ML20212J022

Summary of 990608 Meeting with NEI Working Group Re NRR Staff on-going Work to Assess Risks Associated with Decommissioning.List of Meeting Attendees & Draft Technical Study of Sf Pool Accidents for Decommissioning Plants Encl

ML20212J022
Person / Time
Issue date:	06/17/1999
From:	Dudley R NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
PROJECT-689 NUDOCS 9906280295
Download: ML20212J022 (120)

Text

E ORGANIZATION: Nucle r Energy Institut) June 17, 1999

.i

SUMMARY

OF MEETING WITH THE NUCLEAR ENERGY INSTITUTE 5

The NEl Working Group on Decommissioning met publicly with the Director of the Office of Nuclear Reactor Regulation (NRR) and other members of the NRC staff on June 8,1999, to provide their thoughts on the staff's on-going work to assess the risks associated with decommissioning. A list of attendees is enclosed. The NEl Working Group was disappointed with the staff's progress R date. They felt that the preliminary results were far too conservative, did not reflect t!& Mual state of decommissioning activities, and that the staff efforts were becoming a much srger effort than needed, which would result in the final results of the effort being excessively delayed. The staff reiterated that the results were very preliminary, and that

. the intent of meeting with NEl, and with other stakeholders, was to receive input and comment while the review effort was in progress. A discussion was held regarding the most efficient and effective way to proceed. In order to expedite participation by the stakeholders in the staff's effo'ts to assess the risks of decommissioning, it was agreed that a workshop should be held in July, where the industry could describe those practices at decommissioning plants that could have bearing on the risk assessment. To prepare for this workshop, a planning meeting was tentatively scheduled. Subsequently, June 21,1999, was selected as the date for this meeting.

The staff will use this meeting to finalize details of the July workshop, and to prcvide more details of their assessment efforts to the stakeholders so that stakeholders could be better prepared for the workshop discussions. To facilitate preparation for the upcoming meetings, the staff has prepared a preliminary draft report summarizing the meth-dology, assumptions, and basic insights of the draft staff's risk review. This draft report is for discussion ourooses

.gnly and does not represent a final NRC product. Work on the report is still in progress, and the report has not received stakeholder input, nor has it received the appropriate reviews required for final issuance. The report may substantially change prior to final issuance. A copy j of the report is enclosed. ORIGINAL SIGNED BY:

Richard F. Dudley, Senior Project Manager Decommissioning Section Project Directorate IV & Decommissioning 9906290295 990617 Division of Licensing Project Management PDR REN9P ER C Office of Nuclear Reactor Regulation Project No. 689

Enclosures:

1. List of Attendees

2. Draft Technical Study

)

/

• For previous concurrences cc wiencis: See next page, , ~ '5 3 -

see atta hed ORC DISTRIBUTION: .uo" E-MAIL SCollins HARD COPY JZwolinski/SBlack SNalluswami TFredrichs Docket File (Project 689)

,03-SRichards SBrown MTschiltz PUBLIC T)7 MMasnik Pray MSatorius OGC CPoslusny MWebb THiltz PDIV-D r/f WHuffman TJohnson DLange RDudley AMarkley VOrdaz RBellamy,RI ACRS GHolahan GHubbard BJorgensen, Rill JHickey (T7-F27)

JMinns SMagruder DWheeler LPittiglio DBSpitzberg,RIV JHannon CCarpenter f[Q [M To receive a copy of this document, indicate "c" in the box OFFICE PDIV-D/Rh ,C PDIV-D/LA C PDIV-D/SC* SPLB:D* PDIV-D/D NAME RDudle)[ C7En$$ob MMasnik JHannon SRichards6 DATE b/ [99 b /G /99 6 /17 /99 6 /17 /99 bd/ /99 DOCUMENT NAN E: G:\PDIV-3\NEl\MTS6-8.wpd OFFICIAL RECORD COPY MWo Gi M P

.- ORGANIZATION: Nucle:r Energy institut3 June 17,.1999

SUMMARY

OF MEETING WITH THE NUCLEAR ENERGY INSTITUTE i

The NEl Working Group on Decommissioning met publicly with the Director of the Office of  !

Nuclear Reactor Regulation (NRR) and other members of the NRC staff on June 8,1999, to I provk.e their thoughts on the staffs on-going work to assess the risks associated with i decommissioning. A list of attendees is enclosed. The NEl Working Group was disappointed l with the staffs progress to date. They felt that the preliminary results were far too conservative, l did not reflect the actual state of decommissioning activities, and that the staff efforts were l

becoming a much larger effort than needed, which would result in the final results of the effort J being excessively delayed. The staff reiterated that the results were very preliminary, and that the intent of m.Seting with NEl, and with other stakeholders, was to receive input and comment while the review effort was in progress. A discussion was held regarding the most efficient and effective way to proceed. In order to expedite participation by the stakeholders in the staff's efforts to assess the risks of decommissioning, it was agreed that a workshop should be held in July, where the industry could describe those practices at decommissioning plants that could have beating on the risk assessment. To prepare for this workshop, a planning meeting was tentatively scheduled. Subsequently, June 21,1999, was selected as the date for this meeting.  :

The staff will use this meeting to finalize details of the July workshop, and to provide more '

details of their assessment efforts to the stakeholders so that stekeholders could be better prepared for the workshop discussions. To facilitate preparation for the upcoming meetings, the staff has prepared a preliminary draft report summarizing the methodology, assumptions, and basic insights of the draft staffs risk review. This draft report is for discussion ourooses 90}Y and does not represent a final NRC product. Work on the report is still in progress, and the report has not received stakeholder input, nor has it received the appropriate reviews required for final issuance. The repod may substantially change prior to final issuance. A copy of the report is enclosed. ORIGINAL SIGNED BY:

Richard F. Dudley, Senior Project Manager  !

Decommissioning Section i Project Directorate IV & Decommissioning ,

Division of Licensing Project Management Office of Nuclear Reactor Regulation Project No. 689

Enclosures:

1. List of Attendees

2. Draft Technical Study

• For previous concurrences cc w/encis: See next page see attached ORC DISTRIBUTION:

E-MAIL SCollins HARD COPY

- JZwolinski/SBlack SNalluswami TFredrichs Docket File (Project 689)

SRichards SBrown MTschiltz

• PUBLIC MMasnik Pray MSatorius OGC CPosiusny MWebb THiltz PDIV-D r/f

' WHuffman TJohnson DLange RDudley AMarkley VOrdaz RBellamy,RI ACRS

- GHolahan GHubbard BJorgensen, Rill JHickey (T7-F27)

JMinns - DWheeler DBSpitzberg,RIV CCarpenter SMagruder LPittiglio JHannon To receive a copy of tnis accument, inoicate v in tne Dox OFFICE ' PDIV-D/Rb N A PDIV-D/LA C PDIV-D/SC* SPLB:D* PDIV-D/D

^

NAME- RDudley[ CJa"meNoM MMasnik JHannon SRichards6 DATE )/ - [99 G/ M /99 6/17 /99 6 /17 /99 b /d /99 DOCUMENT NAN E: G:\PDIV-3\NEl\MTS6-8.wpd 1 OFFICIAL RECORD COPY

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q e nC84 9 )

p k UNITED STATES l g j e

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20666-0001 J

• • • • • $ June 17, 1999 i

ORGANIZATION: Nuclear Energy Institute

SUMMARY

OF MEETING WITH THE NUCLEAR ENERGY INSTITUTE

{

The NEl Working Group on Decommissioning met publicly with the Director of the Office of j Nuclear Reactor Regulation (NRR) and other members of the NRC staff on sune 8,1999, to '

provide their thoughts on the staff's on-going work to assess the risks associated with decommissioning. A list of attendees is enclosed. The NEl Working Group was disappointed  !

with the staff's progress to date. They felt that the preliminary results were far too conservative,  !

did not reflect the actual state of decommissioning activities, and that the staff efforts were becoming a much larger effort than needed, which would result in the final results of the effort being excessively delayed. The staff reiterated that the results were very preliminary, and that the intent of meeting with NEl, and with other stakeholders, was to receive input and comment while the review effort was in progress. A discussiun was held regarding the most efficient and effective way to proceed. In order to expedite participation by the stakeholders in the staff's efforts to assess the risks of decommissioning, it was agreed that a workshop should be held in July, where the industry could describe those practices at decommissioning plants that could have bearing on the risk assessment. To prepare for this workshop, a planning meeting was tentatively scheduled. Subsequently, June 21,1999, was selected as the date for this meeting. 4 t

The staff will use this meeting to finalize details of the July workshop, and to provide more I details of their assessment efforts to the stakeholders so that stakeholders could be better l prepared for the workshop discussions To facilitate preparation for the upcoming meetings, j the staff has prepared a preliminary draft report summarizing the methodology, assumptions, and basic insights of the draft staff's risk review. This draft report is for discussion ourposes only and does not represent a final NRC product. Work on the report is still in progress, and .

the report has not received stakeholder input, nor has it received the appropriate reviews I required for finalissuance. The report may substantially change prior to finalissuance. A copy of the report is enclosed.

I N Richard F. Dudley, Senior Project Managar Decommissioning Section Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation l Project No. 689

Enclosures:

1. List of Attendees

2. Draft Technical Study cc w/encts: See next page i

1 Nucle:r En:rgy in:titut3 Projict No. 689 cc: Mr. Ralph Beedle Ms. Lynnette Hendricks, Director Senior Vice President Plant Support and Chief Nuclear Officer Nuclear Energy Institute Nuclear Energy Institute Suite 400 '

Suite 400 1776 l Street, NW 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, Maryland 20852 Washington, DC 20006-3708 J

Mr. David Modeen, Director Mr. Michael Meisner Engineering Maine Yankee Atomic Power Co.

Nuclear Energy Institute 321 Old Ferry Road Suite 400 Wiscassett, Maine 04578-4922 1776 i Street, NW I Washington, DC 20006-3708 Mr. Ray Shadis Friends of the Coast Mr. Anthony Pietrangelo, Director P. O. Box 98 Licensing Edgecomb, ME 04556 l Nuclear Enerriy institute '

Suite 400 Mr. David Lochbaum 1776 l Street, to, 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 Informetion Resource Service Westinghouse Electric Corporation 142416* St. N.W. - Suite 404 i P.O. Box 355 Washington, DC 20036 Pittsburgh, Pennsylvania 15230 l Mr. Peter James Atherton l Mr. Jim Davis, Director P.O. Box 2337 1 Operations Washington, DC 20013 i 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

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, I i

DECOMMISSIONING PUBLIC MEETING ATTENDANCE LIST

\

l June 8,1999 l l

l NUCLEAR ENERGY INSTITUTE Lynnette Hendricks Paul H. Genoa l EEB1 Albert Machiels SERCH LICENSING /BECHTEL '

Althea Wyche

$_Qf l A. E. Scherer l

MYAPC Mike Meisner l

NRC l

Sam Collins Chet Poslusny Stuart Richards Anthony Markley John Zwolinski Stewart Brown Larry Pittiglio l John Hannon l

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I Enclosure 1 i

j I

L L1

DRAFT DRAFT Technical Study of Spent Fuel Pool Accidents for Deg@ning Plants l

DRAFT l Technical Working Group:

Group Leader.Vonna Ordaz, DSSA/SPLB l

Members: Goutam Bagchi, DE Christopher Boyd, RES Michael Cheok, DSSA/SPSB Edward Connell, DSSA/SPLB Tanya Eaton, DSSA/SPLB Edward Ford, DIPM/lOMB Christopher Gratton, DSSA/SPLB Ken Heck, DIPM/lOMB Diane Jackson, DSSA/SPLB -

Glenn Kelly, DSSA/SPSB Larry Kopp, DSSA/SRXB James O'Brien. DIPM/lOLB Jason Schaperow, RES Joseph Staudenmeier. DSSA/SRXB Edward Throm, DSSA/SPSB t

Enclosure 2  ;

j

DRAFT Technical Study of Spent Fuel Pool Accidents for Decommissioning Plants ggyntents EXECUTIVE

SUMMARY

. . . . ... ...... . . .. .. .... .... . ...... . .1 1.0 . 5 Introduction . . . . . . . . . . . . . D R A. F T . . . . . . . . . . . . . . . . . . .

2.0 Deterministic Evaluation of Spent Fuel Heatup . . . . . . ... . . ........ . .7 2.1 Fire Protection Evaluation of the Phenomena of a Zirconium Fire . . . . . . . . .7 2.2 Spent Fuel Heatup Analyses ........... ... ....... .... . .... 10 2.3 Potential for Criticality . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.0 Risk Assessment of Spent Fu k at Decommissioning Plants . . . . . .. 23 3.1 Beyond Design Basis Spent Fuel Pool Accident Scenarios (intemal Event % ~. 26 3.2 Beyond Design Basis Spent Fuel Pool Accident Scenarios (Extemal Events) 32 3.3 Probabilistic Assessment insights . .... .. .. .. ... . ... . 45 4.0 Supporting information . . . . . . ..... .. ... . . . .... .. ... ... 48 4.1 Maintenance Rule . ... . . . ... . .. ........ ....... 48 4.2 Quality Assurance . .. ...... ... ...... .. ............. 50 4.3 Water-basin Type Independent Spent Fuel Storage installations ....... . 51 4.4 Design Basis Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 5.0 References . . . . . . . . . . . . . . . . . . .... ..... ... ...... ....... .. ..... 56 6.0 Acronyms . . . . . . . . . . . . . .. . ... ........ .. .. ......... . 60 Appendix 1 Assumptions for Spent Fuel Pool Probabilistic Risk Assessment . . . . . . A1-1 Appendix 2 Decommissioning Plant Spent Fuel Pool Event Trees and Fault Trees. . . . . A2-1 Appendix 3 Assumptions for Spent Fuel Heatup Time Estimate . . . . . ... . .. . A3-1 Appendix 4 Structural Integrity of Spent Fuel Pool Structures Subject to Tomados and High Winds . .. .. .. . . . ... . . . A4-1 Appendix 5 StructuralIntegrity of Spent Fuel Pool Structures Subject to Heavy Load Drops .. . .. ., .. .. . .. .... . A5-1 Appendix 6 Aircraft Crashes into Spent Fuel Pools . ... . .A6-1

List of Tables Table 3.1-1 Loss of Spent Fuel Pool Cooling Initiating Events . . . . . . . . . . . . . . . . . . . . . 37 ,

Table 3.1-2 Event Probabilities / Frequencies Used in the SFP Risk Analysis . . . . . . . . . . . . 38 Table 3.1-3 Spent Fuel Pool CoolingMAh{s Frequency of Fuel Uncovery (per year). 45 List of Figures Figure 2.2-1 Adiabatic Heatup Time m Rbjs:c gn n Time for a PWR. . . . . . . . . . . . . . 20 Figure 2.2-2 Adiabatic Heatup Time W.'ShM own Time for a BWR. . . . . . . . ..... . . . 20 Figure 3.0-1 Representative Spent Fuel Pool Model . . . . . . . . . ..................36 DRAFT

EXECUTIVE

SUMMARY

After a nuclear power plant is ut down and the reactor is defueled, the traditional accdont sequences that dom reactor risk are no longer applicable. The predominant source of risk remammg at permanently shutdown plants involves accdonts associated with spent fuel stored in the spent fuel pool. Previous NRC-sponsored studies have evaluated unlikely severe accdont scenarios that involve drainage of the spent fuel pool coolant and shielding water that uncoi storage configurations and de- e,K tDe spent fuel assemblies could heat up sh to asperg temperature at which rapid runaway oxidation of the zirconium fuel cladding (zirconium fire) might occur and result in cladding failure and a large offsite release of radioactive materials from the spent fuel. . When evaluatmg the acceptability of license exemption requests from regulations in the areas of emergerny preparedness and insurance for permanently shutdown plants, the staff has EE:::Q- lbility of the spent fuel to a zirconium fire accdont.

Since the exemptions were grFnt on a plant-specific basis, they have resulted in different analyses and criteria used for the bases of the exemptions, in some cases, the staff has requested heatup evaluations of the spent fuel cooled only by air A temperature of 565 *C ,

based on the onset of clad swelling was used as a conservative limit to ensure no radiological I release occurred. However, differences in licensee and NRC evaluations of spent fuel heatup

' phenomena have resulted in questions from industry stakeholders regarding the NRC's j technical and regulatory bases for evaluating specific spent fuel pool severe accident scenarios for decommissioning plants, d

To increase the efficiency and effectiveness of decommissioning requirements, the staff acknowledges that a predictable l risk-informed approach for rulemaking and for the review of 4

exemptions during the interim period prior to the completion of rulemaking needs to be l established. This need was discussed with the Commission during a meeting on  !

March 17,199g. On the basis of this need and the lack of a clear technical basis for evaluating severe accidents in spent fuel pools, the staff formed a technical working group to assess the existing technical and risk information on spent fuerl pool accidents at decommissioning plants.

The working group plans to develop a risk-informed technical basis to establish a predictable method for reviewing exemption requests and for followup rulemaking activities. The working group was also tasked to identify the need for any research in areas of large uncertainty. The staff considers that such an approach will contribute to safety and reduce unnecessary regulatory burden, as well as, increase public confidence.

The working group performed generic calculations and assessments of the heatup of the fuel and cladding to determine the potential for zirconium oxidation and ignition. The working group also performed a probabilistic analysis for spent fuel pool accidents at decommissioning plants.

If the decay heat of the spent fuelis sufficient to raise the temperature of the fuel and clad to the point that self-sustained zirconium oxidation would occur, then the heat of the oxidation reaction would further increase the temperature to cause the ignition of the zirconium. A zirconium fire would raise the fuel temperature to.jufficiently high temperatures that the radionuclides in the

' fuel could be released.

The most extensive work to date was in support of Generic Safety issue (GSI) 82, "Beyond Design Basis Accidents for Spent Fuel Pools." (Ref. 3]. The working group used the studies in 1

References 1 and 2 as the starting point for its assessment in support of GSI 82. The working group noted that charactensbcs ofrzirconium a zirconsum fire had not been well defirwd in previous S fires bum with little flame, generate little studes. A isterature review fo smoke, but attain high temper . ym temperatures for zirconium were reported at temperatures between 1281 *C and 1800 *C The heatup analyses performed in support of GSI #? reported zirconium oxidabon temperatures between 850 - 900 *C. The working group astames inat the latter temperature range _is for the onset of oxidahon that precedes runaway oxidation or ignition. The wosi a c oted that the previous SFP studies provided good

insights into the phenomena c lim oxidation. The studies identified that the initiation of a zirconium fire was highly dependent on decay power and fuel storage configuration. In reviewing the studes, the wortung group identified that operating reactor spent fuel management prachces may affect the calculated decay time in the spent fuel heatup analysis recessary to preclude a zirconium fire. Some of these prachces include the increase in fuel bumup, which leads to higher decay power, gpl storage racking, which could reduce heat removal.

The working group's preliminary results indicate that on a generic basis, the decay time required to maintain cladding temperature below the self-sustaining zirconium oxidation temperature uring air cooling only may be longer than the generic studies performed for operating reactors.

- Decay time is generally defined as the length of time elapsed since reactor shutdown for the most recently discharged fuel. However, previoas plant-specific emergency preparedness exemptions are unaffected because they were approved using analyses that reflected the actua conditions at the particular plant. The working group has identified two potential deterministic criteria for assessing the potential for a zirconium fire, which are described below.

One potential criterion for allowing the reduction of existing regulation with respect to emergency preparedness (and possibly other regulatory areas) is the determination that the decay h I sufficiently low that air cooling is adequate to maintain the clad temperature below the point o self-sustained zirconium oxidation. The working group's preliminary estimates, using generic, near-bounding thermal-hydraulic spent fuel heatup assumptions, indicate that 3 to 5 years of decay time may be needed to reach a point at which air cooling of the fuel is sufficient to prev self-sustained zirconium oxidation. The working group notes that a plant-specific analysis, using actual parameters such as decay heat and spent fuel pool configuration, should yield shorter time estimates. This type of analysis has been relied upon for several previous exemption requests using a criteria based on a maximum fuel temperature of 565 *C, which is the temperature for the onset of clad swelling. However, the working group recognized that the criteria to demonstrate that clad swelling could not occur was conservative. The working group reviewed the temperature criteria used in the spent fuel analysis, and the preliminary results indicate that a maximum allowable temperature of 800 *C may be acceptable if certain boundary conditions are met. These conditions would include criteria such as demonstrating that the maximum temperature including uncertainties remained below the temperature limit, including a model for higher temperature effects, and demonstrating that a release of the radionuclides in the gap between the clad and the fuel was not a concern. The temperature limit was base

. the lowest temperature for the onset of self-sustaining zirconium oxidation identified by the GSI 82 studies.

The second potential criterion for allowing the reduction of existing regulation with respect to emergency preparedness is the determination such that sufficient time is available after th 2

is uncovered that mitigative actions, and if necessary, offsite protective measures could be taken for the public without proplantwng. The working group performed genenc, bounding calcuisbons to correlate the &-  ; me fuel smce final shutdown to the heatup time of the fuelfrom uncovery to runawad oxidahon The calculations were conservatively based on the heatup time for the hottest rod to heat up from 30 to 900 *C assuming adiabatic condibons (no heat is lost). The working group recognizes that the conditions are not realistic because some heat removal would occur. However, it would encompass additional events such as a piece of flat material falli i a _byliding wall or roof on top of fuel assemblies. This type of calculation was used to su :prdvious plant-specific emergency preparedness exemptions. The working group's preliminary generic results indicate that at 2 years of decay time for a boiling-water reactor, and 2.5 years for a pressurized-water reactor, about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> will be available for mitigative actions and offsite protective measures before onset of runaway zirconium oxidahon. The staff previously determined that 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> was sufficient time to take mitigative achons and, if q,Afpif protective measures without preplanning in its evaluation of a plant-specrfic emergency p{reparedness exemption. However, furth is needed to determine whether a generic time to take protective actions without preplanning is appupiiate for all plants (or a group of plants) or whether a site-specific time should be determined for each plant. The staff would need to coordinate with the Federal Emergency

, Management Agency in this evaluation. In addition, a more realistic calculation using plant-specific parameters or not using adiabatic conditions could yield shorter estimates of the decay time needed to achieve a given heat up time.

In the study, the working group identified that decay heat from Branch Technical Position (BTP)

ASB 9-2 [Ref. 4] should not be used as the basis of decay heat inputs for fuel heatup calculations in spent fuel pools. BTP ASB 9-2 is only valid for times less than 10,000,000 seconds (115 days). The draft ANS-5.1/N18.6 - 1973 standard [Ref. 5), which is the bases for BTP ASB 9-2 may be used to calculate decay heat if the applicable uncertainties are included. However, BTP ASB 9-2 is non-conservative in the approximate time range of 1 to 4 years after shutdown compared to best estimate ORIGEN computer code calculations.

The working group reviewed the potential for criticality and the preliminary results indicate that criticality is not a concem for decommissioning plants. Several scenarios were evaluated for PWRs and BWRs and were not considered credible. The only exception that could be identified was if Boraflex was significantly degraded and a raw water reflood of the spent fuel pool occurred assuming that the plant credits Boraflex for subcriticality margin.

The working group also performed a preliminary probabilistic analysis of the initiating events that could lead to fuel uncovery at decommissioning plants. The analysis considered a wide range of initiating events. An important assumption in this type of analysis is the amount of redundancy and diversity of spent fuel pool heat removal systems, spent fuel pool makeup

. systems, and their support systems. On the basis of information gathered by the working group on site visits to four decommissioning plants, the system configurations analyzed had significantly reduced levels of redundancy and diversity in these areas relative to operating plants. In the analysis, it was assumed the spent fuel had 1 year of decay time and the spent fuel pool support systems were representative of plants that had been shut down for 2 or more years. The working group notes that no decommissioning plant today matches the conditions assumed in this risk assessment. Also, the conditions assumed in the working group's risk 3

l assessment do not apply to operstmg plants because of differences in areas such as support systems and personnel availability. The analysis calculated the frequency of fuel uncovery.

