ML20236L135
| ML20236L135 | |
| Person / Time | |
|---|---|
| Issue date: | 10/31/1987 |
| From: | Roberts J NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | |
| Shared Package | |
| ML20236L133 | List: |
| References | |
| NUDOCS 8711100149 | |
| Download: ML20236L135 (8) | |
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1 BURNUP CREDIT CONSIDERATIONS IN DRY SPENT FUEL STORAGE LICENSING i
l John P. Roberts i
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J Irradiated Fuel'Section Fuel Cycle Safety Branch Division of Industrial and Medical Nuclear Safety Office 'of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission i
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l; 8711100149 871027 i
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a Burnup credit in criticality design of spent fuel storage racks has been allowed in reactor basin spent fuel storage at some pressurized water reactors (PWRs) for
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a number of years.
Such storage occurs under strict administrative, procedural, and design controls.
A paoer on this topic by Walter Brooks of NRC's Office of Nuclear Reactor Regulation (NRR) is scheduled for a later session of this meeting.2 q
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In recent years, dry spent fuel storage cask vendors, including Westinghouse,
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Combustion Engineering, Nuclear Assurance Corporation, General Nuclear Systems,
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Inc., and Nutech Engineers, Inc., have expressed continuing interest in' allowance for burnup credit in criticality design.
However,.until 1986 we.had discouraged a
such interest on two grounds.
First, few utilities' reactor spent fuel pools are designed to allow credit for burnup.
Thus, there are no authorized proce-dures for fuel assembly identification in pitce at most facilities with respect j
to burnup credit.
Without such a procedural basis, it is difficult to envision l
granting a Part 72 license for the receipt and storage of spent fuel in an onsite
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dry storage cask installation.
Moreover, prior to 1986 we had not identified any active interest on the part of utilities in burnup credit for dry storage.
Second, we could not justify expending our limited resources on reviews involving burnup credit analysis unless future dry storage licensing actions were expected.
However, interest by vendors has been persistent.
In November 1986 an ad hoc session at the American Nuclear Society's Winter Meeting was organized and author-ized on burnup credit for dry storage and transportation casks.
Presentations were given on this issue and on techniques for measurement of burnup and k eff' Convinced that there was some utility interest, at least on the part of Virginia Power and the Duke Power Company, we reexamined our position.
Later developments have justified our reconsideration.
In September 1987, Duke Power Company sub-mitted a letter of intent to apply for a license under 10 CFR Part 72 for dry storage of spent fuel.3 A modified Nutech Engineers, Inc., NUH0MS design expected to employ burnup credit will be submitted in conjunction with the application.4 On October 7, 1987, NMSS staff met with Virgiriia Power and General Nuclear. Systems, Inc., to discuss submittal of a dry storage cask design model CASTOR X employing burnup credit for use at Virginia Power's Surry Independent Spent Fuel Storage Installation.5 1
In January 1987, at the Institute of Nuclear Material Management Spent Fuel Storage Seminar IV, I had noted that we would accept dry storage license appli-cations which employed designs allowing credit for burnup in criticality design, but that it was unlikely that we would accept a ' topical report for such.6 In addition, we had begun efforts in fiscal year'1987 within our office, Nuclear Material Safety and Safeguards (NMSS), to obtain funding to study this issue.
Through NRC's Office of Nuclear Regulatory Research funding was obtained, and work has been initiated at Lawrence Livermore National Laboratory (LLNL) to study this issue through 1988 with the objective of devel ?ing a technical.
position for reviewing calculational methodologies for allowing burnup credit.
i It is our intent that such a technical position would'also be acceptable with respect to transportation cask designs which may be certificated under 10 CFR j
Part 71.
This is necessary:
first, for there not to be differing review criteria for acceptance within NMSS, and second, because some dry spent storage casks now under Part 72 review may be subsequently reviewed for certification under 10.CFR Part 71 for offsite transportation.
The reverse may also occur, since a rule-j making to amend 10 CFR Part 72 to allow for certificated storage casks for use I
under a general license at reactor sites is being initiated as directed by the Congress in Sections 133 and 218(a) of the Nuclear Waste Policy Act of 1982.7 Vendors of some certificated transportation cask types could apply and qualify for 20 year storage cask certification under this rule.
Indeed, submittal of a dual purpose cask design for storage design approval under existing 10 CFR Part 72 and for certification under 10 CFR Part 71 for transportation is being considered.
A design is being proposed for submittal by Combustion Engineering under a contract to be negotiated with the Department of Energy.8 When I agreed to submit this paper, I expected that our work on examining the technical considerations involved in determining how to address burnup credit j
in cask criticality design would have been well underway at LLNL, Unfortunately, 1
funding for this effort was not received by LLNL until late September.
- Thus, it is possible now only to speak in general terms.
l For cask designers and vendors the advantage of allowance for burnup credit is j
clear - significant cost reduction.
A recent stut.y at Sandia National I
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Laboratories approaches cost reduction from a transportation l cask viewpoint i
However, it is clear that the~' reference to cask capacity. increases from 25~or 10 26 assemblies to 31 assemblies for 100 ton class casks types,is applicable-i for at-reactor-site spent fuel storage.
