ML20209E045

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Summary of 990621 Meeting with Util Re Status of Staff Efforts to Assess Risk Associated with Storage of Spent Fuel Pools at Sites in Decommissioning Phase.Agenda,List of Attendees & EPRI Rept Re SFP Seismic Failure Provided
ML20209E045
Person / Time
Issue date: 07/09/1999
From: Dudley R
NRC (Affiliation Not Assigned)
To:
NRC (Affiliation Not Assigned)
References
PROJECT-689 NUDOCS 9907140133
Download: ML20209E045 (27)
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5 NUCLEAR REGULATORY COMMISSION f WASHINGTON, D.C. 20666-0001 July 9,1999 ORGANIZATION Nuclear Energy Institute

SUBJECT:

SUMMARY

OF MEETING WITH THE NUCLEAR ENERGY INSTITUTE AND PUBLIC STAKEHOLDERS j i

On June 21,1999, the NRC staff met with representatives of the Nuclear Energy Institute (NEI) l and public stakeholders to discuss the status of the staff's efforts to assess the risk associated with the storage of scent fuelin spent fuel pools at sites in the decommissioning phase. The i agenda and attendance list are provided as Enclosures 1 and 2, respectively.

After opening remarks by the NRC, representatives from NEl, Lynnette Hendricks and Mike Meisner, proposed an agenda for a decommissioning workshop to be held in mid-July 1999. The goals of the workshop will be to develop an analytical framework for determining the level of risk associated with decommissioning facilities and to identify basic procedural and/or design commitments necessary to support the analytical basis. During the discussion of the agenda, Stuart Treby, Assistant General Counsel for Rulemaking and Fuel Cycle, stated that since these efforts are destinnd to result in rulemaking, care must be taken at the workshop to facilitate public involvemem Treby also stated that the workshop would be appropriate for collection and dissemination u mlormation upon which the proposed new rules would be based, but that it would not be appropriate to negotiate or develop a consensus on what criteria should compose the proposed rules. The cession concluded when the NRC agreed to provide proposed modifications to the agenda proposed by NEl.

Next, Glenn Kelly, of the NRC, summarized the approach of the staff's preliminary probabilistic risk assessment of spent fuel pools at decommissioning reactors. Mr. Kelly described in detail the event trees and coaditional probabilities associated with severa! different possible scenarios. Since there is little or no automated equipment at these spent fuel pools, operator action is a major factor in recovering from postulated faulted conditions. For many scenarios analyzed, operators may have up to 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> to take action to mitigate the event prior to an offsite release. A major area of discussion between the NRC and NEl related to what probability to assign for the operator failing to take adequate corrective actions to mitigate events when a long period of time is available to respond.

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NEl representatives provided the NRC with additionalinformation in the form of a draft Electric Power Research Institute (EPRI) report regarding spent fuel pool seismic failure frequency.

The report is provided as Enclocure 3.

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Mr Peter James Atherton raised questions regarding the potential for criticality induced by severe accidents in spent fuel pools. The NRC responded by stating that a criticality event coupled with a zirconium fire scenario had a much lower probability of occurrence than a 9907140133 990709 PDR REVGP ERONUNRC

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I l zirconium fire scenario without a criticality event. Furthermore, the consequences of a l zirconium fire coupled with a criticality event would only have a negligible increase in l consequences. Thus, based on information reviewed to date, criticality is not considered a l . significant risk concern.

r Richard F. Dudley, Senior Project Manager Decommissioning Section Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation Project No. 689

Enclosures:

1. Agenda j
2. List of Attendees
3. EPRI Report cc w/encls: See next page l

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I j Nuclxr En rgyinstituts _ Proj:ct No. 689 r.

l l cc: Mr. Ralph Beedle Ms. Lynnette Hendricks, Director l Senior Vice President Plant Support and Chief Nuclear Officer Nuclear Energy institute Nuclear Energy Institute Suite 400 Suite 400 1776 l Street, NW l 1776 i Street, NW Washington, DC 20006-3708 Washington, DC 20006-3708 Mr. Alex Marion, Director Mr. Charles B. Brinkman, Director Programs Washington Operations Nuclear Energy institute ABB-Combustion Engineering, Inc.

Suite 400 12300 Twinbrook Parkway, Suite 330 1776 i Street, NW Rockville, MD 20852 Washington, DC 20006-3708 Mr. David Modeen, Director Mr. Michael Meisner Engineering Maine Yankee Atomic Power Co.

Nuclear Energy Institute 321 Old Ferry Road  ;

Suite 400 Wiscassett, ME 04578-4922 l

1776 i Street, NW "

Washington, DC 20006-3708 Mr. Ray Shadis Friends of the Coast Mr. Anthony Pietrangelo, Director P. O. Box 98 Licensing Edgecomb, ME 04556 Nuclear Energy Institute Suite 400 Mr. David Lochbaum 1776 i Street, NW Union of Concerned Scientists Washington, DC 20006-3708 1616 P St. N.W.

Suite 310 Mr. Nicholas J. Liparulo, Manager Washington, DC 20036 Nuclear Safety and Regulatory Activities Nuclear and Advanced Technology Mr. Paul Gunter Division Nuclear Information Resource Service Westinghouse Electric Corporation 142416* St. N.W. - Suite 404 P.O. Box 355 Washington, DC 20036 Pittsburgh, PA 15230 Mr. Peter James Atherton Mr. Jim Davis, Director P.O. Box 2337

, Operations Washington, DC 20013 1

Nuclear Energy Institute Suite 400 Mr. H. G. Brack 1776 I Street, NW Center of Biological Monitoring Washington, DC 20006-3708 P.O. Box 144 Hull's Cove, ME 04644 Mr. Paul Blanch Energy Consultant 135 Hyde Road '

West Hartford, CT 06117 i

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l June 21,1999 Decommissioning Meeting l Agenda ,

1. Introduction 1:00 - 1:05pm Richard Dudley
2. Opening remarks 1:05 - 1:15pm Stu Richards -
3. Discussion of agenda for July workshop i (including objectives and outcomes)

