and things like 16 that.
17 And there is a certain number of rods that you 18 cannot consolidate you know, minimum number that has to be r
4 19 there, and also a maximum number.
20 MR. MOELLER:
Why a minimum number?
21 MS. GILL:
The criticality they came up with that, 22 you know, if you have less than this particular number, 23 you're going to have enough moderation of the neutron where 24 you can have criticality.
25 MR. ORTH:
It is an undermoderated case that Heritage Reporting Corporation O
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they're basing it on, but that led to a question by me, is 2
that although you have a minimum number listed as a safety s
3 number, how much fewer does it take before you get K-4 effective of one, for example?
Is there anybody got that 5
number?
6 MS. GILL:
I ov.. ' c, but I'm sure the, you know, 7
the utility and Terao who reviewed it, you know, went 8
through that.
But I don't have the answer.
9 MR. ORTH:
Are we a factor of two off, or are we 10 two rods off, for example?
11 MR. CRAIG:
I don't know the answer to that 12 question.
We'll get back to you.
13 MS. GILL:
Yes.
I think they did the calculation 14 with the particular number of 113 per fuel assembly and 226 15 pe.- the canister, and you get K-effective of.85.
And if 16 ot go below that maybe you exceed the.95 limit that they 17 have.
But I don't know.
18 The other things that the utility has to assure 19 before it consolidate this fuel, that you don't have any 20 leaky or damaged rods that are going to be consolidated, and 21 all the shielding and all the activity is all done five feet 22 under the water level above the fuel assembly top.
23 MR. PARRY:
How are the defective rods detected?
24 MS. GILL:
They have what they call a sipping, and 25 they go through seeing which is a leaky fuel, and visual Heritage Reporting Corporation O
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inspection, too.
(')
2 MR. PARRY:
Rod by rod, or fuel assembly --
3 MS. GILL:
I think #uel assembly by fuel -- but I 4
can't swear to it.
Our reactor systemu branch is the one 5
that follows with that.
6 MR. ORTH:
Does this mean that no assembly with a 7
leaking rod then wiJ' be consolidated, or would they be 8
taken apart and somebody try to sort out the leaky -- the 9
bad rods?
10 MS. GILL:
I think that they would not consolidate 11 any leaky fuel assemblies.
But somehow we have to assure 12 that, or, you know, go through the analysis.
13 MR. SHEWMON:
But it's the individual element that 14 leaks, not the whole subassembly, and the question is if you
)
15
.have --
16 MS. GILL:
Can you remove that rod.
17 MR. SHEWMON:
-- eight by eight or 17 by 70, only 18 one in that subassembly is leaking, then what happens to the 19 other 90 percent of the fuel?
20 MS. GILL:
I vould say that if you have a fuel 21 assemblies -- this is my side now -- if you have a fuel 22 assembly with one leaky rod, you can probably pull it out 23 and put a plugged rod in there.
But --
24 MR. PARRY:
That goes back to my point.
25 MR. CRAIG:
As far as I'm aware of the Heritage Reporting Corporation O
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demonstration of the product, the demonstrations to date 2
have not attempted to consolidate any fuel assembly which 9
3 contain leaky rods.
4 MR. MOELLER:
What percent of our -- of the 5
assemblies discharged from a typical plant are totally not 6
leaking?
lgjlll 7
MS. GILL:
I do not know the answer to that.
8 MR. SHEWMON:
The failed fuel rate is down around 9
a few times 10 to the minus four.
10 MR. MOELLER:
Oh, okay
. it a very low.
11 MR. SHEWMON:
So the odds wculd be pretty good.
12 MR. MOELLER:
Thank you.
13 Cliff?
14 MR. SMITH:
Well, I was just going to ask you, you 15 mentioned one reactor where this is being -- been 16 demonstrated.
17 MS. GILL:
Prairie Island.
18 MR. SMITH:
Prairie Island.
Is that the only one?
19 MS. GILL:
This is the only one that NRR has been 20 involved in.
21 MR. SMITH:
I see.
22 MS. GILL:
Before that there were few done, Surry 23 and Millstone 2 and I don't think anybody went to Surry.
24 MR. CRAIG:
You're going to get to that on Slide 25 21.
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MR. SMITH:
Well, I guess what I was looking for (J~)
2 is someone's going to talk about how successful this whole 3
process gets.
4 MS. GILL:
I guess I could talk about Prairie 5
Island, but I couldn't talk about the others because we 6
weren't aware of them and we didn't attend.
7 MR. MOELLER:
And now the container, once the rods 8
-- and the rods are not open, they remain sealed, and you're 9
consolidating them.
Let's say you want to take them out.
10 Can you readily remove them from the container?
11 MS. GILL:
I don't know the exact answer to that, 12 but I would assume you can flip the top guide, too, and you 13 can.
But I couldn't swear to it.
14 MR. CRAIG:
I would think it would he pretty
()
15 difficult to take out a single rod.
I'm sure you could do 16 it, but they're just crammed in --
17 MR. MOELLER:
Yes.
18 MR. CRAIG:
-- the assembly, and it's just 19 friction that's there, an.d you put two tops on them so they 2
have natural circulation to cool them.
So it might be a 21 little bit difficult.
I -- you could, I suppose, push out 22 one rod, but you get into a -- you' re going to get --
23 possibly run into some problems doing that, because the rods 24 are bowed, etc., as they go in, and that's the purpose of 25 the transition canister, to got the right geometry going in Heritage Reporting Corporation
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to get to the consolidation canister.
.()
2 I don't know that anybody's looked at that at this 3
point.
4 MS. GILL:
Okay.
Since 1982, which was one of the 5
questions asked, consolidation demonstration has been 6-Oconee, Ginna, Surry, Millstone 2, and Prairie Island.
We 7
have witnessed Prairie Island, and I think Millstone 2, was 8
that?
Millstone 2, in our branch.
9 MR. MOELLER:
And again, under whose auspices were 10 the others done?
Was it NRR, but another branch?
11 MS. GILL:
We weren't even involved in the others.