=

i to the frequency of a zirconium fire. A zirconium Frequency of fuel uncoveryis fire may result only from fuel E a limited penod after final shutdown The working M7u}ing group's preliminary results mdicate that a seismic event may not be the largest contributor to risk; there are several other credible initiators for decommissioning plants including: intemal fires; loss of coolant inventories; cask drop; loss of offsite power; tomado missiles; loss of pool f site visits, current decommissioning plant cooling; and aircraft impact. gq

's ' bed herein, the working group made a configuratsons, and probabilisach

' preliminary estimate that the frequency of fuel uncovery is about 1E-5 per year. No single initiating event is dominant for uncovering the fuel and the frequency of many of the initiating events is heavily influenced by human error.

'ng group reviewed the implementation of the In support of the risk a5:::=Qg maintenance rule and quality assurance ( A) programs for decommissioning plants and the design of water-basin inoependent Spent Fuel Storage Installations (ISFSis). Implementation o the maintenance rule [Ref. 6) has been essentially the same as for operating facilities; however, the population of systems, structures, and components (SSCs) within the scope of the rule is reduced. For Appendix B QA programs, licensees generally retain their approved Appendix B QA programs [Ref. 7] through the decommissioning phase. Changes to the QA program commitments are submitted for NRC approval through the 50.54(a)(3) [Ref. 8] change control process. In addition, the working group reviewed the regulatory guidance for the design of a water-basin ISFSI,' which is licensed under 10 CFR Part 72 [Ref. 9). The concept of defense in depth is maintained in the design of the water-basin ISFSI by multiple means to protect the clad as the fission product barrier and to alert personnel to the event that could cause a breach in the clad. . Many systems and controls were identified to maintain defense in depth to protect against the release of radioactivity.

Since the Commission meeting on March 17,1999, the staff has held three public meetings with stakeholders to discuss the mission of the working group, the outline of the team's effort, the preliminary results as described herein, and the plans for the stakeholders' further involvement. (

'On June 7,1999, the working group presented the preliminary results to the Nuclear Energy Institute and the other stakeholders for discussion purposes only and stated that it would be premature to apply any findings to the regulatory process at this time. This preliminary draft of portions of the study, which include the scope of the study, the assumptions used in the deterministic and probabilistic analyses, and basic insights is being released to facilitate a public workshop on the study to be conducted in July of 1999. After the workshop, the working group plans to involve outside technical organizations with the report for an independent, technical, quality review to further refine the working group's assumptions and results. The review process is expected to be completed in December 1999. The working group will consider the information shared in the workshop and the results from the independent, technical, quality review process before completing the final report of the risk-informed technical basis for reducing existing regulatory requirements with respect to emergency preparedness, insurance, and possibly other requirements at decommissioning plants in March 2000. Depending on the outcome of the workshop in July 1999, the staff may shorten the schedule for completing the final report.

4

_ . ... .. i.. j

1.0 Introduction As the number of a decommig g Dr ts increase, the ability to address regulatory issues generically has become more m m i. After a nuclear power plant is permanently shut down and the reactor is defueled, the tradihonal accident sequences that dominate operstmg reactor risk are no longer applicable. The predominant source of risk remaining at permanently shutdown plants involves accidents associated with spent fuel stored in the spent fuel pool.

Previous NRC-sponsored stuM11uated unlikely severe accident scenarios that

' involve drainage of the spent fuel pool coolant and shielding water that uncover the spent fuel.

Given certain combinations of spent fuel storage configurations and decay times, the spent fuel assemblies could heat up to a temperature at which rapid runaway oxidation of the zirconium fuel claddmg (zirconium fire) might occur and result in cladding failure and a large offsite release of radioactive materials from tg g When evaluating the acceptability of i decommissioning licensee exemphon requests from regulations in the areas of emergency preparedness and insurance for permanently shutdown plants, the staff has assessed the susceptibility of the spent fuel to a zirconium fire accident.

Since the exemptions were granted on a plant-specific basis, they have resulted in different analyses and criteria used for the basis of the exemptions. In some cases, the staff has requested heatup evaluations of the spent fuel cooled only by air. A temperature of 565 *C based on the onset of clad swelling was used as a conservative limit to ensure no radiological 4 release. This criterion was used because of national laboratory studies that had identified the potential concem for a significant offsite radiological release from a zirconium fire which may oc:ur when all water is lost from the spent fuel pool. Zirconium is used in an alloy to make the hollow rods, called cladding, that hold the fuel pellets. The material is also used in boiling water reactors as a channel box around each fuel assembly. If the decay heat of the spent fuel is

' sufficient to raise the temperature of the fuel and clad to the point that self-sustained zirconium oxidation would occur, then the heat of the oxidation reaction would further increase the temperature to cause the ignition of the zirconium.

However, differences in licensee and NRC evaluations of spent fuel heatup phenomena have resulted in questions from industry stakeholders regarding the NRC's technical and regulatory bases for evaluating specific spent fuel pool severe accident scenarios on decommissioning plants.' The staff acknowledged that a predictable, risk-informed approach for requesting and granting exemptions is needed. This need was discussed with the Commission during a )

meeting on March 17,1999. As a result, the staff formed a technical working group to develop a risk-informed technical basis to establish a predictable method for reviewing exemption request and for followup rulemaking activities. The working group was also tasked to identify the need for any research in areas of large uncertainty. The staff considers that such an approach will contribute to safety and reduce unnecessary regulatory burden, as well as, increase public ,

confidence. l The working group performed a generic deterministic calculations and an assessment of the ]

previous spent fuel pool studies. Section 2.0 provides a review of the available technical information for deterministic evaluations of the heatup of spent fuel when only air cooling is available. The' working group's assessment includes the characteristics of a zirconium fire, the spent fuel failure criteria, existing spent fuel heatup analyses, available generic code i evaluations, heatup calculation uncertainties and sensitivities, critical decay times for reaching l s

l sufficient air coolmg, an estimated heatup time of uncovered spent fuel and the po enbcality as a result of a zirconium fire event.

i ninary probabilistic analysis of the initiating events that The worlong groJp also perfor could lead to fuel uncovery and the consequences t dof a zirconium calculated fire at decommissi the frequency i

. plants. The analysis considered a wide range of initiat ng even s v An assessment of spent fuel pool risk with no release frequency from a re associated operating reactor 7th long decay times introduced new analytical issues for the staff. Risk to the public depends on the event probability and consequences. Fo permanently shut down, the probability of some events (such as a seismic eve However, due to changes in equipment and personnel, other events may become more probable in a decommissioning plant. The consequences and, therefore, the may be reduced though gpive actions. The working group has considered a controls on certain risk parameters that could affect the safe the initiating event probabil, storage of the spent fuel. Section 3 provides the preliminary results of the risk assessm which evaluated intemal initiating events (e.g., intemal initiator frequencies, event tree trees, equipment failure conditional probabilities and human errors), and extemal events (e.g.', seismic, tomado, and aircraft).

The availability and reliability of systems, structures and components at decommissioning p could affect the probability and consequences of events, As supporting information, the w group considered the effects of the Maintenance Rule [Ref. 6] and Quality Assuran programs at decommissioning plants, which are documented in Sections 4.1 and 4.2, respectively. Fusther, the working group reviewed the design considerations and syster water-basin type independent spent fuel storage installations (ISFSis) [Ref. 9), as docume in Section 4.3. . Since a zirecnium fire is considered as an event beyond the design bases nuclear power plan, the working group provided information in Section 4.4 that includes e that are within the design bases of nuclear power plants.

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2.0 Deterministic Evaluation of Spent Fuel Heatup The most severe accident postulated for SFPs is the complete draining of water from the pool. If sufficient heat exists in the spent fuel, the clad will heat up, swell, and burst. The breach in the clad could result in the release of radioactive gases present in the gap twt;;;ca the fuel and the clad, called "a gap release." If the fue t gjo heat up, the temperature of the zirconium clad will reach the point of rapid oxida ~. fThis reaction of zirconium and air is exothermic. The energy released from the reaction combined with the decay energy can cause the reaction to become self-sustaining and lead to the ignition of the zirconium, or a " zirconium fire." The increase in heat from the oxidation reaction could also raise the temperature in adjacent fuel assemblies and cause the propagation of the oxidation reaction. The fuel temperatures reached after the onset of rapid oxidation are sufficient to release fission products trapped in the fuel. At lower tempera , rfy.t: lad rupture and gap release may occur.

At decommissioning plants, fire protection equipment may be the only water source available to provide a large flow rate for the mitigation of a zirconium fire. Therefore, the working group reviewed the phenomena of a zirconium fire in Section 2.1 from a fire protection perspective. In Section 2.2, the working group reviewed the characterization of a spent fuel heatup from a fuel j analysis perspective. Existing studies regarding zirconium oxidation are included with respect to i the applicabi!4ty to spent fuel pools at mrtussioning plants. Calculations performed by the working group augment the insights tror rih Safety issue (GSI) 82, "Beyond Design Basis  ;

Accidents in Spent Fuel Pools," studies to provide an updated, deterministic estimate of generic shutdown decay time,s. Section 2.3 addresses the potential for criticality fcr a zirconium fire event.

2.1 Firo Protection Evaluation of the Phenomena of a Zirconium Fire Fire protection requirements for permanently shutdown plants (those covered under 10 CFR 50.82), are specified in 10 CFR 50.48(f), which requires that the licensee maintain a fire protection program to address the potential for a fire-induced release of radioactive materials.

Draft guidance on the fire protection program for permanently shutdown plants is provided by l d

Draft Regulatory Guide 1069, " Fire Protection Program for Nuclear Power Plants During i Decommissioning and Permanent Shutdown." The draft guide contains specific guidance on the l level of fire protection to be provided for structures, systems and components (SSCs) that are j necessary to provide protection of the spent fuel.  ;

2.1.1 Comparison of Reported Zirconium MetalIgnition Temperatures I i

For loss of coolant events that are not the result of an e::temal fire (e.g., seismic or human )

error), the working group conducted preliminary research on the combustion characteristics of zirconium and zirconium alloys. The working group reviewed ignition temperatures for zirconium and similar metals reported and found that the ignition temperature of combustible metals that undergo heterogeneous combustion (such as zirconium) is highly dependent upon fuel i geometry (e.g., particle size). The ignition temperature is defined as the surface temperature j just prior to rapid temperature runaway and is brought about by an exothermic oxidation reaction l between the solid metal and its gaseous environment. Throughout this section, the point at I which the cladding undergoes a rapid, self-sustaining exidation in air, is referred to as the l

7 i

r i i ld not be igneon temperature. For example, Mellor [Ref. 3] reports that bulk z rcon um cou Ignited at temperatures below 1300 *C. The Nabonal Fire Protection Ahnen (NFPA)

Protechon Handbook [Ref. 4) reports an ignition temperature of 1400 *C in oxygen, and Co

[Ref. 5] reports that ignation tests with single 8-mm long sechons of unirradiated does not ignite at temperatures below 1600 *C. The oxidation temperature of zirconium L3pfs.1 and 2)is 850 *C to g50 *C. This reported by Sandia Nabonal Laborate tition temperatures for zirconium for all other oxidabon temperature range is lower IDr combustion sources that the working group reviewed. The working group believes that the oxidsbon temperature reported by SNL was the onset of oxidation, that would lead to ignit but not the temperature at which rapid, runaway oxidation or ignition occurs as reported by t referenced combuston manuals.

Lto occur in air exists. However, the existing The potenbal for a self-igniting bulk z' ' ' 'u r: onium cladding is considered probable when combustion literature indicates that '

some combinaten of the following conditions are available: (1) an oxide-free surfaca exposed a high oxygen conceGetion in a high pressure environment, (2) the presence of fine zirconium, and (3) fuel assembly loading configuration and decay heat load. Some of the studies that were cited in this report were conducted in pure oxygen environments and prod high ignition temperatures.

2.1.2 Mechanisms, Causes, and ChaNiibcIof a Zirconium Fire incidents involving zirconium fires and explosions [Ref, 7] have frequently involved material relatively high surface-to-volume ratios such as powders, machine tumings, grinding residu and sponges. The particio size is typically below 10gm; however, the exact relationship on particle size, size distribution, temperature, pressure, humidity, and surface characte well defined. Furthermore, the sensitivity of ignition temperature to particle size, rnass, or additior.s of impurities to zirconium, such as lead or manganese, increases its likelihood for auto-ignition.

Markstein [Ref. 8] reported that ignition of metals is always preceded by slow oxidation, w occurs on the metal surface or on and within a protective oxide layer. Once rapid oxidation begins, the self-sustaining reaction proceeds quickly. The ignition properties of metals h been linked to their low-temperature oxidation properties. However, ignition temperatures are not well defined for those metals that form a protective oxide layer, such as zirconium.

After ignition, the combustion process may continue to take place on the metal surface or and within a usually molten oxide layer covering the metal. Altematively, the reaction may occur in the surrounding vapor, which is characterized by a high buming rate and the presence luminous reaction zone that extends some distanco from the metal surface. Surface bumin has been observed if the oxide is more volatile than the metal. Vapor-phase buming occurs only if the metalis more volatile than the oxide, but may be suppressed by the form protective oxide layer. The heat of formation of zirconium-oxide reported by Sandia Kcal/ mole is in agreement with the value of 1097 kJ/mol reported by Glassman [Ref. 9]. T adiabatic combustion temperature reported by Glassman for stochiometric combustion of zirconium-oxide is 4300 K at 1 atmosphere (atm).

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1 Markstein's research on the combustion of metals shows that metal ignmon is strongly I dependent on total gas pressure as well as oxygen partial pressure. Above enhcal oxygen I concentrabons, this research revealed that the manner in which zirconium wves bumed, while j suspended vertically, depended largely on the wire diameter. In a related experiment, Markstein recorded zrconium rod ignition temperatures that were measurad in oxygen at vanous pressures and found that exposure of surface to oxygen, combined with the details of the rod's rupture characteristics pla ant role forignition of zirconium.

4 For zirconium metals, the propagation rate of the combustion zone is largely dependent upon oxygen percentage at 1 atmosphere. In a pure oxygen environment at 0.1 atmosphere, the propagation rate was 1 cm/sec; at 1.0 atmosphere (such as in a SFP building) the propagation rate increased to 3 cm/sec. In addition, the rate of buming on the surface of molten-or.ide metal mixture is limited by the diffusion of a DL 9h the surrounding atmosphere.

The zirconium cladding is continuously submerged under water and exposed to cyclical heating and cooling and radiation exposure before it is removed from the reactor vessel and stored in the SFP. The relationship between these varying exposure conditions and the combustion properties of zirconium, have not been addressed in any of the literature that could be located by the working group at this time. '

The NFPA handbook [Ref.14) reportMN zirconium bums with very little flame and

' generates little smoke but attains high temperatures. Eyewitness accounts of actual fire experience with N-Reactor irradiated fuel cladding hulls at the DOE Hanford Reservation from 1967 through 1968 reported that buming cladding had the appearance of buming charcoal, with the cladding surface glowing red with an occasional small flame. Zirconium dusts, fines, or powders are pyrophoric and bum rapidiy with an intense white glow [Ref.15]. j l

At the request of industry representatives, on Thursday, May 13,1999, the working group ]

contacted two employees of Teledyne Wah Chang, a zirconium manufacturer, by telephone to gain more information on zirconium fires. As a result of this call, some references were sent to {'

the working group that discussed zirconium cladding. The results of both the call and the references enabled the working group to determine that the information presented by Teledyne l Wah Chang was consistent with the working group findings previously mentioned in this report.

2.1.3 Mitigating Actions At decommissioning plants, fire protection equipment may be the only water source available to provide a large flow rate for the mitigation of a zirconium fire. The working group has reviewed available information published by SNL [Ref.1,2], Department of Energy (DOE) [Ref.11), NFPA

[Ref. 4,14], and the United Kingdom Atomic Energy Authority (UKAEA) [Ref.13]. Assuming

. that an initiating event causes the draindown of the spent fuel pool, anql following the event, the fire protection water supply is operable, SNL recommends the use of fire hoses to provide an emergency water spray to maintain fuel cooling on a temporary basis until the pool leak can be

. repaired. SNL notes that the required spray rates would be approximately 100 gallons per minute (gpm) and that the exposure to an individual located 50 feet from the pool would be approximately 200 Rem per hour. The working group does not believe the mitigative actions proposed by SNL are the optimum actions to take because of personnel exposure. DOE 1 9

[Ref 11] states that fires exposing masseve pieces of zirconium can be extinguished with wa based on limited tests conducted by industrial Risk insurers. If a large water supply is available, the working group would recommend considershon of ccareairs a portable monitor nozzle, which is a common piece of fire-fightmg equipment, that plant personnel can set up at a much greater destance from the spent fuel pool than a manual hose line, which does not require contmuous manning, thus reducing is personnel radiation exposure. A portable

< ndent on the water supply) with an effective monitor nozzle can discharge over range of more than 100 feet. Many licensed nuclear power plants have portable monitor nozzles available either on site or from the local municipal fire department.

For fires involving zirconium powder or fines, all investigators [Ref.12 and 13) suggest using Class D extinguishing agents such as sand and graphite or inert gases, such as argon. If prevMed in sufficient quantibes, delivmp tqtheAase of the fire, this method of extinguishment '

may also be effective on bulk zirconio of. M As an altemative to the use of unborated water, the working group considered the use of high expansion foam. A high expansion foam generator can also be operated with minimum manning once placed in operation. One significant advantage of high expansion foam over plain water is the low water requirement. High expansion foam has an expansion volume (foam-to-water) ratio in the range of 200:1 to 1 13Another advantage is that high expansion foam will be lost relatively slowly through the lesili 4tWsgent fuel pool. This method may completely submerge the fuel assemblies in foam, extinguishing a potential zirconium fire and limiting the potential for airbome release. Many nuclear power plants have high expansion foam equipmen available either on-site or from the local municipal fire department. The effectiveness of high expansion foam on bulk zirconium fires is not known at this time.

' 2.2 Spent Fuel Heatup Analyses Spent fuel heatup analyses model the decay power and configuration of the fuel to character

' the thermal hydraulic phenomena that will occur in the spent fuel pool and the building following a postulated loss of water accident. This subsection reviews the existing studies on spent fue heatup and zirconium oxidation, the temperature criteria used in the analyses, and how it applies to decommissioning plants. The working group also provides generic estimates involving deterministic spent fuel heatup caiculations for two criteria that have been used for j

previous plant-specific exemptions.

2.2.1 Spent Fuel Failure Criteria The working group determined that the most severe damage would be caused by rapid, runaway zirconium oxidation. This would lead to significant fission product release even after the gap activity has become insignificant. The onset of rapid oxidation may occur as low as 800 *C [Ref. 4). Runaway o'xidation can raise clad and fuel temperatures to approximately 2000 *C, which corresponds to the melting temperature of zirconium. The release of fission

products trapped in the fuel can occur at fuel temperatures of approximately 1400-1500 *C.

Runaway oxidation starting in a high powered channel could also propagate through radiative and convective heat transfer to lower power assemblies because of the large heat of reaction in zirconium oxidation.

10

Several deerent fuel failure cntana have been used in NRC-sponsored spent fuel pool accident studies. Benjamin, est. al. [Ref.1] used the onset of runaway clad oxidation as the fuel fadure entonon in NUREG/CR 064g. This cntenon was enticized because clad rupture can occur at a much lower temperature causing a gap release. The consequences of gap release can be significant if the radioactive iodine has not yet decayed to insignificant amounts. SHARP (Spent Fuel Heatup Analytical Response P ) ations [Ref. 2] used the onset of clad swelling as a entenon. The onset of clad a to gap release occurs at approximately 565 *C, which corresponds to the temperature for 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> creep rupture [Ref. 3]. A cladding temperature of 570 *C is used as a thermal limit under accident conditions for licensing of spent fuel dry storage casks.

The next temperature threshold that may be of concem in spent fuel pool accidents is the moltmg temperature of the aluminumpp i constituent of some types of the spent fuel storage rack's boron poison plates. AWa rr alts at approximately 640 *C. No evidence was found that boron carbide will dissolve in the aluminum forming an eutectic mixture at a temperature below the melting point of aluminum, but an exhaustive literature search was not performed. Eutectic formation is important for in-vessel severe accidents. If it is possible for the molten material to leak from its stainless case, melting and relocation of the aluminum in the boron carbede-aluminum composite may cause flow blockages that increase hydraulic resistance. No realistic evaluation ofpy ' relocation of aluminum or aluminum / boron carbide eutectic has been performed.LT enltial for criticality is discussed in Section 2.3 of this study.

Another concem is the structural integrity of the fuel racks at high temperatures. Steel and zirconium form an eutectic mixture at approximately g35 *C. Several eutectic mixtures known from severe accident research [Ref. 5) may be important in SFP accidents. Steel and boron cartade form a eutectic mixture at approximately 1150 *C. Also, the steel racks may not be able to maintain structural integrity because of the sustained loads at high temperature. Loss of rack integrity may affect the propagation of a zirconium fire.

If the gap radioactive inventory is significant, then the cladding temperature must be kept below 565 *C.' If the consequences of aluminum / boron carbide relocation and possible criticality are  ;

acceptable, then 800 *C may be a reasonable, deterministic acceptance temperature if I uncertainties are less than the margin to 800 *C, and the effects of the higher temperatures on the material are modeled. Otherwise, the temperature must be lower than the aluminum or eutectic melting point.

l 2.2.2 Evaluation of Existing Spent Fuel Heatup Analyses In the 1980's, severe accidents in operating reactor spent fuel pools were evaluated to assess i the significance of the results of some laboratory studies on the possibility of self-sustaining l

, zirconium oxidation and fire propagation between assemblies in an air-cooled environment, and also the increase in the use of high density spent fuel storage racks. This issue was identified as GSI 82. SNL and Brookhaven National Laboratory (BNL) used the SFUEL and SFUEL1W computer codes to calculate spent fuel heatup. While decommissioning plants were not addressed in the study, many of the insights gained from these studies are applicable to decommissioning plants.

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Y More recently, BNL c;i-;+1 a new computer code, SHARP, that was intended to provide a sunpillied analysis method to model plantWk spent fuel configurabons for spent fuel heatup calad=hans at decommissioning plants. Some of this work was built on the assumption used by SNL and BNL studies in support of GSI 82.

2.2.2.1 SFUEL Senes Based Analyse') RAFT Extensive work on the phenomena of zirconium oxidation in air for a spent fuel pool l j

configuratson was performed by SNL and BNL in support of GSI 82. SNL investigated the heatup of spent fuel, the potential for self-sustaining zirconium oxidation, and the propagati6n to{

agacent assemblies [Ref.1,6]. SNL used SFUEL and SFUEL1W computer codes to analyze the thermalhydraulic phenomena, assuming complete drainage of the SFP water. In L studies on the phenomenology of zirconium-NUREG/CR4982 [Ref. 4], BNL exter dp Ah air oxidation and its propagabon in sp ah . Ess emblies. The SFUEL series o codes include 1.

' all modes of heat transfer, including radiation. However, radiation heat transfer may have been underestimated due to the assunied fuel bundle arrangement, in NUREG/CR-0649 [Ref.1], SNL concluded that decay heat and configuration are important parameters. SNL found that key configuration variables are the baseplate hole size, downcomer width, and the availability of open space air.(km. They also found that building ventilation is an'important configuration variable. U Fl The draft SNL report [Ref. 6] investigated the potential for oxidation propagation to adjacent assemblies. If decay heat is sufficient to raise the clad temperature to within a few hundred degrees of oxxisbon, then the radiative heat from an adjacent assembly that did oxidize could raise its temperature to the oxidation level. The report also discusses small-scale experiments evolving clad temperatures greater than 1000 *C. SNL hypothesszed that molten zirconium material would slump or relocate towards the bottom of the racks and consequently would not be involved in the oxidation reaction. NUREG/CR-4982 did not allow oxidation to occur temperatures higher than 2100 *C to account for the zirconium melting and relocation.

Otherwise, temperatures reached as high as 3500 *C. It was felt that not cutting off the oxidation overstated the propagation of a zirconium fire because of the fourth power temperature dependence of the radiation heat flux. The SFUEL series of codes did not model melting and relocation of materials.

In NUREG/CR-4982, BNL reviewed the SFUEL code and compared it to the SNL small-scale >

experiments and concluded that SFUEL was a valuable tool for assessing the likelihood of sel sustaining clad oxidation for a variety of spent fuel configurations in a drained pool. SNL reported the following critical decay times in NUREGICR-0649 based on having no runaway oxidation.

