ML20211C009
| ML20211C009 | |
| Person / Time | |
|---|---|
| Site: | 07200022 |
| Issue date: | 08/20/1999 |
| From: | Resnikoff M AFFILIATION NOT ASSIGNED |
| To: | |
| Shared Package | |
| ML20211B978 | List: |
| References | |
| 97-732-02-ISFSI, ISFSI, NUDOCS 9908250140 | |
| Download: ML20211C009 (11) | |
Text
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'FROM i RADIDACTIVE lJASTE M46GErENT PHONE NJ. ; 212 620 0518 Aug. 20 1 99 12:11PL1 F2/3 1
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UNITED STATES OF AMERICA
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BEFORE THE U.S. NUCLEAR REGULATORY COMMISSION ATOMICSAFETY AND LICENSING BOARD f
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In theMatter of
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PRIVATEFUEL STORAGE, L.L.C.
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(Irid ~=t nt Spent Fuel
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Docket No. 72-22-ISFSI f
StorageInstallation
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i August 20,1999 DECLARATION OF DR. MARVIN RESNIKOFF IN SUPPORT OF l
STATE OF UTAH'S SECOND AMENDED CONTENTION Q 1, Dr. Marvin Resnikoff, decla:e under penalty of peajury that:
- 1. I am the Senior Associate at Radioactive Waste Management Associates, a private consulting firm based in New York City. On November 20,1997 and January 16,1998,1 prepared declarations which were submitted to the Licensing Board by the State of Utah in support of its contentions regarding Private Fuel Storage, L.L.C.'s proposed Independent Spent Fuel Storage Installation. I assisted in the preparation of State of Utah's original Contention Q, which wts submitted on November 23,1997. A statement of my qualifiestions was attached to the November 1997 declaration. I also prepared a declaration in suppcc of the State of Utah's Amended Contention Q (July 22,1999), which was subsequently withdrawn.
- 2. I am familiar with Private Fuel Stomse's ("FFS's") license application and Safety Analysis Report in this prceeeding, as well as the applications for the storage and transportation casks PFS plans to use. I am also familiar with NRC regulations, guidance documents, and envimnmental studies I
relating to the transportation, stonge, and disposal of spent nuclear power plant fuel, and with NRC decommissioning requirements.
- 3. I assisted in the preparation of the State of Utah's Second Amended Contention Q. N technical facts presented in the Second Amended Contention Q. are true and correct to the best of my knowledge, and the conclusions drawn from those facts are based on my best professionaljudgment.
- 4. If Second Amended Contention Q is admitted for litigation, I would testify regarding my
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9908250140 990820 PDR ADOCK 07200022 C
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'FROM I RAD 10 ACTIVE.uASTE MANAGEMENT PHOtE NO. : 212 620 0518 Aug. 20 1999 12:12PM P3'3 opinion of the inadequacy of the cask stability provided in the Holtec HI-STORM Topical Safety Analysis Report (" TSAR *), Rev.9, provided to the State by Hollec under cover letter dated July 27, 1999. Second Amended Contention Q provides a summary of the testimony I would give, based on the information that has been provided to date. I would expect to be able to expand upon and refine my testimony, after having an opportunity to review the calculations that underlie the inf tion providedin the TSAR.
Dr. M'a5in'Rii,raofi July 22,1999 1
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.n,..soveceu, go,1,,issi2.sim ISG-12 Buckling ofIrradiated Fuel Under Drop Conditions
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Dry Cask Storage l News and Information l NRC Home Page l E-mail Spent Fuel Project Office Interim Stafr Guidance - 12 1ssue: Buckling ofIrradiated Fuel Under Bottom End Drop Conditions i
Discussion:
Fuel rod buckling analyses under bottom end drop conditions have traditionally been performed to demonstrate integrity of the fuel following a cask drop accident. The methodology described by Lawrence Livermore National Laboratory (LLNL) to analyze the buckling ofirradiated spent fuel assembly under a bottom end drop in their report UCID-21246 is a simplified approach. It assumed that buckling occurred when the fuel rod segment between the bottom two spacer grids reached the Euler buckling limit. The weight of fuel pellets was neglected in the analysis; only the weight of the cladding was considered. Material properties for unirradiated cladding were used. The buckling analysis also neglected the stiffness of the pellets which could have been fused or locked to the cladding. It assumed the total weight of the cladding to be on top of the fuel rod segment between the bottom two spacer grids. In addition, it also assumed that the fuel rod segment between the bottom two spacer grids was pin-connected. The restraint and lateral suppon of the fuel basket stmeture to the fuel assemblies were ignored in the analysis.
