ML20210L074
| ML20210L074 | |
| Person / Time | |
|---|---|
| Site: | 07200022 |
| Issue date: | 08/05/1999 |
| From: | Marco C NRC OFFICE OF THE GENERAL COUNSEL (OGC) |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| CON-#399-20713 ISFSI, NUDOCS 9908090031 | |
| Download: ML20210L074 (45) | |
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Augiisi S,I1999 i
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 99 AUG -6 Pl2.I9 BEFORE THE ATOMIC SAFETY AND LICENSING BOMtD AO2 '
In the Matter of )
)
PRIVATE FUEL STORAGE, LLC ) Docket No. 72-22-ISFSI
)
(Independent Spent )
Fuel Storage Installation) )
NRC STAFF'S RESPONSE TO STATE OF UTAH'S REQUEST FOR ADMISSION OF LATE-FILED AMENDED UTAH CONTENTION O l
INTRODUCTION Pursuant to 10 C.F.R. s 2.714(c), and the Atomic Safety and Licensing Board's " Order (Granting Filing Extension Motions and Setting Schedule For Responses to Request For Admission of Late-Filed Contention)"(Board Order), dated July 27,1999, the staff of the Nuclear Regulatory Commission (Staff) hereby files its response to the " State of Utah's Request For Admission of Late-Filed Amended Utah Contention Q" (Late-Filed Contention Q), filed July 22,1999. Forthe reasons set forth below, the State's Late-Filed Contention should be rejected.
BACKGROUND The State of Utah's original proposed Contention Q (" Adequacy of the ISFSI Design to j Prevent Accidents") asserted that "[t]he Applicant has failed to adequately identify and assess potential accidents, and, therefore, the Applicant is unable to determine the adequacy of the ISFSI design to prevent accidents and mitigate the consequences of accidents as required by 10 C.F.R.
Q 72.24(d)(2)." Utah Contention Q at 114. The basis for this contention addressed the Applicant's 9908090031 990805 PDR ADOCK 07200022 C PDR S]O
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accident analysis of a cask drop event. See SAR i 8.2.6. The Board rejected Utah Contention Q in 1
~ l its initial ruling on contentions, on the grounds that:
this contention and its supporting bases fail to establish with I specificity any genuine material dispute; impermissibly challenge the Commission's regulations or rulemaking-associated generic determinations; lack materiality; lack adequate factual or expen opinion support; and/or fail properly to challenge the PFS application. l
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l See Private Fuel Storage, LLC. (Independent Spent Fuel Storage Installation), LBP-98-7,47 NRC '
142,195 (1998). In addition, the Board found no basis for a portion of the contention penaining to the Intermodal Transfer Point (ITP), stating that "the basis for the contention concems purponed accidents involving storage casks rather than shipping casks, the latter being the casks that would be handled at the ITP." Id. at 195 n.17.
On July 22,1999, the State submitted Late-Filed Contention Q. In support of that contention, the State relies upon a recently issued document, Interim Staff Guidance-12 (ISG-12), entitled
" Buckling of Irradiated Fuel Under Bottom End Drop Conditions," which discusses various deficiencies the Staff identified in a Lawrence Livennore National Laboratories (LLNL) report, UCID-21246. See Late-Filed Contention Q at 1. The State asserts that the Applicant relies on the -
LLNL Report in its analysis of potential accidents that may damage the integrity of the spent fuel i cladding, and that a new analysis is required in light of ISG-12. Id. at 2,9.
For the reasons set forth below, the Staff opposes the admission of late-filed Utah Contention Q.
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3 DISCUSSION A. Legal Standards for Late-Filed Contentions.
The criteria to be considered when determining the admissibility of a late-filed contention are set fonh in 10 C.F.R. 6 2.714(a)(1)(i)-(v).8 It has long been held that the first factor, good cause for lateness, carries the most weight in the balancing test. See State ofNew Jersey (Department of Law and Public Safety's Requests 1 ated October 8,1993), CLI-93-25,38 NRC 289,295 (1993). Funher, in instances, such as here, where a contention purponedly is based on the existence of a document recently made publically available, an imponant consideration in assessing good cause for lateness is the extent to which the contention could have been submitted prior to the document's availability.
See Public Senice Co. of New Hampshire (Seabrook Station, Units 1 and 2), ALAB-737, 18 NRC 168,172 n.4 (1983); Private Fuel Storage, L.LC. (Independent Spent Fuel Storage Installation), LBP-98-29,48 NRC 286,292 (1998).
' The five factors are:
(i) Good cause, if any, for failure to file on time.
(ii) The availability of other means whereby the petitioner's interest will be protected.
(iii) The extent to which the petitioner's panicipation may reasonably be expected to assist in developing a sound record.
(iv) The extent to which the petitioner's interest will be represented by existing panies. l (v) The extent to which the petitioner's panicipation will broaden the issues or delay the proceeding.
l 10 C.F.R. I 2.714(a)(1).
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In evaluating the five lateness factors, two factors -- the availability of other means to protect
' the petitioner's interest and the ability of other panies to represent the petitioner's interest - are less l t
1 important than the other factors, and are therefore entitled to less weight. Texas Utilities Elec. Co.
(Comanche Peak Steam Elec. Station, Units 1 and 2), CLI-92-12,36 NRC 62,74 (1992). With respect to the third factor (the' potential contribution to the development of a sound record), the
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petitioner is obliged to " set out with as much particularity as possible the precise issues it plans to cover, identify its potential witnesses, and summarize their proposed testimony." Commonwealth Edison Co. (Braidwood Nuclear Power Station, Units 1 and 2), CLI-86-8,23 NRC 241,246 (1986),
quoting Mississippi Power and Light Co. (Grand Gulf Nuclear Station, Units 1 and 2), ALAB-704, 16 NRC 1725,1730 (1982). In addition to the showing that a balancing of the five factors favors '
intervention, a petitioner must also meet the requirements for setting forth a valid contention.
10 C.F.R. 5 2.714(d)(2).
B. The State Has Failed to Establish Good Cause For the Late Filine of Contention O.
The State contends that it has good cause for the late filing of its contention because the State's witness for this contention, Dr. Resnikoff, discovered ISG-12 within a reasonable time ofits public issuance, and he and the State's attomeys have taken a reasonable amount of time to prepare the new contention. Late-Filed Contention Q at 6. Specifically, the State asserts that Dr. Resnikoff discovered the document on the NRC's web site on July 2,1999, and that the State submitted the contention slightly more than a month after the document was made publicly available. Id. l These assertions do not demonstrate good cause for filing Contention Q late. First, th'e State does not provide the date when it first leamed of the information detailed in ISG-12. Indeed, as set i
forth below, ISG-12 is based on the Staff's review of the LLNL document -- dated October 1987 --
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and no reason has been shown that Dr. Resnikoff could not have assisted the State in formulating its original contention based on his own review of the LLNL report. In this regard, the record establishes that the State has been aware ofissues closely related to ISG-12 for quite some time. Second, the State does not demonstrate that cenain ponions of its new contention are dependent upon the information contained in ISG-12, such that the State could not have raised the issues in the contention with any degree of specificity prior to the Staff's issuance of ISG-12.
Interim Staff Guidance-12 discusses a particular methodology used to analyze fuel rod buckling following a cask bottom end drop accident. See ISG-12, attached to Late-Filed Contention Q as Exhibit 2. In particular,ISG-12 states that the methodology described in LLNL report UCID-21246, neglected the weight and stiffness of the fuel pellets. In ISG-12, the Staff indicated that it had calculated a buckling load using LLNL methodology, factoring into the equation the irradiated propenies for the material and the weight of the fuel pellets. Using the most vulnerable fuel assembly (a 17x17 Westinghouse fuel assembly), the result was 13.86 g, rather than 82 g as had been calculated by LLNL. See ISG-12, at 1. In other words, the fuel assemblies might not be able to withstand as great an end drop force as had previously been anticipated. Accordinglf, ISG-12 .
recommends that if the methodology used in the LLNL repon is used to assess fuel integrity for the cask end drop accident, the analysis should include the weight of fuel pellets and irradiated material properties.2 2
The Staff funher observed, however, that the LLNL methodology is a simplified approach
-- that is, there are several bounding assumptions in this approach which make the results unrealistically low for predicting cladding failure. ISG-12, at 2. The use of realistic methodology, which would not contain such a large margin to actual failure, may yield acceptable results, even taking into account fuel pellet weight and irradiated material propenies. See id.
e Notwithstanding the recent issuance of ISG-12, the State of Utah and its witness, Dr. Resnikoff have long been aware of the methodology detailed in the LLNL report, and they have expressed their concerns with the adequacy of the LLNL report over the past 17 months. First, by letter dated February 27,1998, Dr. Resnikoff expressed his concerns regarding the LLNL repon to the . Director of the Spent Fuel Project Office.' In his letter, Dr. Resnikoff stated that, "the most vulnerable fuel cannot withstand a 63 g force in the most adverse orientation (Holtec TSAR, p. 3.5-1) but a force considerably less." In particular, Dr. Resnikoff expressed concerns regarding irradiated material propenies, such as cladding ductility and yield stress. Funber, regarding the weight of the fuel pellets, Dr. Resnikoff stated that "LLNL's calculation for most vulnerable fuel also does not take into account the weight of the fuelitself, only the g force without the additional weight of the fuel."