NEl 1:15 - 2:15pm l

Questions 2:15 - 2:30pm I

4. Technical explanation of NRC 2
30 - 3:30pm risk review i

Glen Kelly / Mike Cheok l

l Questions 3:30 - 4:00pm

5. Public Comments 4:00 - 4:30pm l

Enclosure 1

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Attendance Decommissioning Public Meeting June 21,1999 NAME ORGANIZATION Larry Kopp NRC i John Hannon NRC Diane Jackson NRC Vonna Ordaz NRC Lany Pittiglio NRC-NMSS/PWM Tim Johnson NRC-NMSS/PWM Duke Wheeler NRC-NRR/PDIV-D Bill Huffman NRC-NRR/PrW-D Anthony Markley NRC-NRR/DiiiP/RGEB John Greeves NRC-NMSS Aby Mohseni NRC-NMSS Mark Rubin NRC Ken Heck NRC-NRR Mike Check NRC-NRR/DSSA Joseph Staudenimeir NRC-NRR/DSSA/SRXB Mark Satorious NRC-OEDO J. C. Shepherd NRC-DWM Sam Nalluswami NRC-NRR/PDIV-D Goutam Bagchi NRC-NRR/DE Jim O'Brien NRC-NRR/DIPM Falk Kantor NRC-NRR/DIPM Serge Roudier NRC-NRR/DIPM Richard Dudley NRC-NRR/DPlV-D Stuart Richards NRC-NRR/PDIV&D Glenn Kelly NRC-DSSA/SPSB Richard Barrett NRC-DSSA/SPSB George Hubbard NRC-DSSA/SPLB Stuart Treby NRC-OGC John Zwolinski NRC-DLPM Alan Nelson NEl Lynnette Hendricks NEl Parviz Moieni Southern California Edison Lanny Dusek Portland General Electric Company Dana Kelly INEEL Martin Somerville BNFL Rita Bowser BNFL Gerry van Noordenner Connecticut Yankee Mike Meisner MYAPC Enclosure 2

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DRAFT l 1

l EPRI TECHNICAL REPORT  ;

Evaluation of Spent Fuel Pool Seismic Failure Frequency j

in Support of Risk Informed Decommissioning Emergency Planning I

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i I

t b Duke Engineering Ed& Services.

nwinn Enclosure 3

1 1

I EXECUTIVE

SUMMARY

The overall objective of this effort is to perform a risk-informed evaluation whether inclusion of "beyond design basis accidents", particularly a Zircaloy oxidation reaction [ fire] accident as the basis for Decommissioning Emergency Planning is warranted.

The annual probability of a Zircaloy cladding fire, resulting from the loss of water from the spent fuel 4 i pool, is estimated in NUREG-1353

  • to have a mean value of 2 x 10 per reactor year for either the PWR or the BWR spent fuel pool. The Spent Fuel Pool (SFP) failure frequency due to seismic, used ,

in NUREG-1353, is documented in NUREG/CR-5176 m. The annual frequency of seismic induced i SFP failure documented in NUREG/CR-5176 was determined by convolving a family of seismic  !

hazard curves with a family of fragility curves. The family of seismic hazard curves was developed based upon preliminary results which were subsequen:ly published in LLNL 1989

  • The family of fragility curves was based on estimates of the seismic capacity of typical BWR and PWR spent fuel  !

pools. Since publication of NUREG-1353 and NUREG/CR-5176, the LLNL seismic haard results  !

were updated in 1993 *. Industry also published seismic hazard results at 61 Nuclear Power Plant ,

(NPP) sites in 1989

  • i Using the methodology to calculate SFP failure frequency due to seismic described in NUREG/CR- i 5176, along with the NUREG-1353 assumptions, the NUREG-1353 SFP release values have been updated based upon use of the LLNL 1993 and EPRI 1989 seismic hazard results. Using the LLNL i 1993 results the annual probability of a Zircaloy cladding fire, resulting from the los3 of water from  !

the spent fuel pool, is estimated to have a mean value of 5.6 x 10" per reactor year for either the PWR or the BWR spent fuel pool. Using the EPRI 1989 results the annual probability of a Zircaloy cladding d

fire, resulting from the loss of water from the SFP,is estimated to have a mean value of 1.8 x 10 per reactor year for either the PWR or the BWR spent fuel pool. On average, use of these updated seismic hazard curves results in a reduction in the SFP failure frequency across the population of plants by a factor of 8 when using LLNL 1993 and about 70 when using EPRI 1989.

The results of this analysis also meet the probabilistic acceptance criteria of Standard Review Plan (SRP) 2.2.3," Evaluation of Potential Accidents." This SRP provides a basis for inclusion or exclusion of potential accidents into the plant design basis. For operating NPPs, emerdency planning is required to ensure the continued protection of the public health and safety in areas around the nuclear facility in the event of a radiological emerg 2ncy. Application of the SRP 2.2.3 criteria provides a basis for elimination of the requirements for off-site emergency planning at decommissioning NPPs.

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TABLE OF CONTENTS

1.0 INTRODUCTION

. ........ .... . ..............................................................................1 2.0 . M ETH O DO L OG Y . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.I SFP Failure Frequency Due to Seismic - NUREG/CR-5176.... . ........ .. ... ........ 3 2.1.1 Generation of the Family of Seismic Hazard Curves ... .. ..... .............. ..... 3  !

2.1.2 Generation of the Family of Fragility Curves............................................. 4 2.1.3 Calculation of the Release Frequency Given SFP Failure. . . .................... 5 '

3.0 RESULTS....................................................................................................I1

4.0 CONCLUSION

S....... .. . ... .. . ......................................................................17

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5.0 RE F E REN C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . ....................................19 LIST OF FIGURES Figure 2.1: Family of Hazard Curves for Vermont Yankee Based on Use of LLNL - 1989 6 Figure 2.2: Family of Hazard Curves for Vermont Yankee Based on Use of LLNL - 1993 8 Figure 2.3: Family of Hazard Curves for Vermont Yankee Based on Use of EPRI - 1989 9 Figure 2.4: Comparison of Alternative Methods to Calculate Beta 10 Figure 3.1: Comparison of SFP Release Frequencies with the NUREG-1353 Figure of Merit 16 LIST OF TABLES i

Table 1.1: Summary of Accident Sequence Quantification from NUREG-1353 2 Table 3.1: Spent Fuel Pool Analysis 11 1

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1.0 INTRODUCTION

i The overall objective of this effort is to provide risk-informed evaluation whether inclusion of "beyond design basis accidents", particularly a Zircaloy oxidation reaction (fire] accident as the basis for Decommissioning Emergency Planning is warranted. This issue was satisfactorily resolved for all plants by NUREG-1353 in 1989. The conclusions remain valid today, because the decommissioning state does not adversely affect the results on which the conclusions were based. Since the publication of NUREG-1353, significant improvements have been made in the seismic hazard results on which the previous conclusions were based in particular, recent work by both the regulator and the industry has reduced the calculated seismic hazard, which is the dominant contributor to the overall spent fuel pool release frequency.