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The weight of pellets and irradiated material properties should be included in any end drop analysis. With these changes, the simplistic method of UCID-21246 may not yield acceptable results. For example, the staff conducted calculations using the same methodology as LLNL repon UCID-21246 except irradiated material properties for the clad, and the weight of fuel pellets are included in the calculations. The most I
vulnerable fuel assembly in the LLNL report, a 17x17 Westinghouse fuel assembly, was chosen for this exercise. Euler buckling loads for the clad were calculated using the following formula:
P,=2El/L2 c
where 6
Eclad = 10.47 x 10 psi j
d d
4 4
Iclad = 1/4 x (r - rj ) = 1/4 x (0.187 - 0.1645 ) = 3.85 x 10-4 in4 o
L = 24 inches The results indicate that l of 2 7/2/9910:55 AM
hnp /www.nrc.Eov/OPNrepons'isg12.htm J
' iso-I? U3uci,:*ing cfIrradiated fuel Under Drop conditions P, = 691b Since the weight of cladding and pellets for the 144 inch-long fuel rod is about 4.98 lb, the buckling load in terms of gravitational acceleration (g) is P,,/W = 69/4.98 = 13.86 g This is consiocrably smaller than the 82 g reported in the LLNL report UCID 21246. However, there are several bounding assumptions in this approach which make the results unrealistically low for predicting cladding failure.
Conclusion:
Analyses of fuel rod buckling performed to demonstrate fuel integrity following a cask drop accident yield results which contain a large margin to actual failure. The calculated onset of buciding does not imply fuel or cladding failure. Where such analyses yield unacceptable results, more realistic analyses of
- dynamic fuel behavior are appropriate and acceptable. If the cladding stress remains below yield strengt the fuel integrityis assured.
Recommendation:
If the analytical approach described in the LLNL report UCID-21246 for axial buckling is used to assess fuel integrity for the cask drop accident, the analysis should use the irradiated material properties and should include the weight of fuel pellets.
Alternately, an analysis of fuel integrity which considers the dynamic nature of the drop accident and any restraints on fuel movement resulting from cask design is acceptable ifit demonstrates that the cladding stress remains below yield. If a finite element analysis is performed, the analysis model may consider the entire fuel rod length with intermediate supports at each grid support (spacer). Irradiated material properties and weight of fuel pellets should be included in the analysis.
The appropriate section of Standard Review Plan, NUREG-1536, should be revised to clearly Reflect analytical approach for fuel rod bucking analyses.
. Approved E. William Brach Date -
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7/2/9910:55 AM 2 of 2
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,na miormanon Notice vsay: Fredicted increase in ruel Kod Claddmg Oxidation Page1of3 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-0001 August 3,1998 NRC INFORMATION NOTICE 98-PREDICTED INCREASE IN FUEL ROD CLADDING 29:
OXIDATION Addressees:
All holders of operating licenses for nuclear power reactors, except those licensees who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vessel.
Purpose:
The U.S. Nuclear Regulatory Commission (NRC) is issuing this infonnation notice to inform addressees ofrecent Westinghouse experience with one ofits reactor fuel designs which has exhibited higher than expected rates ofoxidation of zircalloy cladding at high burnups. It is expected
' that recipients will review the information for applicability to their facilities and consider action as appropnate, to avoid similar problems. The material and discussion contained in this information notice are not NRC requirements; therefore, no specific action or written response is required.