Thus, the State has been familiar with the LLNL repon for at least the past 17 months, and has expressed particular concern with respect to the repon's treatmeut of fuel pellet weight and irradiated material propenies. ISG-12 concerns these same matters, the only difference being that ISG-12 addresses these matters as they penain to the LLNL end drop analysis only.
Further, Dr. Resnikoff's concerns over fuel pellet weight and irradiated material propenies -
of spent fuel have been the subject of ongoing correspondence between the State's witness and the Staff for the past year, following the Staff's receipt of Dr. Resnikoff's letter of February 1998. The Staffinitially responded to Dr. Resnikoff's letter on November 19,1998.' At that time, approaching See Letter from Marvin Resnikoff, Ph.D., Radioactive Waste Management Associates (RWMA), to Charles Haughney, NRC, dated Febmary 27,1998, attached hereto as Exhibit 1.
Dr. Resnikoff's letter indicates that carbon copies were sent to Diane Curran and Connie Nakahara, attorneys for the State of Utah.
d Letter from Mark S. Delligatti, NRC, to Dr. Marvin Resnikoff, RWMA, dated November 19,1998, attached hereto as Exhibit 2.
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the issue as a side drop concern, the Staff indicated its view that the LLNL report delineated
" irradiated fuel cladding longitudinal tensile strength values" and "used the proper weight value in the analysis of the side drop orientation." Dr. Resnikoff responded by letter dated December 31, 1998, in which he stated that the NRC letter did not fully answer his concerns, and he expressed concerns that the LLNL repon failed to consider the brittleness ofirradiated fuel cladding and the effects of dynamicloading.5 The Staff responded to this communication by letter dated February 17, 1999,in which it invited Dr. Resnikoff to submit comments during the public comment period of the HI-STAR 100 storage cask 10 C.F.R. Pan 72, Subpan K rulemaking process.6 On March 26,1999, the State of Utah, with the assistance of Dr. Resnikoff, submitted comments on the proposed rule to add the Holtec HI-STAR 100 Cask System to the list of approved spent fuel storage casks under 10 C.F.R. Pan 72. 7 The State, in its comments, raised several issues penaining to the cladding integrity of the fuel on impact. See Attachment to Exhibit 5 at 2-6. Therein, the State assened that the LLNL repon is deficient because it does not account for irradiation and embrittlement, which lower the impact resistance of the fuel assemblies. See Attachment to Exhibit 5 at 5. The State also took issue with LLNL's " assumption that the fuel within the cladding behaves as a rigid rod." .
In sum, the issuance ofISG-12 did not provide the State with any information that it did not l
already have, or could not have reasonably discovered, and does not excuse the lateness of this late-5 Letter from Marvin Resnikoff to Mark S. Delligatti, dated December 31,1998, attached hereto as Exhibit 3.
- Letter from Mark S. Delligatti to Marvin Resnikoff, dated February 17,1999, attached hereto as Exhibit 4.
7 See Letter from Denise Chancellor, Assistant Attorney General, OfE e of the Attorney General, State of Utah to Secretary, NRC, dated March 26,1999, attached hereto as Exhibit 5.
filed contention. The State has long been aware of the LLNL report, and has previously expressed
- concern to the Staff about deficiencies it perceived in the repon. The state has not shown that it could not have formulated a contention expressing these concerns prior to July 1999. Accordingly, the State i
has not shown good cause for filing its contention late.'
Moreover, even assuming that the State could not have discovered the deficiencies in the LLNL methodology prior to the issuance ofISG-12, it is apparent that the State's contention is not wholly dependent upon the information contained in ISG-12 and, thus, some issues could have been raised earlier. Thus, Late-Filed Contention Q, apan from its discussion of the LLNL analysis,(a) appears to constitute a reply to the Applicant's argument, made in response to the State's original contention, that the canister could act as a replacement for the cladding, and (b) reassens the State's l I
original basis for its former Contention Q concerning lifting accidents during transpon. See Late - )
Filed Contention Q at 4-6. This effort to reassert issues that were raised and resolved previously should be rejected. First, the basis for former Contention Q, like the entire contention, has already been rejected by the Licensing Board. See LBP-98-7,47 NRC at 197 & n.17. The State does not explain how the defects in the LLNL methodology add anything new to that basis. Second, as to the -
State's apparent reply to the Applicant's argument that the canister could act as a replacement for cladding, the State had an opportunity to raise this matter previously upon receiving the Applicant's pleading, but did not pursue the issue at that time. The State's inclusion of this concem in its late-filed contention does not at all depend upon the Staff's issuance of150-12, conceming deficiencies 8
I Indeed, the State in its Late-Filed Contention Q, states its belief that ISG-12 appears to have been issued "in response to the very issues raised by the State's contention." See Late-Filed Contention Q at 4 n.2. Regardless of the origin of the Staff's concems identified in ISG-12, however, it is clear that the State's assenion undercuts its claim of good cause for lateness, in that it demonstrates the State has been aware of these issues for some time.
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found in the LLNL methodology, and the State's delay in raising this matter is thus altogether unsupported by a showing of good cause.
i Accordingly, for the reasons set fonh above, the State has not established good cause for lateness with respect to its Late-Filed Contention Q.
C. The Other Late-Filine Factors Do Not Favor Admission of Contention O With respect to the four other factors specified in 10 C.F.R. l 2.714(a)(1), the Staff submits that those factors weigh against the admission of Late-Filed Contention Q. Regarding factors two and four, while State's interest may not be represented by existing panies with respect to the issues I
raised in Late-Filed Contention Q, other means are available to protect the State's interest with respect to the issues. The State will have an opportunity to comment on the Staff's Safety Evaluation Repons and Cenificates of Compliance for the HI-STORM and TranStor casks during the nilemaking process to amend 10 C.F.R. Pan 72, Subpan K to add these casks to the list of casks acceptable for use by a general licensee. See 10 C.F.R. ll 72.214. Thus, the StMe does have another !
avenue to protect its interest with respect to the fuel rod buckling analyses.' I With respect to factor three, whether the State's participation may be expected to assist in .
developing a sound record, the State has identified Dr. Marvin Resnikoff, who supponed the !
contention originally. Late-Filed Contention Q at 6. While such identification may have sufficed 1 I
for the original Contention Q, which was timely filed, late-filed standards require more. 1 I
Specifically, the State has not provided a summary of Dr. Resnikoff's expected testimony. Without
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' As the Licensing Board is aware, factors two and four carry less weight than the three other factors specified in the regulation. See Commonwealth Edison Co. (Braidwood Nuclear Power Station, Units 1 and 2), CLI-86-8,23 NRC 241,245 (1986); Private Fuel Storage, LBP-98-7, 47 NRC at 208.
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a summary of what Dr. Resnikoff would testify to in support of this contention, this factor must be j
viewed as weighing against the contention's admission. See Braidwood, CLI-86-8,23 NRC at 246; i
Private Fuel Storage, LBP-98-7,47 NRC at 208-09.'
With respect to the fifth factor, the admission of this contention will broaden the issues and will commensurately delay the proceeding. First, inasmuch as original Contention Q has been eliminated from the proceeding in the Board's original ruling on contentions, there remain no other contentions related to the accident analysis of a cask drop that the State would seek to litigate. Thus, J
J admission of this contention would broaden the issues in this proceeding and will inevitably delay the proceeding as well. See Private fuel Storage, LLC. (Independent Spent Fuel Storage i
Installation), LBP-99-6,49 NRC 114,119 (1999).