The purpose of this document is to describe the methodology and results of the seismic technical analysis used to demonstrate the above conclusions are valid. NUREG-1353

" Regulatory Analysis for the Resolution of Generic Issue 82 Beyond Design Basis Accidents in Spent Fuel Pools", dated April 1989 is considered a valid framework for this analysis.

Given the NUREG-1353 framework, the Spent Fuel Pool failure frequencies due to seismic was updated using more current seismic hazard results.

Table 4.7.1 of NUREG-1353 summarizes the frequency of spent fuel damage resulting from accident sequences which can result in the loss of water from the Spent Fuel Pool (SFP) either through drainage or boiling as a result ofloss of cooling. As described in Reference 1, the seismic event contributes over 90% of the PWR spent fuel damage probability, and nearly 95%

for the BWR. However, since publication of NUREG-1353, revisions have been made to the published seismic hazard results at those sites previously evaluated for SFP failure frequency.

In particular, revisions to the Lawrence Livermore National Laboratory (LLNL) seismic hazard results at 69 Eastern United States (EUS) sites was published in 1993. In addition, Electric Power Research Institute (EPRI) hazard results are also available at 61 EUS sites.

NUREG-1353 is considered a valid framework to calculate release frequencies at these sites.

SFP accident frequencies for other scenarios (Missiles, Aircraft crashes, etc.) as shown in Table 1.1, which is a verbatim copy of Table 4.7.1 in NUREG-1353, are considered valid for  !

  • his analysis. Only the SFP failure frequency due to seismic is updated. The SFP failure frequencies due to seismic used in the NUREG 1353 analysis are from NUREG/CR-5176. l Updates of the SFP failure frequency will be based on the methodology and inputs described in l NUREG/CR-5176. There are, r this analysis is in essence a NUREG-1353 analysis with new  !

seismic hazard curves uxd to calculate spent fuel pool failure frequencies. I Using the 1989 and 1993 Lawrence Livermore National Laboratory (LLNL) seismic hazard results at 69 sites east of the Rocky Mountains, and the 1989 EPRI results at 61 sites east of the Rocky Mountains, the SFP failure frequency at each site is calculated. The reduction in SFP failure frequency due to the use of the 1993 LLNL results and the 1989 EPRI results is quantified. Given the NUREG-1353 framework, and the updated SFP failure frequencies, release frequencies are calculated for each of the 69 sites. The mean annual probability of a Zircaloy cladding fire, due to loss of water from the spent fuel pool, is also calculated.

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l Table 1.1: Summary of Accident Sequence Quantification from NUREG-1353 Table 4.7.1 Summary of SFP Accident Frequencies l

l l I PWR Freauency BWR Freauency l

Accident Sequence Best Estimate Upper Bound Best Estimate Upper Bound (per R-year) (per R-year) (per R-year) (per R-year) l Structural Failures 1.0 E-8 1.0 E-7 1.0 E-8 1.0 E-7

1. Missiles 6.0 E-9 2.0 E-8 6.0 E-9 2.0 E-8
2. Aircraft crashes
3. Heavy Load Drop 3.1 E-8 3.1 E-7 3.1 E-8 3.1 E-7 l l

3.0 E-8 5.0 E-7 3.0 E-8(1) 5.0 E-7(1)

Pneumatic Seal Failures Inadvedent Drainage 1.2 E-8 1.0 E-7 1.2 E-8 1.0 E-7 1.4 E-6 6.0 E-8(2) 1.4 E-6

. Loss of Cooling /Make-up 6.0 E-8(2) 1.5 E-7 2.4 E-6 1.5 E-7 2.4 E-6 TOTAL 1.8 E-6 6.7 E-6 Seisntic Structural Failure Conditional Probability of Zircaloy Cladding Fire Given Loss of Water (High 1.0 0.25 Density Storage Racks)

NOTES:

1. BWRs do not,in general, use pneumatic refueling cavity seals, but other pneumatic seals are used in the transfer canal.
2. Includes beyond design basis seismic induced loss of cooling and make-up.

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2.0 METHODOLOGY.

! 2.1 SFP Failure Frequency Due to Seismic - NUREG/CR-5176 The methodology to calculate SFP failure frequency is described in NUREG/CR-5176. SFP failure frequency due to seismic is calculated by convolving the seismic hazard distribution with the seismic fragility of the SFP. The convolution process to numerically integrate the  :

family of seismic hazard curves with the family of fragility curves is described in Reference 6.

2.1.1 Generation of the Family of Seismic Hazard Curves The following assumptions and recommendations from Reference 5 were used to generate the family of seismic hazard curves for this analysis:

1. A lognormal distribution was assumed for the distribution of the uncertainty in i probability of exceedance at each acceleration value. The parameters of the lognormal

' distribution (i.e., median and logarithmic standard deviation (p)) were calculated by using the 50* and 95* percentile values.

2. Given the median and the 95th percentile the logarithmic standard deviation (p) is calculated (p = (In(x,3 /w x )/1.64)). p can also be calculated from the natural log of the ratio of the 85th percentile to the median. Given p, the probability of exceedance (X,)

can then be calculated at various percentiles (X, = x, 'e* p). Z is the standard normal variate.

3. Because it is possible to get probability of exceedance values greater than 1.0, the lognormal distribution is truncated at X,, (the 99 percentile). The lognormal

, distribution was normalized to get a new distribution with cutoff at X99.

4. The range of hazard represented by the truncated (lognormal) distribution at each acceleration was discretized into eleven discrete values of the hazard with subjective probabilities of 0.03,0.05,0.07,0.12,0.15,0.16,0.15,0.12,0.07,0.05, and 0.03.

, Figures 2.1,2.2, and 2.3 show the family of hazard curves generated for the Vermont Yankee site based upon use of the above process. Figure 2.1 is based upon use of the 1989 LLNL results and an estimate of p based on use of the 95* percentile as described in (2) above. Figure 2.2 is based upon use of the 1993 LLNL results and an estimate of p based on use of the 85*

percentile as described in (2) above. Figure 2.3 is based upon use of the 1989 EPRI results and an estimate of p based on use of the 85* percentile as described in (2) above. In Figures 2.2 and 2.3, is estimated using the 85* percentile because 95* percentile results are not available.