700. days PWR,6 kW/MTU decay power per assembly, high density rack,10.25" pitch, 5" '

orifice,1 inch from storage wall 280 days PWR, same as above but for i foot from storage wall i

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180 days BWR,14 kWMTU decay power per assembly, cylindrical baskets, 8.5" pitch,1.5" onfice unknown BWR, high density rack, SFUEL1W code was limited to computation of BWR low density racks.

High density racks with a 5-inch orifMAbdat representative of current storage practices.

A cnhcal decay time for high density BWR racks was not provided due to code limitations. Low density and cylindrical storage rack configurations are no longer representative of spent fuel

' storage.' All currently operating and recently shutdown plants have some high density racks in their spent fuel pools. For an assembly in a high density PWR rack, with a 5-inch onfice, a decay power below 6 kWMTU did not result in zirconium oxidation. All of these estimates were based on perfect ventilation (i.e., unli , -temperature air) and bumup rates of 33 GWDMTU. Currently, some. ed to bum up to 62 GWDMTU and some BWRs up to 60 GWDMTU. For fuel bumup of 60 GWDMTU, the working group estimates the decay time for a bundle to reach 6 kWMTU will increase from 2 years to approximately 3 years.

Therefore, the working group expects the difference between critical decay times for PWRs and BWRs to decrease and that the decay time would now be longer than the SNL estimate for high density PWR racks. The SNL calculations also do not appear to have included grid spacer loss coefficients which can have a significa ff ice the resistance of the grid spacers is greater than the resistance of a 5 inch orifice ~ s rio mixing between the rising air leaving the fuel racks and the relatively cooler air moving down into the pool. Including the grid spacer I resistance and accounting for mixing will result in the critical decay power to be less than 6 kWMTU. The SNL calculations may have understated the effective radiation heat transfer f

heat sink due to the assumed fuel geometry in the calculations. A more realistic fuel configuration pattem in the SFP would give a better estimate of the radiation heat sink and raise the critical decay power needed for significant oxidation.

While the studies in support of GSI 82 provide useful insights to air-cooled spent fuel assemblies, it is the opinion of the working group that they do not provide an adequate basis for exemptions. The studies were not meant to establish exemption criteria and lack sufficient information for all the parameters that could affect the decay time. Additionally, the reports are based on bumup values at that time. Since bumup values have increased, the results may not be directly applicable to today's spent fuel.

The preliminary results and the characteristics described in the studies may assist in assessing issues for decommissioning plants. However, the calculated decay time values do not represent current plant operational and storage practices.

2.2.2.2 SHARP Based Analyses in NUREG/CR-6451 [Ref. 7], BNL investigated spent fuel heatup that could lead to a zirconium 1 fire at permanently shutdown plants. BNL developed a new computer code, SHARP, to I calculate critical decay times to preclude zirconium oxidation for spent fuel. The code was intended to study thermal hydraulic characteristics and to calculate spent fuel heatup to temperatures up to approximately 600 *C, SHARP is limited to low temperatures since it lacks models for radiation heat transfer, zirconium oxidation, and materials melting and relocating.

13 i

SHARP also lacks modelin0 f or grid spacer losses and neglects mixing between the risin air and the falimg cooler air in the SFP. BNL reported the following generic critical deca using the SHARP code.

17 invinhe PWR, high density rack, 60 GWDMTU bumup; 10.4" pitch; 5" orifice months BWR, high if r LM GWD/MTU bumup; 6.25" pitch; 4" orifice

-7 The decay times are based on a maximum cladding temperature of 565 *C. The parame listed with the entical decay times are generally representative of operating practices. Cu fuel bumups, however, have increased to values higher than those used by BNL.and pe ventilabon was assumed, which could lead to an underestimation of the critical decay tim ked, validated or venfied by the staff. The The SHARP code was not significantQy. La entical decay times are much shortertiMEINc!Ie calculated in NUREG/CR-0649 and NUREG/CR-4982, particularly when the lower cladding temperature used for fuel failure higher decay heats used in the earlier analyses are taken into account. This appear driven in part by the fact that the decay heat at a given bumup in the SHARP calculatio significantly lower than what is used in the SFUEL calculat decay times.. The staff has determinq ttaejode will be used as a scoping tool by the staff.

It is not adequate for use as a technkaf Ibf licensees without further code modifications and verification. NUREG/CR-6541 was intended to be an assessment to steer rulemaking activities. The report was neither intended nor was it structured to provide a basis for exempbons for decommissioning plants. The working gmup does not rely on this study heatup analyses information due to the code that the decay time conclusions were base 2.2.3 . Heatup Calculation Uncertainties and Sensitivities The phenomenology needed to model spent fuel heatup is dependent on the chosen temperature success criteria and the assumed accident scenario. Many assumption modeling deficiencies exist in the current calculations. The working group reviewed t to assess the impact of the modeling assumptions. Some of these uncertainties for the series codes are discussed further in NUREG/CR-4982. For cases of flow mixing, decay

' bundle flow resistance, and other severe accident phenomena, additionalinformation is' provided below.

Calculations performed to date assume that the building, fuel, and rack geometry remain This would not be a valid assumption if a seismic event or a cask drop damaged some of t l fuel racks or the building. Rack integrity may not be a good assumption after the onset '

significant zirconium oxidation due to fuel failure criteria issues discussed in Sect building may also be hot enough to ignite other materials. Assuming that the racks r is the most optimistic assumption that can be made about the rack geometry. Any damage the racks will reduce the ability to coof the fuel.

All three SFUEL, SFUEL1W, and SHARP codes assumed perfect ventilation or chimney conditions,' that is an unlimited amount of fresh, ambient-temperature air is available. This i assumption would be valid if the building failed early in the event or if large portions and ceilings were open. If the building does not fail, the spent fuel building ventilation flo 14 l

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i would dictate the air flow available. Mixing between the nomg hot air and the descending cooler air was not included in the codes.

The spent fuel building ventilation flow rate is important in determining the overall budding {

energy balance. Air flow through the buildmg is an important heat removal mechanism. Most of the air would recirculate in the buildin Lat drawn under the racks would be higher than ambient temperature and, therefore, I "Ni mval would occur. Airflow also provides a ,

source of oxygen for zirconium oxidation. Ser:.nivity studies have shown that heatup rates increase with decreasing ventilation flow, but that very low ventilaten rates limit the rate of

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' oxidaten Other oxidabon reachons (fires) that occur in the building will also deplete available oxygen in the buddmg. GSI 82 studies concluded that perfect ventilabon was worse than no ventilabon because the oxidation reaction became oxygen starved with no ventdabon intermedete ventdabon rate results a 4 pecu mented and could be worse than the perfect venblabon case. > M~

A key fuel heat removal mechanism is buoyancy-driven natural circulation. The calculated air flow and peak temperatures are very sensitive to flow resistances in the storage racks, fuel bundles and the downcomer. The downcomer flow resistance is determined by the spacing between the fuel racks and the wall of the SFP. The storage rack resistance is determined by the onfice r,ize at the bottom entran f ndle. Smaller inlet orifices have higher flow

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resistance. As shown by SFUEL a culations, changes in the rack-wall spacing J and th onfice size over the range of designs can shift critical decay times by more than a year.

TM fuel bundle flow resistance is determined by the rod spacing, the grid spacers, intermediate flow mixers and the upper and lower tie plates. SFUEL and SHARP calculations apparently have neglected the losses from the grid spacers, intermediate flow mixers and the tie plates.

These flow resistances will be higher than those from the rack inlet orifice. Therefore, inclusion of the addebonal flow resistance may extend the entical decay time by a year or more.

NUREG/CR4982 concluded that the largest source of uncertainty was due to the natural circulation flow rates.

The extent of flow mixing will determine the air temperatures at the downcomer and bundle inlet.

- The inlet air temperature and mass flow rates are important in determining the peak cladding temperature. The SFUEL and SHARP calculations assume a well-mixed building air space.

The downcomer inlet temperature is set equal to the building temperature. This assumption

- neglects the mixing that occurs between the hot air rising from the bundles and the cooler air descending down the spent fuel pool wall. Computational fluid dynamics calculations performed by the NRC's Office of Research (RES) using the FLUENT code indicate that perfect mixing is not a good assumption. The mixing that occurs between the cool air flowing down into the pool and the hot air flowing up out of the fuel bundles can significantly increase peak cladding temperatures. Even using different turbulent mixing models can affect the peak temperatures by approximately 100 *C. The calculations indicate that fully 3-dimensional calculations may be

'needed to accurately predict the mixing because unrealistic flow topologies in 2-dimensional

- approximations may overstate the mixing.

' Radiation heat transfer is important in zirconium oxidation calculations. Radiation heat transfer can affect both the onset of a zirconium fire and the propagation of a fire. Both the SFP loading pattem and the geometry of the. fuel racks can strongly affect the radiation heat transfer between

. adjacent bundles. Sirnple gray body calculations show that at clad temperatures of 800 *C, a 15

temperature difference of 100 *C between adjacent bundles would cause the radiabon heat flux to exceed the cribcal decay power of 8 kW/MTU. Therefore, the temperature ddrerence that could be mantained by adjacent bundles is highly constrained by the low decay heat levels.

SFUEL calculabons performed by SNL and BNL included radiabon heat transfer, but the radsabon heat transfer was under predicted smce the spent fuel placement is two-dimensional and the hottest elements are in the ~ f te, pool with cooler elements placed progressively toward the poolwalls. Heat transfer flotter and cooler assemblies has the potential to be significantly higher if the fuel bundles were intermixed in a realistic loading pa#em.

At temperatures below 800 *C, the heat source is dominated by the decay heat. SNL and BNL found that, for high density PWR racks, that 6 kW/MTU was the critical decay heat level for a zirconsum fire to occur in configurabons resembling current fuel storage practices At the fuel bumups used in the calculabons, this c acpy heat level was reached after two years.

However, reliable decay heat calcu het available to the staff. As a result, decay heat calculations from NUREG/CR-5625 will be used as the basis of a regulatory guide for calculating fuel assembly decay heat inputs for dry cask storage analyses. These decay heat calculations are conses, tent with the decay heat used in SFUEL calculations. Extrapolation of the decay heat calculations from NUREG/CR-5625 [Ref. 8] to current bumups indicate that approximately 3 years will be needed to reach a decay heat of 6 kW/MTU.

Several licensees have proposed usihMknh Technical Position ASB 9-2 decay heat model for use in SFP heatup calculations. Using ASB 9-2 decay heat with a "k factor" of 0.1 produces non-conservative decay heat values in the range of 1 to 4 years after shutdown.

ASB 9-2 explicitly states that it is good for time less than 10,000,000 seconds (115 days). The basis of ASB 9-2 is the 1973 ANS draft decay heat standard. The standard gives k factors to use that are beyond 10,000,000 seconds. The working group has found that a k factor of 0.2 will produce conservative decay heat values compared to ORIGEN calculations for the range of 1 to 4 years after shutdown At temperatures below the onset of self-sustaining oxidation, the heat source is dominated by the decay heat of the fuel. When zirconium reaches temperatures where air oxidation is significant, the heat source is dominated by oxidation. The energy of the reaction is 262 kcal per mole of zirconium. In air, the oxidation rate and the energy of the reaction is higher than zirconium-steam oxidation. Much less data exists for zirconium-air oxidation than for zirconium-steam oxidation. A large amount of data exists for zirconium-steam oxidation because of the large amount of research performed under the emergency core cooling system (ECCS) research program {Ref. 9]. If all of the zirconium in a full 17x17 PWR fuel bundle fully oxidizes in air over the period of an hour, the average power from the oxidation is 0.3 MW. The critical decay heat as determined with SFUEL is approximately 2.7 kW for the bundle. The oxidation power source would amount to approximately 60 MW if the whole core was buming. A

- 20,000 cubic feet per minute (CFM) air flow rate is needed to support that reaction rate based on 100-percent oxygen utilization. The SFUEL oxidation rate was modeled using several parabolic rate equations based on available data. SFUEL had limited verification against SNL experiments that studied the potential of zirconium fire propagation. BNL found that although they could not find a basis for rejecting the oxidation rate model used in the SFUEL, uncertainties in oxidation of zirconium in air could change the critical decay heat by up to 25-percent. It was found that the onset of runaway zirconium oxidation could occur at temperatures as low as 800 *C. Different alloys of zirconium had oxidation rates that vary by as 16

4 much as a factor of four. Apparently, it was found that oxidation in air was worse than oxidation in pure oxygen. This suggests that the nitrogen concentrabon can have a significant impact on the oxidation rate. Since the relative conoontrabon of oxygen and nitrogen varies as oxygen is consumed, this causes addstional uncertamty in the oxidabon rate. The oxidabon was cut off at 2100 *C in the BNL calculations in support of GSI 82. This was done to simulate zerconium clad relocabon when the molting point of reached if the oxidabon was not cut off, temperatures could be as high as 3 determined that the propagation to adjacent bundles was over predicted if no cutoff temperature is used due to the fourth power dependence of temperature on the radiabon heat fluxes.

The combustion literature discussed in Sechon 2.1 also shows that there is a large range in the temperature for zirconium ignition in air. Evidence cited from the literature states that bulk zirconium can not ignite at temperatu zip t.it an 1300-1600 *C. It is known from the extensive ECCS and severe acodent d Tp rograms that zirconium-steam runaway oxidabon occurs at temperatures below 1300 *C. Since oxidation in air occurs more rapidly than oxidation in steam, temperatures in this range are not credible for the onset of runaway oxidation in air. Correlations listed in Reference 10 give ignition temperatures for small zirconium samples in the range of runaway oxidation computed by the SFUEL series codes when the geometry factors calculated from zirconium cladding are input into the correlations.

Reference 11 appears to be the only f ton / hat is applicable to zirconium oxidation in j sustained heating of fuel rods. In the. ;d test, sections of zirconium tubing were oxidized e

at temperatures of 700 *C,800 *C and 900 *C for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The average oxidation rate tripled for i each 100 *C increase in temperature. This is consistent with the change in oxidation rates predicted by the parabolic rate equations examined in NUREG/CR-4982. The zirconium combustion literature reviewed for ignition temperature did not discount or provide attemate oxidation rates that should be used in the SFUEL calculations.

As documented eariier, current operating practices bum fuel to higher levels than used in the evaluations. The BNL and SNL studies in support of GSI 82 represented operating practices of the 1980's with bumup levels around 33 GWD/MTU. In NUREG/CR-6451, BNL used bumup levels of 40 and 60 GWD/MTU for BWRs and PWRs, respectively. While these values are closer to current operating practices, they stiil underestimate bumup values. The decay heat used in the SHARP analyses is significantly lower than that used in the SFUEL analyses. Given that bumup is an important parameter for determining the critical decay time, this is a significant i change. The increase in bumup level will increase the critical decay time needed to ensure that  !

air cooling is sufficient to maintain the zirconium cladding below the oxidation temperature.

The BNL and SNL studies in support of GSI 82 represented storage practices of the 1980's when plants were starting to convert to high density storage racks. The studies did not address high density BWR racks, and the high density PWR racks in the reports were not as dense as the designs used by many plants today. The higher density racking currently used will decrease the air flow available for heat removal. Therefore, lower decay heat values are needed to ensure that air cooling is sufficient to maintain the zirconium clad below the oxidation temperature.

s 17 I

S

2.2.4 Critical Decay Tunes to Reach Sufficient Air Cooling Cribcal decay time is defined as the length of time after shutdown when the most recently discharged fuel temperature will not exceed the chosen fuel failure entena when cooled by air only. Earbor calculabons using the SFUEL have determined a enbcal specsfic decay heat of 6 kWMTU is needed for the onset of 1Wponium oxidation. This 6 kWMTU estimate calculated using SFUEL in a high Ipo rack configuration is reasonable and is based on the best calculations to date. However, this estimate is based on perfect ventilation conditions in the building.

For high bumup PWR and BWR fuel, the working group estimates that it will take approximately 3 years to reach the critical decay heat level cited in NUREG/CR-4982. Better modeling of flow mixing and accounbng for the grid tia alete flow resistance could reduce the critical decay power level and increase the ' 5is} time beyond 3 years, but this may be counterbalanced by increased radiation heat transfer from realistic fuel bundle loading. Other assumptions such as imperfect ventilation could extend the critical decay time for the onset of a

. zirconium fire by 1 to 2 years. Accounting for imperfect ventilation, the working group estimates that genoncally it will take approximately 3 to 5 years to reach the entical decay time for current plant conditions. A plant-specific calculation using actual plant decay heat and spent fuel configuration should yield shorter timt 3.RA FT Previous EP exemptions using this analysis method for critical decay times were based on plant-spinXic calculations. .The working group's generic results do not affect the previous plant-specific exemption determinations.

2.2.5 Estimated Heatup Time of Uncovered Spent Fuel

- The staff recognizes that the decay time necessary to ensure that air cooling was adequate to remain below the temperature of self-sustaining zirconium oxidation is a conservative criteria for the reduction in emergency preparedness. Using the fact that the decay heat of the fuelis reducing with time, credit may be given, if quantified, for the increasing length of time for the accident to progress after all water is lost from the SFP. The working group sought to quantify the decay time since final shutdown such that the heatup time of the fuel after uncovery was adequate for effective protective measures without preplanning. The fuel heatup time was used in two previous plant-specific exemptions.

While the staff has not determined that this criterion is appropriate for generic use, it does provide a measure of time that may be bounding for all heatup scenarios, including scenarios such as having a flat material fall on top of an assembly. The heatup time calculated by this method assists the staff in determining the amount of time available for onsite mitigative actions and, if necessary, offsite protective measures. The staff would need to work with the Federal Emergency Management Agency to evaluate whether a generic time for all plants (or a group of

. plants) or if a site-specific time should be determined for each plant.

The heatup time of the fuel depends on the amount of decay heat in the fuel and the amount of heat removal available for the fuel. The amount of decay heat is highly dependent on the bumup. The amount of heat removal is dependent on several variables that may be difficult to confirm on a plant-specific basis. As such, the working group assumed no heat removal was 18

9 available. The worlang group calculated the time required to heat a single fuel rod from 30 *C to 900 *C without considering any heat losses at various times following permanent shutdown.

This is very conservative because some cooling, through conduchon, convechon or radiabon, would occur and increase the time to heatup. Figures 2.2-1 and 2.2-2 show the increase in fuel hoatup time that is required to heat a fuel pin adiababcally from 30 *C to 900 *C as the decay time following final shutdown increasg of the conservative nature of the calculation, more than the calculated time would mer if the scenario were to occur. Additionally, a i more-realistic or plant-specific calculation should yield a shorter decay time. I This type of calculation was used to support two previous plant-specific emergency  !

preparedness exemptions. A 10-hour and a 14-hour period were determined to be sufficient time after the fuel is uncovered such that mitigative actions and offsite protechve measures I could be taken without proplanning. y  !!. exemptions were based on plant-specific conditions, and therefore the working.gri De wric results do not affect these previous J exemption determinations.

For the calculatior.s, the working group used a decay heat per assembly and divided it equally among the pins. It assumed a 9X9 assembly for the PWRs and a 17x17 assembly for the {

BWRa.. The design values used for the calculation are in Appendix 3. Decay heats were

]

computed using an extrapolation of tK cay r tables in NUREG/CR-5625 [Ref. 8]. The decay heat in NUREG/CR-5625 is b EN calculations. The tables in NUREG/CR-5625 for the decay heat extends to bumups of 50 GWDMTU for PWRs, and 45 GWDMTU for BWRs. The working group recognizes that the decay heat is only valid for values up to the maximum values in the tables, but the functional dependence of the decay power with respect to bumup for values in the table indicate that extrapolation may provide a reasonable estimate of the decay heat for bumup values beyond the limits of the tables. The BWR decay heat was calculated using a specrfic power of 26.178 MWMTU. The PWR decay heat was calculated using a specific power of 37.482 MWMTU. Both the PWR and BWR decay heats were calculated for a bumup of 60 GWDMTU and include an uncertainty factor of 6 percent.

k l

19 l m

PWR Adiabatic Heatup 20

~

18

$16

.E -

/

v14 E12 ~

10 '

$8

~

$6 ~

4  :  :  :  :  :

200 400 600 800 10001200140016001800 Shutdown Time (days)

Figure 2.2-1 Adiabatic Heatup Time vs. Shutdown Time for a PWR BWR Adiabatic Heatup 22 S

b 16 54 1 h 12 ,

910

/

[

N8 6 / t 4

f l

200 400 600 800 1000 1200 1400 16001800 j Shutdown Time (days) i l

)

Figure 2.2-2 Adiabatic Heatup Time vs. Shutdown Time for a BWR 20

2.3 ~ Potential for Criticality 2.3.1 Evaluation of the Potential for Criticality There are several scenarios in which the potential for criticality from spent fuel pool accedents exists:

DRAFT

1. A compression of the stored assemblies could result in a more optimum geometry (closer spacing) and thus creating the potential for criticality. This could situation could exist for BWR pools which contain no soluble boron. For PWR pools with soluble boron, this would not be a concem because of the additional reactivity holddown of the boron.

However, for this scenario to occur, it must be assumed that the storage racks are somehow removed from the p ag.jppliow/or the closer spacing of the fuel assemblies.

The working group does not c aMidl&Fiisito be feasible.

2. If the stored assemblies are separated by neutron absorber plates (e.g., Boral or Boraflex), loss of these plates could result in a potential for enticality for BWR pools. For PWR pools, the soluble boron would be sufficient to maintain suberiticality. The absorber plates are generally enclosed by a stainless steel cover plate and the tolerances within the cover plate oul and to prevent any appreciable fragmentation and movement of the enclosed " r baterial. The totalloss of the cover plate is not considered to be feasible.

However, Borafiax has been found to degrade in spent fuel pools due to gamma radiation and exposure to the wet pool environment. Because of this, the NRC issued Generic Letter 96-04 to all holders of operating licenses, conceming Boraflex degradation in spent fuel storage racks. Each addressee that uses Boraflex was >

requested to assess the capability of the Boraflex to maintain a 5% suberiticality margin and to submit to the NRC proposed actions to monitor or confirm that this 5% margin can be maintained for the lifetime of the storage racks. Many licensees have subsequently replaced the Boraflex racks in their pools or have reanalyzed the criticality aspects of their pools assuming no reactivity credit for Boraflex. Licensees that rely on Boraflex for reactivity holddown will continue to monitor for poss51e Boraflex degradation.

3. If sufficient fuel rod cladding damage occurred such that many of the fuel pellets became loose in the pool, the pellets might be rearranged into a configuration that could lead to a potential for criticality. If the storage rack remained intact, that would tend to keep the pellets in separate regions and there would probably be no potential for criticality.

However, if there is large-scale c: adding damage as well as damage to structural parts of the pool, this could allow a large fraction of the pellets to reassemble into a minimum leakage configuration in one region causing a potential for criticality. Since any structural damage would cause the storage rack material as well as any absorber material to be admixed with the fuel pellets, the working group does not consider the assembly of a critical mass of fuel pellets alone to be a credible event.

21

2.3.2 Evaluabon of the Potential for Criticality From Personnel Actions in Response to an Accident Without modershon, fuel at current eniid nnent limits (no greater than 5 wt% U-235) cannot achieve enhcality no matter what the configuration. If it is assumed that the pool water is lost, a reflooding of the storage racks with uggJgier or fire-fshting foam may occur due to personnel actions. However, both %- lMR storage racks are designed such that they remain subcriticalif moderated by unborated water in the normal configuration. The phenomenon of a peak in reactivity due to low-density (optimum) moderation (fire-fighting foam) is not of concem in spent fuel pools since the presence of relatively weak absorber material such as stainless steel plates or angle brackets is sufficent to preclude neutronic coupling between assemblies. Therefore, personnel actions involving refilling a drained spent fuel pool would not create the potential for a eM" "iA. FT 2.3.3 Potential for Criticality Summary Most scenarios that could result in inadvertent criticality of stored spent fuel assemblies are not considered to be credible for the reasons given above. The only feasible scenario appears to be due to long-term Boraflex degradation. However, per NRC Generic Letter 96-04, Boraflex monitoring will continue for decommipcajrp alaats which use Boraflex in their storage racks, thereby avoiding any potential for criticelifyn - l

' It should be mentioned that most of the fuel stored in spent fuel storage racks has been removed from reactor cores because it has lost sufficient reactivity to be useful in producing a self-sustaining chain reaction and in producing power. In addition, the reactivity of stored assemblies has continuously decreased even further during long-term storage due primarily to Pu-241 decay and Am-241 growth. Therefore, it would appear difficult to achieve criticality with stored spent fuel assemblies.