12 MR. CRAIG:
I don't know the answer to your 13 question, since they took place sometime previously, and the 14 demonstration eo far have been done under the provisions of
()'
15 10 CFR 50.59.
I -- it's my understanding that in general 16 project managers were aware, and I assume the NRR technical 17 branches were where the demonstrations were taking place.
18 MS. GILL:
The demonstration at Prairie Island, 36 19 fuel assemblies were consolidated, each had 179 fuels, 14 by 20 14.
Average consolidation rate was two assemblies per 20 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />, and exposure was eight man-rem total.
22 And from the 6444 spent fuel rods handled, only 23 one was bent, and it was at a 30-degree angle, and it's only 24 at the upper 18 when you' re trying to put it into the 25 transition canister.
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During the process with no physical planning
()
2 damage, and the rod consolidation's still an evolving 3
ecology, and is still experimentally --
4 The dry spent fuel storage capacity's going to be 5
' talked about in detail by MMSS by Mr. Robert Johnson.
6 The review a'reas that NRC have to go through is 7
the structure on mechanical, physical, thermal, and 8
radiological characteristic to ensure conformity to the 9
design reviewed and accepted by.NMSS and plant technical 10 specification.
Fuel and heavy load handling is done by NRR 11 by our branch, and to ensure spent fuel integrity and 12 equipment required for safe and shutdown, and decay heat 13 removal are not jeopardized by postulated load drop.
14 We nave,the NUREG-0612, which is handling of heavy
()
15 loads, and it's our branch responsibility.
16 That concludes my presentation.
17 MR. MOELLER:
If this is successful, I presume we 18 could anticipate many utilities adopting the practice.
19 MS. GILL:
Right.
20 MR. MOELLER:
Okay.
21 MS. GILL:
Right.
As a matter of fact, Prairie 22 Island already came proposing to do the rest of their --
23 MR. MOELLER:
Okay.
Didn't one of the Nuclear 24 Waste Policy Act versions require DOE to provide interim 25 storage space, and then I gather that was withdrawn?
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MS. GILL:
That's right.
/~N 2
MR. MOELLER:
Okay.
%s/
3 MR. ROBERTS:
You'll notice our ambitious title 4
here is "Recent Progress in Dry Storage Licensing," but I 5
think a good number of the people here were here about a 6
year and a half ago when we talked'about what was going on 7
in dry storage, and there has been, I think you'll find, a 8
fair emount of progress since then.
9 MR. MOELLER:
What are the competing factors or 10 the advantages and disadvantages that decide whether you go 11 toward dry storage or rod consolidation?
12 MR. ROBERTS:
Probably cost.
Dry storage, if you 13 were a utility on the basis of cost, you'd rerack until you 14 could not presumably rerack any further because that's A)
, enerally the least costly.
(_
15 g
16 I'm not sure of the exact figures because 17 obviously it's still evolving, but it appears that rod 18 consolidation would be less costly than dry storage.
Some 19 people may give you an argument on that -- the vendors --
20 and then dry storage at the reactor site, and as I will 21 discuss, there are a fair array of different types of 22 approaches to dry storage that have evolved in the last 23 several years.
24 And there are also some new technical developments 25 that are continuing to come along and will affect dry Heritage Reporting Corporation
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storage and cost as well.
()
2 MR. MOELLER:
Go ahead, go ahead.
3 MR. SMITH:
John, I was just going to add to what 4
Dade said, though, if you have the lab, utility and you have 5
the lab and given the fact that dry spent fuel storage has 6
been with us awhile and mechanically a little simpler, I'm 7
sure that all comes into play.
8 MR. ROBERTS:
Well, in a way, it does.
In fact, 9
our most recent applicant is Duke Power at Oconee and for a 10 number'of reasons -- they had previously done a 11 demonstration as noted of consolidation -- for a number of 12 reasons involving operational and timing and so forth, they 13 did choose to go to dry storage.
9 14 So it's, I think one of these things that we'll A
(~)
15
.see a lot of variation between various utilities and a lot 16 of different outcomes, and I don't think anybody can really 17 predict exactly how it will all phase in.
18 MR. SMITH:
And the next thing will be that 19 they'll want to consolidate and dry fuel store, isn't it?
20 MR. ROBERTS:
Well, as a matter of fact, I'm going 21 to be talking briefly about that, too.
22 MR. MOELLER:
I quickly, and I don't claim to be 1 23 mathematician, I divided it out and it looked like 250 24 millirem per assembly to consolidate the rods on the basis 25 of the Prairie Island experience.
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What are the exposures associated with dry
()
2 storage, with loading the assemblies and
--?
3 MR. ROBERTS:
Loading a cask is I think less than 4
a man-rem, and that --
5 MR. MOELLER:
How many assemblies, then?
6 MR. ROBERTS:
Twenty-one to 26 assemblies.
7 MR. MOELLER:
Okay, so it's pretty --
8 MR. ROBERTS:
What you' re doing is you have the 9
cask in the water there in the pool area and you load it and 10 you place the top on, and so it's a remote operation.
11 Just going back a ways to refresh our memories 12 here, in November 1980, was when 10 CFR Part 72 was issued 13 in final form by the Commission.
That rule was specific to 14 the storage of spent fuel outside the reactor base.
And it
()
15 was a broad rule.
It covered storage both at reactor sites 16 and at separate sites wet or dry.
17 Now, at the time it was passed back during the 18 Carter era, the assumption was that there would be large 19 water pool storage at separate sites by DOE as the interim 20 storage licensee.
That changed considerably and in fact 21 when the Nuclear Waste Policy Act of 1982 came out, it said 22 the utilities are responsible for the interim storage, with 23 some minor provisions like Federal interim storage as a last 24 resort, and that comes under our 10 CFR Part 53, 25 But in fact, there was a concurrent development i
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going on technically back in the late '70s and so forth, DOE
(}
2 was already doing some R&D with dry storage out at the 3
Nevada test site, concrete silo and drywell work.
4 Of course, Canada had dono dry storage, White 5-Shell and later Chantilly.
There had been dry storage for 6
many years at DOE sites of non-LWR fuel, from the ' 60s on.