In sum, the Staff submits that the State has failed to establish good cause for the late filing of Contention Q, given the State's awareness over a year ago of the deficiencies with the LLNL i
methodology and the nature of the remainder of the issues raised in the State's contention, which are not dependent upon a discovery of deficiencies with the LLNL methodology. Further, the Staff submits that the State's lack of good cause for filing this contention late has not been overcome by _
a compelling" showing that the factors specified in 10 C.F.R. f 2.714(a)(1) favor its admission.
State ofNew Jersey, CLI-93-25,38 NRC at 296. For these reasont :e-filed Contention Q should be rejected.
The Staff recognizes, however, that if the contention is assumed to reflect Dr. Resnikoff's ,
views and his expected testimony, this factor would weigh in favor of the contention's admission.
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D. The Admissibility of Late Filed Contention O The Staff submits that the State has not set forth an admissible contention in accordance with the Commission's regulations in 10 C.F.R. I 2.714. Specifically, the issued raised by Dr. Resnikoff most appropriately should be raised in the rulemaking proceedings on the Holtec cask systems rather J
than in this ISFSI proceeding. See PFS, LBP-98-7,47 NRC at 186 (the Commission's regulatory scheme establishes "a separate cask design approval process under rulemaking procedures and cask 1
design approval prior to licering of the PFS facility."). Since the State seeks to litigate a mstter that is about to be considered in nilemaking, the contention is inadmissible. See PFS, LBP-98-7, 47 NRC at 179. Indeed, as discussed above, the State has raised similar issues in the Holtec HI-STAR 100 storage cask rulemaking proceeding. 1 Second, while the State asserts that the analysis described in the LLNL report should be redone, the contention erroneously a:,sumes that the analysis has not been redone; in fact, the analysis has been redone for the HI-STORM cask. Thus, by letter dated June 8,1999, Holtec submitted a revised section ofits TSAR for the HI-STORM storage cask to respond to ISG-12."
In Holtec's new analysis for fuel rod buckling, "the weight of the pellets is conservatively assumed ,
to be attached to the cladding for all discussions and evaluations." See HI-STORM TSAR Rev. 7 at 3.5 7. The analysis concludes that fuel rod integrity is maintained in the event of a hypothetical accident condition leading to a 45 g design basis deceleration. . ." See id. at 3.5-9. Therefore, a new
" ' See 12tter from Bernard Gillian, Project Manager, HI-STAR /HI-STORM Licensmg i Project to NRC, dated June 8,1999. j l
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analysis has been performed such that late-filed contention Q does not present a genuine dispute of material fact with the Applicant, at least with respect to the HI-STORM cask.i2 Finally, the State's assertion that "[t]he Applicant must not only address lifting accidents while onsite at the ISFSI, but at the intermodal transfer site or during transport on either rail or highway, where significant damage could occur during an accident with potential resulting release of nuclear material" should be rejected. See Late-Filed Contention Q at 5. This matteris repeated verbatim from the State's original Contention Q, and should be rejected for the same reasons that the Board rejected it originally. See LBP-98-7,47 NRC at 195 n.17.
2 The Staff notes, however, that PFS should revise its SAR to include a reference to Holtec's June 8,1999, submittal,in which Holtec provides a new analysis in response to ISG-12. This may be done as part of PFS' expected revision ofits application to reference the final Holtec and/or TranStor cask system that is approved in rulemaking.
CONCLUSION For the reasons set fonh above, the State's Late-Filed Contention Q should be rejected as failing to satisfy the Commission's requirements for the admission oflate-filed contentions and as failing to state an issue that is appropriate for litigation in this proceeding.
Respectfully submitted, Dch Catherine L. Marco Sherwin E. Turk Counsel for NRC Staff Dated at Rockville, Maryland this 5* day of August 1999 i
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UNITED STATES OF AMERICA 00CKETED NUCLEAR REGULATORY COMMISSION USNPC BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 99 AUG -6 Pl2 :19 In the Matter of )
) Oij PRIVATE FUEL STORAGE L.L.C. ) Docket No. 72-22-IQSIl , y
)
(Independent Spent )
Fuel Storage Installation) )
CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF'S RESPONSE TO STATE OF UTAH'S REQUEST FOR ADMISSION OF LATE-FILED AMENDED UTAH CONTENTION Q'" in the above captioned proceeding have been served on the following through deposit in the Nuclear Regulatory Commission's internal mail system, or by deposit in the Nuclear Regulatory Commission's intemal mail system, as indicated by an asterisk, with copies by electronic mail, or by deposit in the United States mail, first class, as indicated by double asterisk, with copies by elec ronic mail as indicated, this 5th day of August,1999.
G. Paul Bollwerk, III, Cb-% nan
- Atamic Safety and Licensing Board Administrative Judge Panel Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington,DC 20555 Washington, DC 20555 i (E-mail copy to GPB @NRC. GOV) Office of the Secretary
- I ATTN: Rulemakings and Adjudications Dr. Jerry R. Kline* Staff Administrative Judge U.S. Nuclear Regulatory Commission .
Atomic Safety and Licensing Board Washington, DC 20555 U.S. Nuclear Regulatory Commission (E-mail copy to .
Washington, DC 20555 HEARINGDOCKET@NRC. GOV)
(E-mail copy to JRK2@NRC. GOV) .
Office of the Commission Appellate Dr. Peter S. Lam
- Adjudication Administrative Judge Mail Stop: 16-C-1 OWFN Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, DC 20555 (E-mail copy to PSL@NRC. GOV)
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- James M. Cutchin, V* Danny Quintana, Esq.* *
' Atomic Safety and Licensing Board Danny Quintana & Associates, P.C.
U.S. Nuclear Regulatory Commission 68 South Main Street, Suite 600 Washington,DC 20555 Salt Lake City, UT .84101 (E-mail to JMC3@NRC. GOV) (E-mail copy to quintana
@Xmission.com)
Jay E. Silberg, Esq.**
Ernest Blake, Esq.
Paul A. Gaukler, Esq. Joro Walker, Esq.**
SHAW, PITTMAN, POTTS & Land and Water Fund of the Rockies /
TROWBRIDGE 2056 East 3300 South, Suite 1 2300 N Street, N.W Salt Lake City, UT 84109 Washington, DC 20037-8007 (E-mail copy to (E-mail copy to jay _silberg, paul _gaukler, joro61@inconnect.com) and ernest _blake
@shawpittman.com)
Denise Chancellcr, Esq." John Paul Kennedy, Sr., Esq.** ,
Fred G. Nelson, Esq. 1385 Yale Ave.
Utah Attorney General's Office Salt Lake City, UT 84105 160 East 300 South,5th Floor (E-mail copy tojohn@kennedys.org).
P.O. Box 140873 Salt Lake City, UT 84114-0873. Richard E. Condit, Esq.**
(E-mail copy to dchancel@ State.UT.US) Land and Water Fund of the Rockies 2260 Baseline Road, Suite 200 Connie Nakahara, Esq." Boulder, CO 80302 Utah Dep't of Environmental Quality (E-mail copy to rcondit @lawfund.org)
-168 North 1950 West P. O. Box 144810 Diane Curran, Esq.* * , j Salt Lake City, UT 84114-4810 Harmon, Curran, Spielberg & Eisenberg (E-mail copy to enakahar@ state.UT.US) 1726 M Street, N.W., Suite 600 Washington, D.C. 20036
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(E mail copy to deurran@harmoncurran.com)
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assr.-
Catherine L. Mark Counsel for NRC Staff i
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RADIOACTIVE WASTE MANAGEMkNT . ASSOCIATES February 27,1998 chases naughney, Director Spent Fuel Project Of5ce, mad Stop 6F18 Nuclear Regulatory Commission Washington, D.C.20555 Re: HohecHI STAR 100 TSAR NRCDocketNo.721008
Dear Charley:
This letter con:4ms the g force that spent fuel cladding can withstand and the use ofthis parameter in safety analyses by Hohec, Sierra Nu:1 car and other cask manufacturers. This issue relates to the Hohee and Transter storageAranspon cask and transpenation caake in ger.:ral. In my opinion the most vulnerable fuel cannot withstand a 63g force in the most .
adverse orientation (Hohec TSAR, p. 3.5-1) but a force considerably less. At the very leut, additionel information should be requested from Holtec before issuing a Cenificate of Compliance for the HI STAR 100 cask. The Commission may also need to fund ,
additiord stuSes to consider this issue as it generaEy relates to transponation accidents invoMng irra6ated fuelassemblies.