Description of Circumstances:
On October 28,1997, Westinghouse notified NRC that modification ofits fuel cladding corrosion model in its fuel rod design code, PAD, to reflect new data on Zircaloy-4 oxidation at high burnup may create compliance issues for its Integral Fuel Burnable Absorber (IFBA) fuel with Zircaloy-4 cladding. The modified code may predict higher fuel temperatures and internal pressures at high burnup conditions. This, in turn, may lead to code results that do not meet the Westinghouse criterion
. prohibiting gap reopening and that do not meet the loss-of-coolant accident (LOCA) criterion in 10 CFR 50.46(b)(2).
The Westinghouse criterion prohibiting gap reopening was approved by the NRC staff for steady-state operation when interna: pressure m the rod exceeds reactor coolant system pressure. The staff approved this criterion in lieu of a criterion requiring that the internal pressure of the fuel rod not
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exceed reactor coolant system pressure. Both criteria have the same purpose, which is to not allow l
separation between the fuel pellet and the cladding late in life; this limits temperature difference between fuel and clad and therefore minimizes maximum fuel temperature. '
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' The acceptance criterion in 10 CFR 50.46(b)(2) requires that the calculated maximum total oxidation i
of the cladding not exceed 0.17 times the total thickness of the cladding before oxidation. Total I
oxidation includes both pre-accident oxidation and oxidation occurring during a LOCA. If this total oxidation limit were to be exceeded during an accident, the cladding could become embrittled. The ladding could then fracture and fragment during the reflood period and lose structural integrity. This c
in turn could compromise the structural soundness and coolable geometry of the core and ultimately the ability to keep the core cooled.
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- Historically, the focus of compliance with 10 CFR 50.46 has been on 10 CFR 50.46(b)(1), " Peak
- Cladding Temperature," which usually is most limiting at the beginning of fuel life. Because the oxidation rate is known to be dependent on temperature, total oxidation was also deemed most severe
- http://www.nrc. gov /NRC/GENACT/GC/IN/1998/in98029.html
-08/20/1999
NRC Infomation Notice 98-29: Predicted increase in Fuel Rod Cladding Oxidation Page 2 of 3 at the beginning oflife (BOL). The contribution of preaccident oxidation to the calculated total oxidation had not previously been thought to be significant, but as measured cladding oxidation thickness in the later stages of assembly senice life increased faster than had been predicted, it became so.
On November 6,1997, the NRC staff, Westinghouse, and the Westinghouse Owners Group (WOG) met in a public meeting to discuss this matter. At that meeting, the WOG stated that it would provide a list of affected plants, the projected dates when each might become vulnerable to potential non-compliance, and details ofits plans to address the issue. The WOG also stated that each affected plant would take appropriate individual actions in terms of reporting pursuant to 10 CFR 50.46(a)(3)ii before the plant reached its projected date of vulnerability.
Westinghouse stated that it planned to perform more detailed assessments for individual plants and to make timely recommendations to each licensee for compensatory actions with regard to the compliance issue. In the longer term, Westinghouse will correct its model in PAD to better account for recent higher burnup oxidation data and will begin using the revised model by August 1998.
The NRC staff found that this approach was adequate to address in the near term the specific problems reported by Westinghouse and that plants with Westinghouse IFBA fuel could continue to operate in compliance with 10 CFR 50.46. The staff noted that the burnup related phenomena, which could result in noncompliance with the midation requirements of 10 CFR 50.46, may not be limited to Westinghouse IFBA fuel but might alTect any Zircaloy fuel used in high bumup applications. The staff also notes that the oxidation-related phenomena discussed in this information notice may affect licensees' compliance with the reportmg requirements of 10 CFR 50.46(a)(3), as well as the performance criteria of 10 CFR 50.46(b).
Discussion:
Westinghouse employs the NRC-approved PAD computer code to evaluate fuel performance. In 1996, Westinghouse found that two cladding-related models in PAD were nonconservative in analyses of fuels at high burnup. It has recently been shown that the effects of these non-conservatisms in the models could lead to nonconservative calculation of post-LOCA cladding oxidation. These analyses are used to show compliance with 10 CFR 50.46 (b), criterion (2).