7 So there was a range of experience already.
8 And in the Nuclear Waste Policy Amendments Act, 9
Congress specifically told DOE under Section 218A to develop 10 dry storage technologies in support of NRC licensing.
And 11 they then proceeded to do that, and have carried on a number 12 of demonstrations specifically at Idaho and have supported 13 licensing at Surry and Robinson II.
14 One of the things that has resulted out of this is
()
15 various vendors have come to us with topical safety analysis 16 reports.
These are, if you will, similar to an FSAR where 17 you've got a final fixed design, we have performed safety 18 reviews on these, issued safety evaluation reports and
'19 letters of approval.
By letters of approval which don't 20 have really a legal standing, if you will, what the letters 21 of approval say is that tho Staff has reviewed these l
22 designs, and will not rereview them in a site-specific 23 license application barring certain circumstances, such as 24 going beyond the parameters of the original review or 25 modifying the design, or possibly a later regulation change.
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We have thus far reviewed and approved five
(~)
2 designs.
Three of these are cask.
The nodular cast iron v
3 CASTOR V\\21 cask, the Westinghouse ferritic steel cask which 4
is the MC-10, and the NAC stainless steel and lead cask.
5 The ferritic steel and s;tainless steel and lead cask havs 6
solid neutron shields.
7 These casks, generally the CASTOR V contains 21 8
PWR assemblies, the Westinghouse 24 and the NAC 26.
9 Generally speaking the fuel is five to ten years old.
10 And I would like to move -- well, let me say also, 11 just before I do show some photographs here -- we have also 12 reviewed concrete module storage.
This is the NUTECH design 13 which is being implaced at Robinson II, and is also a design 14 that Duke is planning to use at Oconee.
And we have also
()
15 recently reviewed a modular vault storage design which was 16 submitted by Fostcr Wheeler Company, FW Energy Applications 17 in concern with General Electric Company of the U.K.
18 This is similar to dry storage design at the Wylfa 19 Plant in the United Kingdom which was initiated in 1972, 20 And I will show this also a little.
21 MR. STEINDLER:
Isn't that process also akin to a 22 rulemaking, without the benefit of all the trappings that go l
23 with a rulemaking?
l 24 MR. ROBERTS:
The topical report review?
25 MR. STEINDLER:
No.
The topical report review Heritage Reporting Corporation
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with the provision that once approved, you no longer elect
(')
2 to reviow it on a site-specific basis, which is what I
%J 3
thought you said.
4 MR. ROBERTS:
Yes. And it's not akin to a rule.
5 It is a fairly common practice actually for a number of 6
different things in tne systen.
7 What we say is this:
We have performed a review 8
on this specific issue, the component being, for example, 9
the cask or the module.
This is not -- that can then be 10 referenced in the safety analysis report for the site-11 specific application.
12 There are other factors involved in the site 13 specific safety analysis that will obviously have to be i
l 14 reviewed including the questions of the interface, namely,
(')
15 does your site and the proposed use you propose to use this i
16 component on match what the Staff has already reviewed in 17 the topical.
If it doesn't, then it has to be reanalyzed 18 and\\or changes have to be made, and they have to be l
l 19 reanalyzed.
l 20 And I will talk about this a little later, because 21 there are two examples here.
The Surry case was fairly a 22 straightforward case where the CASTOR V was reviewed and it 23 was basically fit into the safety analysis report, and the 24 other specific aspects of the site were reviewed and there,
25 was no appreciable change to the CASTOR V at all, l
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In the case of the Robinson II situation, the
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2 NUTECH design the topical, the design was subsequently 3
modified to match the site and reanalysis was performed.
4 MR._STEINDLER:
Do you believe it is up to the 5
applicant to determine and in a sense certify that the 6
conditions of his site match the criteria for the review of 7
that particular cask, or whatever?
8 MR. ROBERTS:
He has to do it, but we review it.
9 That's what it amounts to.
In other words, he has to prove 10 his case.
-11 (slide) 12 MR. ROBERTS:
This is a CASTOR V cask, about 13 roughly 20 feet long by about eight feet in diameter.
14 Behind it are some earlier versions of nodular cast iron l
()
15 casks employed in the Federal Republic of Germany.
This 16 particular photograph was taken at the Craft Werk Union site 17 in Muelheim, where fabrication of casks occurred.
18 You'll notice the older casks have a square 19 configuration, the fins are also cast on.
The CASTOR V is 20 circular and it's cast in a permanent mold.
Generally, the 21 process involves the multiple pouring of 110 tons of l
22 material in roughly 150 seconds.
The material is then f
23 cooled for about eight days, and then goes through several i
i 24 weeks of heat treatment, about seven.
25 The cask is then fabrication is subsequently Heritage Reporting Corporation
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begun, and you'll notice in this case CASTOR V, the fins are (j
2 machined on..
The entire cask is rotated basically and the 3
fins are machined on, and the entire cask fabrication 4
process'and testing take roughly nine months.
And-5 essentially 80 percent of that time is fabrication testing.
6 This particular cask is out at Idaho as part of 7
the DOE demonstration.
And this cask, the American company 8
involved is General Nuclear Systems, Inc., which is a 9
partnership between Chem Nuclear Systems and Gesellschaft 10 Fuer Nukleur-Service, which is a West German firm.
11 MR. MOELLER:
What is the protrusion at the top?
12 MR. ROBERTS:
What, this?
13 MR. MOELLER:
No, on the circular one over to the 14 right, see the part up on top, what does that do?
()
15 MR. ROBERTS:
Oh, that's another cask.
16 MR. MOELLER:
Oh, I'm sorry.
17 MR. ROBERTS:
Yes.
These cask colors actually are 18 also, I believe red is for the Wurgassen reactor, and I 19 forget, blue and yellow, I believe, are Stadte a:id Obrigheim 20 units, but I forget which is which.
21 MR. MOELLER:
Does this open, then, at the other 22 end or this end or?
23 MR. ROBERTS:
Not, it opens at this end.
You can 24 see here the lid and I believe there are 44 bolts.