22

3.0 Risk Assessment of Spent Fuel Pool Risk at Decommissioning Plants in its p-A+r' risksassessment (PRA) Policy Statement, the Commissen stated that PRAs and associated analyses should be used in regulatory matters, where practical within the  :

bounds of the state-of-the-art, to reduce unnecessary conservatism assocated with current regulatory requirements and staff k-informed regulaten the staff uses insights denved from PRAs in combination lity' Wna tic system and engineering analyses to focl) licensee and regulatory attenbon on issues commensurate with their importance to safety.

Therefore, the level of risk assocated with the spent fuel pools is one factor considered in deceion making regardmg exemption requests for emergency preparedness, insurance, and i safeguards requirements at decommissioning plants. Risk insights are used to complement the staff's tradsbonal deterministic decision malang process, and are not used in place of deterministic evalustens or requiremi$A FT The working group's preliminary deterministic evaluations (see Section 2) indicate that zirconium cladding fires could not be ruled out for spent fuel that had been transferred from reactors up to five years previously, based on a simplified, conservative analysis. To assess the risk in the time from shutdown to one year, the working group performed a simplified preliminary PRA, which modeled intemal and extemal initiating events to assess the potential risk associated with spent fuel pools at decommissioning ; er plants. The preliminary results suggest that i there may be non-negligible risk assa spent fuel pools at decommissioning plants during the period analyzed in the risk assessment (one month after removal of last fuel from the i

reactor to one year after removal of the last fuel.)  !

The working group modeled a decommissioning plant's spent fuel pool cooling system based on the sled-mounted systems that are used at many current decommissioning plants. Information about existing decommissioning plants was gathered by decommissioning project managers and during recent visits to four sites covering all four major nuclear steam supply system vendors (General Electric, Westinghouse, Babcock & Wilcox, and Combustion Engineering). In addition, the working group used several previously published deterministic and probabilistic l evaluations of potential risks at both operating and decommissioning plants to assist in this 1 study.

1 PRAs are systematic methods of examining a complicated system (e.g., a nuclear power plant)

M identify strengths and weaknesses in the design or estimate the level of risk associated with operation of the plant or system. The traditional event tree, fault tree method' was used when developing the risk assessment. The working group chose one representative spent fuel pool l configuration (See Figure 3.0-1) for the evaluation except for seismic events, where the PWR and BWR spent fuel pool designs (i.e., the difference in location of the pools in PWRs and BWRs) were specifically considered. The working group looked at three cases as boundary conditions for its risk evaluation. Case 1 assumes that the hypothetical spent fuel pool and its  ;

support systems are available and have equipment (including instrumentation and power

' Event trees (ETs) are logic diagrams that analyze a sequence of events by modeling the outcome of success or failure of these events. Fault trees (FTs) are logic diagrams that model how a particular event can fail (e.g., a pump can fail to perform its function if it loses electric power, its water supply is cut off, thL motor breaks, or it is turned off).

23

sources) available similar to that found by the wortong group in its visits to four decom plants. In Case 1, transfer of the last fuel from the reactor to the spent fuel pool have occurred one year previously. Case 2 is the same as Case 1, except that the tran the last fuel is assumed to have occurred one month previously. This is a bounding cas it is not expected that a utility would disable all the support systems normally used to p t fuel transfer. In fact, it may not be possible spent fuel pool cooling within one mo f

i to remove the heat load in the pool one month for the sled-mounted systems current after the last transfer. However, Case 2 does provide a bounding value to help determine th impact that higher decay heat loads (and therefore less recovery time) have on the fuel uncovery. Case 2 is bounding in another way in that the frequency of fuel uncovery fo Case 2 is highly dependent on the decaying heat load, while the analysis of Case 2 assumes that the head load stays constant (i.e., heat load at one month after transfer from the reac the entire year.

DRAFT For Case 3, it is assumed that the spent fuel pool and its support systems are configured in a manner that is slightly better than the minimum allowed by current NRC regulations, and the working group believes that no prudent utility would so configure its spent fuel pool system.

However, none of the assumptions in the " minimal state" are precluded by current NRC regulations. Case 3 assumes that the last fuel from the reactor was transferred one year ago.

The working group evaluated Case 3 M wip determine if there were a need for additional L/ M P regulation in this area.

Note that the time available,for operator action for each of the three cases is based on wor l group calculations. In addition to rapid draining events, the working group considered p i

heatup after a loss of pool cooling followed by bulk pool boiling as a possible way t.: unco fue!. It takes a very long time to uncover the fuel if inventory is lost in this manner due to the large amount of water in a spent fuel pool, the large specific heat of water, and the large heat of vaporization for water. Simple calculations for a Millstone 1-sized spent fuel pool a conservative decay heat assumptions show that it will take longer than one day to heat the water to the boiling temperature and take almost five hours to boil off every foot of water at six monthe after shutdown. Therefore, it would take about five days to reach the point of fuel uncovery if the only mechanism to lose water is through heatup and boiloff. At one year after shutdown it would take longer than five days to uncover the fuel in this manner.

Appendix 1 discusses the assumptions in Cases 1,2, and 3. Table 3.1-3 summarizes the preliminary calculations of frequency of fuel uncovery for each of these cases and for all initiators analyzed. In order to perform a risk-informed evaluation of spent fuel pool accidents (beyond design bases), various references (INEL-96/0334; NUREG-1275, Vol.12; NURE 1353; Draft for CcMment NUREG " Postulated Accidents for Permanently Shutdown Reactor Draft " Risk Analysis for Spent Fuel Pool Cooling at Susquehanna Electric Power Station";

NUREG/CR-6451; NUREGICR-4982; NUREG/CR-5281; and NUREG/CR-5176) were reviewed dealing with spent fuel pools and decommissioning plants to develop a list of potential initia that could result in a loss of spent fuel pool cooling. Table 3.1-1 lists the intemal and extemal

' 24

initiatmg events8found to be potenhally important by qualitative screening processes in the above references.

• Each of the spent fuel pool accident sequences can end up in one of two ways, known as end states. The first end state is for sequences that do not result in signsficant offsste doses (i.e., the fuelin the spent fuel pool does not reture whe e runaway oxidaten of the zirconium clad will occur.) The is where the fuel pool has been dramed of water (as the result of loss of inventory and subsequent loss of cooling capability, or directly ,

from loss of spent fuel pool cooling capability) and the fuel has heated up to the point that there ,

. is runaway zirconium cladding oxidation (i.e., ignition) that is called a " zirconium fire." From the i perspective of offsite consequences, the working group only concemed itself with the zirconium fire eri state, because there has to be an energetic source (e.g., a large high temperature fire) to transport the fission products offsit ' r.trLhave potentially significant offsite consequences.

I The working group visited four decommissioning nuclear power plants to gather information on the as-operated, as-modified spent fuel pools, their cooling systems, and other support systems.  ;

The working group investigated the mitigation systems available as well as potential recovery actions by certified fuel handlers at the facilities. The working group also considered the availability of alarms, instrumentation ire E, staffing, electric power sources, maintenance  ;

personnel, fire protection systems, m Iurces, redundancy, diversity, and equipment  !

location with respect to the spent fuel pool.

Section 3.1 describes beyond design bases intemal event accident sequences that the working group's preliminary estimates show lead to significant offsite releases from spent fuel pools (i.e.,

a sequence of equipment failures or operator errors that could lead to a zirconium cladding fire i and release of radionuclides offsite). Section 3.2 discusses beyond design bases extemal event accident sequences. Sechon 3.3 provides the working group's insights from this preliminary risk l evaluation Appendix 1 lists the major assumptions made in the risk assessment. Appendix 2 details the event trees and fault trees used in the risk assessment. Appendix 4 discusses the integrity of spent fuel pools duririg tomadoes and high winds. Appendix 5 discusses the robustness of spent fuel pool integrity given heavy load drops. Appendix 6 discusses the potential effect from and frequency of aircraft crashes into spent fuel pools.

The risk from sabotage is not normally evaluated in a PRA, in part because such acts are not random events nor are they subject to easy analytical evaluation. The working group risk analysts willidentify to the NRC safeguards staff the structures, systems, and components that 2 Internal initiating events are events that begin within the confines of the nuclear power plant and cause plant disruption. Two examples are inadvertent closure of the spent fuel pool cooling system suction valves and a pipe break in the spent fuel pool cooling system. Extemal events are those events that begin outside the confines of the nuclear power plant. Two examples are seismic events and hurricanes. There are a few events that begin inside plants, such as intemal floods and fires, that have been characterized in some PRAs as external eventL.

25

are most important in helping assure that the spent fuel pools do not represent an undue the public.

3.1 Boys 6d Design Basis Spent Fuel Pool Accsdent Scenarios (Internal Events) i yalvj tj developing event trees and fault trees to The startmg point for this risk assess model the initiating events and syst po tent failures that lead to fuel uncovery (these trees are provided in Appendix 2). Working group estimates for initiating event frequencies considered potentially risk significant are given in Table 3.1-1. The table identifies the each frequency estimate (they are generic and not plant-specific). Working group est preliminary conditional failure probabilities of mitigating systems and components (b and passeve) are given in Table 3.1-2. The table identifies the sources for the condition pobability estimates.

DRAFT The following are the dominant scenarios (i.e., a description of how the working group mo the sequences with the highest expected frequency of fuel uncovery) developed by the group's preliminary risk assessment:

Dominant Scenarios Loss of Offsite Power from PlantMand Grid Related Events The initiating event in these scenarios is the loss of offsite power caused by hardware failures or design deficiencies within the plant, problems with the offsite power grid, human errors, or localized weather-induced faults such as lightening. With power lost, there is no effective heat removal process for the spent fuel pool. If power vWe not restored quickly enough, the poo heat up and boil off inventory until the fuel is uremt94 (if there is inadequate makeup). In the scenarios, the diesel-powered fire pump is avaikk W ::oolant makeup in case the ac power is not recovered if the diesel fails, and if offsite powei ". aot recovered in a timely manner, offsite recovery us g fire engines is a possibility. For the case of 1-year-old fuel (i.e., the last fuel w removed froci the reactor one year ago),127 hours0.00147 days <br />0.0353 hours <br />2.099868e-4 weeks <br />4.83235e-5 months <br /> is available for this recovery action. For 1-month-old fuel, 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br /> is available.

Even given r3covery of offsite power, the operator has to restart the fuel pool cooling pumps.

Failure to do this or failure of the equipment to restart will necessitate other operator recovery actions. Again, considerable time is available.

4 Case 1: Frequency of fuel uncovery (FFU) = 1.3x10 per year (Case 1 is the as-found case with the last fuel transferred one year previously) in this case the dominant sequences (i.e., combinations of events that have to occur to lead to fuel uncovery) are as follows:

' i) (loss of offsite power) x (the diesel-driven fire pump fails to start) x (offsite power is no recovered before fuel uncovery) x (no recovery help from offsite sources such as fire engines) f This sequence of events contributes to 60% of the FFU.

26

+

ii) (loss of offsite power with offsite power recovered, but the diesel-driven fire pump fails to start) x (operator fails to restart cooling system and electric fire pump) x (failure of recovery actions using offsite sources) = 29%

iii) (loss of offsite power with power recovered and the diesel-driven fire pump is available) x (the cooling syste f is 19festart, either due to hardware failures or operator errors) x (failure of re sctio.1s such as use of fire pumps or offsite sources) = 11%

Case 2: FFU = 4.2x104 per year (Case 2 is the as-found cne with the last of the spent fuel transferred one month previously)

The differences between Case 1 andM heIdriven by less time being available for recovery of failed equipment for Case 2 where there is higher decay heat.

l) = 25%

ii) = 69%

iii) = 6% DRAFT Case 3: FFU = 8.0x10-5 per year (Case 3 assumes the spent fuel pool and its support systems are configured in a much more minimal manner than found during the working group's visit to decommissioning plant sites. It assumes that the last spent fuel was transferred one year previously.)  ;

With no fire protection equipment available, and if offsit'e powcr is not recovered, the only recovery available is from offsite sources such as fire engines.

l) (loss of offsite power is recovered) x (the operator fails to restart the spent fuel pool .

cooling system) x (failure of recovery help from offsite sources such as a fire engine) i

= 95%

ii) (loss of offsite power and no recovery) x (failure of offsite recovery sources) = 5%

Loss of Offsite Power from Severe Weather Events These scenarios are similar to the loss of offsite power events caused by plant-centered and grid related events. However, initiating event frequencies are smaller (0.007 per year versus 0.08 per year) while the probability of not recovering offsite power is also much larger (0.02 for 1-year fuel,0.1 for 1-month fuel, versus 0.001). Note that all initiating event frequencies and offsite power recovery probabilities are based on generic data, not plant-specific data.

27

Case 1: FFU = 1:4x104 per year l} (loss of offsite power occurs with no recovery) x (the diesel-driven fire pump fails to start or run) x (no recovery from offsite sources such as a fire truck) = 97%

ii)(offsite poweris recovered)f il qJhe spent fuel pool cooling system to re-start and run) x (operator fails to stest -dliven fire pump or electric fire pump for makeup) x (failure of offsite sources) = 2%

Case 2: FFU = 9.4x104 per year Even with less time for recovery of failed equipment (because decay heat is much higher in Case 2 than Case 1), the two scenari i dominate.

l) = 97%

ii) = 2%

Case 3: FFU = 1.4x104 per year With no fire protection system to proMSacSp source for inventory makeup, the preliminary risk assessment shows scenario (ii) becomes more important.

1) (loss of offsite power with no recovery) x (no recovery from offsite sources) = 51%

ii) (loss of offsite power with recovery) x (operator fails to restart cooling system) x (no recovery from offsite sources) = 49%

Internal Fire The risk from this initiating event appears to be dominated by one scenario that involves a fire that is not suppressed in the building containing the spent fuel, and that is large enough to either fail the offsite power feeds, or the fuel pool cooling pumps. Spent fuel would become uncovered if the operator fails to refill the pool using the diesel-driven fire pump and if recovery using offsite sources (e.g., fire engines) fails. Note that because of the fire, recovery of offsite power or cooling pumps is assumed to not be likely and therefore was not modeled.

r ase 1: FFU = 8.6x10-'per year The initiating event frequency of 8.9x108 per year is based on generic data on pump-initiated fires and electrical cabinet-initiated fires.

Case 2: FFU = 1.0x104 per year The initiating event frequency from electrical cabinet and pump initiated fires is similar to above.

However, in this case, cutting and welding are assumed to be prevalent because the plant is assumed to be actively in the process of physically dismantling the support systems to the spent fuel pool and other plant systems or components. Using generic data, the initiating event g

28

frequency of 0.04 per year was used. Since this scenario assumes the last fuel was transferred from the reactor one month previously, it was assumed that the piutslity of fire suppressson was also greater because the building will more likely be occupied. The probability that a fire is not suppressed and that there is an effect on spent fuel pool cooling function is 0.05 for this case, versus 0.1 for Case 1.

Case 3: FFU = 9.0x104 peryeDRAFT This scenario is similar to Case 1, except the probability of failure of fire suppression is doubled (no fire fighting equipment available) and recovery of cooling or inventory is not possible, since no fire pumps are available. The only recovery is through offsite sources (e.g., fire engines).

Loss of Cooling DRAFT The initiating event frequency includes the loss of coolant system flow from the failure of pumps or valves (See Figure 3.0-1), from piping failures, from an ineffective heat sink (e.g., loss of heat exchangers), or from a local loss of power (e.g., electrical connections.) Operational data from NUREG-1275, Volume 12 shows that the frequency of loss of spent fuel pool cooling events in which a temperature increase of more than 20'F occurred can be estimated to be on the order of two to three events per 1000 react - a 12e data also showed that, for the majority of events, the duration of the loss of ens than one hour. Only three events exceeded 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, with the maximum duration being 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />. There were four events where the temperature increase exceeded 20 F, with the maximum increase being 50'F.

For loss of cooling events, there is a lot of time for operator recovery. In the case of 1-year-old fuel,127 hours0.00147 days <br />0.0353 hours <br />2.099868e-4 weeks <br />4.83235e-5 months <br /> is available. For 1-month-old fuel, the recovery time is 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br />. The preliminary evaluation is that the risk of fuel uncovery is small.

Case 1: FFU = 1.5x104 per year l} (loss of cooling) x (failure of control room alarms) x (operator failure to notice pool steaming and level drop during walkdowns) >99%

ii) (loss of cooling) x (failure to recover the regular fuel pool cooling system) x (failure of fuel pool makeup using fire pumps) <1%

iii) (loss of cooling) x (alarms fail but operator notices signs of loss of fuel pool cooling during walkdown) x (failure of onsite and offsite recovery actions) <1%

29

p Case 2: FFU = 1.7x10-7 peryear For this case the wndow for successful operator action becomes shorter because of increased decay heat loads, and therefore, recovery actions become relatively more important..

1) = 90% '

DRAFT ii) = 9%

iii) = 1%

' Case 3: FFU = 1.7x104 per year The control room alarms are assumebMM less reliable. There is no makeup capability through the fire pumps because they are not required to be operable.

l) = 91%

ii) = 7%

iii) = 2%

DRAFT Loss of Coolant inveatory This initiator includes loss of coolant inventory from events such as those resulting from ~

configurabon control errors, siphoning, piping failures, and gate and seal failures. Operational data provided in NUREG-1275, Volume 12 show that the frequency of loss of inventory events in which a level decrease of more than one foot occurred can be estimated to be (on the order of) less than one event per 100 reactor years. Most of these events are as a result of operator error and are recoverable. NUREG-1275 shows that, except for one event that lasted for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, there were no events that lasted more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Eight events resulted in a level decrease of between one and five feet, and another two events resulted in an inventory loss of between five and 10 feet.

Using the information from NUREG-1275, it can be estimated that 6% of the loss of inventory events will be large enough and/or occur for a duration that is long enough so that isolation of the loss is required if the only system available for makeup is the spent fuel pool makeup system. For the other 94% of the cases, operation of the makeup pump is sufficient to prevent fuel uncovery.

Case 1i FFU = 2.9x104 per year I) (a "small" loss of inventory event) x (level not restored by m'keup pumps or fire pumps due to operator error or hardware failure) x (offsite recovery, sv.c as fire engines, unsuccessful) = 49%

30 u

(

ii) (a "large" loss of inventory event) x (operator does not isolate loss) x (level not restored by fire pumps due to operator error or hardware failure) x (offsite recovery, such l as fire engines, unsuccessful) = 23%

iii) (a "small" loss if inventory event) x (failure of control room alarms) x (failure of the operator to notice condition di r s) = 16%

iv) (a "large" loss of inventory event) x (failure of control room alarms) x (failure of the l operator to notice condition during walkdowns) = 11%

1 Case 2: FFU = 6.0x10-5 per year l This case is the same as Case 1 exo cparator error events become more dominant due '

to the shorter window within which op Muut act to recover. The 'large" break also becomes more important because less time is available for the operators to respond.

4 l} = 47%

ii) = 51%

iii) = 0.8% DRAFT iv) = 0.5%

Case 3: FFU = 1.3 x 10d per year l} (A "small" loss of inventory) x (failure of control room alarms) x (the operator fails to notice condition during walkdowns) = 39%

ii) (a "small" loss of inventory) x (level not restored by makeup pump or fire pumps) x (offsite recovery fails) = 36% -

iii) (a "large" loss of inventory) x (failure of the control room alarms) x (operator fails to notice condition during walkdowns) = 10%

iv) (a "large" loss of inventory) x (loss is not isolated or makeup pump fails) x (offsite recovery fails) = 8%

Heavy Load Drops The working group investigated the frequency of dropping a heavy load in or near the spent fuel pool, and investigated potential damage to the pool from such a drop. Details of this evaluation can be found in Appendix 5. Based on discussions with structural engineers, the working group assumed that only spent fuel shipping casks had sufficient weight to catastrophically damage the pool if dropped. Other heavy loads are assumed to be treated using the guidelines presented in NUREG-0612, and are of low likelihood to be moved over the spent fuel pool.

31

For drop frequency, it is eshmated that the likelihood of a load drop into the spent fuel pool is in 4

the range of 2.0x104to 1.5x10 per year for a non-single failure proof load handhng system.

For a smgie failure proof load handling system (or a plant conforming to4 the NUREG 0612 guidelmes), the range is reduced by about a factor of 10 (2.0x10* to 1.5x10 per year). Once the load is dropped, the next question is whether the drop did significant damage to the spent fuel pool.

DRAFT For the failure of the pool wall, it is assumed that the load physically travels over the wall 2 percent (0.02) of the time (10 percent of the 5 to 25 percent of the total load path) and it is assumed a conditional one-in-ten (0.1) chance of significant damage given load drop. The 4

working group estimates a failure rate in the 2.0x104 to 3.0x10 per year for the non-single 4

failure proof system and 8.0x10* to 2.0x10 per year for the single failure proof system (or a plant conforming to the NUREG-061pgip (The value given in NUREG/CR-4982 for wall failure was 3.7x104 per year.) Th-n - - _ _ _ are bounded by other more likely initiating events.

For failure of the pool floor, the working group assumes one-in-ten events (0.1) will result in per

. significant damage. The working group estimated a failure rate range of 2.0x104 to 1.5x104 year for the non-single failure proof system and 2.0x104 to 2.5x104 per year for the single failure

-0612 guidelines). The upper ends of these proof system (or a plant conforming t estimates represent potentially impo ors to spent fuel pool risk at decommissioning plants.

3.2 Beyond Design Basis Spent Fuel Pool Accident Scenarios (Extemal Events)

Seismic Events Spent fuel pool structures at nuclear power plants are constructed with thick reinforced concrete 1/4 inch thick. The walls vary from 4.5

. walls and slabs lined with thin stainless steel liners 1/8 to to 5 feet in thickness and the pool floor slabs are around 4 feet thick. The overall pool dimensions are typically about 50 feet long by 40 feet wide and 55 to 60 feet high. In boiling water reactor (BWR) plants (BWR-3s, BWR-4s, and BWR-5s) the pool structures are located in the reactor building at an elevation several stories above the ground In pressurized water reactor (PWR) plants and BWR-6s, the spent fuel pool structures are located outside the containment structure supported on the ground or partially embedded in the ground. The location and supporting arrangement of the pool structures determine their capacity to withstand seismic loads beyond their design basis. The dimensions of the pool structure are generally derived from radiation shielding considerations rather than structural needs. Spent fuel structures at operating nuclear power plants are inherently rugged in terms of being able to withstand loads substantially beyond those for which they were designed. Consequently, they have significant seismic capacity.

In its preliminary risk analysis, the working group assumed that the spent fuel pools are robust for seismic events less than three times the safe shutdown earthquake (SSE). It is assumed that the 32

A high con 6dence, low probabihty of failure (HCLPF)* value for the spent fuel pool integrity is 3 x SSE. For the majonty of sites and plants, 3 x SSE is in the peak ground accelershon range of 0.4 to 0.5 g's (where g = the accoloration of gravity or 980 cm per sec .) Seismic hazard curves from the latest Lawrence Lsvermore Nabonal Laboratory Study (NUREG-1488) show that, for most plants, the mean frequency for seismic accelerations of 0.4g to 0.5g is on the order of or less than 2x104 per year. These haza y are consistent with those from the Electnc Power Research Institute (EPRI) studi w 4726 and EPRI NP-6395-D). )

Using the definition of HCLPF, the working group applied a mathematical shortcut

• to get the frequency of a sessmic event that will challenge spent fuel pool integrity 2x104 per year x 0.05 51x104 per year.

DRAFT Given such an earthquake, it is assumed that the spent fuel pool structure will fail, the pool will i drain, there will be no recovery, and fuel uncovery frequency will equal 1x104 per year. For the l cases (95% of the time) where the pool is intact, it is assumed that some part of the spent fuel I pool cooling components will fail and cooling will be lost. Recovery from offsite sources may be l possible (although there would be considerable damage to the area's infrastructure at such accelerations).' The estimate for the contribution from this scenario is  ;

earthquake also 1E-6 per year. D preliminaEA FT l The preliminary estimate for the total seismic contribution to the frequency of spent fuel uncovery for a plant with a spent fuel pool HCLPF three times its SSE and sited at most sites east of the Rocky Mountains is about 2x104 per year.