The first deficient model deals with fuel rod gap pressure. For the last several years, Westinghouse plants have used high-duty fuel rods, with IFBA and Zircaloy-4 cladding in their core designs. The IFBA rods have a boron coating on the UO2 pellet surface. Westinghouse discovered that for higher burn up IFBA fuel, the rod internal pressure buildup attributed to helium released from IFBA was higher than the buildup previously modeled by the PAD code. Westinghouse revised the PAD model to account for increased helium release from IFBA rods and the increased rod pressure buildup resulting from this helium release.
The second deficient model is the Zircaloy-4 cladding oxidation calculation in the PAD code.
Westinghouse corrected the corrosion model for Zircaloy-4 cladding material to address the accelerated levels of corrosion actually being measured for high bumup fuel rods. The measured corrosion levels were higher than had been calculated u;ing the previous oxidation model. Using the corrected corrosion model, Westinghouse interpreted the PAD results to indicate that the degraded thermal conductivity of the cladding due to the higher oxidation levels produced an increase in fuel cladding temperatures and consequent higher clad creep rates. These higher creep rates could, in tum, lead to gap reopening, which would be contrary to a Westinghouse design criterion. In addition, Westinghouse concluded that with potential gap re-opening, degraded thermal conductivity of the fuel pellets due to high bumup further elevated the local fuel temperature.
The accompanying higher stored energy level and the high pre-LOCA oxidation level could, as early as the second half of the second duty cycle, make this higher burnup fuel more limiting with respect 08/20/1999 http://www.nrc. gov /NRC/GENACT/GC/fN/1998/in98029.html
Y NRC Information Notice 98-29: Predicted Increase in Fuel Rod Cladding Oxidation Page 3 of 3 to the LOCA criterion of 10 CFR 50.46(b)(2) than the analysis of record for BOL fuel. Westinghouse further indicated that the gap reopening is a concern not only for IFBA rods, but also for gadolinia rods which contain gadolinia powder mixed homogeneously with UO2 pellets. The gadolinia degrades the thermal conductivity of the fuel pellets, resulting in a higher operatmg temperature of the fuel.
Westinghouse stated that exceeding the criterion prohibiting gap reopening did not directly lead to clad failures. However, fuel rods with gap reopening could be more vulnerable to swelling and rupture during LOCAs and could challenge the 17 percent oxidation limit. Therefore, high burnup or high duty-fuel rods with a tendency toward gap reopening would be more vulnerable under LOCA conditions. Licensees and fuel vendors with other types of Zircaloy clad fuels may wish to consider the relevance of this information to the oxidation models in use for their specific fuels in light of this new experience, which suggests that oxidation levels at high burnup may be more severe than previously expected.
This information notice requires no specific action or written rgsponse. If you have any questions about the information in this notice, please contact one of the technical contacts listed below or the appropriate Office ofNuclear Reactor Regulation (NRR) project manager.
orig /s/'d by David B. Matthews FOR Jack W. Roe, Acting Director Division of Reactor Program Management Office ofNuclear Reactor Regulation Technical contacts: Kulin Desai, NRR 301-415-2835 E-Mail: kdd@nrc. gov Frank Orr, NRR 301-415-1815 E-Mail: fro @nrc. gov Edward Goodwi' RR 301-415-1154 E-Mail: efg@nrmov
'(NUDTX'S Accession Number 98D7290f2~9) http://www.nrc. gov /NRC/GENACT/GC/IN/1998/in98029.html 08/20/1999 k
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STATE OF UTAH OFFICE OF THE ATTORNEY GENER AL oF ?gg p
J AN GR AH AM ATTORNEY GENER AL CAnot CLAwsON Rito RcHARos PAL,AER DEPAuus seems,osarei cw Depwy Ammy cameral cewtor swr June 9,1998 VIA FAX: (609) 797-0909 and FIRST CLASS MAIL
- Mr. Gary Tjerland HoltecInternational 555 Lincoln Drive
'Marlton,NJ 08053 re: Revisions to HI STORM and HI-STAR Cask Licensing Applicat ons
Dear Mr. Tjerland:
As you recall, the State of Utah and Holtec International entered into a Confidentiality j
Agreement on November 7,1997, to enable cenain employees of the Utah Department of Environmental Quality and its contractors to have access to and review proprietary information submitted by Holtec to the U.S. Nuclear Regulatory Commission. The State has now been admitted as a party to the Private Fuel Storage, LLC (PFS) ISFSI licensing proceedings (Docket No. 72-22), and as PFS is proposing to use Holtec as a cask vendor, the status of the HI STAR and HI STORM cask license applications are inextricably linked to the PFS licensing proceeding.