As can be seen, there is a significant reduction in the uncertainty between Figure 2.1 and Figures 2.2 and 2.3. At about 2 g, the uncertainty in Figure 2.1 ranges from about 5 x 10-5 to 4

about I x 10'", whereas in Figure 2.2 the uncertainty ranges from about 4 x 10 to 2 x 10-".

The Figure 2.3 results (EPRI) behave similar to the Figure 2.2 results.

Figure 2.4 compares estimates of p (logarithmic standard deviation) at each of the 69 sites using the LLNL 1989 results. As can be seen, based upon use of the 95* percentile is equal to or lower than p based on use of the 85* percentile in all cases. Use of p based on 85*

percentile estimates for the Vermont Yankee site would result in a wider uncertainty band than

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based on the use of the 85* percentile would result in SFP failure frequencies about a factor of 2 higher than values calculated based on use of the 95* percentile. As described earlier, LLNL 3 1993 does not contain 95* percentile results. Therefore, LLNL 1993 SFP failure frequencies based on use of the 83* percentile to estimate p are in general about a factor of 2 higher in probability than SFP failure frequencies would be if 95* percentiles were available to estimate p.

2.1.2 GeneraJon of the Family of Fragility Curves l

in previous SFP failure frequency analyses by Brookhaven National Laboratory (BNL)*,

l estimates of the seismic fragility of spent fuel pools were assigned by comparing with i published PRAs. For the Millstone 1 BWR, the seismic fragility developed for the Oyster Creek reactor wasjudged appropriate for the Millstone I spent fuel pool. For the Ginna spent fuel pool, the fragility of the Zion plant auxiliary building shear walls was used.

In NUREG/CR-5176, one objective was to develop realistic estimates of the seismic capacity l I

of typical BWR and PWR spent fuel pools. To accomplish this task, a detailed evaluation was performed of the Vermont Yankee (BWR) and Robinson (PWR) spent fuel pools. Structural drawings, the Final Safety Analysis Report (FSAR) and spent fuel pool reports (References 8 and 9) were reviewed. Based on this review, the seismic capacity of these two representative (

l spent fuel po.ils, a BWR and a PWR, was determined. For BWRs, the SFP fragility is defined by:

1 I The median fragility (x3a ) = 1.4g The random uncertainty pa = 0.26 The uncertainty in location pu = 0.39 For PWRs, the SFP fragility is defined by:

i The median fragility (x3o ) = 2.0g The random uncertainty p, = 0.28 The uncertainty in location pu = 0.40 i

l As described in Reference 6, the uncertainty in the median is described by the following j equation: 7 I . .

4 = A , esT (1) )

l where:

4 = uncertainty in the median,

= the " median median" fragility (1.4 for BWR,2.0 for PWR),

f A,,,

4

= the lognormal standard deviation of the a distribution, and l Du G = the standard normal variate.

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I Five curves, described by ( values ranging from -1.28, -0.58,0.0,0.58, and 1.28, are used to  !

define the uncertainty in median SFP fragilities. The basis for five fragility curves is described in Reference 10.

The desired discrete value for a fragility curve is then:

1 d

a = a, en (2) i l where: l l

a = acceleration value at a given failure frequency defined by Z, l s a, = a median fragility,

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i p, = the lognormal standard deviation of the random uncertainty about the l median, and j i

! Z = the standard normal variate.  !

l Equations 1 and 2 can be combined such that the failure frequency at given accelerations (a),

l usually those which describe the hazard curve, can be readily calculated. The final equation is.  ;

F,(a) = N(In(a/A,,, es"t)/ p,) (3)

F,(a) =

N(Z) (4) l where, (

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F,(a) = the fraction of eanhquakes to fail the SFP at acceleration a, and N(Z) = the area under the normal curve up to point Z.  !

1 2.1.3 Calculation of the Release Frequency Given SFP Failure l

The methodology to calculate the release frequency given SFP failure is described in NUREG- i i

1353. Table 1.1 of this report is a duplicate of Table 4.7.1 of NUREG-1353. As can be seen in Table 1.1, the annual probability of a SFP failure for a PWR is described by the sum (1.5 x d

10 ) of the SFP failure frequencies associated with Structural Failures, Pneumatic Seal Failures, Inadvertent Drainage, and Loss of Cooling Make-up plus the Seismic Structural i Failure. The annual probability of a release is the product of the annual SFP failure frequency i and the conditional probability of Zircaloy cladding fire given loss of water. For PWRs, the l conditional probability of the Zircaloy cladding fire is considered to be 1.0. Values less than 1.0 for a PWR are supponed by Table 4.5.1 in NUREG-1353. For a BWR, the process is ev.actly the same with the exception that the conditional probability of Zircaloy cladding fire given loss of water is 0.25. Values less than 0.25 for a BWR are supported by Table 4.5.1 in NUREG-1353.

Using this approach, SFP seismic failure frequencies were calculated at each of the 69 sites using the LLNL results and at 61 sites using the EPRI results.

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3.0- RESULTS Based un the methodology described in Section 2 the annual probability of a SFP failure due to seismic was calculated. The SFP failure frequency was calculated based on reproduction of the LLNL 1989 results, and use of the LLNL 1993 results, and the EPRI 1989 results. The annual probability of a release based upon the alternative SFP failure frequencies was also calculated.