It's 25 actually double lidded though.
The outer lid is nine Heritage Reporting Corporation
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millimeters, and there is a -- I'm sorry -- nine centimeters
()
2 and a 29-centimeter inner lid.
3 Well, while I'm talking about it, I have some 4
photographs later, but basically there is a redundant 5
pressure sensing device that will sense the pressure in the 6
inner lid area here, which is an over pressure.
There's 7
under pressure here, and it's all inert helium cover gas.
8 And the seals are metallic seals, aluminum over inconel with 9
inconel spring.
10 These seals then, should either one of these fail, 11 there would be a flow of helium into the interior, or should 12 the outer one fail, a flow of helium out. And the pressure 13 sensing device would detect it.
So basically, you get no 14 release.
()
15 MR. MOELLER:
Where the assemblies are, it's not 16 filled with helium?
17 MR. ROBERTS:
Yes.
18 MR. MOELLER:
Oh, it is?
19 MR. ROBERTS:
Yes.
I will get into that a little 20 later. But all dry storage to date is under inert gas cover.
21 MR. MOELLER:
And is that for corrosion or what?
22 MR. ROBERTS:
Yes.
And there is a particular 23 other factor involved here.
It has been found that 002 can 24 go to U308 under oxidation conditions, and that expands as 25 it undergoes the change, goes to U308, it's a phase change.
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And this can rupture cladding.
Now, one of the designs I'll
()
2 be showing here later -- there's continuing research work on 3
that, and that's one of the technology points I wanted to 4
mention -- but that's why all storage is presently under 5
inert gas.
6 And there is'some work going on to study the 7
mechanisms, and has actually been going on for several years 8
to study this mechanism and to see if there is a threshold 9
or a point at which oxidation affects are so low under 10 temperature conditions that one would not be concerned.
11 MR. STEINDLER:
You say that that's helium they 12 use?
13 MR. ROBERTS:
Yes.
14 MR. STEINDLER:
What's the leak rate of one of
()
15 these monstrosities for helium?
16 MR. ROBERTS:
10-6 millibar leader per second.
So 17 that effectively, it holds up for over the 20 year storage l
18 license period.
l 19 MR. STEINDLER:
It might have been easier to use 20 another inert gas.
21 MR. ROBERTS:
Well, in fact, nitrogen is also used 22 in another design here.
Helium was the initial one, and in 23 the initial emplacement of the lids and so forth, helium 24 sniffing is done to insure that the seal is proper.
25 (slide) l Heritage Reporting Corporation
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MR. ROBERTS:
This particular design is the NUTECH
()
2 horizontal modular storage system, or HUHOMS design.
It is 3
the topical report that I mentioned we reviewed.
And the 4
length of a module is roughly 19 feet by 12 feet high. Wall 5
thickness are about three feet thick here.
The inner 6
canister which contains 7 PWR assemblies is about 38 inches 7
in diameter and about 180 inches long.
This can vary, 8
depending on the type of fuel.
9 The average dose at the surface is about 4 10 millirem per hour.
I say average because obviously there 11 are some baffles here for air circulation through and out.
12 This particular design as I said was employed by CP&L for 13 its license application at Robinson II.
And that particular 14 design, they will be loading fuel in August.
They had
()
15 fabricated three modules within the protected area, and they 16 will be loading in August.
l 17 The design differs somewhat from the topical 18 report design and I might mention this design is quite 19 similar to what is being requested by Duke Power Company at 20 Oconee, except that their design involves a larger canister 21 with a containment of 24 PWR assemblies.
22 And I would mention they also asked for burn-up credit in the cr,ticality design analysis. And that's a i
l 23 24 point I want to get back to later.
25 The Robinson design that they had available for l
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transfer of this 7 PWR assembly, and indeed what it was
[\\
2 sized for, was an IF-300 cask.
Consequently, the transfer s) 3 canister could not be moved in through from the front, 4
because the IF-300 obviously is solid at the bottom of the t
5 cask, and there was no way of providing some kind of 6
hydraulic ram.
Consequently, they had a small opening at 7
the back, and a hydraulic ramp and a docking collar here, 8
the IF-300 moves off docks, the ram grapples the canister 9
and pulls it through, rather than a push system.
10 The Duke system would follow the original design, 11 being a push system.
12 Again, this is a -- there is lead and steel 13 shielding at the ends here, and it's double welded shut, and 14 again helium sniff tested.
Helium is the cover gas in that
(
15
, design.
16 MR. SHEWMON:
These aubassemblies go in without 17 separation and repacking?
18 MR. ROBERTS:
No, there is --
19 MR. SHEWMON:
Or is there consolidation?
20 MR. ROBERTS:
Yes.
They are intact fuel 21 assemblies.
22 Okay, this is the last design we have most 23 recently reviewed.
It is the Foster Wheeler design that I 24 mentioned.
And you'll notice all of these designs are 25 passive.
That is, the cask sits out in the open air, or the Heritage Reporting Corporation
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module sits out'in the open air, and the air, it's a natural
,Pi 2
convection system.
%)
3 This is also.
This is the design that owes some 4
of -its originality to the Wylfa design and this is a passive 5
system of flow.
In fact, if I might try and illustrate that 6
a little later here.
7
.You have an inlet barrier system.
These are fuel 8
tubes in an outlet..And this has, as I say, been in 9
operation on the coast of Wales since about '72.
There were 10 originally three modules, and there are now about five.
11 It's been quite successful.
And we feel that there was a 12 strong basis in experience.
13 There are electronic interlocks in this fuel 14 handling machine design to assure that once you perform a
()
15 particular action and lock on to load and unload, and you've 16 got fuel in and out vertically of the machine, that there is 17 no movement.
Naturally, the whole thing is seismically 18 designed to meet an SSE and Part 100 and also Reg Guides 19 160\\161 criteria.
But it's up to.25G.
20 In general, I mention this because in general, the 21 small module, relatively speaking, the CASTOR V cask is 22 obviously about 117 tons loaded, as a matter of fact, but 23 the smaller modules, the seismic criteria is not really that 24 important when you're considering a cask is being moved, 25 going to be picked up and dropped and slapped down, and Heritage Reporting Corporation O
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tornado, missile impact are some of the things we consider 2
here.