Such. increase'd storage capacity directly drives down cost' even without consideration o'f pot'ential' reduced cask fuel basket costs due io lower poison material costs.
'Given the potential. cost savings,1which,.although their magnitude may be subject
' to discussiun, are clearly real, the next question that must be addressed
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involves safety.
Can the use of credit for b'urnup in atcriticality design be a safe practice?
-l Clearly, in the case of PWR reactor basins, the' answer is positive.
- However, conditions for spent fuel storage in. dry casks-differ'from reactor'. basin storage.
Obviously, the difference to which I refer :is not during the storage phase of cask deployment.
With double-sealed or welded and/or double-Tidded ' construction c
for storage casks, invasion of water into dry cask cavities is precluded.and, it is arguable, not credible.
Thus, for dry casksLduring the storage phase,.
one can generally expect keff < 0.4.
For cask loading and unloading operations, the difference in conditions does raise potential issues that must be adequately. addressed. When water:is being introduced into a cask in a fuel unloading operation,its' density as a steam, vapor, and water mix varies as it fills and cools the fuel in the cask.
During that process there is no guarantee that any boron would be or would remain dis-solved in the steam and water introduced,:as it ~does in PWR reactor basin water (to a level of about 2000 ppm boron).
Moreover, since the state and density of the moderator is. changing and can.be
.j a fraction of.a gram per cubic centimeter, fission reactions are going. through' j
a phase'from dominant-fast to dominant-thermal neutron reactions.
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Does this change matter?
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Ifjone were not taking credit for burnup for.' casks.with closely packed assemblies
.the answer.is generally not.11 I would' note that we have found.this to.be true:
in criticality reviews of submitted storage casks, for example, the recently approved Westinghouse MC-10 cask design.12 However, if one were taking. credit for burnup,.the question needs to be examined, although it can be. argued for q
specific designs and unloading processes that this issue will~not arise-because
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13-steam density will be too low to have a.significant effect on k,7f.
Fissiori products that may be absorbers'in a fully thermalized system could be-scatterers for a less thermalized spectrum.
How significant this matter is,.as
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well as its' time dependence over a 20 year storage period, needs to be' deter-
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mined if one proposes to take credit for fission products in criticality design j
analysis.
Fi.sion product time dependence has, of course, been previously addressed for reactor basins and recently for dry casks.14 I have raised the fission product question simply to demonstrate that there are analytical questions that need to be addressed or readdressed with respect to-burnup credit in cask criticality design.
In raising, examining'and answering' 1
or dismissing potential questions, we may also note that the definition of burnup credit, or rather what may be allowed for credit, is flexible.
One can approach this subject, as has been recognized for many years,15 by considering three pro-cesses:
(1) uranium depletion, (2) plutonium build-up and (3)-fission product generation.
For discussion let us assume that' spent fuel in dry storage has decayed at least five years.
Then we might assume that, effectively, only the last of these processes continues to have a significant time dependence for longer-lived isotopes if we conservatively ignore Pu-241 decay.
Alternatively, one could ignore the presence of fission products altogether and define allowance for burnup credit in terms of uranium depletion and plutonium build-up.
One could also conservatively assume a11' plutonium is Pu-239.16 One problem raised with respect to these approaches is that they beg a more fundamental question in that they implicitly assume that we have an accepted value for fuel burnup. While an average value is certainly-determinable based-on fuel assembly reactor histories, is it a dependable parameter for use in criticality design? Burnup varies axially within each assembly fuel rod.
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i variation is large at the top and bottom ends of the active fuel region of a rod.17 In results from an Electric Research Power Institute / Babcock & Wilcox Cooperative Program on PWR Fuel Rod Performance report, a figure of burnup versus fuel rod length illustrated this.
The burnup.for fuel (taken from Oconee Nuclear Station E0C-2) was at 12 inches from the top of the active fuel l
region about half the average burnup cited, 24,500 MWD /MTU.18 That value drops rapidly with position to the top of the fuel rod. What is the significance of this? We have a sizeable region of fuel rod with a relatively l
low burnup.
If the thickness (i.e., fuel rod length) of the area at the upper extreme of fuel assemblies is sufficient, then it constitutes a fuel volume l
where designing with credit for a higher average burnup value may not be accept-able.
Unpublished calculations indicate that for the assumptions of zero burnup and an infinite (horizontal) slab of about 4 percent enriched PWR fuel assemblies, i
an upper active fuel rod length of about 6 inches is sufficient for criticality.19 j
Earlier published work for 2.7 percent enriched oxide fuel corresponds to these results.20 I cite this value only as a very general guide to show that the i
process of defining burnup for credit requires careful examination for spent fuel storage casks.
This we shall proceed to do in developing a technical position.
This process of development does not mean that, in the interim, we shall not accept license applications for designs employing burnup credit.
On the con-trary, we believe that the experience gained in addressing the technical aspects of analyzing for burnup credit for specific designs will enhance our generic position development efforts.