Tornadoes The worlung group performed a preliminary risk evaluation of tomado threats to spent fuel pools (details are in Appendix 4). It was assumed that very severe tomadoes (F4 to F5 tomadoes) would be required for significant damage to a PWR or BWR spent fuel pool. The working group then looked at the frequency of such tomadoes occurring and the conditional probability that if such a tomado hit the site, it would seriously damage the spent fuel pool or its support systems.

To do this the working group examined the frequency and intensity of tomadoes in each of the continental United States. The working group determined, based on the buildings housing the spent fuel pools and the thickness of the spe'1t fuel pools themselves, that the probability of a tomado causing a catastrophic failure of the spent fuel pool is negligible and can be ignored. The working group assumed that an F2 to F5 tomado would be required for possible significant

• A HCLPF is the peak acceleration value at which there is a 95% confidence that less than 5% of the time the structure, system, or component (SSC) will fail. The HCLPF value is frequently used as part of a process of demonstrating that an SSC is sufficiently rugged for its intended purpose.

' The working group independently confirmed that the short cut is accurate within about a factor of two, which is well within the uncertainty inherent in hazard curves for a site and the fragility estimates for the spent fuel pool.

33

damage to a spent fuel pool support system (power supply, heat exchanger or makeup water supply). The estimated frequency of fuel uncovery in the spent fuel pool from such tomadoes in the range of 2x10-7 per year for Cases 1 and 3 and in the range of 1x104 per year for Case 2.

Overall, the hkehhood of significant spent fuel pool damage from tornadoes is bounded by other more likely catastrophic spent fuel pool failure and loss of cooling modes.

Aircraft DRAFT The working group analyzed the likelihood of aircraft crashing into a nuclear power plant site and sonously damaging the spent fuel pool or its support systems (details are in Appendix 6). The genenc dats provided in DOE-STD-3014-96, " Accident Analysis for Aircraft Crash into Hazardous Facilities," U.S. Department of Energy, October 1996, were used to assess the hkahhood of an aircraft crash into or r " iiissioning spent fuel pool. Aircraft damage I or affect the availability of nearby support can affect the structural integrity of systems, such as power supplies, heat exchangers and water makeup sources, and may also affect recovery actions.

The working group's preliminary estimate of the frequency of significant pool damage is as follows:

The frequency of significant PWR speMilamage resulting from a direct hit is based on the point target area model for a (100 x 50) foot pool with an probability of 0.3 (large aircraft penetrating 6-ft of reinforced concrete) that the crash results in significant damage, if 1-of-2 aircraft are large and 1-of-2 crashes result in spent fuel uncovery, then the value range is 2.7x104to 4.0.x104 per year.

The frequency of a significant BWR spent fuel pool damage resulting from a direct hit is the same as that for the PWR,2.7x104 to 4.0x104 per year. Mark-l and Mark-Il secondary containments generally do not appear to offer any significant structures to reduce the likelihood of penetration although a crash into one of four sides of a BWR secondary containment may have a reduced likelihood of penetration due to other structures being in the way of the aircraft. Mark-lit secondary containments msy reduce the likelihood of penetration as the spent fuel pool may be considered to be protected on one side by additional structures.

These frequencies of catastrophic spent fuel pool failure are bounded by other initiators.

' The working group finds the frequency of significant spent fuel pool support system unavailability caused by aircraft crashes to be as follows: l The working group modeled loss of a support system (power supply, heat exchanger or makeup water supply) based on the DOE model. The crash consideration includes the area the wing sweeps as the plane skids and the skid length. The target area is 400 x 200 x 30-foot area with a conditional probability of 0.01 that one of these systems in the target area is hit. The frequency range is 7x104 to 1.0x104 per year. These values are significantly lower than other failure modes that can fail spent fuel pool support systems.

The working group also modeled loss of a spent fuel pool cooling support system (power supply, l

heat exchanger or makeup water supply) based on the DOE model where the concem was about 34

httmg a 10 x 10 x 10-foot structure. In this case the results gave a frequency, range of 7.3x104 to 1.1x104 per year with the wing and skid area. For the point target area model, the point estimate is 7.4x10 to 1.1 x104 per year. These values are significantly lower than other potential support system failure modes.

DRAFT DRAFT DRAFT 35

Figure 3.0-1 Representatwo Spent Fuel Pool Model r

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Table 3.1-1 Loss of Spent Fuel Pool Cohng initiating Events NTIATNG EVENT FREQUENCY fporveed SOURCE OF FREQUENCY Loss of Offsite Power-Plant 0.08 Genene frequency obtened from contered and grid related operabonalevents at nuclear power events - DRAFT plants (INEL-96/0334 and NUREG/CR-5496)

Loss of Offsde Power- Events 7.0E-03 Generic frequency obtained from inibated by severe weather operational events at nuclear power '

plants (NUREG/CR-5496)

Intemal Fire 9.0E-03 (1 yr after Genenc frequencies obtained using

] 7 the methodology provided in the EPRI 0.04 Tm@oafterslutdown)

L Fire-Induced Vulnerability Evaluation report (EPRI TR-100370s)

Loss of Pool Cooling 3.0E-03 Generic freewncy from operating experience (NUREG-1275 Vol.12)

Loss of Coolant inventory 0.01 Generic frequency from operahng nRAFT experience (NUREG-1275 Vol.12)

Seismic Event 2.0E-05 Estimate for a 0.4 to 0.5 g earthquake, obtained from NUREG-1488 Cask Drop 2.5E-06 Estimate based on industrial data (NUREG-0612 and other studies - see Appendix 5 for more detail)

Aircraftimpact 4.0E-08 Estimate based on DOE-STD-3014-96 and other studies (see Appendix 6 for more detail)

Tomado Missile 5.6E-07 Estimate based on data from the National Climatic Data Center- see Appendix 4 for more details) 37

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Table 3.1-3 Spent Fuel Pool Coolmg Risk Analysis Frequency of Fuel Uncovesy (per year)

INITIATING EVENT D W iT CASE 2 CASE 3 Loss of Offsste Power- Plant 1.3E-06 4.2E 06 8.0E-05 centered and grid related events Loss of Offsite Power- Events 1.4E-06 9.4E-06 1.4E-05 initiated by severe weather intamal Fire ' DRM7T 1.0E-06 9.0E-06 Loss of Pool Cooling 1.5E-07 1.7E-07 1.7E-05 Loss of Coolant Inventory 2.9E-06 6.0E-05 1.3E 04 Seismic Event 2.0E-06 2.0E-06 2.0E-06 Cask Drop 2.5E-06 2.5E-06 1.5E-05 Aircraft impact MO@ '

4.0E-08 4.0E-08 Tomado Missile - 53ED7 5.6E-07 5.6E-07 Total 1.2E-05 8.0E-05 2.7E-04 3.3 Probabilistic Assessment insights

1. The frequency of spent fuel pool fuel uncovery appears to be driven by the degree of defense-in-depth, redundancy, and diversity in the sled mounted spent fuel pool cooling systems (including makeup capability and support systems) at decommissioning plants that replaced operating reactor spent fuel pool cooling systems and backup systems

_( e.g., residual heat removal system).

2. Preliminary results indicate there is the possibility of a factor of 20 or more increase in the frequency of fuel uncovery if utilities choose to reduce their spent fuel pool cooling capabilities much below that found at the four decommissioning sites visited by the working group in preparation for the spent fuel pool risk assessment.

3. One year after the last fuel is transferred from the reactor to the spent fuel pool, given appropriate procedures, instrumentation, equipment, and operator awareness, there appears to be sufficient time for operators to respond to most loss of spent fuel pool cooling initiators and terminate the event before there is a potential for offsite dose consequences. This is not necessarily true for heavy load drop and very large seismic events that have the potential to rapidly drain the spent fuel pool and uncover the fuel. If one of these initiators were to occur during the first year or two after the last fuel was transferred from the reactor to the spent fuel pool, it appears that there would be only five to seven hours available for local emergency response. This may be too short for an effective evacuation to preclude some early fatalities.

45

L t

4. Lack of action levels higher than the ' alert" level at decommissioning plants may need to be reevaluated. The working group is evaluatmg whether the timing of a warning should be commensurate with the potg@uences of having a particular release at a decommissioning plant.

5. In performing its preliminary risk assessment, the working group evaluated seismic-induced risk at decommissioning plant spent fuel pools. The working group is evaluating whether the risk is sufficiently low to allow exemptions for emergency preparedness in the area of seismic events if the HCLPF of the spent fuel pool can be shown to be at least c basis, utilities seeking exemptions might three times the site's SSE. Onqprg nfidence that the as-built spent fuel pool has follow the efforts described beldwtal a sufficiently robust HCLPF.

Walk down the pool structure and its vicinity and note:

a. physical conditions such as cracking, spalling of concrete, signs ofleakage or leaching and separation of pool walls from the grade surface,

b. arrangement and layout of supporting columns and shear walls, assessment of other loads from tributa l h @Mpping of any existing structural cracks, pool, as-built dimensiorir's ntfr g

c. adjacent structures that can impact the pool structure above and below the grade surface, supporting arrangement for superstructure and crane and potential for failure of the superstructure and the crane, the weight of the heaviest object that can drop in the pool structure and the corresponding drop height.

Calculate the seismic capacity of the pool structure. Typically such a calculation consists of the following:

a. review existing layout drawings and structural dimensions and reconcile the differences, if any, between the as-built and as-designed information and consider the effects of structural degradation as appropriate,

b. from design calculations, determine the margin to failure and assess the extrapolated multiple of SSE level that the pool structure should survive, determine whether or not design dynamic response analysis including soil-l structure interaction effects are still applicable at the capacity level seismic event; if not, conduct a new analysis using properties of soil at higher strain levels and reduced stiffness of cracked reinforced concrete,

c. determine the loads from pool structure foundation uplift and from impact of pool structure with adjacent structures during the capacity level seismic event,

' determine loads from the impact of spent fuel rack on the pool floor and the side walls and determine the loads from dropping of heavy objects from the collapse of superstructure or the overhead crane,

d. determine a list of plausible failure modes; failure of side walls due to the worst loading from the capacity level earthquake in combination with fluid hydrostatic and sloshing head and dynamic earth pressure as appropriate, failure of the pool floor slab in flexure and bending due to loads from the masses of water and the 46

)

I 1

spent fuel and racks, local failure by punching shear due to impact between I structures and the spent fuel racks or dropping of heavy objects, e.

D. RAFT the calculations to determine the lowest structural capacity can be based on I ultimate strength of reinforced concrete structures due to flexure, shear and j punching shear. When conducting an yield line analysis, differences in flexural I yield capacities in two orthogonal directions and for the negative and positive bending moments influence the crack patterns and several sets of yield lines may l have to be investigated to obtain the lowest capacity. For heterogeneous i materials, the tradition 'ya ulysis provides upper bound solubons; l consequently, conside hs r eeded to determine the structural capacity j based on the yield lines that approximate the lower bound capacity.

l l

6. Heavy load drops are potentially the most significant loss of spent fuel pool cooling initiators. A heavy load drop has the potential to fail the spent fuel pool, such that it would drain rapidly and would be incapable of being reflooded from onsite or offsite sources. l The risk perspective from zirconium fires during their period of vulnerability can be significantly improved by control of movement of heavy loads near or over the spent fuel .

P

DRAFT l

l l

l l

1 l

l i

l l

47 i

~ 4.0 Supportmg information i

4.1 Maintenance R* DRAFT 4.1.1 Identdication of Maintenance Rule Concepts at Decommissioning Plants I f

Maintenance Rule (MR) concepts at decommissioning plants are expressed in 10CFR50.65,

• Requirements for monitoring the effectiveness of maintenance at nuclear power plants."

also provides the fundamental regulatory requirements for maintenance rule and scoping o MTimiss#oning status plants (DSPs). Scoping structures, systems, and compone.9ts requirements have been establisheo . ... n ; clear power plant for which the licensee has submitted the certifications specified in GO.82(a)(1), this section only shall apply to the extent that the licensee shall monitor the performance or condition of all structures, systems, and components associated with the storage, control, and maintenance of spent fuelin a safe condition, in a manner sufficient to provide reasonable assurance that such structures, systems, and components are capable of fulfilling their functions..."

Normally, a DSP will have already submitted the 50.82(a)(1) certification; this will satisfy the status statement for a plant as quoted ic;from a practicable maintenance rule viewpoint we only need worry about how the DS e p rovides for: "... monitoring the performance or condition of all SSCs associated with the storage, control, and maintenance of spent fuelin a safe condition...."

4.1.2 Identification of Potential Systems, Equipment, Functions at Decommissioning Plants The Office of Nuclear Reactor Regulation (NRR) staff has conducted workshops and provided a

" staff support member" for three onsite inspections of MR implementation at shutdown plants.

Consequently, by consuNation at the workshops and precedent at the three sites inspected, a scoping list of expected SSCs has _been developed. Among those generic SSCs that ma the scope of the Maintenance Rule for DSPs are the following:

• SFP cooling and cleanup (SFPCC) e SFP structure and any connecting piping system seals e SFP building

• Radiation monitors above or in the SFP e Standby service water system (i.e., the portion that is heat sink for SFPCC heat exchangers)

• For boiling water reactors, the residual heat removal system can also be used for SFP cooling

• Leak detection system (this system detects leakage from the SFP, the transfer pool, and reactor well pool liners. Portions of this system that detect leakage from the SFP could be under the scope of the Maintenance Rule) ,

48 e

e Heating, ventilation and air conddioning sy% above the SFP e SFP waterleveli e SFP emergency or normal makeup water supply e SFP crane (i.e., may need to be included as a result of accident scenarios involving dropped fuel bundles, which could cause cladding damage and potenhal radishon exposure to personnelin the SFP area.)

e Standby auxiliary ac power sys .t power supply to SFPCC pumps) e SFP transfer tube penetration seals or bellows, pneumatic air system which inflates the seals, and transfer tube gate seals (i.e., failure of these seals could cause a SFP drain down to a few feet above top of active fest, the area around the SFP would become a very high radiation area, and plant staff would need to evacuate the SFP area).

This list will vary depending on the circumstances of each individual DSP and is only intended to be a representative list.

MMT The MR requires that SSCs be monitored for reliability and availability by the licensee. MR policy recognizes that the longer a plant is shutdown the less likely that fuel is vulnerable to zirconium oxidahon. Consequently, the stringency of reliability and availability standards are graded accordingly.

4.1.3 Evaluation of the Maintenance Rule Oversight at Decommissioning Plants Background For implementation of the MR For Decommissioning Status Plants r

The maintenance rule was effective July 1996, and amended August 1996, to include nuclear plants which have submitted the certifications specified in 10 CFR 50.82(a)(1) (i.e.,

decommissioning status plants). As amended, the rule requires licensee's to monitor the performance or condition of SSCs associated with the storage, control and maintenance of spent fuel in a safe condition and in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions.

Subsequent to the August revision of $50.65, the associated RG 1.160, " Monitoring The Effectiveness Of Maintenance At Nuclear Power Plants," and NUMARC 93-01,' Industry Guideline for Mon;toring the Effectiveness of Maintenance at Nuclear Power Plants,' were revised to address the amended rule provisions for decommissioning status plants.

(NUMARC 93-01 is the industry guideline for monitoring the effectiveness of maintenance at nuclear power plants and has been endorsed by the NRC's Regu: story Guide 1.160.)

Overview of implementation of the MR for Decommissioning Status Plants o implementation of the maintenance rule for decommissioning status plants has been fundamentally the same as for operating facilities; however, the population of SSCs within scope were reduced.

49

Relative to the scope of SSCs that would be applicable for a decommissioning status plant boonsee's need to focus on the functions necessary for the preservation of spent fuel in a condition. For decommissioning statug Ilyexplicit process related to risk ranking of

e necessary.

SSCs, delineated in NUMARC g3-01, may not Expectations For implementation of the MR for Decommissioning Status Plants Licensees for decommissioning status plants will continue to establish SSC performance measures and condition monitoring as well as the requisite goals and preventive maintenanc activities consistent with the requireirgfgg.

An SSC would be considered to be in the (a)(1) catergory when the performance or con

' the SSC does not meet established goals and corrective actions are required.

An SSC would be considered to be in the (a)(2) category when it has been demonstrated perfonnance or condition of the SSC is being effectively controlled through the per appropriate preventive maintenance.

Licensee's Periodic Evaluation on Eggs of the MR for Decommissioning Status Plants Effectiveness of the maintenance rule implementation process will be evaluated by licensee's a peiicdicity not to exceed 24 months between evaluations. Balancing reliability and a are part of the periodic evaluations. Preventive maintenance safety assessments are don

. ongoing basis.

4.2 - Quality Assurance Part 50 of Title 10 to the CFR, Appendix B quality assurance (QA) requirements apply to all activities affecting the safety-related functions of those SSCs that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and sa public. These activities include designing, purchasing, fabricating, handling, shipping cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, and m Safety-related structures, systems, and components are those SSCs that are relied upon remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary, (2) The capability to shut down the reactor and maintain it in a safe shutdown condi or, (3) The capability to prevent or mitigate the consequences of accidents which cou in potential offsite exposures comparable to the applicable guideline exposures set forth in $50.34(a)(1) or $100.11.

Upon docketing of certifications required by $50.82(a)(1) for permanent cessation of and permanent removal of fuel l rom the reactor vessel, the license under Title 10 of t 50 l

l longer authorizes operation'of the reactor or emplacement or retention of fuel into the reactor vessel. After the reactor fuel has been permanently removed from the vessel and placed in the l spent fuel pool, potential offsite releases are the primary safety consideration.

Subsequent to permanent relocation of all fuel to the spent fuel pool, licensees may apply the

$50.5g regulatory change control process g$if SSCs that no longer perform safety functens associated with maintaining ti= reactor coola(nt pressure boundary j of shutting down the reactor. Reclassification of safety-related equipment through the $50.5g  !

process may considerably reduce the number of SSCs subject to Apperdx B QA requirements.

Licensees generally retain their approved ; ir.84A programs through the -

decommissoning phase. Changes to QA i iiorimitments are submitted for NRC approval through the $50.54(a)(3) change control process. Although considerable simplificaten in the management, administraten, and oversight of the QA program may be possible, the eighteen  !

Appendix B criteria are genetally applicable to any equipment or structure that performs a safety funchon Beyond decommissioning activities, licensees can use their approved Appendix B programs (following review for applicabilityMng and transportation of radioactive '

material (Part 71) and for storage of spent nuclear fuel and high-level radioactive waste (Part 72).

1 An altemate approach, planned by one licensee, would be to submit a new quality asse e nce  ;

program specifically applicable to decommissioning activities. If such a program were suumitted, i the staff would review it for conformance with Appendix B criteria in accordance with Standard ,

Revow Plans (SRP) 17.1/17.2. Pursuant to $50.34(b)(6)(ii), this review would extend to the j determination of hgg Appendix B requirements would be satisfied through licensee commitments to applicable regulatory guides and industry standards. The staff would review for acceptability any proposed attematives to NRC-endorsed industry standards.

Changes to technical specification (TS) administrative controls ($50.36(c)(5, 6)) related to quality assurance are revewed in accordance with applicable standard review plans, regulatory guides,  ;

industry standards, and other regulatory guidance, such as the proposed standard technical l slJecifications for permanently defueled Westinghouse plants (NUREG-1625). Relocation of TS  !

administrative controls to the licensee's QA program description are reviewed using the guidance contained in Administrative Letter 95-06.  ;

For design basis events which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in $50.34(a)(1) or $100.11, the quality assurance requirements of Appendix B to 10 CFR Part 50 apply.

4.3 Water-basin Type Independent Spent Fuel Storage Installations A spent fuel storage facility built separately from a reactor facil.ty would be licensed under the i regulations of 10 CFR Part 72 [Ref.1). This facility would be called an independent spent fuel storage installation (ISFSI). There are two types of ISFSis, a dry cask facility and a water-basin facility. Many decommissioning plants isolate the SFP ares from the plant so that decommissioning activities can proceed vdth minimalimpacts on the SFP area. The concept of a

" spent fuel pool island" is similar, although not identical, in nature to an ISFSt. The working group explored the design considerations and controls for an ISFSI to identify items that should be  !

consistent between the two types of facilities.

51 l i

Reguiste'f Guide 3.4g [Ref. 2] geees a method to comply with the design requirements for a water-besin ISFSI by endorsing, with provisions, American Natenal Standard institute /American Nuclear Society standard (ANSI /ANS) 57.7-1981, ' Design.Critena for an independent Spent Fuel Storage instalisten (Water Pool Type)" [Ref. 3]. The function of the ISFSI is to provide interim protective custodial storage of spent fuel. By the standard, it is expected that the facility would be staten or reprocessing facility; however, completely independent or adjacent to a no no direct means to transferring fuel assem: es om the nuclear facility to the storage installabon can exist. This basic function is the same for a SFP at a nuclear power plant. 'However, the SFP is required to have the capability to be connected to the facili+y to receive spent fuel directly from

)

the reactor.-

SFPs at nuclear power plants were desigth,N,hrklicenseo ln accordance with 10 CFR l Part 50 regulaton [Ref 4]. Without a demonstrated substantial increase in the overall pwi. den of public health and safety or the common defense and security in accordance with 10 CFR 50.109 [Ref. 5], the staff can not require the modification of or addition to systems, structures, components or design D of a facility' RAFT While the requirements from the Part 72 standard for a design of an ISFSI can not be applied to SFP facilities that are already built, it is recognized that the design considerations and controls may provide insights on what should considered important for decommissioning plants. An evaluation of Part 72 considerations and controls will also provide consistency within the agency where possible.

4.3.1 Design Considerations There are four overall design considerations in ANSI /ANS 57-7-1981. Each of these are described below as they applies to spent fuel pools at decommissioning plants, e Overall Design Consideration A: Short-lived, high specific activity radionuclides, particularly of iodine and xenon are no longer present in significant quantities in spent fuel that has aged for more than one year since discharged from the reactor core.

' This design consideration describes the natural decay process of radionuclides in spent fuel following the discharge from a nuclear reactor. The statement also applies to soent fuel at decommissioning plants.

o Overall Design Consideration B: The storage of aged spent fuel is a low hazard potential activity. Very little of the radioactivity present is available in a dispersible form and there is no mechanism present to cause the release of radioactive materials in significant quantities from the installation.

This criteria is used to qualitatively assess offsite consequences. An NRC Backfit Review Panel l [Ref.13) determined that a significant loss of spent fuel pool water accident that results in fuel uncovery is a reasonable credible accident to consider for exemptions to some 10 CFR Part 50 regulations, such as emergency planning. If the decay heat is sufficient to increase clad temperature to the point of self-sustaining oxidation, a zirconium fire could occur (see Section 2).

. If a fire occurred, significant quantities of radioactivity could be released (see Section 3.3). For

' fuel uncovery events, considering high bumup fuel (up to 60-62 GWD/MTU) and high density 52

spent fuel racking in the storage pools, a zirconium fire is possible beyond one year following permanent shutdown of the plant.

• Overall Desen Consideration C: Decay heat is not a signWicant design considershon in an ISFSI because the spent fuel shall have aged a minimum of one year since descharge from the reactor ca'*- DRAFT j For a nuclear power plant, the capability to remove spent fuel decay heat is a design cnteria for j the SFP cooling system. General Design Criterion 61 requires, in part, that the system shall be designed with residual heat removal capability that reflects the importance of decay heat. SFP cooling and makeup systems at nuclear E-:T -

• is re designed to remove the decay heat from a whole core of freshly discharged sp er M At decommissioning plant, the amount of cooling that is needed reduces as the decay heat decreases.

o Overall Design Consideration D: Because of the low potential for airbome radioactive materials above a spent fuel pool, the buildi housing such pools need not be designed for confinement. DRAET The buildings for spent fuel pools are not required to be designed for confinement of airbome radioactive material. This design consideration is consistent with the design criteria used for spent fuel pools at nuclear power plants.

The ANSI /ANS standard states that the sole function of the ISFSI is the interim protective custodial storage of spent fuel. While this function is the same function for a spent fuel pool at a nuclear power plant, some of the design requirements are different due to the different emphasis that is needed for the storage of the fuel in each facility, in the design of a spent fuel pool, the decay heat removal capability must be sufficient to handle a whole core of freshly discharged spent fuel. In contrast, a water-basin ISFSI cannot accept spent fuel until it has decayed at least one year. Another significant difference in the operation of a water-basin ISFSI and a spent fuel pool at a decommissioning plant is that an ISFSI is expected to routinely receive spent fuel shipments, while at a decommissioning plant, except for final removal from the pool, minimal fuel movement is expected. As such, the decay heat load at an ISFSI will continuously fluctuate, I while the heat load at a decommissioning plant spent fuel pool will generally only decrease. With distinctions such as these is mind, the working group believes that not all of the controls required for a water-basin ISFSI would be necessary for a spent fuel pool at a decommissioning plant.