The "63 g" force for most vulnerable fuel is based on an ardysis of the more ductile uninadated, not irrndisted, cladding. Despite the title of the Lawrence Livermore Natiend Laboratcry repon on which Hohec relies (" Dynamic Impact Effects on Spent Fuel Assemblies," UCID 21246, October 1987), the LLNL repen does not deal with
" spent fuel" assemblies, only with non irradiated fuel usemblies. As you are aware, -
irradition within a reactor makes fuel assemblies more brittle and less resistant to impact.
"Cladoing duct 111ty decreases and yield stress increases with increasing neutron fluence."
(" Assessment ofthe Use ofExtended Burnup Fuelin Light Water Power Reactors "
Partelle Pacific Nenhwen Labs, NUREG/CR-5009, p. 2 5, February 1988). .
LLNL's calculation for most vulnerable fuel also does not take into account the weight of the faelitseg only the g force without the ad6tional weight of the fuel. This considerable ad6tierA weight is an additionalinterrd force. LLNL unimes fuel pellets remainin a rigid array in a high impact accident and will not impan a force to the cladding. This is obviously not correct. -
NRC staff should ask Hoftec and Sierra Nuclear to address this issue in their TSAR's. If no available studies analyze irradiated fuel cladding in high impact accidents, the NRC should knd additional studies to address this issue.
Marvin ResnikoT, Ph.D. + Senior Aucciate 526 West 26 h St.,Rm. 527 + bY, NY 10001 + 212420 0526 + Tax 212420-0518 + email radwas:eCrwma.com
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I vnb these comme to be in:!uded in Hohec's NRC docket and to be considered in the Stars safety evahndon report. Pieue send me a copy of the rds dr:A safety eveadon repen for the Hohac cuk so that we may provide com:nects. Ifyou have quesdons, fee) free to can.
ec: D Cunan '
oerely, C Nakahara
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION
- s WAsMING7oN, D.c. meMen November 19, 1998 -
Dr. Marvin Resnikoff, Senior Associate Radioactive Waste Management ha~4=ta= .
526 West 26* Street, Room 517 '
New York, NY 10001 -
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Dear Dr. Resnikoff:
I am responding to your February 27,1998, letter regarding your concems related to the
- Weturat integrity of spent fuel cladding under hypothetical accident conditions in spent fuel casxs. In his March 11,1998, letter,' Charles J. Haughney, at the time, Acting Director, Spent Fue1 Project Office, Ir.dicated the Nuc! ear Regutatory Commission (NRC) staff was review your concems and would report their findings to him to report directly to you. I apologize for the delay in responding to you; however, Mr. Haughney is currently serving in another office and severallicensing actions took precedence in a!!ocation of limited stati resources for completing the review. The staff has now completed its review of your concems regarding the Lawrence Livermore National Laboratory (LLNL) Report UCID 21246, " Dynamic impact Effects on Sp Fuel Assemblies," dated October 20,1987, and determined that the LLNL report appeared to use sufficiently conservative data in the characterization of spent fuel cladding properties. The staffassumptions.
and also found that the LLNL report conclusions appeared to be based on acceptable analy in particular, you stated tht the LLNL report does not address irradiated fust cladding, only unirradiated fuel cladding, in ntuality, Table 3 of the report defineates irradiated cladding longitudinal tensite strength values. This table Indicates that Irradiated cladding has a greater strength value than unirradiated cladding. The LLNL report analysis used the values of unitrad:ated cladding strength, which is acceptable.
In your letter, you also stated that the LLNL report did not take into account the weight of the fuel assembly in the side drop orientation evaluation. In actuality, the fuel weight was delineated in Table 4 of the report and used appropriately in the analysis in Appendix A of the ;
report. Thus, the LLNL report used the proper weight vafue in the analysis of the side drop orientation.
The NRC is committed to ensuring the safe operation of dry spent fuel storage and transpo '
casks. The NRC staff wilt continue to evaluate industry data and analysis on spent fuel '
cladding properties in hypothetica1 accident conditions for these caska.
Please note that your letter has been placed Iri all applicable dockets (i.e.,721008,721014, 719261,721023, and 719268) and your questions and concerns w1!! certainly be considered in the staff's safety evaluations of the pertinent cask designs. You wi!! also have an opport to comment v., the draft safety evaluation report for each cask design during the public comment period of fedetal rufemaking to incorporate that cask into Part 72 to Title 10 of the Code of Federal Regulations.
= . - -
2 L h*
~~
4 M. Resnikoff -
I trust this responds to your concems. If you have additiona1 questions or wish to discuss this matter further, please contact me at (301) 415-8518.
j 1
Sincerely,'
{
~
ORIGINAL S1GNED B[fs/
Mark S. Demgatti, Senior Project Manager Spent Fuel Ucensing Section Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards Docket Nos.: 72 1008,72 1014,71 9261,
- 72 1023,71 9268 -
l
O O
EXHIBIT 3 1
Dr. M;rvin Resnikoff, Senior Associate Febnery 17, 1999 Radioactive Waste Management Associat:s 526 West 26* Street, Rm. 517 Now Ycrk, NY 10001 -
SUBJECT:
RESPONSE TO YOUR LETTER DATED DECEMBER 31,1998
Dear Dr. Resnikoff:
~ ~ ' "
I am responding to your December 31,1998, letier that the Nuclear Regu!atory Commission (NRC) received in late January 1999. We appreciate your continued interest, not only in the proposed Private Fuel Storage, LLC. Independent spent fuel storage instaffation on the reservation of the Skull Valley Band of Goshute Indians, but.atso in the certification of dual-purpose cask system technologies. In a letter to the staff dated February 27,1998, you had expressed your concems regarding irradiated fuel cladding briftfeness and dynamic loading in dual purpose casks. Our November 19,1998, response to you indicated that the staff had considered your concerns during the review of the Hottee international HI STAR /HI STORM en . systems and planned to consider them for the BNFL Fuel Solutions' TranStor cask
- cicm. ,
k The preliminary Safety Evaluation Report (SER) and Certificate of Comp!Iance for the Holtec
{
H1 STAR 100 storage cask is currently available for public comment, as part of the rulemaking j process required to amend 10 CFR Part 72 to include the cask as one available for general use pursuant to Subpart K. The public comment period is open untit March 29,1999. If after )
reviewing the SER you have additional concems, we would appreciate your submitting comments as part of the Hi STAR 100 rutemaking. This would enable us to specifically 3 document the resolution of your comments as they apply to the Holtee submittal. Similarly, when the TranStor rulemaking is undertaken later in 1999, your comments would be usefulin that process, as we!!.
If you have any questions regarding the rulemaking process, please contact Mr. Earl Easton, Section Chief of the Transportation and Storage Safety Section, Licensing and inspection Directorate, Spent Fuel Project Office. Mr. Easton can be reached at (301) 415 8520.
Sincerely, original signed by Eric J. Imeds for /s/
Mark S. De!!igatti, Senior Project Manager Spent Fuel Licensing Section Spent Fuel Project Office
- Office of Nuclear Materlaf Safety and Safeguards -
l 4
I i
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EXHlBlT 4 ;
I
I, w, ~
h RADIOACUVE; WASTE MANAGEMcNT ASSOCIATES
.D-M 31,1998' '
. Mark $ DcI11gstti, Senior Project Manager ,
. Spent Fuel.Licenalag Section ,
~
NMSS ..
US NuclearRegulatory Commission-w..Mnston,DC 20555 .
~
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Thank you for your November 19 response to my February 27 Ictier. Your letter did not
- 2 fully answer my concerns, so T11 t:y once more -
Ertttseness f. _..
From severs! NRC contractor repona, h !s my understanding that Irradiated fuel cladding is more brin!c than unindisted hel dams nis shculd aher the consequences of transponation or ISFSI accident involving Impact. Yoistated that Irradiated fu cladding has 's greater strcogth value' than unitradited fuel cladding, but e does not1 address my concerns about brittleneu. It does obt Holtec and SNC r. bat thh important tween inadisied drinetion be, appear and unitradiated het that NRC st clad 6ng. $1mply u'inj s uniradiated clad 6ng'sse' cgth istbi Hohee and SNC SAR's'ma'y
~ ~
et be acceptable. , ,
. . .g
. l
.... \
eight Is taken into accoun! !n the p KL repon and '
I am awarch.a! the fuel anembly w' static,'th'at 1(6ti ll the Holtec'SAR,but thelokdinSis thifu'el Weight is assaned to be cyenly disinhted along the cladding. ni model Is ess a y a beam Setwb two supports.