So that but this is a sufficiently large structure
(}
3 that we are looking here at.
4 Now, I would mention that at the time of 5
application, they did propose for storage under air because 6
this is quite a low temperature situation.
The maximum fuel 7
temperature that you would see is a temperature of about 150 8
degrees C.
9 And frankly, we could not see our way clear to say 10 that in-air storage was acceptable.
They had designed for 11 dual use of nitrogen cover gas, and that's what they are 12 using now.
There is, however, work on going at Battelle and 13 Northwest and elsewhere.
14 And there is some evidence that at temperatures at
()
15 least around perhaps 135 degrees C, that you can get a 16 situation where the UO2 in oxidizing goes to -- what is it -
17
- U409, I guess, and it in essence seems to bypass the U308.
18 Bypass may not be the right word, but it goes through
'19 without a phase change, apparently.
20 If that turns out to be a characteristic that we 21 can nail down in the future, and it turns out that there is 22 not this type of phase change that can cause clad rupture 23 and the rate of oxidation at these temperatures is 24 sufficiently low, then it is potentially at least possible 25 that we might have in air storage.
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MR. MOELLER:
Potential for what?
\\
2 MR. ROBERTS:
In-air storage, rather than using an s
3 inert cover gas.
4 MR. SHEWMON:
These are end tight?
5 MR. ROBERTS:
Yes, these are --
'6 MR. SHEWHON:
Let me back up.
Let me finish the 7
question.
This is intact fuel.
8 MR. ROBERTS:
Yes.
9 MR. SHEWMON:
So there's a zirconium zircoloid 10 barrier around all of the UO2, but you'[re not concerned 11 about the oxidation of the zirconium.
You're assuming 12 there's a rupture in the zirconia cladding, and it's that 13 volume change on oxidation of the U02 fuel or whatever that 14 you' re concerned about?
()
15 MR. ROBERTS:
Let me answer that in steps.
16 We do concern ourselves about the oxidation of the 17 zircoloid, but that doesn't appear to be significant.
18 However, one of the things that is very difficult to do is 19 to determine what do you mean when you say, the cladding is 20 absolutely intact and that there are not pin holes, and that 21 there would not be any oxygen leakage into the fuel pellet 22 system.
23 We have for older fuel such as we're looking at 24 which is generally five to ten years old, sipping is not 25 that good a measure.
There is an ultrasonic technique if Heritage Reporting Corporation
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the fuel is still in the pool, Brown-Boveri, for example, 2
which is quite effective, and that has been employed in fact
()
3 at Surry.
4 But -- I'm kind of getting ahead of myself in some 5
of the technical areas -- but basically, what we have a 6
concern about is that wo don't necessarily want to require 7
somebody to say that all their fuel is absolutely intact, 8
because,for storage unde.: inert gas, there's no reason to do 9
so, to put it bluntly.
hhy?
10 But if you go into a question of storage under 11 air, where you've got oxygen present, now you get into that 12 area, and generally speaking, people don't want to set up 13 the process of having you test every fuel assembly or fuel 14 rod.
So that's where this comes into play.
()
15 If it would be obviously much easier if the design 16 and the conditions were such that you were assured that the 17 oxidation was not a problem.
18 And frankly, there is another reason for this work 19 being good work which is being funded by DOE and others, and 20 that is that just in the general idea of fuel handling, for 21 example, the dry consolidation for MRS at the repository, 22 knowing and having firmly established the rates of oxidation 23 and their affects on fuel assemblies and fuel rods where 24 there may be p.3rforations in the fuel, I think has some 25 potential fut*na value.
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Also, it's another area where in an accident
()
2 situation, you might say to yourself, or in a situation 3
where you were leading, and you wanted to lay fuel over for 4
a period of time in air, it would be a valuable area. And 5
that's why this research work continues to be done.
6 MR. SHEWMON:
Is the research work using 7
completely just UO2 pellets, or is it using spent fuel?
8 MR. ROBERTS:
It is using both, both irradiated 9
and non-irradiated fuel pellets.
Work has been going on now 10 for some considerable time.
11 MR. SHEWMON:
Okay, but there is stuff that's 12 moved around some during the irradiation, although these are 13 low energies and so it would be worthwhile making sure that l
14 you got the same kind of phase changes or whatever you want
()
15 to call them --
16 MR. ROBERTS:
That's is the point.
That's what 17 they have been trying to establish.
18 MR. STEINDLER:
This is an entirely passive 19 system, it has no auxiliary fans, or requires --
20 MR. ROBERTS:
No.
This, as far as the fuel is 21 concerned, this is a passive system.
But because of when 22 yo's are loading and unloading and you want to maintain 23 pressure differentials, yes, there is ventilation, there are 24 fans, there are filters in multiple stages to protect the 25 various areas.
So this is different than say the concrete Heritage Reporting Corporation
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module storage or the cask storage.
2 MR. MOELLER:
But do you turn those fans off when
()
3 you've finished?
4 MR. ROBERTS:
No.
5 MR. MOELLER:
You don't.
6 MR. ROBERTS:
They are continually operational.
7 MR. STEINDLER:
I'm a little confused.
You say 8
you don't turn them off?
9 MR. ROBERTS:
That's right.
10 MR. STEINDLER:
Does that mean they' re required in 11 order to maintain the temperature of no more than 150 12 degrees?
13 MR. ROBERTS:
No, no, no.
My point was that this 14 whole area, the cooling of the fuel is strictly natural
()
15 convection.
But the other is for purposes of maintenance of 16 the radiological safety.
17 MR. ORTH:
Btfore we leave all these various 18 casks, do you have any financial data, like what's the price 19 per element for some of these fancy designs, or whatever?
20 MR. ROBERTS:
Okay.
I'll give you a kind of top 21 of the head of $1 million per cask.
But I think that's a 22 little difficult to say because people are getting 23 competitive.
And also with some of the newer designs, some 24 of the designs like the concrete design and so forth, I 25 think potentially the cost would be lower.