In a meeting (open to the public) between Holtec and the NRC Staff on May 28,1998, i
it became apparent that Holtec must submit revisions to its HI-STORM and HI STAR cask license applications. It is our understanding that Holtec will submit the first revision to the HI-STORM application to the NRC by June 15,1998.
.The purpose of this letter is two fold. First, the State requests that any revised cask license applications or amendments, and other relevant information, such as responses to the Staff's Requests for Additional Information, that Holtec or its representatives submit to the NRC also be sent to the State of Utah. Please use my name as the contact person. Second, the State would like to establish a procedure for using the existing Confidentiality Agreement to allow the State to access proprietary information that Holtec submits to the hTC.
Currently the definition of Confidential Information is limited to the proprietary information 180 East 300 South. Bth Floor, P.o. Bos 140873, $ sit Lake City, Utah 84114 0873 Telephone:(801) 366 0200 Facsimile: (801) 366 0282
Mr. Gary Tjerland
. HoltecInternational
[
June 9,1998 Page 2 -
the State now has m its possession. One method would be to add an addendum to the existing agreement each time Holtec sends additional proprietary information to the State. If you feel that adding addenda to the existing Agreement is an unworkable approach, the State is willing -
to enter into a separate agreement that would allow the State to timely receive proprietary information that Holtec submits to the NRC.
Please do not hesitate to contact me at (801) 366-0286 to discuss the most expeditious way by which the State may be kept appraised of ongoing developments in the Holtec cask licensing proceedings.
Sincere Denise Chancel r Assistant Attorney General s
cc: -
Docket No. 72-22 Service List
)
g a 1 O Hollec Center,555 Lincoln Drive West, Mariton, NJ 08053 HOLTEC Telephone (609) 797-0900 Fax (609) 797-0909 INTERNATIONAL HY OVERNIGHT MAIL July 27,1999 Ms. Denise Chancellor Assistant Attorney General State of Utah 1
160 East 300 South,5"' Floor P.O. Box 140873 Salt Lake City, UT 84114-0873
Subject:
Revision 9 to III STORM 100 Topical Safety Analysis Report
References:
Private Fuel Storage, LLC; Holtec Project No. 70651
Dear Ms. Chancellor:
Enclosed please find one uncontrolled copy of Revision 9 to the HI-STORM 100 Topical Safety Analysis Report (TSAR). Revision 9 also includes changes made in Revisions 7 and 8 during the final licensing interactions between Holtec and the NRC over the past couple of months. A summary of changes for Revisions 7,8, and 9 is included behind the cover letter in the front of Volume 1. Revision 9 completely replaces Revision 6, currently in your possession. Therefore, your copy of Revision 6 should be destroyed or returned to Holtec at your earliest convenience.
Please note that TSAR Revision 9 is non-proprietary, therefore, a revision to the Confidentiality Agreement between Holtec and the State of Utah is not necessary.
In accordance with our previous agreement, Holtec will invoice the State of Utah $1,500 for this two-volume TSAR document under separate correspondence to cover costs associated with the copying and shipment of these large documents.
If you have any questo. ple ; e mtact me at (609) 797-0900, extension 668.
I Sincerely, j
./
^
Brian Gutherman Project Manager
Enclosure:
One uncontrolled copy of Revision 9 to HI-STORM 100 TSAR J
ae s6 Holtec Center,555 Lincoln Drive West, Marlton, NJ 08053 HOLTEC Telephone (609) 797-0900 Fax (609) 797-0909 iNTERNATlONAL Ms. Denise Chancellor State of Utah Document ID PFS018 Page 2 of 2 Document ID: PFS018 cc:
Max Deleng, PFS (w/o encl.)
Jay Silberg, Shaw, Pittman, Potts, and Trowbridge (w/o encl.)
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