But this model may not bound the pb'ysicil situati'on. Iris side Impact, the cladding and the fuel are distinct beams. Under impact the fuel pellets Would be expected to break' their faed configuration and sente .the cla'ddling'yith fdtcV. .Thu, dynamic loading is not coruidered in the ILNL report ad may.b'e Importani.'It does not appear that NRC staff are querying Hohee and SNC about @ 1m'oortant distinctI6n between static and dynamic ta. u_ to ,.
year.
Thank you for reconti,dering these luues. And best wishes for ,
7
'/ '
/
Marvin Resnikoff. Ph.D. + Senior'Assoelste' .
526 West 26th St.,Rm. 517 + NY, NY 10001 + 21242H526 + Fax 21242M518 t email rndwamecrwms.com
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.- . . _ . . _....... E __ .. _ _ . -
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EXHIBIT 5 i
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1
STATE OF bTAH '
CFFICE CF THE ATTOR N E) G E 4 E ll A L ,
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J AN GR AH AM ATToRNE Y GENER AL i
Jawss R So*Em Rito Rcwastos Patusm OtPavus SacaanGovea- CM Dem anomer Generas Dreew of Pm Po.cy & Co%ig.g March 26,1999 -
Secretary U.f. Mu:! ear Regulatory Commission Sent via Federal Express &
11555 Rockville Pike E mail: sot @nre. gov Rockvile MD 20852 2739 Attn: Rulerna. kings and Adiudications re: Comments on Proposed Rule to add Holtec Hi Star 1000 Cask System to the List of Approved Spent Fuel Storage Casks.
l 1
Dear Secretary:
In response to 64 Fed. Reg.1542, January 11,1999, the State of Utah submits ,
comments on the Preliminary Safety Evaluation Report and Proposed Certificate of l Compliance for the Holtec HI STAR 100 Storage Cask. These comments have been prepared with assistance from Marvin Resnikoff, Ph.D., Radioactive Taste Management Associates. .
Sine y, Qenise Chancellor Assistant Attorney General Ene:
- 4 0 E a s: 3 0 0 S,c ute 5tm F8 cot P o Bei 1408'3 Sa'? Late cery ,U t a a 8J1*J 08?3
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Comments from the State of Utah Preliminary Safety Evaluation Report and Proposed Certificate of Compliance HI-STAR 100 Storage Cask March 29,1998 The State of Utah. with assistance from Marvin Resnikoff, Ph.D. of Radioactive Waste Management Associates, submits these comments on the preliminary Safety Evaluation Report (SER) and proposed Cenificate of Compliance (CoC) for the Holtec Hi-STAR 100 irradiated fuel storage cask, NRC Docket No. 72 1008. 64 Fed. Reg. 1542 (1999).
The Hl STAR 100 is an all metal cask with an outer metal overpack that encloses a sealed helium filled canister (MPC) containing irradiated fuel. Although the SER and CoC under review here address storage only. the HI-STAR 100 is designed for both storage and transponation of spent nuclear power plant fuel.
The State has a three-fold interest in the adequacy of the SER and CoC for the H1 STAR 100 storage cask. First. its design is vinually identical to the design of the H1 STAR 100 transponation cask which Private Fuel Storage LLC. (PFS) proposes to use to transpon spent fuel to its proposed independent spent fuel storage installation (ISFSI)in Utah. The only difference between the storage cask and transponation cask is the fact that the transponation cask uses impact limiters and must satisfy hypothetical accident conditions under 10 CFR Pan 71. Second. PFS plans to use the H1-STAR 100 storage cask's intemal welded canister (multi purpose canister or MPC) to transpon and store fuel. The MPC will hold the irradiated fuel during transponation to the Private Fuel Storage .
facility. After arrival at the PFS facility, the MPCs will be stored in the HI STORM 100 concrete overpack at the proposed PFS facility. Third although PFS intends to use the HI STORM-100 cask for storage under normal conditions,in case of an accident, the all-metal cask Hi STAR 100 cask will be used as a storage backup.
Dese comments address the conclusions of the SER, as well as the assenions made by the cask manufacturer, Holtec Intemational. in the Technical Safety Analysis Reports (TSARS) for the Hi STAR and Hi-STORM casks. The HI-STAR 100 storage TSAR is Holtec Repon HI 941184 (NRC Docket No. 72-1008). The H1-STAR 100 transportation TSAR is Holtec Repon HI-951251 (NRC Docket No. 71-9261). The HI STORM 100 norage TSAR is Holtec Repon HI 951312 (NRC Docket No. 72-1014).
1
y Csmments on Preliminary SER, Hi STAR 100 Storage Cask Page:
y state sf Utah F
The State is in the process of finalizing a confidentiality agreement with Holtee that will
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allow the State access to the Holtec proprietary version of Hi -STAR 100 SAR, Revision 9 and HI STORM 100 TSAR, Revision 5. The State will submit additional comments, as a proprietary and confidential submittal, after it has received and reviewed Holtec proprietary documents.
General Comments The HI STAR 100 design shoulif not be approved,' because Holtec has not provided reasonable assurance that the cladding and cask will retain their integrity under normal, off normal and accident conditions. Moreover, Holtec does not correctly calculate health j arracts under bounding accidents. Nor has Holtec evaluated the impact of a sabotage !
event. Finally. the TSAR and SER do not provide assurance the cask and cladding will I retain their integrity under thennal conditions that exist at ar. ISFSI. Rather than addressing these deficiencies, the NRC's SER has glossed them over. These issues are crucially imponant to protecting the public health and safety. and therefore must be addressed before the Holtec CoC can be issued.
Specific Comments Cladding Integrity Under Impact According to the Hi STAR 100 storage TSAR (Sec. 3.5). the Hi-STAR 100 system is designed to withstand a maximum deceleration of 60 g while a Lawrence Livermore National Laboratories repon shows that the most vulnerable fuel can withstand a deceleration of 63 g in the most adverse orientation (side drop).' Holtec therefore assens that fuel rod integrity will be maintained under all accident conditions. In the preliminary SER (at I l-6), the NRC Staff concurs that "there is reasonable assurance that the cladding ' '
will maintain confinement integrity during a design basis drop."
In our view this analysis is incorrect. Holtec and the NRC Staff have not demonstrated a
- reasonable assurance that the cladding will maintain its integrity. -
Holtec's analysis does not provide reasonable assurance for the following reasons: (1)it does not take into account the possible increase in rate of oxidation of cladding of high bumup fuel; (2) Holtec relies for its analysis on a Lawrence Livermore National Laboratories (LLNL) repon that fails to distinguish the effects of reactor irradiation on
' UCID-21246. Dynamic 1menet Efrects on Soent Fuel Assemblies. Chum. Win, Schwanz (October 20.
1987)(LLNL Repon).
seas.cnun rre..m.no sex. ru a 4 An .w 3 erg m . . ,
. State tf Utah fuel assemblies; and (3) Holtee also relies on the LLNL Report's incorrect assumption that fuel assemblies act as a static rigid rod. The first factor (increased rate of oxidation) increases the likelihood that fuel cladding may rupture. and. when the three factors are taken together, they compound the likelihood of a release ofradioactive materials during a foreseeable drop accident at the proposed ISFSI.
- 1. Increased Rate of Oxidization.
The NRC's Information Notice TN 98 29, entitled " Predicted Increase in Fuel Rod Cladding Oxidation"(August 3,1998), provides new information, not considered in the Holtec TSAR, that calls into question Holtee's accident analysis. IN 98 29 discusses the oxidation rate of fuel cladding for high burnup fuel based on reported experiences with Westinghouse's fuel assemblies. NRC advises recipients to " review the information for appucaemty to their facilities and consider action as appropriate. to avoid similar problems." Nuclear Regulatory Commission. IN 98 29, Predictedincrease in Fuel Rod Cladding Oxidation (August 3. I998) at 1.
The Notice reports that in October of 1997. Westinghouse notified NRC that modification ofits fuel cladding corrosion model in its fuel design code to reflect new data on Zirealoy-4 oxidation at high bumup "may create compliance issues for its Integral Fuel Bumable Absorber (IFBA) fuel with Zircaloy-4 cladding." Id. at 1. As noted in tlie information Notice.