Table 3.1 provides the overall results of the analysis. The site numbers in Table 3.1 are ordered the same as the Nuclear Power Plants (NPPs) listed in Table A.1 of NUREG-1353 with the exception that no western US NPPs were included in this analysis. Column 2 contains the code ,

specifying the plant at the site to be either a BWR or a PWR. Column 3 presents the SFP failure (

frequency based on use of the LLNL 1989 results and estimates of p based on the natural log of l the ratio between the 85* percentile and the 50' percentile. Column 4 presents the SFP failure frequency based on use of the LLNL 1989 results and estimates of based on the natural log of the ratio between the 95* percentile and the 50* percentile divided by 1.64. The column 4 SFP I failure frequencies are on average about a factor of 2 lower than the column 3 SFP failure frequencies. Column 5 presents the SFP failure frequency based on use of the LLNL 1993 results

)j J

-_and estimates of p based on the natural log of the ratio between the 85* percentile and the 50*

percentile. Column 6 presents the SFP failure frequency based on use of the EPRI 1989 results and estimates of p based on the natural log of the ratio between the 85* percentile and the 50* J percer, tile. Column 7 quantifies the reduction in SFP failure frequency based upon use of the 1993 LLNL results. Column 7 is the ratio of the LLNL 1989 (Column 4) and LLNL 1993 (column 5)

SFP failure frequencies. Column 8 quantifies the reduction in SFP failure frequency based upon I use of the 1989 EPRI results. Column 8 is the ratio of the LLNL 1989 (column 4) and EPRI 1993 (column 5) SFP failure frequencies. Columns 9,10, and 11 are the overall SFP release frequencies )

at each site. I As shown in Table 3.1, the average reduction in SFP failure frequency across the population of EUS sites was about a factor of 8 when the LLNL 1993 results were used and over a factor of 70 when the EPRI 1989 results were used relative to the SFP failure frequency when the 1989 LLNL I results were used. At some sites the SFP failure frequency increased slightly when the LLNL 'l 1993 results were used. This was due to use of the 85* percentile to estimate the logarithmic standard deviation p from the LLNL 1993 results.

Figure 3.1 is a plot of the annual probability of a release at the population of EUS sites based on the LLNL 1993 results and tla EPRI 1989 results. As can be seen, all NPPs are on the order of 10'

' or less based upon the LLNL 1993 results with an overall mean annual probability of 5.6 x 10.

4 Two plants are slightly above the figure of merit (2.0 x 10 ) presented in NUREG-1353. All 4

NPPs are less than 10 based on the EPRI 1989 results (overall mean = 1.8 x 10). In general there is excellent agreement between the LLNL and EPRI release frequency results with the 4

exception of those LLNL NPP results that exceed 104. All NPPs that exceed 10 based on LLNL 1 seismic hazard results are soil sites.

I1

{~

4f ffs[ pup i y 9 7 7

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c n 8-7 0- 0- 0-8 0-7 0-8 0-7 0-7 0-7 0-7 0-8 0-7 0-7 0-7 8 a e 0- 0-1 e u RI 7

E E E E E E E E E E E E E E E 1 l e q 2 6 6 5 5 8 9 9 0 0 3 3 2 8 R e P r

8 6 3 8 7 8 5 6 6 9 2 5 5 8 7 F E 1 1 2 3 1 5 1 1 1 2 5 1 1 1 5 y

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e q N 0 5 2

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1 2 7 7 8 3 3 2 4 8 8 7 3 2 3 0 6 5 F 5 1 5 2 2 1 8 3 2 5 3 5 1 7 1 2 5 1 6 -

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0 E E E E E E E E E E E E e 4 5 l

0 5 4 0 0 9 0 5 1 8 R e L r L 9 2 3 8 2 0 0 6 6 8 4 7 3 0 3 1 5 1 8 F 5 1 7 3 1 5 2 9 6 4 2 1 5 8 1 8 1 1 4 9

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l 5 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0-e -

u 6 I 8

x E E E E E E E E E E E E E E E F R =

0 7

0 2

0 6

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0 5

0 0 0 0 0 0 0 0 0 0 t P a 4 8 9 9 4 8 8 8 2 1 n E t e 3 1 8 4 2 8 4 1 1 1 5 2 2 3 8 e B 2 p 1 S 0 3 5

9 /

x 7 6 7 7 7 7 6 7 8 7 7 6 8 7 6 8 7 7 7 5 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0-1 -

3 5 L 8

x E E E E E E E E E E E E E E E E E E E 0 0 0 0 0 0 0 0 0 0 0 0 0 0 l

e N =

a 6 1 7 6 2 8 2 8 1 9 3 2 8 8 0

9 0

7 0

5 0

2 0

b L 1 t

3 1 3 9 5 3 7 3 a L e 1 1 2 2 2 5 4 8 3 5 1 T B 0

9 5 8 /x 6 6 6 6 6 5 6 7 6 7 6 6 7 7 6 7 6 6 6 5 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0 0- 0- 0-4 L 9

x E E E E E E E E E E E E E E E E E E E 0 0 0 0 0 0 0 0 0 0 0 0 0 N = 8 1 2 4 1 0 0 5 3 3 0 0 0 0 0 0 L

L t a

5 1 7 1 1 2 8 1

8 6 3 2 8 8 5 A 6 4 3 8 e 6 3 6 5 6 1 4 1 B

0 9 5 8

x 5 6 5 6 6 5 6 6 5 7 6 5 7 6 6 6 6 6

/

0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 6

- 5 0- 0- 0- 0-3 L 8

x E E E E E E E E E E E E E E E E E E E 0 0 0 0 0 0 0 0 0 0 0 0 0 N =

a 1 4 0 6 0 1 3 9 8 1 6 3 9 0

1 0

0 0

2 0 0 0 L 1 4 1 L t e 1 1 1 1 2 9 7 1 1 4 5 1 9 1 6 2 2 7 4 B

1 2

2 R R 2 2 2 1 2 1 1 2 2 2 2 1 2 2 2 1 2 W

B P W 1 1 e

1 it 1 2 3 4 5 6 7 8 9 0 1 2 3 4 5 6 7 8 9 S 1 1 1 1 1 1 1 1 1 1

l 4t{If!tfi, f L

l

_ e 8 7 8 7 8 7 7 8 7 7 7 8 7 7 7 7 7 8 0- 0- 0- 0- 0- 8 s

m 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0-1 a E E E E E E E E -

1 e 5 4 8 6 E E E E E E E E E E E l

e a 2 5 8 1 0 1 5 0 0 6 0 0 0 6 3 0 0 8 e 8 9 4 8 7 0 5 4 0 4 6 0 7 5

_ R r 7 1 4 1

8 F 1 3 3 1 6 1 4 2 9 1 2 2 1 3 8 4 y -

e s

c n 0-7 7 0-8 0-6 0-7 0-7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 8 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0-0 a e

e E u E E E E E E E E E E E E E E E E 0-E 0- 0-E 0- 0- -

E 1 l q 3 7 5 5 0 5 0 0 3 8 0 0 3 0 E -

e 3 4 2 4 3 8 0 8 0 8 0 e 8 2 3 7 8 4 R

F r 1 2 9 1 2 1 5 2 1

3 4 6 2

6 6

3 6

6 3

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9 4

8 3

5 9

y

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0-6 0-6 0-6 0-6 7 0-6 6 7 7 6 7 6 6 6 6 6 6 7 6