And there are Heritage Reporting Corporation
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also some other concrete cask designs that are coming into
(}
2 us, I think.
3 MR. ORTH:
That's why I said, per element, not per 4
cask.
5 MR. ROBERTS:
Yes, I understand.
There are 21 6
assemblies in, for example, there are generally 21 to 26 7
assemblies.
That's where we get into the potential for 8
things like consolidation might make these casks more 9
effective if you were storing twice as many fuel assembly 10 rods.
11 But we generally don't get, you know, cost is not 12 our consideration because --
13 MR. ROUSE:
My name is Lee Rouse.
I'm Chief of l
14 Fuel Cycle Safety Branch, and John and I've been involved in
()
15 this dry storage from the very beginning as far as NRC is 16 concerned.
17 The cost you mentioned, DOE occasionally comes out 18 with some cost comparisons of the difforent technologies.
19 And I think right now, they're using a unit of per KG, and 20 they're comparing somewhere on the order of $50 perhaps up 21 to $100 for KG for dry storage, hoping to see rod 22 consolidation coming in maybe somewhere from 30 to 50.
But i
23 it's sort of a moving target right now, as John says.
24 Some f the dry storages are beginning to bring l
25 that number down perhaps closer to the 50.
Don't think Heritage Reporting Corporation
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m ' ve.
~. men any numbers on rod consolidation, yet.
But it's
()
i
.a that sort of neighborhood.
3 MR. ROBERTS:
Going back then, we have performed 4
two license reviews, Surry and Robinson.
They were both 5
licensed in the summer of ' 86.
And this April, we received 6
an application from Duke Power Company for its Oconee site.
7 And as I mentioned before, Robinson and Oconee would be 8
along the NUTECH design, the concrete module.
Surry is the 9
CASTOR V.
10 (slide) 11 MR. ROBERTS:
Let's see if this shows up fairly 12 well.
Probably not too well, unfortunately.
I think the 13 next two are a little bit better.
But this is a CASTOR V 14 cask being moved out to a single pad thus far constructed at i
()
15 Surry.
The pad is 230 feet long by 32 feet wide and it's 16 about three feet thick reinforced concrete.
It has to be 17 that thick basically to hold up the series of a 2800 ton 18 cask.
Right now, there are six casks out on the pad.
19 (slide) 20 MR. ROBERTS:
This is the cask beginning to be l
l 21 moved up into position.
22 (slide) 23 MR. ROBERTS:
And finally, this is the cask in 24 position.
And you'll notice this is the electrical j
25 connection that comes down from -- this is basically about a Heritage Reporting Corporation O
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500 pound steel rain hat that's put over the top of the cask 2
-- this electrical connection is plugged in and a doubly
(}
3 redundant wiring goes through to the alarm system.
4 As I said, the pressure sensor itself is doubly i
5 redundant.
And there is an overhead light that will flash 6
if you started to see leakage in the interstitial lid space.
7 And then a board could be observed which would tell you 8
which cask, if you were going to see that.
9 I frankly don't think we're going to see it with 10 these metal seals in place.
And that's Surry.
11 MR. STEINDLER:
Before you leave that, heat 12 dissipation is strictly by conduction out through the 13 surface of the concrete?
14 MR. ROBERTS:
It's cast iron.
It's strictly --
()
15 MR. STEINDLER:
It's strictly conduction?
16 MR. ROBERTS:
Conduction.
Yes, the maximum fuel 17 clad temperature is about 370 degrees in the inner most 18 assembly.
19 MR. STEINDLER:
C?
20 MR. ROBERTS:
C.
Technically, the calvulation you 21 should see on the surface of the cask of about 30 millirem 22 per hour but because they tend to put older fuel around the 23 outer rim of that, it's roughly about half that, I believe.
24 And as you can see, it's not too warm to get next to.
25 (slide) l l
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MR. ROBERTS:
Upcoming license applications.
We
(}
2 expect early next year license applications again from CP&L 3
for Robinson and Brunswick.
The Robinson demonstration 4
license that they got, which was a real license, but it was 5
just for a small number of units inside the protected area.
6 This next one will be outside the protected area for a much 7
larger number of assemblies, and they have not specified 8
that. I'm assuming it would be to their advantage to go to 9
the life of the plant as Duke did.
10 Brunswick is a BWR, and again, this would be a 11 variation.
I'm assuming that since CP&L has, I believe, 12 some rights to the NUTECH design that they will g,, with 13 NUTECH.
But that remains to be seen.
14 We are meeting in July with SG&E.
Calvert Cliffs
()
15 had recently gone out with an RFP for two vendors, and they 16 have told us that they will be applying at the beginning of 17 next year,.
They have not yet told us which design they have 18 chosen.
19 And we have a letter in from TVA that in 1992, 20 they will be coming in for storage at Sequoyah.
21 Initially, back quite a number of years ago, TVA 22 was interested in dry storage rather strongly, but they've 23 had other issues to concern them.
24 MR. MOELLER:
And all of the heat is wasted, isn't 25 it?
It would be nice if you could have one in your backyard Heritage Reporting Corporation
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and heat your house?
(V']
2 MR. ROBERTS:
As I mentioned, we'd approved five 3
designs.
We have other designs in right now from Combustion 4
and Sharing Transnuclear, NAC, and the next NUTECH module 5
design.
6 The NAC cask is a variation on their other earlier 7
proved design, and it would store consolidated fuel.
This 8
is potentially for VEPCO Surry, and the idea would not be to 9
store consolidated fuel in the pool, but to store 10 consolidated fuel in the cask, and I think basically they're 11 looking at this option to see what is involved and what the 12 costs are.
Whether they're actually going to go ahead with 13 it, remains to be seen.
But we do have this design in and 14 we are looking at it now.
()
15 Some of the other vendors have been interested in 16 burn-up credit in the criticality analysis. And as you're 17 aware buta-up credit is allowed in I think it's roughly a 38 dozen PWR pools, now.
20 We have not looked at this issue before.
We will 21 be looking at it, and are looking at it in the Oconee case 22 because the design involves it.