The modified code may predict higher fuel temperatures and intemal pressu .s at high burnup conditions. This, in tum. may lead to code results that do not n 'et 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). i Id.- Although the problem was initially discovered by Westinghouse with relation to Zirca!cy 4 fuel. the Information Notice notes that the bumup related phenomena, whiih could result in noncompliance with the oxidation requirements of 10 CFR 50.46, may not -
be limited to Westinghouse IFB A fuel but might affect any Zircaloy fuel used in high bumup application." IN 98 29 at 2. Thus, the experience at Westinghouse is also germane to any high bumup fuel that may be stored in Holtec casks--not just to Westinghouse fuel. -
According to IN 98 92, the increased oxidation of the cladding is a function of the fuel bumup. Oxidization may cause the cladding to become effectively thinner, decreasing its structural integrity. This thinner cladding due to oxidization also lowers the 'g' impact force at which fuel cladding will shatter. Holtec's TSAR relies on the premise that fuel cladding will not shatter for any foreseeable drop, This premise is based on the assumption that it would take a side drop of greater than 63g to damage the cladding.
Our spreadsheet calculations, presented below, shm. that the g loading for high burnup
m a nges..m ryc.
, State af Utah 4"
fuel with oxidized cladding approaches 45g. The NRC Staff should not approve the t
Holtec application unless and until Holtec has factored the information in IN 98 29 into its calculations. The clear implication of fN 98 29 is that the lift height of the Hi STAR 100 cask must be reduced to lower the g forces on the cladding.
. Effects of Changing Variables Table 4 in Lynamic Impact Efects on Spent Fuel.4ssemblies A B C D E F G Rod array 17xl? 17xl? 17xt7 17xl?
17xl7 17xl? 17x17 Assembly weight (Ib) I450.00 1450.00 1450.00 ' 1450.00 1450.00 I450.00 1450.00 8ofrods 264.00 264.00 264.00 264.00 264.00 264.00 264.00 Fueled length (in) 144.00 144.00 144.00 144.00. 144.00 144.00 144.00 W of spacers (N) 7.00 7.00 7.00 7.00 7.00 7.00 7.00 L = (fueled length /N- 24.00 24.00 24.00 24.00 24.00 24.00 24.00 E (psi) 1.04E+07 1.04E+07 1.04E+07 i.30E+07' l.04E47 1.04E+07 1.04E+07 c .ym -
8.05E44 3.05E+04 8.05E+04 8.05E44 4.50E+04 4.50E+04 8.05E44 t (in) - 0.02 0.02 . 0.02 0.02 0.02 0.02 0.05 ro(in) 0.19 0.19 0.18 0.19 0.19 0.19 0.21
. ri(in) 0.16 0.16 0.16 0.16 0.16 0.16 0.16 A (in2) 0.02 0.02 0.02 0.02 0.02 0.02 0.05
! = (1/4'3.14(ro*4- 3.85E 04 3.85E 04 3.09E-04 3.85E-04 3.85E 04 3.85E 04 9.37E-04
' W (Ib) 0.84 0.84 0.69 0.84 0.84 0.84 1.78 w (Ib'in) 0.04 0.04 - 0.04 0.04 0.04 0.04 0.04 r (in) 0.18 0.18 0.17 0.18 0.18 0.18 0.19 pressure (Ib) 2250.00 l187.80 2250.00 2250.00 2250.00 1187.80 2250.00 os (psi) 8787.50 4639.02 10472.14 8787.50 8787.50 4639.02 4675.00 M (Ib in) 2.32 2.32 2.32 2.32 2.32 2.32 2.32 ob (psi) 1128.70 1128.70 1373.17 1128.70 1128.70 1128.70 $ 19.50 P tib) 68.56 68.56 55.00 85.69 68.56 68.56 166.87 ga 81.93 81.93 80.06 102.41 81.93 81.93 93.71 gy. 63.54 67.21 50.81 63.54 32.08 35.76 145.96 A: Values from Westinghouse specimen (Dynamic impact Effects... Table 4)
B: Pressure changed to a lower value ( value in An Assessment of the Risk )
C: Thickncss of fuel cladding decreased due to oVdation by 17*b. Column A's thickness is reduced by 17%
D: E Modulus changed to higher salue (value in An Assessment of the Risk )
E: Yield stress lowcred to half the original value F: Yield stress lowcred and pressure lowered (E and B)
G: Doubling the thickness Note: .
DIE = Dynamic impact Effects.
AAR = An Assessment of the Risk.
AAR's E modulus was expected to be lower, as they took irradiated zircaloy into account. However. it was not. after conversion.
o AAR DIE E modulus .1.30E+07 1.04E+07 psi Pressure 1187.80 ! 2250.00 psi
qEaluuseuuCKm3D9 h1Rd.trJU-5$1/AR ivu storage C.ta age r
, - State cf Utah n
. 2. Irradiated and Unirradiated Fuel Assemblies.
Holtec's TSAR for the Hl STAR 100 storage cask relies for its estimate of g force that will damage fuel cladding upon a 1987 repon by LLNL.: The LLNL Repon fails to take into account the increased brittleness o ' firradiated fuel assemblies.) Because the irradiated fuel assemblies may have been embrittled. they would also be less resistant to impact. During the course of a fuel assembly's life, subatomic particle bombardment.
including neutron flux, significantly decreases the assembly's ductility and increases the assembly's yield stress, thereby embrittling the fuel assembly. " Cladding ductility decreases and yield stress increases with increasing neutron fluence."'
Furthermore, the proposed Hi-STAR 100 will store only irradiated fuel assemblies; thus, 6 Applicant cannot rely on LLNL's analysis because the LLNL does not account for irradiation and embrittlement, which lower the impact resistance of the fuel assemblies.
These facts are significant when coupled with the increased oxidation rate reponed in IN 98 29 because increased oxidation could tangentially cause an increase in cladding embrittlement.' ThusclN 98 29 compounds the LLNL's error in disregarding the brittle characteristics ofirradiated fuel cladding. {
- 3. Fuel Assemblies Do Not Act as a Rigid Rod.
Holtee's calculations rely upon LLNL's erroneous assumption that the fuel within the cladding behaves as a rigid rod. Thus. Holtec merely used a static calculation instead of {
taking into account the dynamic loading upon impact. The LLNL Repon specifically states, "It is imponant to emphasize that the g loadings shown in Figs. 6 and 7 are static loadings."' This assumption is incorrect. Instead of a homogenous, rigid rod, the fuel rod consias of fuel pellets stacked like coins within thin tubing. In any impact scenario, the fuel assembly does not act as a rigid rod: rather,it acts as a dynamic system with the fuel impacting the inside of the cladding and creating a greater likelihood of cladding rupture.
Holtee has not shown that the assumption of a rigid rod is conservative. The thinner ' {)
cladding due to the increased oxidation serves to compound t.his effect because a smaller g force would be required to rupture the assembly. The NRC staff should not approve the
~
2 LLNL Repon.
3 See e.g. UCID-21246. Table 4, which makes no distinction betw een Young's modulus and yield strength of a range of fuel assemblies. ,
' " Assessment of the Use of Extended Bumup Fuel in Light Water Power Reactors." Banelle Pacific Northwest Lab, NUREG/CR 5009 (February 1988) 8 Thin cladding acquires brinle characteristics at a faster rate than thicker cladding during fuel life. See IN 98 29 at 2 ("If this total oxidation limit were to be exceeded during an accident, the cladding could become embrittled.").
- LLNL Repon.
L_ _ _ - - ---
- .;po mgsmeerfun n. e..w .w n.v - ,.
, wie erUtah a
Holtec application without a showing by the applicant that its calculations are conservative.
In sum, the newly discovered findings at Westinghouse. as recognized in the NRC's Notice, and the other concems discussed above. raise significant questions about the adequacy of Holtee's accident analysis.
Health Impact of Accidents The calculated health impacts under hypothetical accident conditions, discussed in Chapter 7 of Holtec's HI STAR 100 TSAR, are not conservative. Three issues need to be mnre fully examined by NRC Staff: the design basis accident, the radiation pathways, and me dose to children.
- 1. Design basis accident.
Holtee's hypothetical design basis' accident condition assumes 100% of the fuel rods are non mechanically ruptured ard the gases and particulates in the fuel rod gap between the cladding and fuel pellet are released to the MPC cavity and then to the external environment. Radiation doses are calculated 100 m from the cask. In the time interval between production of Rev. -l and Rev. 6 of the TSAR. the NRC Staff reguested Holtee to conduct the dose calculations in conformance with the nnal version of NUREG 1536.'