_ 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0-

_ a e E E E E E E E E E 0-

_ 9 l e uq 3 5 1 5 5 5 5 5 8 E E E E E E E E E E E e 6 3 5 0 6 4 5 5 9 5 3 4 R re 8 4 2 9 8 1 4 8 1 0 4 1 6 6 4 5 0 -

6 1 1 1 4 F 1 1 2 2 1 5 4 3 5 1 1 3 2 1 2 5 9

1 i

o E 8 9 5 2 0 7 5 9 3 4 6 8 4 t

9 0 7 1 5 9 0 8 a 9 / 7 5 2

2 0 9 5 0 1 8 6 1

1 5 1 1 1 0 4 R 8 1

4 1 5 1 1 4 1 9 4 2 1 3 4 1 1

1 1 2 1

L i

s 3 s 9 y io L 6 5 0 8 8 8 3 8 0 8 3 3 -

l 7 t a /

7 9 8 8 3 1 2 S. 4 -

c 9 6 5 0 3 2 5 0 2 0 5 0 5 4 s R 8 1 2 1 1 1 5 7 6 5 1 l

A L 1 2 o

P o 9 8

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0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- -

e 8 E E E E u 6 I x 5 E E E E E E E E E E E E E E F R 0 0 0 0 0 0 0 0 t

n P

E t

=

a e

4 1

7 3

5 4

6 3

2 6

9 1

1 3

2 1

0 5

2 0

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0 6

9 0

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5 _

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1 3

9 /x

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0-8 0-7 0-6 0-7 0-7 0-7 0-7 0-6 0-6 0-7 0-7 0-6 0-6 0-6 0-7 0-7 0-6 0-7 0-6 0-7 0-3 5 L 8

x E E E E E E E E E E E E E E E E E E E E E 0 0 0 0 0 0 0 0 0 0 0 0 e N 0 0 t

h c

L L t

=

a e

8 3

7 9

2 2

3 1

3 1

5 3

8 3

2 1

1 1

8 1

9 4

7 4

3 1

5 2

0 2

1 0

9 4

0 5

1 0

0 1

0 4

3 0

4 1

0 3

2 T B -

0 9 5 6 6 6 6 6 7 6

8 /x 6 6 6 6 7 6 6 6 6 6 6 6 6 -

5 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- -

9 0-4 L x E E E E E E E E E E E E E E E E E E E E -

N = 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 .

a 5 7 5 1 8 9 0 3 2 5 9 9 5 4 5 3 2 9 L 9 6 _

L t e 2 1 5 1 1 9 2 1 2 1 2 3 4 6 3 2 6 1 1 5 B

0 -

9 5 8 /x 6 6 5 6 6 6 6 6 6 6 6 7 6 6 6 6 5 6 6 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 5

- 5 0- 0- 0- 0- 0- 0- 0- 0-3 L 8

x E E E E E E E E E E E E E E E E E E E 0 0 0 0 0 0 0 0 0 0 0 0 0 E -

N = 9 4 9 8 5 9 1 2 3 8 7 0 0 0 0 0 0 0 _

L a 1 1 8 5 1 1 9 4 4 3 4 7 3 2 t

L e 1 1 1 2 2 6 6 5 9 5 7 1 3 2 1 B -

1 2 --

= n 2 R R 1 2 1 2 2 1 2 2 1 1 2 2 1 1 1 2 2 1 2 W W 1 1 B P e 0 1 2 3 4 5 6 7 8 9 0 1 it 1 2 3 4 5 6 7 8 9 0 -

S 2 2 2 2 2 2 2 2 2 2 3 3 3 3 3 3 3 3 3 3 4 -

y e

s c

n 7

0-7 0-7 0-7 7 8 7 7 7 8 7 8 7 7 7 7 7 7 8 7 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0-1 a e u E E E E E E E E E E E E E E E E 1 e q 0 0 0 5 0 5 E E E E le 3 0 1 5 6 1 6 0 0 1 0 4 3 0 e 5 4 0 9 5 2 1 2 7 2 1 9 2 5 0 5 5 9 2 6 R r 3 3 1 1 1 6 3 2 6 2 3 2 4 5 F 1 1 2 1 6 2 y

s e c n

7 0-7 0-7 0-7 0-7 0-7 6 7 7 7 6 7 7 7 7 7 7 7 7 7 7 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0-0 a e E E E E E E E E E E E E E E E 1 e u 0 0 E E E E E E q 8 0 8 0 4 0 0 5 5 5 0 0 0 0 0 le e 0 5 3 1 3 3 0 6 3 0 7 1 1 9 6 7 2 0

9 0

1 5

3 0

9 R r 4 4 5 6 3 3 5 7 8 9 5 3 F 1 1 1 1 4 4 6 2 4 y

e c 6 6 7 7 6 7 6 7 7 7 6 7 7 6 6 7 7 6 7 6 6 s n 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0-a e E E E E E E E E E E E E 9 e u 5 5 3 E E E E E E E E E q 0 1 8 4 0 0 3 5 3 0 5 5 9 0 5 0 9 le e 9 4 1 6 9 3 5 7 9 6 4 0 7 9 0 4 8 0 3 1 5

3 R r 1 4 4 8 6 5 6 6 7 2 5 7 2 F 1 1 4 3 5 4 1 2 9

io 8 1 1 E 0 6 0 8 7 0 0 4 7 0 7 5 0 7 0 5 0 8 t a / 3 3 4 9 9 2 6 5 6 4 0 7 5 9 9 0 5 6 9 9 6 0 R 2 1 1 2 2 2 2 1 1 1 0 4 6 4 2 8 1 1 L

i s 3 s

y o 9 L 3 7 8 4 it 2 8 5 3 5 8 3 9 8 1 6 7 5 9 6 8 5 l

a 7 a /

9 7 4 0 1 6 6 1 1 0 0 0 2 0 5 6 0 0 4 0 5 6 n R 8 1 1 1 1 A L l

o 0 o 9 5 P 8 /

x 7 7 7 8 7 7 6 8 8 7 8 9 8 7 7 0 7 8 8 7 l 5 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 1 0- 0- 0- 0-e - -

u 6 I 8

x E E E E E E E E E E E E E E E E E E E E F R =

0 0

0 9

0 5

0 5

0 5

0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 t P a 0 1 0 1 0 6 4 6 0 5 6 0 4 9 1 n E t e 2 1 2 4 4 1 1 7 2 1 6 6 7 3 3 9 1 4 9 1 ep B 4 1