4.3.2 Defense-in-Depth The defense-in-depth concept is used in the design of a water-basin ISFSI. The primary barrier l is the fuel clad. Several secondary barriers are provided to control radioactivity. The pool water ,

itself retains radioactive particulate material and provides plant personnel shielding. Suberiticality l la provided by the design of the spent fuel storage racks and design features of the handling system. The pool clean up system is provided to remove dissolved and suspended radioactive i material. A clean water environment is important to minimize corrosion, coolant cleaning and  !

water chemistry monitoring are means to ensure a quality environment is maintained. A sufficient source or sources of water should be available for routine makeup and be of sufficient capacity for event mitigation activities. Indication of loss of water, including level indication and radiation monitors, are important for early indication to provide personnel time to take compensatory SL

s f

achons. Further, proper load handling (,e..'u4 nre needed to ensure the safety of the fuel. The working group beheves that defense in depth shmid be maintained for spent fuel pools at decommissioning plants to protect from accidents that could lead to a radiological release. For l an operahng reactor, defense in depth is considered to be maintained through fission product bamers. These barners include the clad, the reactor pressure boundry, and containment. In a spent fuel pool, the only fission product ba@@ifains is the fuel clad., which is similar to the water-basin ISFSI.

4.3.3 Accdent Scenarios ts. The accu $ents ange from those that ISFSis are designed to withstand a specty ly consM$ered "possible', but could have are expected to occur frequently to those the potential maximum impact on the environment. The latter of the accedents include events such as criticality, total loss of water, and a dropped cask. The loss of water accident analysis has only been evaluated for the effects from shine. Part 50 of Title 10 of the CFR provides regulations that dictate design criteria and accidents for spent fuel pools. The working group believes that a loss of water accident with $M@aTror self-sustained zirconium oxida valid consideration for severe accidents at spent fuel pools.

4.3.4 Systems and Controls The ISFSI design emphasizes the importance of controlling radioactivity. The following SSCs and control programs that are identifieH$ in the guidance as important for the operation of a water-basin ISFSI and may be considered important for spent fuel pools at decommissioning plants.

• SFP water cooling and cleaning system o SFP water makeup system j

e SFP waterlevelinstrumentation e Radiation monitoring e Heating ventilation and air conditioning system (HVAC) e On-site and off-site electrical power, instrumentation and controls, and communications

• Water chemistry program o SF cask handling program

• . SF handling program

• . Radioactive waste treatment 4.4 Design Basis Events Design criteria for the storage of spent nuclear fuel requires that fuel storage and handling l

systems be designed to assure adequate safety under normal and postulated accident conditions. In' addition, these systems are to be designed with appropriate containment, confinement, and filtering systems, and be designed to prevent significant reduction in the l

coolant inventory under accident conditions. Design guidance for fuel storage facilities includes l

preventing the loss of SFP coolant from the pool that would result in the fuel becoming uncovered, protecting the fuel from mechanical damage, and providing the capebility for limiting potential offsite exposures in the event of a significant release of radioactivity from the fuel.

54 l

u__ _ .

i-Unless mingative measures are taken, loss of water from the SFP during the initial decay penod of about 3 to 5 years could cause overheating of the spent fuel and result in damage to the fuel I claddmg integrity and could result in release of radioachve materials to the envronment. Events considered in the spent fuel storage facil ' n with the potential to cause significant inventory

-loss from the SFP include: FT

=

Earthquakes Loss of coolmg Siphoning Dropping of heavy loads-3g

+

Other extemal events that could cw sig wii.= E tie inventory of the pool Fuel storage facility features designed to prevent significant inventory loss from the SFP include a seismically qualified SFP structure, redur. dant, and in some cases, seismically qualified SFP i

cooling systems (some with safety related y r ips), seismically qualified and safety-related SFP coolant makeup systems, anti:all t action, local and remote level indication, local and remote radiation alarms, pool structure designs that plan for the effects of a dropped spent fuel cask without significant leakage from the fuel storage area.

Even a small amount of mechanical damage to the spent fuel stored might cause an offsite radiological release if no dose reduction features, such as iodine removal, are provided.

Mechanical damage resulting in radiological releases from the spent fuel can occur from the following events:

Dropping fuel assemblies during fuel handling operations

• Dropping objects onto the stored fuel

=

lmpacting the stored fuel with intemally or extemally generated missiles Fuel storage facility features designed to prevent mechanical damage to the stored fuel include a pool structure design that prevents missiles from impacting the stored fuel, and physical design features and administrative controls are used that minimize the possibility of damaging the stored fuel by dropping a fuel assembly or dropping a heavy object onto the stored fuel.

For SFPs not located within the reactor containment, without adequate protective features, radioactive material could be released to the environs as a result of either loss of water from the SFP or mechanical damage to the fuel within the pool. Should radioactive material be released from the stored fuel, design requirements to provide a controlled leakage building surrounding the SFP with the capability to limit releases of radioactive material address this concem.

55

5.0 References References for the Executive Summary and Section 1.0

1. Benjamin, A.S., McCloskey, D.J., NTand Dupree, S.A, " Spent Fuel Heatup Following Loss of Water During Storage," NUREG/CR 0649 (SAND 77-1371),

March 1979.

2. Sailor, et. al., " Severe Acculents in Spent Fuel Pools in Support of Generic Safety issue 82", NUREG/CR-4982 (BNL-NUREG-52093), July 1987. -

3. U.S. Nuclear Regulatory Commissig.$tgp depry Analysis for the Resolutico of Generic Safety issue 82, 'Beyond Design B sti s Mc6 dents in Spent Fuel Pools," NUREG-1353 April 1989.

4. U.S. Nuclear Regulatory Commission, " Residual Decay Energy for Light-Water Reactors for Long-Term Cooling," Branch Technical Position ASB 9-2, Rev.2, July 1981 from the

' Standard Review Plan for the Revi f a fat LAnalysis Reports for Nuclear Power.

5. American Nuclear Society, Draft AP  % 5, " Decay Energy Release Rates Following Shutdown of Uranium-Fueled Thermal Reactors," October 1973.

6. U.S. Code of Federa/ Regulations, " Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Section 65, Part 50, Title 10, " Energy."

7. U.S. Code of Federal Regulations, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," Appendix B, Part 50, Title 10, " Energy."

8. U.S. Code of Federal Regulations, " Conditions of Licenses," Section 54, Part 50, Title 10,

" Energy."

9. U.S. Code of Federal Regulations, " Licensing Requirements for the independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste," Part 72, Title 10, " Energy."

References for Section 2.0

1. U.S. Code of Federal Regulations, " Changes, Tests and Experiments," Section59, Part 50, Title 10, " Energy.'

2. Benjamin, A.S., McCloskey, D.J., Powers, D.A., and Dupree, S.A, " Spent Fuel Hestup Following Loss of Water During Storage," NUREG/CR-0649 (SAND 77-1371),

March 1979.

3. Sailor, et. al., " Severe Accidents in Spent Fuel Pools in Support of Generic Safety issue 82", NUREG/CR-4982 (BNL-NUREG-52093), July 1987.

4. U.S. Code of Federal Regulations, " Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Section 65, Part 50, Title 10,

• Energy."

5. U.S. Code of Federal Regulations,

• Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," Appendix B, Part 50, Title 10, " Energy."

6. U.S. Code of Federal Regulations, " Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste," Part 72, Title 10, " Energy."

References for Section 2.1

1. Benjamin, A.S., McCloskey, D.J., Powers, D.A., and Dupree, S.A. " Spent Fuel Heatup i

Following Loss of Water During Storage," NUREG/CR-0649 (SAND 77-1371),

March 1979.

56

2. Benjamin, A.S.,

• Dynamic Modeling of Physical Phenomena for Probabilistm Assessment of Spent Fuel Accidents," Sandia National Laboratories, SAM 097-2870C, November 1997.

3. Mellor, A.M., " Heterogenous ignition of Metals: Model and Experiment", Aerospace and Mechanical Sciences Report No. 8@R4[3 Tant NsG-641, Department of Aerospace and Mechanical Sciences, Princeton University, Princeton, New Jersey, October 1967.

4. Tapscoctt, R.E., Fire Protection Hancibook,18" Edition, Section 4, Chapter 16, " Metals,"

National Fire. Protection Association, Quincy, MA, February 1997.

5. Cooper,

• Review of Zirconium-Zircaloy Pyrophoricity," RHO-RE-ST-31P, Rockwell Intemational, Rockwell Hanford op

6. Abbud-Madrid, A., Branch, M.C., . .9c.F;grg,ftpland, WA," November

, T.J., and Daily, J.W., Ignition of Bulk1984.

Metals By A Contmuous Radiation Source in A Pure Oxygen Atmosphere," Flammab!!ity and Sensitivity of Materials in Oxygen-Enriched Atmospheres: 6* Volume, ASTM STP 1197, Philadelphia,1993.

7. Levitz, N.M., Kullen, B.J., and Steir(10 .glanagement of Waste Cladding Hulls. Part I. Fyrophoncity and Compaction *, Mt , Argonne National Laboratory, February 1975.

8. Markstein, G.H., " Combustion of Metals," Technical Report CAL-86-P, Comell Aeronautical Laboratory, November 1962.

9. Glassman, I., Combustion, Third Edition, Academic Press, New York, NY,1996.

10. Department of Interior, "The Explosivity of Titanium, Zirconium, Thorium, Uranium and Their Hydrides," NYO-1562, U.S. Atomic Energy Commission and Department of Interior, Osirridge, TN, June 1951.

11. U.S. Department of Energy, " Primer on Spontaneous Heating and Pyrophoncety," DOE HDBK-1081-94, Washington, DC,1994.

12. National Fire Protection Association, NFPA 482, " Standard for the Produchon, Processing, Handling and Storage of Zirconium," Quincy, MA,1996. '

13. Bulmer, G.H. " Recommendations on the Safe Handling or Zirconium Metal and Zirconium Alloys," United Kingdom Atomic Energy Authority, Risley, Warrington, Lancashire,1969.

14. Tapscoctt, R.E., Fire Protection Handbook,17* Edition, National Fire Protection Association, Quincy, MA.

15. , Schulz, Wallace, 'ARH-2351 Shear-Leach Processing of N-Reactor Fuel-- Cladding Fires," Separations Chemistry Laboratory, Research and Development, Chemical Processing Division, Atlantic Richfield Hanford Company, Richland, WA, February 15,1972.

References for Section 2.2

1. Benjamin, A.S., McCloskey, D.J., Powers, O.A., and Dupree, S.A,

• Spent Fuel Heatup Following Loss of Water During Storage," NUREG/CR-0649 (SAND 77-1371),

March 1979.

2. Nourbakhsh, et. al., " Analysis of Spent Fuel Heatup Following Loss of Water in a Spent Fuel Pool", NUREG/CR-6441 (BNL-NUREG-52494), May 1998.

3. Smith, C. W., " Calculated Fuel Perforation Temperatures, Commercial Power Reactor Fuels", NEDO-10093, September 1969.

4. Sailor, et. al., " Severe Accidents in Spent Fuel Pools in Support of Generic Safety lesue 82", NUREG/CR-4982 (BNL-NUREG 52093), July 1987.

57

4 1

5.

"MELCOR 1.8.1 Computer Code Manual," Volume 2- Reference Manuals and Pmgrammers' Guides, June 1991.

6.

Peano, et. al., "The Potential for Propagation of a Self-Sustaining Zirconium Oxidsbon Following Loss of Waterin a Spent Fuel Pool" Sandia National Laboratory draft report prepared for the U.S. Nuclear Reg @Nission, January 1984 7.

Travis, R.J., et.al., "A Safety and Regulatory Assessment of Genenc BWR and PWR Permanently Shutdown Nuclear Power Plants,' NUREG/CR-6451 (BNL-NUREG-524 August 1997.

8.

Hermann, et.al., " Technical Support for a Proposed Decay Heat Guide Using SAS2H/ORIGEN-S Data", NUREG 8 eptember 1994.

)

U.S. Nuclear Regulatory Commissi ,

pe xlium of ECCS Research for Realistic J

9. I LOCA Analysis", NUREG-1230, December 1988.

10. Cooper, T. D.,

• Review of Zirconium-Zarcaloy Pyie,,t.e,Uty," RHO-RE-ST-31P, Ro intomsbonal, November 1984.

iy phoricity and Recommendaticas for

11. Kullen, et. al.,'" An Assessment of 7 em a onal Laboratory, November 1977.

Handling West Hulls", ANL-77-63, s References for Section 3.3 1.

U.S. Nuclear Regulatory Commission, " Residual Decay Energy for Light-Water Reactors for Lon9-Term Cooling,' Branch Technical Position ASB 9-2, Rev.2, July 1981 from the

" Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power

2. Sailor, et. a1., " Severe Accidents in Spent Fuel Pools in Support of G2neric Safety

.lssue 82", NUREG/CR-4982 (BNL-NUREG-52093), July 1987.

3. " Code Manual for MACCS2," NUREG/CR-6613, May 1998.

4. U.S. Nuclear Regulatory Commission,

• Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants," NUREG-1150, December 1990.

References for Section 3.3.2 1.

Sailor, et. al., " Severe Accidents in Spent Fuel Pools in Support of Generic Safety issue 82", NUREG/CR-4982 (BNL-NUREG-52093), July 1987.

2. Travis, R.J., et.al., "A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Nuclear Power Plants," NUREG/CR-6451 (BNL-NUREG-52498)',

August 1997.

3. " Code Manual for MACCS2," NUREG/CR-6613, May 1998.

4. " Calculations of Reactor Accident Consequences Version 2, CRAC2: Computer Code,"

NUREG/CR-2326, February 1983.

- 5.

. Charles E. Ader, U.S. Nuclear Regulatory Commission, memorandum to Robert C. Jones, U.S. Nuclear Regulatory Commission, "Offsite Consequence Evaluation of Steam

! Generator Tube Leaks and induced Steam Generator Tube Ruptures," July 24,1996.

6. Charles E. Ader, U.S. Nuclear Regulatory Commission, memorandum to Robert C. Jones, U.S. Nuclear Regulatory Commission,"Offsite Consequence Evaluation of Steam Generator Tube Leaks, Steam Generator Tube Ruptures, and Induced Steam Generator Tube Ruptures," August 30,1996.

58 1

i References for Section 4.3

1. U.S. Code of federa/ Regulations, " Licensing Requirements for the independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste," Part 72, Title 10,

• Energy."

2. U.S. Nuclear Regulatory Commissi@,filh%tkeTdent Spent Fuel Storage Installation (Water-Basin Type)," Regulatory Guide 3.49, December 1981.

3. American National Standard Institute / American Nuclear Society standard (ANSl/ANS) 57.7-1981, " Design Criteria for an independent Spent Fuel Storage Installation (Water Pool Type)," issued on February 19,1981.

4. U.S. Code of federa/ Regulations, ggpnsing of Production and Utilization Facilities," Part 50, Title 10, "Energ: .

5. U.S. Code of Federal Regulations, "Backfitting ' Section 109, Part 50, Title 10, " Energy."

6. U.S. Nuclear Regulatory Commission, " Operating Experience Feedback Report -

Assessment of Spent Fuel Cooling," NUREG-1275, Volume 12, February 1997.

7. U.S. Nuclear Regulatory Commissi' , U :.Q;0800, Standard Review Plan forthe Reviewof Safety Analysis Reports rar Power Plants, Light Water Reactor Edition, Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup System," Rev.1, July 1981.

8. EDO, U.S. Nuclear Regulatory Commission, memorandum to the Commission,

" Resolution of SF Storage Pool Action Plan issues," July 26,1996.

9. U.S. Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nudear Power Plants, Light Water Reactor Edition, Section 15.7.4, " Radiological Consequences of Fuel Handling Accidents," July 1981. i 10, U.S. Nuclear Regulatory Commission, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Light Water Reactor Edition, Section 15.7.5,

" Spent Fuel Cask Drop Accidents," NUREG-0800, Rev. 2, July 1981.

11. U.S. Nuclear Regulatory Commission, " Control of Heavy Loads at Nuclear Power Plants, '

Resolution of Generic Technical Activity A-36," NUREG-0612, July 1980. l

12. U.S. Nuclear Regulatory Commission, " Spent Fuel Storage Facility Design Basis (for Comment), Regulatory Guide 1.13, August 1975.

13. Samuel Collins, Office of Nuclear Reactor Regulation Director, letter Mr. Michael Meisner, President Maine Yankee Atomic Power Company, dated November 6,1998.

l 59

6.0 Acronyms ANSI American Natmal Standard institute ANS American Nuclear Society ASB NRC Auxiliary Systems Branch @@Tms Branch)  ;

' atm s'.mosphere BNL Brookhaven National Laboratory -

j i

BTP Branch Technical Position BWR Boiling Water Reactor DRAFT l CFD TComputational Fluid Dynamics CFM Cubic Feet per Minute CFR Code of Federal Regulations DOE Department of Energy DRAFT DSP Decommissioning Status Plant ECCS , Emergency Core Cooling System EPRI Electric Power Research Institute ET- Event Tree .

FAA Federal Aviation Administration FFU Frequency of Fuel Uncovery FT Fault Tree -

gpm Gallon (s) per Minute GSI Generic Safety issue GWD Gigawatt-Day HCLPF_ High Confidence / Low Probability of Failure HVAC Heating, Ventilation, and Air Conditioning INEL. Idaho National Engineering and Environmental Laboratory ISFSI Independent Spent Fuel Pool Installation kW Kilowatt LLNL- Lawrence Livermore Nationat Laboratory MR Maintenance Rule MW: Megawatt - .

MWD Megawatt-Day MTU Megaton Uranium NEl National Energy Institute NFPA National Fire Protection Association

'NRC- Nuclear Regulatory Commission 60

NRR NRC Office of Nuclear Reactor Regulabon .

PRA Probabilistic Risk Assessment PWR Pressurized Water Reactor QA Quality Assurance DRAFT RES NRC Office of Research

• RG Regulatory Guide SF Spent Fuel -

SFP Spent Fuel Pool SFPCC Spent Fuel Pool Cooling and Cleaning System SNL Sandia National Laboratory SRP Standard Review Plan j SSC Systems, Structures, and ComknMFT (

SSE Safe Shutdown Earthquake TS Technical Specification l

UKAEA United Kingdom Atomic Energy Authority 1 1

f-l I

':\

I l 61 1

t .

3 -

Appendix 1 Assumptions Made in the Spent Fuel Pool Probabilistic Risk Assessment Case One and Two:

Case One and Two are based on the as-found spent fuel pool and spent fuel pool cooling and support systems during the four site visitsM)eplE{g group member.

1. The spent fuel pool has one level monitor, one temperature monitor, and radiabon monitors and alarms in the control room The instrumentation is maintained in an operable state. Annunciators are in computers, which can track trends.

2.

JRAFT The certified fuel handlers are forn Br senior reactor operators who know the facility, the surrounding community, and facility maintenance personnel well.

3. The site has operable two fire pumps (one diesel-driven and one electrically driven from offsite power)

DRAFT

4. There is limited makeup capability (with respect to volumetric flow) with the exception of fire pumps, which can provide makeup via 2.5 inch diameter hoses.

5. The certified fuel handlers enter the spent fuel pool area once or twice per shift (8 to 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts).

6. The site is staffed during the day shift Monday through Thursday with maintenance people, health physics staff, QA/QC staff, fuel handlers, and administrative staff. Nights and weekends there is a skeleton staff.

7. It is known to the site personnel (through guidance and / or training) that backup sources to replenish the spent fuel pool inventory can be fro the fire protection pumps or offsite sources, such as from fire engines.

8. The spent fuel pool water was clear and the fuel was observable. The control room monitors the spent fuel pool level via a camera that can zoom in on a measuring stick in the peol that can alert operators to level changes. The measuring stick is about three to four feet long.

9. There is little fire protection equipment in the spent fuel pool area and it all is manual.

10. Overhead cranes have stops to help prevent heavy loads from being moved over the spent fuel pool.

A1-1 i .. .. .

Case Three:

Case Three is based on spent fuel pool and spent fuel pool cooling and support systems that are slightly better than the minimum required by NRC regulations.

1. The spent fuel pool has one level M %rJperature indicator in the control room.

They are not required to be maintained in an operable state. The radiation monitors and associated alarms in the control room are only required to be operable when fuel is being moved. Annunciators are analog, are not trended, and need not work. In the PRA model, the control room / spent fuel pool instrumentatum is assumed unavailable 10% of the time (i.e., the fraction of time dg he component isin a failed state is 1 x 104)for Case 3, which is 20 tinits h r unavailability than assumed in Case 1 (i.e., 5 x 10).

2. The certified fuel handlers are not former utility employees who know the facility, but are individuals who meet the minimug;gyirements for a certified fuel handler.

3. There is limited makeup capability (with respect to volumetric flow). The fire pumps have been removed from the site.

4. The certified fuel handlers rarely enter the spent fuel pool area. It was assumed in the PRA model that the HEP is five times higher than Case 1 (i.e., the Case 3 HEP is 0.05.)

5. The site is staffed during the day shift Monday through Thursday with a minimal number of maintenance people, health physics staff, QA/QC staff, fuel handlers, and administrative staff. Nights and weekends there is a skeleton staff. Recovery probabilities are assumed to be lower for Case 3 than Case 1.

6. Overhead cranes had stops to help prevent heavy loads from being moved over the spent fuel pool. Cranes are operated by non-nuclear trained operators.

7. The spent fuel pool water clarity is not maintained. Control room cannot monitor the spent fuel poollevel directly.

8. There is no requirement that fire protection equipment for the spent fuel pool area be maintained operable. Therefore, no credit is taken for the fire pump's ability to provide inventory makeup or fire suppression.

A1-2

1 1

Assumphons independent of Case One Two, or Three:

l 1. An F4 to F5 tomado would be required if significant damage were to be possible to a PWR or BWR spent fuel pool.

2. An F2 to F5 tomado would be req @A foTsible significant damage to a spent fuel pool support system.

3. Shipping cask handling is the dominant heavy load operation.

4. Spent fuel casks will be the only yieved over the spent fuel poolwith sufficient mass to significantly dam ge pool.

5. Crane operators will follow safe load path procedures when moving heavy loads near the spent fuel pool. l

6. Spent fuel pools are robust and wil s smic events less than three times the safe shutdown earthquake (SSE). t

)

7. The staff used generic loss of offsite frequencies, j

8. Pumps or valves are manually operated and aligned. Every action must be  !

accomplished by a certified fuel handler. l I

g. The effects of criticality were not considered in the risk evaluation. Its potential for I impact on risk is considered to be very low.

10. One year after the last of the reactor fuel is transferred to the spent fuel pool there is no longer any day-to-day NRC onsite oversight. We did not attempt to quantify the effect of this assumption.

11. Extemal flooding is assumed not to be a significant contributor to loss of spent fuel pool l cooling. l

12. The utility has removed the emergency diesel generators and other support systems  !

such as residual heat removal and service water that could provide spent fuel pool 1 cooling or makeup prior to the plant being decommissioned. l l

13. The highest action level of offsite waming for decommissioned plants is an alert.

l

14. The spent fuel pool cooling system (sled mounted) and any support systems all run off i the same electrical bus.

15. The utility has procedures for small leaks from the spent fuel pool or for loss of spent fuel I pool cooling system. J I

I A1-3  !

i i

16. Tne only significant technical specdication applicable to spent fuel pools is the requirement for radiation monitors to be operable when fuel is being moved.

17. There are no large pipes that go into the spent fuel pool that either extend mote than a few feet down into the pool or that penetrate the spent fuel pool at a level near the top of the fust DRAFT

18. The evacuation is assumed to begin three hours prior to the release of fission products to the environment.

19. Ninety-five percent of the populatigggpyniles of the de is assumed to evacuate at a speed of four miles per hour.

D. RAFT A1-4 l

i

n.._..