The accident analysis in the final version of NUREG-1536 increased the amount of radioactivity to the MPC cavity by 5 orders of magnitude and would have placed doses at 100 m over the EPA's limit of 5 rem. In Rev. 6. Holtee responded to the NRC's request by changing the method of calculating doses to incorporate an extremely small cask loAage rate, rather than assuming 100% of the cask cavity was released to the extemal environment. TSAR. Rev. 6 at 11.2-15. Thuc. Holtec's new analysis increased the raaroactivity released to the cask cavity by 5 orders of magnitude, but the leakage rate reduced the amount released from the cask cavity to the environment by more than 5 -
orders of magnitude. The net effect of this sleight of hand, was to change the design basis accident, so as to reduce the doses to the thyroid and whole body at 100 m. In essence, the NRC staff has allowed the applicant to change the definition of a bounding accident to one that involves 100% fuel rod cladding rupture, with the cask lid intact,i.e.. .
only slight leakage from the cask. The design basis accident no longer represents a loss-of-con 6nement barrier accident.
Holtec's attempt to change the design basis accident for storage casks is not only inappropriate, but is unsupported. We strongly disagree that the slight cask leakage,,1.5 x
' Nuclear Regulatory Commission.* Standard Review Plan for Dry Cask Storage Systems." NUREG 1536, January 1997.
i I
7 u....amanr.ca. c.m m....v. m ....... ...A . , .
-a State cf Utah -
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> 10d em8/s, constitutes a bounding accident.- A scenario that could lead to a greater release rate is a welding error that allows helium to leak from the MPC if a cask is dropped.
Leakage of helium will allow the maximum cladding temperature to rise and the fuel rod cladding to rupture. In this case, the percentage of fuel rods that rupture may be less, but the leakage rate from the cask cavity would be greater than assumed by Holtec.
- 2. Radiation pathways excluded In Chapter 7, Holtee has calculated the radiation dose to an adult 100 m from the accident, duc solely to inhalation of the passing cloud. Other relevant pathways, such as '
direct radiation from cesium and cobalt-60 deposited on the ground, resuspension of I deposited radionuclides, ingestion of contaminated food and water and incidental soil ingestion, are not considered, in violation of 10 CFR 72.24(m). 1
- 3. Dose to children not considered
{
Contrary to the standards in 10 C.F.R. Pans 72 and 20. Holtec has not calculated the dose to children. These standards prescribe dose limits for "an individual outside the controlled area"(10 C.F.R. { 72.24m)). and " individual members of the public"(10 C.F.R. f f 20.1301. 20.1302). For purposes of the Pan 20 dose standards. the regulations define " individual" as "any human being." and " member of the public"_ as any individual except when that individual is receiving an occupational dose." (Emphasis added). The concept of "any individual" clearly includes people other than adult men. i.e.. children. i Nor does the Atomic Energy Act limit its protection against undue risk to adult males. In fact. NRC regulations already make special exception for the dose to a minor (10 CFR {
20.1207) and the dose to an embryo / fetus (10 CFR f 20.1208) within restricted areas.
Funher, Regulatory Guide 3.51. " Calculational Models for Estimating Radiation Doses to Man from Airbome Radioactive Materials Resulting from Uranium Milling Operations,"
also calculates the dose to children and infants by adjusting the organ size. breathing rate and dose conversion factors.
Children are more vulnerable to radiation than adults because of their higher surface-area-to volume of organs ratio.' Other contributing factors include the fact that children have higher soil ingestion rates than adults.' Children also have reduced ingestion and inhalation rates compared to adults;'8 nevenheless, the dose to children under a design basis accident is likely to be significantly higher than the dose to an adult. Thus. in order to satisfy the regulations and the Atomic Energy Act,it is necessary to determine whether
' Intemational Commission on Radiological Protection,"1990 Recommendatiens of the International Commission on Radiological Protection," ICRP 60,1991. Pergamon Press.
' EPA," Risk Assessment Guidance for Superfund: Volume ! Human Health Evaluation Manual (Pan B, Development of Risk based Preliminary Remediation Goals," EPA!540/R.92/003, December 1991.
7 Eckerman, KF et al," Health Risks from Low Level Environmental Exposure to Radionuclides," Federal Guidance Repon No.13, Pan I Interim Version. prepared fo the EPA '1998.
euauuaculdamnagamnum ivu wrne . .g >
, state dUtah the dose limits are satisfied for children. In addition. children are at a higher risk than adults of developing cancer because children live longer than adults and their cells grow
- more rapidly than adults' cells.
Note that it is not the regulation of'25 mrlyr or 100 mrly or the EPA accident dose limits of 5 rems that are deficient. Rather. it is the NRC Staffs methodology in calculating exposures to children.
Sabotage Event' '
We disagree that an accident involving 100% fuel rod cladding rupture with slight lid !
Skage is a bounding accident. See wpra, discussion on design basis accident. We urge NRC stafito consider the effect of a sabots 3e event with an anti tank missile. The NRC already considers the impact of a tomado missile and explosion; but a tomado missile, like an 8" diameter steel rod striking the cask at 126 mph," does not have the impact of an anti tank missile. Similarly, an explosion with an extemal pressure of 300 psig does not have the impact of an anti tank missile. The lack of a comprehensive assessment of the risks of sabotage and terrorism against nuclear waste facilities and shipments is well established. Terrorists have shown that they are capable of exploiting the weak interfaces associated with transportation as well as causing tremendous damage to static structures such as the World Trade Center and the Oklahoma City govemment building. As NRC Staffis aware. German regulatory authorities have imposed an additional condition on casks. namely, that they be able to withstand the impact of a 1-ton missile impacting a cask at the speed of sound. By this condition. German casks are able to withstand the impact of ajet engine striking a cask. NRC staff could impose additional conditions on
{
dry storage casks and ISFSis. e.g.. the CoC could require that an ISFSI be designed with an earthen berm to reraove the line-of sight. ,
.l Since the early 1980s. the NRC has relied on and has poorly interpreted an outdat'ed set of experiments carried out by Sandia and Battelle Columbus Laboratories that measured the .
release of radioactive materials as a result of cask sabotage. In one of the Sandia experiments, a GE IF-200 truck cask containing one unirradiated fuel assembly was anacked with an M3Al, a military " shaped charge." Although the results " demonstrated that casks could indeed be breached by military explosives and that a considerable fraction of spent fuel could be released by such an attack,"82the NRC responded to Sandia's findings by concluding that since only 2/1,000,000 of the total fuel weight was released in inhalable form, the " average radiological consequences of a release in a heavily populated urban area such as New York City would be no early fatalities and less
" Holtec HI STAR 100 Storage TSAR, Table 2.2.5.
12 Halstead. Robert J. and James Dasid Ballard," Nuclear Waste Transportation Security and Safety issues; The Risk of Terrorism and Sabotage Against Repository Shipments." prepared for the Nevada f Agene3 or Nuclear Projects.
Carson City. Nevada. October.1997, p.25.
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that one (0.4) latent cancer fatality."" But this analysis is highly deficient. making a large number of questionable assumptions rega.rding evacuation and failing to include several significant radiation pathways. such as direct gamma exposures from deposited radionuclides. A more recent analysis" of a trahsportation accident in a rural area for a cask holding 14 (not 24) PWR fuel assemblies shows a cost of $620 million and a recovery time of 460 days for the cleanup operation. This cleanup is only to a level that would reduce doses to 500 mrem / year. Based on this accident scenario in a rural setting, a properly conducted, realistic accident scenario in an urban area can be expected to show billions of dollars in cleanup costs and lost revenues. An' urban accident would also cause a large number ofadverse heafth effects. Deficient as the analysis of a sabotage event during transportation is, the NRC Staffhas never estimated the economic and safety implications of a sabotage event at a fixed storage facility.
S ?!.".C :. .s never explained why it considered the Sandia experiment indicative of what could occur in any type of terrorist attack, no matter the circumstances. Following the publication of these Sandia results, the NRC proposed elimination of many of the safety requirements for shipments of spent fuel aged more than 150 days, such as, no armed guards for the shipments in highly populated areas. no advance notice to the NRC or local law enforcement officials. and no periodic communication between escorts and a communications center." "At least 32 parties submitted more than 100 pages of comments in response to the notice." to which the NRC never publicly responded. See supra. Halstead and Ballard. note 12.
In the intervening years since the Sandia experiments. anti-tank weapons with much greater accuracy and penetrating power have been manufactured and widely distributed.
These devices could release much more radioactive material. The NRC suspended action J on the rule making, but it inappropriately continues to use the unrevised conclusions in the proposed rule as a basis for its policies on terrorism and sabotage of nuclear '
shipments.