S 0 3 5 9 /

x 7 7 6 7 6 7 5 7 7 7 6 7 7 7 7 7 7 7 7 7 7 1

5 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0-3 5 L 8

x E E E E E E E E E E E E E E E E E E E E E 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 e N =

a 5 0 0 6 2 7 2 1 8 7 6 1 6 4 1 2 7 4 6 9 4 l

b L 2 3 L t e 2 4 1 3 1 4 5 2 1 3 6 8 4 2 2 3 4 7 3 a

T B 0

5 9 x 6 6 6 7 6 6 5 7 7 6 6 7 7 6 6 8 7 6 7 6 6 8 / 0- 0- 0- 0- 0- 0- 0- 0 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0-5 -

4 L 9

x E E E E E E E E E E E E E E E E E E E E E 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 N = 8 3 5 1 5 4 2 2 4 9 3 6 2 8 9 9 3 9 8 6 2 L a L t e 1 4 1 7 7 2 2 5 5 2 1 6 4 4 6 9 2 4 2 4 2 B

0 9 5 8 /

x 6 6 6 7 5 6 5 7 7 6 6 6 7 6 5 7 7 5 7 6 6 5 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0-3 L 8

x E E E E E E E E E E E E E E E E E E E E E 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 N = 5 1 6 2 1 5 2 0 5 5 2 1 7 4 0 1 5 0 2 6 9 L a L t e 4 8 2 9 1 4 2 7 7 5 2 1 6 8 1 2 4 1 4 9 3 B

1 2

2 R R 2 2 1 2 1 1 1 2 2 1 2 1 2 2 2 2 2 2 2 1 2 W

B P W

e 1 2 3 4 5 6 7 8 9 0 1 2 3 4 5 6 7 8 9 0 1 1 it S 4 4 4 4 4 4 4 4 4 5 5 5 5 5 5 5 5 5 5 6 6

y _

s e c n

7 0-8 0-7 7 7 7 7 7 _

0- 0- 0- 0- 0- 0- .

a e 1

1 e u E E E E E E E E _

q 0 5 0 0 6 0 2 _

le e 1 2 1

5 5 7 1 0 8 R r 2 7 1 3 2 2 F 1 1 y

e s

c n

7 0-6 0-7 0-7 0-7 0-7 0-6 6 7 0- 0- 0-0 a e E E E E E E E E E 1 e u q 8 5 8 0 0 3 le 5 5 1 e 4 5 4 5 5 0 6 2 6 R r 2 1 2 4 5 2 2 1 5 F

y e

s c

n 7

0-6 0-6 0-7 0-6 0-7 0-6 7 6 0- 0- 0-a e E E E E E E E E E 9 e u q 0 5 6 0 5 0 5 0 le e 5 6 9 6 8 1 0 0 1

9 R r 6 1 1 6 5 8 3 9 F 1 9

o 8 0 0 3 5 0 9 0 9 8 it E 2 a /

9 5 5 6 8 3 3 5 3 R 8 2 5 9 2 3 4 1 7 L

i s 3 s

y o 9 L 3 5 l

7 it 1 1 2 7 2 7 2 a a /

9 5 1 9 1 4 2 1 0 8 n R 8 1 1 A L l

o 0 o 9 5 P 8 /

x 8 7 0 7 8 8 8 7 l 5 0- 0- 1 0- 0- 0- 0- 0-e -

u 6 I 8

x E E E E E E E E F R =

0 0

0 4

0 3

0 0

0 0

0 6

0 0 3 5

t P a n E t e 6 1 5 2 2 6 5 1 5

e p B 1 S 0 -

3 5 9 /

x 8 6 7 7 7 8 6 6 7 1

5 0- 0- 0- 0- 0- 0- 0- 0- 0-3 5 L 8

x E E E E E E E E E -

0 0 0 0 0 0 0 0 7 l

e N =

a 8 4 4 0 0 3 5 1 0 b L -

L t e 9 1 8 3 4 5 2 1 9 a

T B 0

9 5 8 /

x 7 6 6 7 6 7 6 7 6 5 0- 0- 0- 0- 0- 0- 0- 0- 0-4 L 9

x E E E E E E E E E 0 0 0 0 0 0 0 0 8 N =

a 0 5 7 1 7 6 9 5 2 L t 5 1 7 5 5 6 2 7 3 L e B

0 5

9 x 6 6 5 7 6 6 6 7 6 8 /

0- 0- 0- 0- 0- 0- 0- 0- 0-

- 5 3 L 8

x E E E E E E E E E 0 0 0 0 0 0 0 0 2 N =

a 4 2 5 3 6 1 8 1 5 L 6 6 2 4 9 6 L t e 1 2 1 B

1 2 e g

R R 2 2 2 2 2 2 a

r 2 2 1 e

W W v B P A e 2 3 4 5 6 7 8 9 1 it S 6 6 6 6 6 6 6 6

, l

. 1

. 1

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W

. "'*** .....,*:= W w f D

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o cr z y E - O y w * .h f g.......w u

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i

4.0 CONCLUSION

S l The following key results are derived from NUREG-1353:

  • The annual probability of a Zircaloy cladding fire, resulting from the loss of water from 4

the spent fuel pool, is estimated to have a mean value of 2 x 10 per reactor year for either the PWR or the BWR spent fuel pool.

  • The seismic event is the dominant contributor to the annual probability of a Zircaloy cladding fire resulting from the loss of water from the spent fuel pool.
  • The risk due to beyond design basis accidents in spent fuel pools, while not negligible, are sufficiently low such that no further risk reductions were warranted.

The SFP failure frequency due to seismic and used in NUREG-1353 is documented in NUREG/CR-5176. Since publication of NUREG-1353 and NUREG/CR-5176 the LLNL seismic hazard results have been updated. Industry also published seismic hazard results at 61 NPP sites.