.. 1 DRAFT Appendix 2 Decommissioned Plant Spent Fuel Pool Event Trees and Fault Trees DRAFT DRAFT A2-1

b i

I NN Wutl Ell 85 m 8eeBIE9m guessump p i

Sale W Spaans Cashg

~ n,- aangi.ty

e. ef - m unsg

.a e.n.ees

. .f D.

8 ftmas SW eymess gues Seuses emusmes 8 """"* DesaurTOR e resgatency 9444s sts em cEd CsF ces ops

1. 1 EQE OK S03 EQECSF OK 1 tee 403 DCEE40.

- 503 EQECSFOC8 On 3 00E.002 .. ,

pgN 30d EQEcsFocsoFe e esE*ese

(

fge*' sm Eaces on i

1 1

l 100E*000 -, . . - .

< i l

kghgg 501 EQEC&iOF9 9 50E407 1

l, I

gN - -..-- - -- 801 EQEFm 100E.005 I

C WWresvPetapp.stof E VT 88 07 26 Ob2849 We**eUP8tA 10 h*P RtG8' AI DEGQeAndt&&M MT6 Quemmen Dee Ob7045 M 08 07 TOTAL CMP *

• enE406 586SWIC NTIATING EVENT CASE 1

e 6 P ,

ruar - M N - g O

~ gung m MW SIERIE punam m 8EE4 Marome M. s

! andaM 0E80MPfDR FEE 0meCY

.Ca.mus. o= - -

e.s.a.u.sr i.ta.n .

94488 ocs CF.

gm pas a CW 301 EQE OK 3 01 EQECar OK 4 ocem UGEL4pr, 301 EQECSFocs OK e

, M.1 DFDC40C

. = = . co c.,oc.o,. . - -

a sane. =

> = =

IE EGE I

1

, I ljNffY J s

gg; - .oac or. . - -

I 1

1

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srestr

-- i . y , , Ar ac w . c o o r.

C - ..nC. En .1. u na.

Quemeramen Dele Ob7bte M 14 31 TOTALCadF*19% 409 seisaalC NTaATNeG tVENT CASE 2 I

I l

- - - _ -__D

V e

4 rea m muisems - -igr mesmum, a p

$4Enl lugns taupity af Symmes Cassig Utse Omens g g casuseriten Casang Pummme usag er cases O s 814uss M essaamm Cuman amuseme emmes 8 ""N 8 N Seets M pm Ces 0E7 OCS OPB 901 EQE OK

- 9 01 EQECSF OK f 20E 0DI UGEL Lur.,

501 EQECSFoca OK S o0E402 0FDC. LOC 1,=a . EQEC ,0C.0,. .

i i

1 4

g 800 EQECSI OK bNiTY

hj 500 EQECssoF8 e 50E407

)

i l

8,g9, ,[ 501 E0EFm t coEcos l

C wvmMVPRA489 340E EVT OS oC 22 DW2899 WN HVPRA i 0 6FP Ri&A AT DECOMWi& Sam wT6 Qwumemen Date Ob4&M 07 59 to total cur e 1 ttE.cus SE1EMIC NTIATING EVEWT CASE 8

i 4

FWW M FWI - -. W utse O e e me a een OEud Ilummary ORAs g S Am W en WW SM4 emme ensus, s =- DEscusaron e rumousscv wBus Pus Pisuus cast OPD Pet car 901 FF OM 90E433 unem

$02 FIR 06P Ost S 00E402 DEP#m 301 FinosPouK ost 19CE40%

out.pcFp ~ ~~

504 FiROSPoun0FD 4S18407 kg g SF# Rtan AT Otcoww:ssioedQ P@NTS C WWinNUPRAGPP t#LA tVT 10 3618 DW1999 WIN NUD8tA 10 Qvenhttsten Dnie 08/199910 as 44 TOTAL CWF e $ $$4407 INTERNAL FlRE INITIATING EVENT CASE 1

== -=== , = 1 .

w .m  ;

.m, , . --===m. y .

FWt CEP OISL OPD l

l Ml M @

l

! l Em 303 FIROSP OK 1

I 1

I DEP m I

e 1

$03 FIROSPOedK Qet I 1

I s oot.coi I ow.ocry h#,gg Son FIROSPoefEOFD t DeEas 8 ~ ^t;; %" M i.in," a r::,," N" 2 "

• • R- at a=== wi.

WTERNAL FIRE NeITIATING EVENT t

M$

I A

W g p NM N D ,

W RE h M ANEW $

D EB O S tage ORAs NongeQE N e P4GLegy

' Ass W en SPP Pisumin emuass #

ghese flammer M pts remus put er ORK 08D

$01 FR OK e80E405 um 803 FIROSP OK 9 00E401 DSP-Fwt 8C3 FIROSPOMK OK l

I.

T M6~'

UNITT i

w r,Ros m a a . m.a agem, SFD Rt&E AT DECOMMi&&sONED PLNsTS C wwuevPsLAsop.3'im EVT 11 03 33 GhnOst WN NvMtA 1 g QuenuncmenDate N1040119302 TOT AL CMP o 9 00E400 INTERNAL Fist $ NTIATING EVENT CASE 3 i

'- .-_ w

b n

W5 ==== Umur p useum N4 a p 4 h h sf Amisery W team h Game h { g asums Lams W Csung spuun lang fte o e Casag pumus e aseecE 0macsurttut e Pummacy LOC N ee OCS CPD ops 301 LOC on g set OK ,

100E404 l ucat4ac j 30Q LMMM ON 2 00E407 OFDEbJc m,C*= ;ggr . loc = .o, - . .-4,,

801 LOCCAA OK

$0s LOCCAAOCS OK 2 20t40a CCSLlos

(( h 501 LOCCRAOctoFD OK 3 00E402 OFDL LOL

=Jg1 .0. LxC--. . . 0.

l g'g"t - ----. Sol LOCCRAINO t 50E407 l

C WWeweUPRA4FP ttoC f VT 10 37 29 cavtome m aavPRA10 67P R4A AT DECOMedi&&schED WTS Querwecaten Due OM049 10 m at foiAL CMF e i sot 407 LOSS OF COOLueG CAST i

4 6

p M M M W Sv5 p g o MW Md Rumuey W MN O m Omsme s 4 h une rue o e-= m amms Las cause asumm e secescE DEacesPTom o e,,,a

, w m M M M a W 1 1

. Loc oK l

l 1

i 1

. LM @

5 soEm ut as400, 801 LOCOCSOFD OK 900Em DFD0 Loc 88 8 14 ages scu LocoCsorooFe soote R' EC.CES T IELDC aol LOCCRA OK SOE LOCCRACCS OK t 10fm2 OC5C10c sol LOCCRAOC$0F0 OK

$ 01f 23 CN AEAMu 170Em1 0FDL10c 808 LOCCRAOCSOFDOFS 21N 000 h# ,

, .m, gg - _ _ _ . . . . . . . .

__. toCC 67P as AT DECOMM44 ONEC PLANTS C weaNupstasFP7&oc EW is on 33 0$MG49 WNWUP8tA 10 a Charedwen Ose cM0se to M e4 TOTAL CWF

• 187t@T LDSS OF COOUNG CASE 2

4

'.A l

- - ., ., - 0 I ,

O 1 -

- O . u. .

C O  % O. =- . =

tac CNA .O OCS OPD WB l

M1W OK ga s terre og l

4 3EE402 1RlE4w sos LOCOC80F0 OK .

_t 00E6 WN1IT i

k g $0d LOCOCSOFOOFB 113E405 I

4 SOE LOCCRA OK

$0E LOCCRAOCS OK 120E401 OCSL LOC

{,N[, 501 LOCCRAOCSOFD OK t 00E*000 DNi1Y hE 500 LOCCRAOCSOFDOFB 3 SOE407 0 . - . . . - -

501 LOCCRAINO 158408 C eNnNUPRAMP 3&OC EVT 9105 34 Dett099 tMS**vPRA 10 Oummerman Dme feiege 1I 0152 TOT AL CWF e t 73840g SFP Ri&A AT DECOMutb680PeEO PLAMi$

LOSS OF COOLeaG CASE 3

4 a

m som p seen omar upaus umums ussuer summe a a

uma er nasumumme tuumme teams esmano Omas smuus g annummas af a=ues Lams 0 cmmend lemmes admammy sammu ensmagny bdumme uns, m g tuumsg Cas. met.y sus

.m Asmus um. .M.m.me .mm GAgh.ee

u. . w e elmosus usame W esame swigs ces at cask OPD CnA esD La ett

$01 tot OK

$02 Losols OK t 10E@1 DILE

$N 630t$onen OK 2 coE 002 CENC4 k= & . to,C.-

- A =

SOE LOCftAcus OK 1 toEm1 Vi$ L Sol LOICftAOtSOWK OK hhr . -.2 DifK E See LOCAAOtSOMKOFD 2 79E.8

[E F 100E.002 it-Lp

$0E LOICRAIND 3 01f 40,

((g'

$tt LoeNLL OK

$11 LONupL OK 102E401 VileV

$12 L@NLLOLOe4K OK 120t402 LC

$t2 LONuotLDMKOFD 141E.00s

.h40Egg

$14 LONLLCRA OK

$11 LONuCRAOIL OK t 02t@1 OlWU y 3 g-~

$14 LONuCRAOILouK OK 120E-002 OWKL3 kgg4d-

"I# $11 LONuCRAOILOHKOFD 1 08(400

$,,L.NaC . , .,,,,,

=gg_.... - . . _ _ _ _ . _ . . . _ . _ . .

C eMassuPnAspp.ssos eVT et4e at 06ste iMmesuPRA10 See Aisn At OtccAsuis&OesD PLANTS Ovantdseien Dele 06/2499 064731 10f AL CMF e 3 seg40s LOSS OF COOLANT INVENTOsty CASEt 3

W 0 .. t 0

. W O. - - - .

Les stL Cuta .c Ost OL Oam Opp 301 LOi OK M LOIOUS OK 100 Eat

=

8an LOosonsk OK

. .m.

m

[**'O tu LQosOasKWD S I.E4 6

- Laen a Sol LOCAAO.$ OK 2 coEm1 m

. 0,.m.

gnar $01 LOpm OK t1 U 0E401 R-4

[,yy* Sol LOC 8tAOSOMMOFD 2 73E407 1cof402 WLGi e

1 out.001 gggg Sol L@CAAfND 3018407 3 $1C LO8su OK

$11 LONLLDL OK 102f m1 Carr -

$1J LONLLOLOWK OK 120E402 OWK401-

[' [g#"

y $13 LQNLLOLOMMOFD 3 sa40s

$14 L@NuC8tA OK

$11 LONLLCftACIL OK 102 Eat OiWu 5 098 003 CR.ALAAF $10 LONLLC8tAOLOWN OK 120(402 OWK101 -

hh9- $17 LONaCRAOLOMKOFD 141Em?

gg,2--- .- 8,no,Naena,ND ,,,,,,

c - - - ... 6" R'6" af ENS &8W %76 Quewkenen Date .mo.

Ob2.o.

4 90 4 3. TOT A6 CMP e 6 006405 LO&$ OF COOLANT .NENTORY CASE 3

E

.I.88V Afd b.IIES 388 W .EE. PD. ==== - -

i t

W C I F

.. f.

A.NI.

th. -,-

. CPD CftA se CIS 2L Oist Los an.L 3 01 LOI OK Sol Lolois OK 2 00E@t DER toi LosossonsK oK

.. 0 Up. I T

ry a w~ ~,an ,m Sol LosCRA OK SOE LosCRAct$ OK 4 00E401 m

ou

{,"f'g- se LosCRAoisonex

. . o0 DsH Y hh#h sod LOCRAN'NO a n4es IEloi sol LosCRAmo 1JeE406 Std LOINLL OK l

$11 LONLLQlt OK 110EIO1 DE8U sta LONLLoeLoesk oK 100E*000 nm ag,5 .t i wNuo,Lo Ko,o ..

rygt

- SH LotNLLCAA OK

$11 L@NLLCRAOIL OK 11 01.0EJ at

'ge $10 LONLLCRAOtLO.K OK

1. 0o0 DNrn l ing- so Lo uCRAo.Lo aro ..0 l M6 mh+ M=6M e6 M # eum SFP Albn AT L4CQhted1&$tQpeED PWT4 C WWeNUPRA4FP-3 tor EVT 00004. Ob7bSS WI488UP Qesenesseen Date Ob2b90 06 De to TOTA 6 CadF e 1296-.#.A.10 LOSS OF Cock. ANT INVENTQfty CASE S

I

• a

' C l

k 6aus ur- umms sus - remer m p massary uume m nun g y Pumer Puuss Pte Puur ftsusumy spunun causs ensuus g D Plum Osuusuf Sunausles Puer a RMame ammog tunnem 0 e me tus Antund Puellaussury ed Amn 8 NN 8 N Seusse _ _ .

ws 50 crut ocs otAL orD W LP1 OK SEG LPtocs OK 4 4er40s M UF 301 LP1Mer*M oK 2 00f402 m

fg" Sod LPtmen*MOFD t 44 Emf l

sei LPioP. oK t 006 403 REC 08Pi

_ 80s LPtoPRoe#K OK 1 toe 401 M DGFP i

g ar ;gg w ,,op.o .. i l

301 LP10G OK l

sot LP10GCCS OK 4 44E 003 OC$Elor j

SfC LP1000CsouK OK t 30E40,  ? 00E402

% 5 TART On#4EM (go0Ee $11 LP1DooCSOMEOFD 3 6st 007

$12 LP10GOPR OK 1 noE-003 REMSP1

$---- $11 LPIDGOPROFD t t0E407 C WenNUPRA4FP.iLPi f VT Os 02 26 0428m NNUPRA 19 SFP 844K AT DENWONED PWT&

ownw Osw obtem ce 0119 TOTAL cup

• 127E-oos Loss op orrsm powEn FROM PLANT CEWTERED EVENTS CASE 1

r 6

W N 884 s P

• m er - ummeimi m resur EinaBG nessumy taas Ombas ammms E O Pts pump amassey annum O 8 PumurPumm lh>emt samm>es Syseme 8 Pass commes emme sud fham Pas to 8 ed eenAuges Piss tauseumy aufIher Gumes cah OPD 90 cret OCs upt J

$01 LPt OK

$C3 LP1CCS OK 1 ME403 ksas.4pr e

OK .

303 LPtoC30asK t 00E402 URIKfr5

1. $04 LP10C80eMOFD 2.3tE@7 30s LP10PR OK f 00Ee3 REOCSP1 sol LP10PRoesK OK 2 60E&

DMK'DGFP 6 00Em2 501 LP10PetoseKOFO 9 00E*000 ecoE402 RECDEE m Lri see LPtDG OK 501 LP1DGOCS OK 7 taE403 OCSE. Lor

$1C LP1Mv'N OK l

.. 1 2 a. N0E@

DCe$TARF"~

$11 LP1"*^=ND 2 mE4os jgogEm?

$13 LP1000PR OK q

{

i cores REC 4&PF e

$13 LP1DGO*AOFD 1 OsE406

[E

._ , . ,,....Tm_,,,,,_,,

C -, x., m . on QueeArmen Dec 06/28/89 08 S2 el TOTA 6 CMF s 4166405 LOSS OF OFF5 RTE power FRoad PLANT CENTERED EVENTS I

CASE 2

(. .

?

l 4 l

l l

I 4

un a usmuse - er umme pumus summe esseur numme,ysum , ,

Pamur Pam PtsPump nuesumy sussen stenmuery unge cese emuses, g ,

Stum Ceuuud thus auf Shun Peer a RMas4 teenMy Spuumm 0 s aus elsitsused PuelW aus shun 8 """ CESCIEP7088 8 NM Sune LP1 00 crut cct cast opp N1 LP1 OK 90l Lt10CS OK 1 seE402 OCEElur esi LPlocacesK OK 2 00E402 Unun#rs gpg , w-e .. 0 l

l 8 01 LP10 Pet OK 100E4DS RECouri see LPioPRouK OK 190E401 DWKM a =f g,=g . utop lo u-o .. 0 SOE LP10G OK 501 LP1DQQCS OK 1 set 4c2 OCSE40r SIC LP1DGOC80teK OK 1 006 000 1 006*000 DW UknN

$11 LPIOGOCS0WKOFD 7 57E405 kE STJ LP1DGOPR OK 100E403 REC 0SPs

$13 LP1DGOPROF0 a 00E 00s kE O 3

C WenNUPRASPP 3(P18 W OS 1110 0S7699 m IO W R& AT EpsstM MTs Guanuhr44sm Dme Ob2490 0810 23 7CTA CW8 e 7 07E406 t063 0F OFF3;TE power FROM PLANT CENTERED EVENTS CASE 3

_._ - - = m - E

=

P y, ~~ g _ _ -  : -

=- ,74 ,

om oc. om cPo t,a ao 3 01 LP2 OK ECG LP20C8 OK 4 44E403 eew SEM LP20C30tfK OK t 00E402 ERNC m 504 LP20C30am4>D IJssans h00E@2

- mop. oK 2 00t402 RECMP3 SOE LP20PROMK OK 1 toe 401 Nr g,=g w momo-o,o . 0 000

,=>

SOE LP2DOOCS OK 4 44Eac3 OCSE-LOP -

$1C LP20GocsoMK ok

? 00E 402 190E 40t OM4IPFP D4GTART

$ 23E40s

$11 LP2DG0C50MMOFD kEM it2 moco. .

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o Appendix 3 Assumptions for Spent Fuel Heatup Estimates The working group calculated the time to heat a single fuel rod from 30 *C to 900 *C without considering any heat losses at various times following permanent shutdown. This is a very conservative, non-realistic type of calculagsome cooling through conduction, convention or radiation would occur and increase the time to heatup.

For the calculations, the working group used a decay heat per assembly and divided it equally among the pins. Design values for a 9X9 assembly for the PWRs and a 17x17 assembly for the BWRs were used as representative asserf ~ . spy heats were computed 'using an

' extrapolation of the decay power tables ire CR-5625. The decay heat in NUREG/CR-5625 is based on ORIGEN computer code calculations. The tables for the decay heat extend to bumups of 50 GWDMTU for PWRs and 45 GWDMTU for BWRs. It is recognized that the decay heat is only valid for values up to the maximum values in the tables, .

but the functional dependence of the deca power with respect to bumup for values in the table indicate that extrapolation may provide a e rMTstimate of the decay heat for bumup values beyond the limits of the tables. The BWR decay heat was calculated using a specific power of 26.178 MWMTU. The PWR decay heat was calculated using a specific power of 37.482 MWMTU. Both the PWR and BWR decay heats were calculated for a bumup of 60 GWDMTU and include an uncertainty factor of 6 percent.

Assumptions: PWR BWR bumup/ assembly (MWD /MTU) 60000 60000 MTU/ assembly 0.47 0.17 assemblies / core 193 764 rods / assembly 264 72 pellet diameter (m) 0.0081916 0.00906 inside clad diameter (m) 0.0083276 0.00925 outside clad diameter (m) 0.0094996 0.01077 fuellength (m) 3.65 3.8 rod peaking factor 1.166 1.166 axialpeaking factor 1.2 1.2 fraction theoretical UO2 0.95 0.95 Years Decay Power per assembly 1 15.5561572 12.54475868 in kW/MTU 1.5 11.4163371055 9.33601650748 2 8.85047 7.345154884 2.5 7.252431291 6.12127S58952 3 6.02689457463 5.21793956972 3.5 5.11496410863 4.60129503226 4 4.341018 4.05752418 4.5 3.68417781163 3.57801496228 A3-1 l

f 4 .

Decay power in watts for one assembly:(kW/MTU)*MTU*1000 Years Decay PowerT.47*E3 Decay PoM.17*E3 1n pr(W) Decay Power (W) 1.5 LJ 11.393884 2132.6089756 2 5365.678439585 1587.1228062716 2.5 4159.7209 1248.67633028 3 3408.64270677 1040.6170202184 3.5 2832.6404500761 887.0497268524 4 782.2201554842 g.gSSF310561 00LO.27846 689.7791106 4.5 Calculate adiabatic heatup time from 30 C to 900 C in hours for one fuel rod: Heatup time (h rs)

Years ggg(h rs) 1.5 6.32 7.7 2 8.15 9.79 2.5 9.94 11.75 3 11.96 13.78 3.5 14.1 15.64  !

16.62 17,72 {

4 20.11 I 4.5 19.57 i I

I 1

1 A3-2 i

L, I

I Appendix 4 Structural Integrity of Spent Fuel Pool Structures Subject to Tomadoes and High Winds Tomado or high winds damage, resulting from missile neration, can affect the structuralintegrity of the spent fuel pool or affect the availability oM systems, such as power supplies, cooling pumps, heat exchangers and water makeup sources, and may also affect recovery actions. A set of site specific evaluations for tomados and high winds was documented in NUREG/CR-5042,

' Evaluation of Extemal Hazards to Nuclear Power Plants in the United States," Lawrence Livermore National Laboratory, December 1987. It is noted that this study was performed to assess core damage frequencies at operating plants. In pRf944, "Tomado Damage Risk Assessment,"

Brookhaven National Laboratory, Septembei 9 , a methodology and assessment of tomado risk was developed.

The National Climatic Data Center (NCDC) in Asheville, N.C., keeps weather records for the U.S. for the period 1950 to 1995 (Ref: http://www.ncc4 a d. These data are reported as the annual average number of (all) tomadoes per 10,00 its per state, and the annual average number of strong-violent (F2 to F5) tomadoes per square mile per state, as shown in Figures A4-1 and A4-2.

Figure A4-1 Figure A4-2 Annual Awrage Nurrber dTamadcas per Awray Annual Nurrber d SiraigMaler:t (F2-F5) 14000 Squere Mies by State.19541995 Ter7 adoes per 10,000 %gsere Mlas by State

^

03 gi 04 !25 j 08 , 01 04 j 03 34 gj M N 82 2 0

7 55 47 $4 35

2 3428 00 30 04 ~

09 {19

't IS 0 I, 07 07 13 03 .*

18 el 33 02 8' l3 D 58 38 n 09 8

. 93 g7 f p 29 e

00 01 1.7 8'

13 to 09

.e . . (p 40 3d .*

• w tp ggM *I 01N '

' e Oc ,

c0 A comparison of the site-specific evaluations (from NUREG/CR-5042) and general regional values from the NCDC database is presented in Table A4-1. The NCDC data was reviewed and a range of frequencies per square mile per year was developed based on the site location and neighboring state (regional) data. In general, the comparison of the NUREG/CR 5042 tomado frequencies for all tornadoes to the NCDC tomado frequencies for.all reported tornadoes shows good agreement between the two sets of data.

A4-1

d Table A4 Tomado and High Wind Data Summary NCDC data NUREG/CFgp .. ,, ,

Frequency Frequency High wind Tornado 1950 1995 1950 1995 Tomado Tomado damage damage average for average for hoquencv strike F2-F5 (por frequency frequency FO F5 (per (per mi - frequency (per year) 8 mi-year) mi8-year) year) (per year) (per year)

Site bM1tAF 41.0xie i.2-2.2xi0 0.247xiod indian Pt. 2 1.00x10d i.90xic

<1.0x10 3 1.2-2.2x10d 0.2 0.7x10d indian Pt. 3 1.00x10d 1.00x10d 1.80x104 d 2.2-3.4x10d 0.7-1.3x10d I.amenck 1-2 1.13x10d 2.30x10d 9.00x104 <1.0x10

( <F1 )

Minstone 3 1.87x10d 1.87xte GRAF Tt0xie 2.5-3.4xic u-i.ixie Low <1.0x10* 2.8-3.4x10d 0.7-0.9x10d Ckonee 3 2.50x10d 3.50x178 1 mi rad.

<3.89x10 4 2.06x10* 1.8-3.8x10d 0.4-1.1x10" Seabrook 1-2 1.26x178 7.75x104 LOSP & RWST

<1.0x10 4 3.4-5.4x10d 1.2-2.0x10d Zion % 1.00x1&* 1.00x104 N.A.