The numerous terrorist attacks of the last several years have graphically demonstrated that the NRC continues to ignore the risks of sabotage at significant peril to the public. ~
The NRC should adopt the specific recommendations of Halstead and Ballard for creating a realistic. up-to-date terrorism risk assessment. Some of the reference parameters Halstead and Ballard suggest are as follows:
- The reference weapon should be portable anti tank missiIes for their ability to permeate the strong cask materials, their range and availability.
" M at 26.
" Sandquist GM et at.*Esposures and Health Effects from Spent Fuel Transportation? Rogers & Associates. RAE.
8339/121. Nosember 29. I98s.
' " Halstead and Ballard at 27.
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, Commenu on Preliminary SER. Hl. STAR 100 Sarrage Cask t' age su
, sinie cf Utah '
1 A 10-year-cooled, medium bum up. Westinghouse PWR assembly should be the reference spent fuel. A Holtec HI STAR 100 storage cask loaded with 24 PWR assemblies of the reference fuel would represent a total radioactivity of about 5.5 million Curies. A terrorist incident resulting in a one percent release would have radiological consequences far greater than those assumed in the HI-STAR TSAR.
The following two scenarios. at a minimum should be considered: "an attack in which the cask is captured, penetrated by one or more explosive devices, and releases a significant amount (at least one percent) ofits radioactive contents; and an attack in whi.:h the cask is perforated by one or more armor piercing rockets or missiles and releases a significant amount (at least one percent) ofits radioactive contents.""
Note,that Halstead and Ballard recommend a 1% release because that is the percentage of unirradiated fuel released in the Sandia sabotage tests." We maintain that a design basis accident should not be the release of2 x 10 8 or less of the cesium inventory, but 1%,
based on the Sandia sabotage tests. Further, it is not simply inhalable-sized particulates that are imponant. Larger sized paniculates will be released and deposited downwind, giving rise to a direct gamma dose.
Thermal Requirements The proposed CoC temperature conditions for the Holtee HI STAR 100 storage cask are not sufficient to guarantee that cladding and neutron shield degradation will be minimized. To reduce high temperatures. NRC staff must incorporate an additional condition into the CoC. a minimum pitch or center to-center distance between casks.
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While Holtec has suggested a pitch of 12' or a 4' spacing between casks, this analysis is likely not based on rigorous calculations. Until the State receives the proprietary calculations from Holtec. it cannot comment with specificity on them. However, based on review ofsimilar proprietary calculations for the HI STORM 100 casks we have ,
reviewed, we are skeptical that the proprietary calculations for the HI-STAR 100 cask are
- rigorous and sufficient.
Under the present regulatory framework. NRC staff and contractors must show that individual casks will not overheat if subjected to normal (average T = 80 *F) and off. ~
normal (average T = 100 *F) teraperatures. If the normal or off normal temperature conditions are satisfied, then the cask may be used in that location. This is similar to the approach for the CoC canhquake and tomado conditions, but with one imponant difference: individual casks may interact with each other, causing temperature conditions above ambient temperature conditions. As a result, the Holite neutron absorbing material
" 14. at sw.
" Sandora). RP et al. An Assessment ofshe SafetyofSpen Turiiransportation in L'rban Environs SANDt2 236s prepared for DOE by Sandia Labs. June 1983.
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Stais cf Utah 4-2, i and the cladding may degrade due to excessive heat. In the Hi STAR 100 TSAR. the presence of adjacent casks and the concrete pad may not be correctly taken into account.
as far as one can determine from Holtec's sketchy nonproprietary analysis. This should be properly addressed in the SER and CoC. -
If the center to-center distance between adjace'n t HI STAR 100 casks is too small, casks may thermally interact with each other, effectively increasing the ambient temperature.
According to Holtec's TSAR, the overpack shell outside surface temperatures are 229 'F and 249 'F under normal and off normal temperature conditions." In the most extreme example, if adjacent casks are in immediate contact, instead of the ambient temperature
~ being 80 'F under normal conditions, it would be 229 'F. As the casks are moved away from each other, at some distance the casks become thermally independent of each other.
Holtec attempts to calculate this distance in Fig. 4.4.5 by assuming a radiative blocking 9 r due to the presence ofother casks. But the situation at an ISFSI is far more complicated.' It is not a blocking factor so much as the presence of adjacent heat sources at 229 'F. The effective ambient temperature will be raised as the casks interact with
- each other. The distance at which casks will act independently of each other must be calculated by Holtec and included in the CoC. For the HI STAR 100 cask, the critical j
temperature is 300 *F for the inner surface of the Holite neutron absorbing material that '
surrounds the metal cask. The maximum temperatures of the Holite under normal and off-normal conditions are 274 'F and 294 'F. respectively. That is. the Hi STAR 100 is already operating with a thin safety margin not accounting for the interaction between casks.
To take into account the interaction of casks, the following factors must be incorporated into the calculation. As a first approximation. Holtec could assume adjacent casks at the same temperature. To = 229 *F. Insolation, average pad temperature. external convective air currents, and wind speed must also be incorporated into the model. The surface temperature of the center cask. Ti . could then be calculated. In the next iteration, the adjacent casks could be taken at temperature Ti and a new temperature for the center cask could be calculated T:. Holtec could then determine whether the series T., Ti , T:is I
converging to some asymptotic value, if the value for the inner surface of the neutron i shield exceeds 300 *F. the casks must be spaced further apart.-
As the situation presently stands, the SER and CoC are deficient. The maximum cladding temperature or temperature of the neutron shield inner surface has not be ~
correctly calculated. NRC staff and Holtec are assuming there is no interaction between the casks. This assumption is not conservative.
" Holtec. Topical Safety Analysis Repon for the Hl. STAR 100 Cask System. Holtec Repon Hl.941184.NRC Docket No. 72 1008. Table 11.1.
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ummenu on trenmmary sER, tilol An sou storce w ry . .
, State cf Utah ,
1 Removable Surface Contamination The TSAR includes Technical Specifications for removable surface contamination." If the smearable contamination exceeds 2200 dpm/1002cm from gamma and beta emitting sources. the Technical Specifications require that the accessible surface be flushed or pressure washed. "If the smearable contamination limits still cannot be reduced to acceptable levels, evaluate and perform altemative actions up to. and including, removal of the MPC from the HI-STAR 100 overpack after removing the spent fuel from the MPC."" These conditions cann,ot be met at the proposed off site Skull Valley ISFSI in Utah. No provisions exist for decontaminating casks under PFS's " start clean, stay clean" philosophy. PFS's proposed policy is to retum casks that are contaminated above regulatory limits back to reactor sites. No provisions would exist at PFS for removing the
!.'"C f.om the HI STAR 100 overpack. In recognition of the conflict between the Tech Specs for the HI-STAR 100 and design of the PFS facility (and possibly other ISFSIs),
the NRC should specify that all users of the HI STAR 100 have the capability to remove i smearable contamination onsite.
Future Rulemaking Procedures The State of Utah strongly disagrees with any proposal by the NRC to approve future additions and revisions to the list of approved spent fuel storage casks as direct final rules. Under such a procedure there would be no proposed rule. Instead, the rule would become final within 75 days after publication unless NRC receives "significant adverse comments on the direct final rule within 30 days after publication." 64 Fed. Reg.
at 1543. On receipt of such significantly adverse comments. NRC would withdraw the rule, vidress the comments, then publish a final rule. First the premise underlying NRC's proposed procedural change -- that " additions and revisions to the list of approved spent fuel storage casks are noncontroversial and routine"- is inaccurate. The above comments show that NRC's approval is not " routine." Moreover, given the past
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problems. such as hairline cracks associated with dry storage casks, it is imperative that future approval or revision to the list of approved casks be subject to adequate and rigorous public scrutiny. Second. a direct final rule reduces to 30 days the period of time for effective public comment. This is an insufficient time period to review and prepare _ l comments that may be "significantly adverse" to cause NRC to withdraw the published final rule. Third, a direct final rule will diminish the public role in commenting and affecting the outcome ofrulemaking procedure. It is more likely that NRC will give due consideration to comments at the propose rule stage: any comments at the final rule stage i
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" Id at 12.3-16.
" Id ,
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' . Comments on Preliminary SER, H1, STAR 100 Storage Cask O Page 13 State cf Utah i
e would have to be "significantly adverse
- for NRC to reverse course and withdraw the direct final rule.
Safety considerations are too imponant for NRC to expedite the approval process at the
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expense ofdiminishing the public's role in commenting on the approval ofspent nuclear fuel casks. l j
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