But our position has been 23 basically to consider the fuel as either fresh or the 24 highest reactivity, because that -- yes?
25 MR. ORTH:
I was just going to say, under the Heritage Reporting Corporation
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conditions of dry storage, which is what we're talking T'T 2
about, what difference does burn-up make?
(/
3 MR. ROBERTS:
Right.
Here's the difference.
When 4
you load and unload these casks, you do it in the pool.
And 5
how do you guarantee -- you either guarantee the boron 6
content of that water or else --
7 MR. ORTH:
It's the wet case, then, not the dry 8
case?
9 MR. ROBERTS:
Yes.
Yes, right, it's not the dry.
10 And in fact, I think you could make some arguments with some 11 of these designs that critict411ty is not credible during the 12 storage phase, but during tPs loading and unloading, you've 13 got a situation where you're introducing water, it's perhaps 14 turning to vapor.
()
15 So we are looking at this now, and we' re looking 16 at this issue.
We have some work on going at Livermore on 17 methodologies.
In particular, we're looking at the 18 distribution of the burn-up.
Rather than using an average
~
burn-up, we're looking at the question of just how fast the 19 20 burn-up envelope drops off, particularly at the upper end of 21 the fuel assemblies, and che last foot or so, whether you 22 could get a slab that was critical, that sort of thing.
23 We' re also looxing at potential m saar.rement 24 systems.
25 So in conclusion, with the advent of dry storage, i
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we've seen a number of different types of dry storage
{}
2 technology, using everything from nodular cast iron, 3
'ferritic steel, stainless steel and lead to concrete 4
modules, to concrete vaults, to questions of storage of 5
consolidated fuel-to questions of burn-up credit, and to 6
questions of storage potentially "nder air.
7 And we expect that there will continue to be new 8
designs and new approaches.
We also would mention, expect 9
to see some changes institutionally.
The MRS rule is being 10 voted on by the Commission now.
So I expect that will be 11 out reasonably shortly.
Literally, they are voting on it 12 now.
13 We also have a proposed rule which we will be 14 discussing tomorrow on certification for dry storage casks.
()
15 This is the type of stand-alone cask such as thq CASTOR V or 16 the Westinghouse MC-10 or the NAC Storage \\ Transport cask.
17 And that concludes my remarks.
If there are any 18 questions?
19 MR. MOELLER:
John?
20 MR. ORTH:
One question.
Have there been any I
l 21 intervenor action going on with respect to dry storage?
22 MR. ROBERTS:
None whatsoever.
It seems to be the 23 least disliked option.
24 MR. STEINDLER:
When you get an application, what l
25 sort of time limit do you see the applicant coming in with?
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How long is he planning to store the fuel?
2 MR. ROBERTS:
I'd say we' re talking roughly 12
.{ }
3 months now.
4 MR. STEINDLER:
Twelve months?
5 MR. ROBERTS:
I think so.
We've been coming up a l'
6 learning curve but I think 12 months, 12 to 15, maybe.
It 7
depends, we will, as you see, be busy next year.
We've got 8
thrse applications we expect to see coming in.
So if you 9
start getting a real crowding affect, then obviously 10 something's going to have to wait.
11 MR. STEINDLER:
What does the applicant do with 12 the fuel after 12 months?
13 MR. ROBERTS:
Wait a minute?
Are you talking 14 about the time to license?
()
15 MR. STEINDLER:
No.
I'm talking about -- I 16 - thought we were misconnected.
17 MR. ROBERTS:
Oh, Twenty year period is the 18 period of storage for under Part 72, then you can renew.
l 19 You can apply for renewal for up to another 20 years.
20 MR. STEINDLER:
And then after 40 years, they must l
21 go somewhere else?
l 22 MR. ROBERTS:
No.
There's no provision in 72, but l
l 23 hopefully we will have a repository by then.
24 MR. STEINDLER:
So there's, in this fairly l
25 inexpensive attractive storage system, there's a possibility l
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of repeated renewals at 20-year intervals?
2 MR. ROBERTS:
Yes.
g-)
\\_/
3 MR. MOELLER:
Other questions for John Roberts?
4 (No response.)
5 MR. MOELLER:
Okay.
I hear no more.
6 And does that wrap up things from the Staff, or 7
are there ocaer presentations?
8 Okay.
All right, well that wraps up, then, I 9
think, our formal sessions ror today.
Unless I hear any 10 calls for questions or additional discussion.
11 I would suggest that we terminate today's formal 12 meeting, and we'll take a short break and then we'll go into 13 executive session to decide what actions we'll take on the 14 basis of what we have heard discussed.
15 And the Committee then will be resuming its formal
/}
16 meeting at 8:00 tomorrow morning in this same place.
17 (Whereupon, at 4:20 p.m.,
the Committee was 18 recessed, to reconvene the following day, Tuesday, June 28, i
l 19 1988, at 8:00 a.m.,
in the same place.)
20 21 22 23 24 25 l
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1 1
CERTIFICATE o
2 3
This is to certify that the attached. proceedings before the 4
. United States Nuclear Regulatory Commission.in the matter of:
FIRST MEETING OF ADVISORY COMMITTEE ON NUCLEAR WASTE 5
Name:
AFTERNOON SESSION
- 6 7
Docket Number:
8 Place:
Washington, D.C.
June 27, 1988 9
Date:
10 were held as herein appears, and ' hat this is the original
.11 transcript thereof for the file of the United States Nuclear 12 Regulatory Commission taken stenographically by me and, 13 thereafter reduced to typewriting by me or under the direction 14 of the court reporting company, and that the transcript.is a 15 true and accurate recor f the f eg in proceedings.
7S 16-
/S/
,F /,
y v.