Our early discussions with license applicants interested in burnup credit have reenforced this belief.5,21 Also, in conducting licensing reviews we shall be able to address procedural issues of fuel iden-tification and handling including k sensitivity to error in assembly selec-eff tion.
In this regard, we also welcome continued development of techniques for measurement of burnup and k Such techniques, in concert with analysis, I
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would address present safety concerns.
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References 1,
Robbins, T.R., "Taking Credit for' Fuel Burnup in Spent Fuel Rack Design and Criticality Safety Analysis," Proceedings Third International Spent Fuel Technology Symposium / Workshop ~, pp. W-197 - W-217, Seattle, Washington, USA, April' 8-10, 1986.
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2.
Nuclear News, Vol. 30, No. 12, p. 52, Mid-September 1987.
3.
Letter to U.S. Nuclear Regulatory Commission, Attention:
Document Control Desk from H.B. Tucker, Vice President, Duke Power Comnany, and William J.
j McConaghy, Vice President, Nutech Engineers, Inc., dat.ed September 14, 1987.
i Available at the NRC Public Document Room, 1717 H Street NW.,. Washington, DC, under Docket No. 72-4.
4.
-Memorandum with proposed Agenda for NUTECH/NRC Meeting of. October 15, 1987,
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telefaxed to John Roberts from W.J. McConaghy dated October 5, 1987.
Available at the NRC Public Document Room, 1717 H Street NW., Washington, DC, docketed under Project No. M-49.
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5.
' Memorandum to Leland C. Rouse, Chief, Fuel Cycle Safety Branch, from John P. Roberts, Section Leader, Irradiated Fuel.Section, dated October 17, 1987.
Available at the NRC Public Document Room, 1717 H Street NW.,
1 Washington, DC, under Docket No. 72-2.
6.
Roberts, J.P., " Dry Spent Fuel Storage - Licensing Update," Proceed' ings l
Institute of Nuclear Materials Management Spent Fuel Storage Seminar IV, Washington, DC, January 21-23, 1987 (Paper republished in the Journal of the Institute of Nuclear Materials Management).
7.
Nuclear Waste Policy Act of 1982, Public Law 97-425, January 7, 1983.
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8.
Letter to H. L. Thompson, Director, Office of Nuclear Material Safety and
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Safeguards, USNRC, from A. E. Scherer, Director, Nuclear Licensing, Combus -
tion Engineering, Inc., dated September 11, 1987.
Available at the NRC i
Public Document Room, 1717 H Street NW., Washington, DC, under Accession l
No. 8709150216.
I 9.
Sanders, T.L., Westfall, R.M., and Jones, R.H., Feasibility and Incentives for the Consideration of Spent Fuel Operating Histories in the Criticality Analysis of Spent Fuel Shipping Casks (SAND 87-0151), August 1987.
l 10.
Ibid, p. xii.
11.
American National Standard Design Criteria for an Independent Spent Fuel Storage Installation (Dry Storage Type) (ANSI /ANS-57.9-1984), Appendix E, i
- p. 50, December 31, 1984.
4 12.
" Topical Safety Analysis Report for the Westinghouse MC-10 Cask for an Independent Spent Fuel Storage Installation (Dry Storage)" (WCAP-10741),
Revision 1, p. 3.3 - 38, December 1984.
Available at the NRC Public Document Room, 1717 H Street NW., Washington, DC, docketed under Project No. M-41.
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" Topical : Safety Analysis Report. forithe Comuustion Engineering' Dry fCap
. Cask for..an Independent Spent Fuel-Storage Installation (Dry Storage)" '
(CENPD.No. 273-NP, Rev. 03NP)..p.' 3-35, August 1987.
Available'at the NRC' Public Document Room,'1717 H Street'NW., Washington, DC docketed under 1
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Project No. M-43..
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14, Sanders, op. cit.,' Appendix B.
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15.
Shappert, L..B.', " Cask Designer's Guide, A Guide for th'e Design, Fabrication,.
u and'0peration of Shipping Casks for Nuclear Applications," p. 192, February 1970.
l 16.
Ibid.,'p.'193.
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17.
"Poolside Examination of PWR Demonstration Fuel Assemblies,and Creep-
- Specimens -- End of Cycle 2 Fuel ' Rod Performance, Key. Phase ' Report No.
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Babcock & Wilcox, p. 2-9, August 1978.
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18; Ibid., p. 1-1.
19.
Personal Communications',:.C. R.:Marotta, USNRC, and W. R. Lloyd, Lawrence cH Livermore National Laboratory.
H 20.
Thompson, T. J. and Beckerley, V. G.,.The Technology ' f-Nuclear Reactor:
o Safety, Paxton, H. C., and Keepin, G. R., Nuclear Safety of Fuel 0utside Reactors, p. 279, The MIT Press, 1974.
21.
Memorandum to Leland C. Rouse, Chief, Fuel Cycle Safety Branch,'from.
3 John P. Roberts, Section Leader, Irradiated Fuel Section, dated October 23, j
1987. 'Available at the NRC Public Document Room, 1717'H Street NW.,
Washington, DC under Docket No. 72-4.
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