Using the methodology to calculate SFP failure frequency due to seismic described in NUREG/CR-5176, along with the NUREG-1353 assumptions, the NUREG-1353 SFP release values have been updated based upon use of the LLNL 1993 and EPRI 1989 seismic hazard results. The average reduction in SFP failure frequency across the population of EUS sites was about a factor of 8 when the LLNL 1993 results were used and over a factor of 70 when the EPRI 1989 results were used relative to the SFP failure frequency using the 1989 LLNL results. Using the LLNL 1993 results, the annual probability of a Zircaloy cladding fire, resulting from the loss 4

of water from the spent fuel pool, is estimated to have a mean value of 5.6 x 10 per reactor year for either the PWR or the BWR spent fuel pool. Using the EPRI 1989 results, the annual >

probability of a Zircaloy cladding fire, resulting from the loss of water from the spent fuel pool, is ,

estimated to have a mean value of 1.8 x 104per reactor year for either the PWR or the BWR spent fuel pool. These results indicate that the mean risk due to beyond design basis accidents in spent fuel pools across the population of EUS sites is essentially negligible.

In addition, NUREG-1353 states that the high confidence low probability of failure (HCLPF) value for SFPs is estimated to be in the 0.5 to 0.65 g range, about three times the safe shutdown carthquake (SSE) peak ground acceleration values for typical EUS NPPs. The SFP niedian capacity is estimated to be in the 1.4 to 2.0 g range.10 CFR Part 100, Appendix III.(c) defines an SSE as:

"that earthquake which is based upon an evaluation of the maximum earthquake potential considering regional and local geology and seismology, and specific characteristics oflocal subsurface material. It is that earthquake which produces the maximum vibratory ground motion for which certain structures, systems, and components are designed to remain functional.

These structures, systems, and components are those necessary to assure:

l l (1) the integrity of the reactor coolant pressure boundary, (2) the capability to shutdown the reactor and maintain it in a safe shut down condition, or (3) f the capability to prevent or mitigate the consequences of accidents which l

could result in potential off-site exposures comparable to the guideline exposures of10 CFR Part 100."

17

The results of this analysis also meet the probabilistic criteria of SRP 2.2.3," Evaluation of Potential Accidents." This SRP provides a basis for inclusion or exclusion of potential accidents into the plant design basis. For operating NPPs, emergency planning is required to ensure the continued protection of the public health and safety in areas around the nuclear facility in the event of a radiological emergency. Application of the SRP 2.2.3 criteria provides a basis for elimination of the requirements for off-site emergency planning at decommissioning NPPs, as explained below.

The probabilistic acceptance criteria for exclusion of accidents, in SRP 2.2.3, is as follows:

"Accordingly, the~ expected rate of occurrence of potential exposures in excess of 10 CFR Part 100 4

guidelines of approximately 10 per year is acceptable if, when combined with reasonable qualitative arguments, the realistic probability can be shown to be lower." As can be seen in Figure 3.1, the LLNL 1993 mean results are on the order of 104 . The results of this analysis are conservative for the following reasons:

1. Loss of cooling / makeup is less probable for decommissioning nhnts because of fewer potential challenges to the fuel pool cooling / makeup system, as well as increased simplicity / reliability of the system.
2. Complete loss of SFP water is assumed given the seismic failure in NUREG-1353, however only a partial loss may actually result.
3. The conditional probability of Zircaloy cladding fire given loss of water for operating PWRs and BWRs has been assumed to be guaranteed and 0.25 respectively (bounding values).

Decommissioning PWRs and BWRs, experience spent fuel decay which immediately and continuously reduces this probability.

Based on the results of this analysis, it is concluded that the probabilistic acceptance criteria for exclusion of potential accidents which could result in radiological release in excess of the 10 CFR Part 100 guidelines is conservatively met and therefore, the need for off-site emergency planning for decommissioning plants is eliminated.

5.0 f

18

6.0 REFERENCES

1. U.S. Nuclear Regulatory Commission (USNRC)," Regulatory Analysis for the Resolution of Generic Issue 82, 'Beyond Design Basis Accidents in Spent Fuel Pools'," NUREG-1353, April 1989.
2. Lawrence Livermore National Laboratory (LLNL)," Seismic Hazard Characterization of 69 Nuclear Plant Sites East of the Rocky Mountains," NUREG/CR-5250, January 1989. -
3. U.S. Nuclear Regulatory Commission (USNRC), " Revised Seismic Hazard Estime' 4 for 69 Nuclear Plant Sites East of the Rocky Mountains," NUREG-1488, October 1993.
4. Electric Power Research Institute (EPRI),"Probabilistic Seismic Hazard Evaluation at Nuclear Plant Sites in the Central and Eastem United States: Resolution of the Charleston Issue," EPRI NP-6395-D, April 1989.
5. Lawrence Livermore National Laboratory (LLNL)," Seismic Failure and Cask Drop Analyses of the Spent Fuel Pools at Two Representative Nuclear Power Plants,"

NUREG/CR-5176, January 1989.

6. Kaplan, S.,"On the Method of Discrete Probability Distribution in Risk and Reliability Calculations", Risk Analysis, Vol.1,1981.
7. Sailor, V.L., K. R. Perkins, H. Connell, and J. Weeks, " Severe Accidents in Spent Fuel Pools in Support of Generic Safety Issue 82", NUREG/CR-4982, Brookhaven National Laboratory, June 1987.

E. Vermont Yankee Nuclear Power Station," Vermont Yankee Spent Fuel Storage Rack Replacement Report", April 1986. ,

9. " Spent Fuel Pool Storage Extension New Column Under Fuel Pool Floor", Calculation No.

HB 102 prepared by EBASCO Services Incorporated for Carolina Power & Lighting, March 1982.

10. Ravindra, M. K., Sues, R. H., Kennedy, R. P., and Wesley, D. A., "A Program to Determine the Capability of the Millstone Nuclear Power Plant to Withstand Seist.iic Excitation above the Design SSE"" draft report prepared for Northeast Utilities, Berlin CT, November 1984.

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n ... . ..

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( July 9, 1999 2

zirconium fire scenario without a criticality event. Furthermore, the consequences of a zirconium fire coupled with a criticality event would only have a negligible increase in consequences. Thus, based on Information reviewed to date, criticality is not considered a significant risk concern.

0 RIG.~ SIGNED BY Wm. Huffman FOR Richard F. Dudley, Senior Project Manager Decommissioning Section Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation Project No. 689

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