17 (Signature typed):
18 Official Reporter 19 Heritage Reporting Corporation 20 21 22 23 24 l
l 25 i
()
Heritage Reporting Corporation l
(202) 628-4888
NRC STAFF PRESENTATION TO T HE O
SUBJECT:
RECENT PROGRESS IN DRY SPENT FUEL STORAGE LICENSING l
l DATE:
JUNE 27, 1988 PRESENTER:
LELAND C. ROUSE 7 30HN P. ROBERTS
)
O PRESENTER'S TITLE / BRANCH DIV.:
BRANCH CHIEF / FUEL. CYCLE SAFETY BRANCH /IMNS SECTION LEADER /lRRADIATED FUEL STORAGE SECTION/IMSB/IMNS PRESENTER'S NRC TEL NO.:
49-23328 49-20608 SUBCOMMITTEE:
O TO BE USED ALL PRESENTATIONS TO THE ACNW BY NRC EMPLOYEES T
O O
O-NRC STAFF PRESENTATION TO THE ACNW
~
SUBJECT:
RECENT PROGRESS IN DRY SPENT FUEL STORAGE LICENSING DATE:
JUNE 27, 1988 PRESENTER:
LELAND C. ROUSE JOHN P. ROBERTS PRESENTER'S TITLE / BRANCH DIv.:
BRANCH CHIEF / FUEL CYCLE SAFETY BRANCH /IMNS SECTION LEADER / IRRADIATED FUEL STORAGE SECTION IMSB/IMNS PRESENTER'S NRC TEL. NO.:
49-23328-49-20608 SUBCOMMITTEE:
TO BE USED ALL PRESENTATIONS TO THE ACNW BY NRC EMPLOYEES i
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Advisory Committee on Nuclear Waste (Brief for Meeting on June 27,1988) b RECENT PROGRESS IN ORY SPENT FUEL STORAGE LICENSING Recent progress in dry spent fuel storage licensing has been rapid.
Five dry storage design submitted for NRC safety review in topical safety analysis reports have been approved for referencing in site-specific license applications for dry storage at reactor sites undeQ CFR Part 72 These include the General Nuclear Systems, Inc. (GNSI), nodular cast iron [ CASTOR V/21 f
{]with a capacity of 21 PWR asseablies, thkliO~ ECH NUH0MS[ concrete and T
stainless steel canister modular system with a capacity of 7 PWR assemblies, andsinceSeptember1987theWestinghous(AC-10feNitic~stee1[ca
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with a capacity of 24 PWR assemblies, tha FW Energy Applications, Inc., Modular Vault PWR or 17(~BWR assemblies per vault module Dry Store with a capacity of 8 !
(approved for up to five modules, and the Nui ear Assurance Corporation
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O steiaiess steel end leed cesk with e cePecity of 28 eWR essembiies.
In addition to water pool storage (as racked approximately 700 MTV capacity) at the General Electric Morris Operation in Grundy County, Illinois, licensed in 1982, there are two at-reactor dry storage installations, licensed in 1986, at the Virginia Powed SurrylNuclear Power Station siteln Surry County, Virginia, j
and at the Carolina P ser and Light Company H. B. Robinson Steam Electric Plant, m
Unit 2, site in Darlington County, South arolina. [ CASTOR V/21/ casks are used attheSurrysitewhileamodifiedh0MS;modularsystemisinplaceatthe Robinson 2 site with initial loading projected for August 1988.
l New dry storage designs continue to be submitted for NRC staff safety review.
In April 1988, NRC received an application from the Duke Power Company (Docket No. 72-4) to apply for a license to store fuel at its Oconee plant site in South Carolina. A letter of intent has also been. received from Virginia Power to amend its license fo' use of a new nodula{ cask f ron cask, the CASTOR X/28-33.j I
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Receipt of a topical safety arialysis report for the CASTOR X/28-33 design is expectedinJuly1988 fro $GNSI.
Future amendments from Virginia Power are
. expected for other cask types also. The Duke application is linked to a new NUTECHfconcretemoduledesignwithacapacityof24PWRassemblies.
Both the Duke and Virginia Power applications rely on burnup credit allowance in l
criticality design.
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Advisory Committee on Nuclear Waste (Brief for Meeting on June 27,1988) psU RECENT PROGRESS IN CRY SPENT FUEL STORAGE LICENSING Recent progress in dry spent fuel storage licensing has been rapid.
Five dry storage design submitted for NRC safety review in topical safety analysis reports have been approved for referencing in site-specific license applications for dry storage at reactor sites under 10 CFR Part 72.
These include the General Nuclear Systems, Inc. (GNSI), nodular cast iron CASTOR V/21 cask with a capacity of 21 PWR assemblies, the NUTECH NUH0HS concrete and stainless steel canister modular system with a capacity of 7 PWR assemblies, and since September 1987 the Westinghouse MC-10 ferritic steel cask with a capacity of 24 PWR assemblies, the FW Energy Applications, Inc., Modular Vault Dry Store with a capacity of 83 PWR or 170 BWR assemblies per vault module (approved for up to five modules), and the Nuclear Assurance Corporation U
stainless steel and lead cask with a capacity of 26 PWR assemblies.
In addition to water pool storage (as racked approximately 700 MTU capacity) at the General Electric Morris Operation in Grundy County, Illinois, licensed in 1982, there are two at-reactor dry storage installations, licensed in 1986, at the Virginia Power Surry Nuclear Power Station site in Surry County, Virginia, and at the Carolina Power and Light Company H. B. Robinson Steam Electric Plant, Unit 2, site in Darlington County, South Carolina.
CASTOR V/21 casks.are used at the Surry site while a modified NUHOMS modular system is in place at the Robinson 2 site with initial loading projected for August 1988.
New dry storage designs continue to be submitted for NRC staff safety review.
In April 1988, NRC received an application from the Duke Power Company (Docket l
No. 72-4) to apply for a license to store fuel at its Oconee plant site in South Carolina. A letter of intent has also been received from Virginia Power to amend its license for use of a new nodular cask iron cask, the CASTOR X/28-33.
I
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Receipt of a topical safety analysis report for the CASTOR X/28-33 desJgn is expected in July 1988 from GNSI. Future amendments from Virginia Power are expected for other cask types also. The Duke application is linked to o new NUTECH concrete module design with a capacity of 24 PWR assemblies. Both the Duke and Virginia Power applications rely on burnup credit allowance in criticality design, i
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