Regional w/o recovery of onsite power GSI A 45 PRAs Local 5.69x104 2.53x10d 3.7-7.5x10d 1.7-2.4x10d ANO 1 5.18x10d 1.53x108 4.37x10d 1.00xigs 5.00x104 3.4-4.7x10d 1.21.5x10 4 Point Beach 12 6.98x10d 5.38x10d 4.11x10d d

<<1.0x104 5.08x10' 3.4-5.4x10 1.2-2.0x10d Quad Cities 1-2 5.18x10d 1.04x10-8 5.44x10d

<<1.0x10 4 1.61x10* 8.4x10d 1.2x10d . ,

St. Lucie 1 6.98x10d. 1.70x10d l 1.20x108 3.30x108 2.54x10* 8.4x104 1.2x10' Turkey Pt. 3 3.37x10d 1.70x10d 5.83x10 8

)

The Storm Prediction Center (SPC) raw data, for the period 1950 to 1995, has been used to develop a data base for this assessment. There have been about 121 F5, and 924 F4, It tomadoes recorded between 1950 and 1995 (an additional four in the 1996 to 1998 period).

is estimated that about 30 percent of all reported tornadoes are in the F2 to F3 range and about 2.5 percent are in the F4 to F5 range.

A4-2

r l

. 4 DOE-STD-1020-94, " Natural Phenomena Hazards Design and Evaluation Cntens for Department of Energy Facilities," January 19g6, Department of Energy, provides some insights  !

into wind generated missiles.

l +

For sites where tomadoes are notg  ;

debris, a 2x4 inch timber plank ing weighg1J lbs is considered as a missile fo) winds and hurricanes. With a recommended impact speed of 50 mph at a maximum height of 30 ft above ground, this missile would break annealed glass, perforate sheet metal siding and wood siding up to to 3/4-inch thick. For weak tomadoes, the timber missile horizontal speed is 100 mp 4o a height of 100 ft above ground and a i vertical speed of 70 mph. A secor is considered: a 3 inch diameter steel pipe weighing 75 lbs with an impact velocity of 50 mph, effective to a height of 75 ft above '

ground and a vertical velocity of 35 mph. For the straight wind rnissile, an 8 inch CMU (concrete masonry unit) wall, single wythe (single layer) brick wall with stud wall, or a 4 inch concrete (reinforced) is considered adequate to prevent penetration. For the tomado missile, an 8 to 12 inch CIMMn"Sie wythe brick wall with stud wall and metal ties, or a 4 to 8 inch concrete (reinforced) slab is considered adequate to prevent ,

penetration (depending on the missile). (Refer to DOE-STD-1020-94 for additional details.)

For sites where tomadoes are considered a viable threat, to account for objects or debris, the same 2x4 inch timber is considered but for heights above ground to 50 ft.

The tomado missiles are (1) the 15 lbs, 2x4 inch timber with a horizontal speed of 150 mph effective up to 200 ft above ground, and a vertical speed of 100 mph; (2) the 3 inch diameter,75 lbs steel pipe with a horizontal speed of 75 mph and a vertical sped of 50 mph effective up to 100 ft above ground; and (3) a 3,000 lbs automobile with ground speed up to 25 mph. For the straight wind missile, an 8 inch CMU wall, single wythe brick wall with stud wall, or a 4 inch concrete (reinforced) is considered adequate to prevent penetration. For the tomado missile, an 8 inch CMU reinforced wall, or a 4 to 10 inch concrete (reinforced) slab is considered adequate to prevent penetration (depending on the missile). (Refer to DOE-STD-1020-94 for additional details.)

The winds associated with hunricanes and other storms are generally less intense and lower in magnitude than those associated with tomadoes. Generally, high winds from wind storms and hurricanes are considered to be the controlling wind level at a higher frequency but at a lower magnitude.

i A4-3

E

~

Recommended values for risk-informed assessmerit of spent fuel pool The tomado strike probabilities for each F-Scale interval was determined from the SPC raw data on a state basis. For each F-Scele, the probability was obtained frem the following equabon, for the point strike pmbability: . DRAFT I

p, , '.In < a >[ x Equation i s A,. .s Yu "h'*

' P. = strike probability for F-Scale (ff;R AFT

<a>r = tomado area, mi8

- A = area of observation, m n(state land area)

Y, = interval over which observations were made, years L = sum of reported tomados in t g pservation The tomado area, <a>r, was evaluated at the mid-point of the path-length and path-wxith intervals shown in Table A4-2.

Table A4-2 Tomado characteristics Damage Path-length Length Path-width Width F-Scale -

and wind speed scale (miles) scale (yards) i 0 < 1.0 0 < 18 yds.

0 Light Damage l (40-72 mph)

Moderate Damage 1 1.0 - 3.1 1 18 - 55 yds.

1 (73-112 mph) 2 3.2 - 9.9 2 56 - 175 yds.

2 Significant Damage (113-157 mph)

Severe Damage 3 10.0 -31.9 3 176 - 527 yds.

3 (158-206 mph) 4 32 - 99.9 4 528 - 1759 yds.

4. Devastating -

Damage (207-260 mph)

Incredible Damage . 5 100 > 5 1760 yds >

5 (261-318 mph) -

The tomado area, <a>1, was then corrected using data from NUREG/CR-2944 (Table 6b, page 19 and Table 7b, page 21) to correct the area calculation for the variations in the tomado intensity along the path length and width.

[

i A4-4

1 Ca I

}

l The central U.S. region (12 states - Kansas, Oklahoma, Iowa, Missouri, Arkansas, lihnois, Indiana, Ohio, Kentucky, Tennessee, Mississippi and Alabama), was used to develop the tomado strike probabilibes for each F-Scale interval. This region has the highest frequency of tomado occurrences and is about equal to the central region defined in NUREG/CR-2944 (region A in Figure 2, page 12.) . DRAFT -

The SPC raw data for each state was used to determine the F-Scale, path-length and path-width characteristics of the reported tomadoes. Equation 1 was used for each state and the summation of the 12 states performed. The results are shown in Table A4-3. 1 h f"% A r"*T

!/I M i i Table A4-3 Point Strike Probabilities F-scale . Tomado strike y

0 1.2x10-5 1 3.5x10-5 2 4.0x10-5

.3. 2.5x104 4 7.7x104 5 6.3x104 Significant pool damage An F4 to F5 tomado would be needed to consider damage to the spent fuel pool from a tomado missile. The likelihood of a tomado of this magnitude even striking the plant is estimated to be 7.7x104 per year, or lower. In addition, the spent fuel pool is a multiple-foot thick reinforced concrete structure housed within another (normally) foot thick reinforced concrete structure (i.e.,

the building housing a spent fuel pool normally is designed to mitigate an F4 or F5 tomado.)

Based on the DOE-STD-1020-94 information, it is very unlikely that a tomado missile would tr able to penetrate the spent fuel pool. The working group concludes that tomado missile strikes j

' from F4 and F5 tomadoes add negligibly to the risk from spent fuel pools at decommissioned plants.

l Support system availability An F2 or larger tomado would be needed to consider damage to a support system, such as

- power supplies, cooling pumps, heat exchangers and water makeup sources. The likelihood of j this magnitude tornado striking the plant is estimated to be 4x10-5 per year, or lower.

A4-5 i

4 Recovery from loss of support systems if the spent fuel pool support system (s) are damaged by an F2 or larger tomado, then offsit recovery is considemd to be available (e.g., fire trucks) to provide makeup to the spent fue The assumed conditional probability of urggppsite recovery is 0.02 for Cases 1 and 3 and 0.1 for Case 2. Combining the frequency of an F2 or larger tomado with failureper of offsste recovery results in a frequency of fuel uncovery in the spent fuel pool in'the range of 8x year from damage to support systems caused by tomadoes for Cases 1 and 3 and in the r of 4x104 per year for Case 2.

The SPC data is summarized in Table A4Q R A FT Table A44 SPC Tomado Summary by State (1950 - 1995)

Frachon in F-range Fujita dama0* scale F3 F4 F5 F4-F5 F2-F5 Total F0 F1 F2 State 0.49 323 129 36 14 0.049 AL 1031 165 364 149 31 0 0.031 0.51 1007 198 298 331 AR 0.08 11 2 0 0 0.000 AZ 160 90 57 2 0 0 0.000 0.10 223 142 58 21 CA 0.001 0.10 441 99 15 1 0 CO 1172 616 5 2 0 0.031 0.42 65 9 29 20 CT- 0.00 0 0 0 0 0.000 DC 1 1 0 11 1 0 0 0.000 0.22 DE $5 20 23 30 4 0 0.002 0.15 2148- 1156 665 293 FL 0.34 266 65 17 0 0.016 GA 1032 147 537 119 74 9 0.052 0.39 1607 478 506 421-IA 0 0 0 0.000 0.06 124 63 53 8 ID 113 39 3 0.031 0.35 1342 431 440 316 IL 108 77 8 0.082 0.44 1038 246 336 263 IN 404 168 54 16 0.030 0.27 KS 2363 1111 610 65 35 3 0.079 0.49 KY 483 79 168 ,133 268 123 16 2 0.014 0.33 LA 1254 225 620 31 8 3 0 0.022 0.30 MA 138 24 72 26 5 0 0 0.000 0.18 MD 172 49 92 17 0 0 0 0.000 0.21 ME 82 21 44 210 57 30 7 0.046 0.38 MI 807 195 308 158 53 28 6 0.036 0.26 MN- 953 372 336 334 109 48 1 0.036 0.36 MO 1367 298 577 369 136 -59 10 0.054 0.45 MS 1268 226 468 42 33 4 0 0 0.000 .0.15 MT 253 174 143 44 26 0 0.038 0.31 NC 687 153 321 A4-6

Fujita damage scale Frachon in F-range State Total F0 F1 F2 F3 F4 F5 F4-F5 F2-F5 ND 830 490 211 91 28 7 3 0.012 0.16 NE 1818 827 585 255 105 42 4 0.025 0.22 NV 49 41 '8 0 0 0

{

0 0.C00 0.00 )

NH 75 24 34 15 2 0 0 0.000 0.23 NJ 127 42 58 23 4 0 0 0.000 0.21 NM 400 261 104 31 4 0 0 0.000 0.09 NY 268 101 106 35 21 5 0 0.019 n.23 OH 733 157 321 166 53 27 9 0.049 0.35 OK 2580 845 808 626 209 83 9 0.036 0.36 OR 49 31 15 3 0 0 0 0.000 0.06 PA 506 93 220 143 26 22 2 0.047 0.38 RI 8 3 4 1 0 0 0 0.000 0.13 SC 516 136 234 100- 31 15 0 0.029 0.28 SD 1172 651 259 197 57 7 1 0.007 0.22 I TN 596 107 241 139 76 29 4 0.055 0.42 TX 5934 2632 1837 1067 317 76 5 0.014 0.25 1 UT 79 53 19 6 1 0 0 0.000 0.09 VA 318 84 132 68 28 6 0 0.019 0.32 j VT 33 7 14 12 0 0 0 0.000 0.36 >

WA 56 24 17 12 3 0 0 0.000 0.27 WI 949 204 378 276 62 24 5 0.031 0.39 WV 87 27 36 16 8 0 0 0.000 0.28 WY 444 247 145 43 8 1 0 0.002 0.12 Total 38459 13776 13251 7834 2553 924 121 0.027 0.30 A4-7

F l o

i i

Appendix 5 Structural integrity of Spent Fuel Pool Structures Subject to Heavy Load Drops A heavy load drop onto the spent fuel pool wall or into the spent fuel can affect the structural integrity of the spent fuel pool. Previously, the staff evaluated heavy loads as Generic. Technical Activity A-36, which resulted in the publication of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants," U.S. Nuclear Regulatory Commission, July 1980. The working group assumed that shipping cask handling is ti c avy load operation. .

In NUREG-0612, the staff evaluated Navy data for cranes and hoists for the period February 1974 to October 1977. There were 74 load drop events identified. Of these,23 percent were attributed to design and maintenance errors, 73 percent to operator (or human) error and 4 percent to rigging errors.

The working group also r'eviewedmore several,D R A conceming recent stud.tes FT heavy load drops:

1. Notice No.125 - March 27,1984, U.S. Department of the interior, Minerals Management Service (MMS), Gulf of Mexico OCS Region The Risk and Safety Analysis UnitbN[ reviewed 50 crane-related accidents for the period January 1,1971 to June 30,1983. The major findings included the following:

• The major contributing cause of crane accidents has been employee negligence and/or error.

• About 44 percent of the crane accidents involved some type of equipment failure due to poor maintenance and/or overloading of the crane.

2. " Findings and Recommendation of the Crane Accident Workgroup" B. Hauser, B. Lewis, W. Rhome, Engineering & Operations Division, October 16,1998 (U.S. Department of the interior, MMS)

The workgroup reviewed 34 incidents for the period 1995 to 1998.

(1) 17 incidents were listed as equipment failure (EF) and 12 listed as human error (HE).

(2) Fatalities were reported in 1 of the 12 HE incidents and in 5 of 17 EF incidents.

(3) Major property damage was reported in 1 of 12 HE incideats and in 6 of 17 EF incidents.

3. " Independent Oversight Special Study of Hoisting and Rigging incidents Within the Department of Energy," October 1996, Office of Oversite, Office of Environment, Safety and Health, U.S. Department of Energy A5-1

4 The Department of Energy (DOE) Occurrence Reporting and Processing System (ORPS) served as the principal information source for incidents relating to hoisting and rigging (H&R) operations. An initial set of 491 occurrence reports, corresponding to the j October 1,1993, to March 31,1996, period, describing incidents related to H&R. There l

were 131 relevant hoisting and rigging incidents between October 1993 and March 1996.

The data is summarized in Table A5-1 and Table A5 2.

Table AS Distribution of hbMbIgging incidents and accidents Accidents as a Accios.7ts as a Equipment Number of Number of incklents as a Ir @ nts AWants Percent of Total

• Percent of Total

• Percent ofincidents*

St* 74 %

Crane 66 49 50$ DRAFT 31 % 38 % 90%

Forklift 40 36 19% 11% 44 %

Other" 25 11 100% 73 %

Total 131 96 100 % p a g--

• Rounded to the nearest whole number.~ ' "

" Includes manual and power-operated. hoists, chainfalls, and block and tackle.

Table A5 Root cause of hoisting an'd rigging incidents by equipment type

• Crane Forklift Other Rcot Cause 20% 23 % 8%

Inattention to Detail 18% 3% 27 %

Work Organization and Planning Procedure Not Used or Used incorrectly - 9% 15% 0%

9% 10% 4%

Policy Not Adequately Defined, Disseminated, or Enforced 5% 5% 19%

inadequate or Defective Design Defective or inadequate Procedure 9% 5% 0%

9% 0% 4%

Inadequate Administrative Control 5% 5% 8%

Defective or Failed Part 3% 3% 12%

Other Management Problem 3% 3% 0%

Other Human Error inadequate Work Environment 0% 10% 0%

- A5-2

e Root Cause Crane Fortdift Other Lack of Procedure 2% 3% 4%

Insufficent Refresher Training 3% 3% 0%

insufficient Practice or Hands-On Experience 5% 0% 0%

Communication Problem ng3c 4% 3% 4%

VI \ ni inadequate Supervision 0% 3% 4%

Error in Equipment or Material Selection 0% 3% 4%

Weather __,_

,0% 3% 0%

No Training Provided W# 0% 0% 4%

• Rounded to the nearest whole number.

Since crane and hoisting operations are govt Administration (OSHA) requirements, accde, id g.S. Occupational enerally Safety available. However, and there is Health only limited data available on the number of operations, or lifts, for developing load drop per operation data.

In NUREG-0612, the staff evaluated the Navy data for the period February 1974 to October 1977. Based on some assumptions regarding the number of cranes and hoist in service, the number of lifts per year, and the 74 load drops identifed in the period studied, it was estimated that frequency of a heavy load drop is in the range of 2.5x104 to 3.0x10d per operation.

In NUREG/CR-4982, " Severe Accidents in Spent Fuel Pools in Support of Generic Safety issue 82," Brookhaven National Laboratory, July 1987, the staff estimated the human error contributor to a heavy load drop as 6.0x10" per lift, based on data conceming human reliability in the positioning of heavy objects.

A more recent study on heavy load drops was performed for the DOE Savannah River Site to quantify human errors," Savannah River Site Human Error Data Base Development for Nonreactor Nuclear Facilities," Westinghouse Savannah River Co., WSRC-TR-93-581, February 28.1994. The likelihood of a heavy load drop, when using cranes or hoists, was quantified using a generic data base. An operation was defined as " lift, move, and setting down" the load. Based on 200 drops in 2,000 crane years of operation and additional crane data from nuclear power plants, the drop rate was estimated as 1.5x10d per operating-hour. No error factor information was provided in the study. ,

if it is assumed that it takes 1-hour for an operation (lift, move, and setting down) and using an error factor of 10 from the nominal value (1.5x10d), then the estimated values for a human error resulting in a heavy load drop are provided in Table AS-3.

A5-3

Table A5-3 Department of Energy Recommended Heavy Load Drop Frequencies l l

Frequency Error Factor Type of Load 1.5x10-5 / operation 10 Standard l Low mean value 1.5x10-4 / operMAFT 10 Typica Nominal mean value 1.5x10-3 / operation 10 Unusual High mean value -

j bat drop frequency is a factor of 2 to 4 lower.

While the DOE nominal mean value for th rf than the values reported in previous evaksti) , he supporting data base is believed to be representative of the expected cask handling operations in a spent fuel pool, or movement of heavy loads near the spent fuel pool. The DOE recommends that the high value be used for ,

unusual, unevenly balanced loads with standard construction or industrial equipment and that )

the low value be applied to standardized I potter present.

3 The number of cask operations per year and the time span during which a drop may impact the spent fuel pool wall or fall into the spent fuel pool need to be considered as part of a risk- -

informed evaluation. The numerical values presented in Table AS-1 have an implied time I

duration of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per operation. l l

For heavy loads not involving cask operations, it is expected that a " safe load-path' procedure will be followed if movement of objects over the spent fuel pool, other than casks, is to be considered, then a description of the assumptions used for the risk-informed evaluation needs to be documented, including items such as weight, number of movements, potential spent fuel pool failure modes (e.g., crush or fail the wall, or puncture the bottom of the pool), and the drop rate l

frequency assigned to the particular movement.{GOES INTO RECOMMENDATION SECTION]

in reviewing the drop per lift data, it was determined that the ranges of values presented in the 1994 DOE study were nearly identical to those used in the 1980 NUREG-0612 study.

Therefore, the working group used NUREG-0612 as the bases for developing estimates of heavy load drop frequencies.

Non-sinole failure oroof load handlina system ,

Figure B-2 (pg. B-16) in NUREG-0612 provides a quantified fault tree for load handling over a

- spent fuel pool for a load handle system that is not single failure proof. The staff estimated the handling system failure rate to be 1.0x10d to 1.5x10-8 per lift. This value was based on a nuclear power plant estimate expected to be in the 1.0x105 to 1.5x10" per lift range, including a factor-of-2 improvement over the Navy data evaluated (reported to be in the 2.5x10.s to 3.0x10d per lift range), and included an assumption that failure of the interlocks and/or failure to follow an approved load path lead to a factor-of-10 reduction in the load handling system reliability (the staff estimated the interlock / load path failure range to be 2.0x10~8 to 1.0x104 per year.). The per A5-4

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lift range from the DOE Savannah River study,1.5x104 to 1.5x10d for this type of load handling, is nearly the same as the expected NUREG-0612 range.

Only some fraction of the load drops will lead to significant spent fuel pool damage. in NUREG-0612 the staff assumed that the only 10 percent of the critical load path was over the spent fuel. The working group estimated that between 2 to10 drops would occur for 200 lifts per year resulting in an estimate of 2.0x104 to 1.0x104 drops per year into the spent fuel pool.

Therefore the working group estimates thgkqij;lr apt b be in the range 2.0x107 to 1.5x10d per year (applying the NUREG-0612 methods # Edds Dm ptions). If the NUREG-0612 assumption on the load handling system reliability is not considered por example, the load ' drop is not related to failure of the interlock or the load path) then it was estimated that the load drop is in the range 2.0x104 to 1.5x104 per year.

Sinole-failure-oroof load handlina system MMT Figure B-3 (pg. B-17) of NUREG-0612 provides a quantified fault tree for load handling over spent fuel pool for a single-failure-proof handling system. The staff estimated the failure of the handling system to be in the range of 4.0x10-7 to 1.0x10" per year. The staff estimated the likelihood of the drop occurring over spend t tg0x10-2 to 2.5x104 per event (using an estimate of between 5 percent and 25 percon he load path - based on the 10 percent estimate of the total path). The resulting range is 2.0x104 to 2.5x104 per year, a reduction from '

the non single failure proof load handling system range of 2.0x10-7 to 1.5x10" per year. The single failure proof handling system reduced a load drop by a factor of 10.

Calculated values for risk-informed assessment of spent fuel pool l The working group estimates the likelihood of a load drop into the spent fuel pool to be in the

- range of 2.0x104 to 1.5x10" per year for a non-single failure proof load handling system. For a single failure proof load handling system, the range is reduced by about a factor of 10 (2.0x104 '

to 2.5x104 per year). An estimate of the likelihood of substantial damage (rapid pool draining) given the drop is on the pool wall or into the pool is provided.

For the failure of the pool wall, it was assumed that the load physically travels over the wall 2 percent (0.02) of the time (10 percent of the 5 to 25 percent of the totalload path) and it was assumed that a one-in-ten (0.1) chance of significant damage given load drop. The working group estimated a failure rate in the 2.0x104 to 3.0x104 per year for the non-single failure proof system and 8.0x10* to 2.0x104 per year for the single failure proof system. (The NUREG/CR-4982 value for wall failure was 3.7x104per yeer.)

For failure of the pool floor, it was assumed one-in-ten events (0.1) will result in significant damage. The working group estimated a failure rate range of 2.0x104 to 1.5x104 per year for the non-single failure proof system and 2.0x104 to 2.5x104 per year for the single failure proof system.

A5-5 1

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insights to be considered In assigning a drop rate to a particular operation, the following information should be considere It appears that the human error contribution to heavy load drops has not changed since the original work documented in NUREG-0612 in 1980. The following conclusions were reported the 1996 DOE, Office of Oversite, study:

1.

Human error is the major cause oMknIrigging (H&R) incidents.

Human error, whether directly associated with supervisors or equipment operators, is the principal cause of H&R incidents. Factors not related to human performance, such as equipment failure and weather, are responsible for only 6 percent of H&R incdents-management (35 percent) and peg (33 percent) collectively account for 68 percent of all H&R incidents, as report into RPS.

2. Management shortcomings and workers' inattention to detail account for a large proportion ofincidents.

Further analysis shows that deficib Nka ning (43 percent) and inadequate definition, dissemination, and enforcement of policy (24 percent) are responsible for two thirds of the incidents attributable to management deficiencies. Inattention to detail (56 percent) and not following procedures (28 percent) account for 84 percent of H&R incidents caused by personnel error. Furthermore, inattention to detail is the most prevalent cause of all 131 H&R incidents, accounting for about one in every five

~ incidents. Additionally, there are no indications that certain root causes are becoming less frequent over time, are being remedied, or are being replaced with other causal factors.

i j

3. Work planning is a significant factor in non-forklift incidents.

. Work organization and planning require more attention in operations involving cranes and "Other" hoisting equipment than when forklifts are utilized. This is evident by the fact that inadequate work planning was the cause of 18 percent of allincidents involving cranes,2.7 percent of the incidents involving "Other" hoisting (i.e., non-forklift) equipment, and only 3 percent of all forklift incidents.

4. Training related deficiencies were not identified as a major problem.

Training-related deficiencies were not identified as a significant problem.

Procedure-related problems, including applying procedures incorrectly, defective or inadequate procedures, or procedures not used, are responsible for 18 percent of crane ,

incidents. They were not found as causal factors for incidents involving "Other' )

j equipment. Communication, lack of procedures, and defective or failed parts cause s

incidents with approximately equal frequency for all equipment type categories, although I it is the greatest for "Other" equipment (e.g., hoists, chainfalls, block and tackle).

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5. The use of mobile cranes by subcontractors is expected to increase, heightenmg the need for effective oversight of subcontractors' safety performance.

Descussions with H&R experts within the DOE (Federal workers, contractors, and subcontractors) indicate that as prM@n-related operations are curtailed and superseded with activities directed at waste management, environmental restoraten, and facility dismantlemeret, the need for stationary or overhead cisnes will be reduced, and mobile units will be in more cranes owned and operated by subcontractors are often used to rial handling tasks of varying complexity, whereas overhead cranes are generally operated by contractors and are used to perform maneuvers that are relatively simple and often routine. Independent evaluatons performed by the Office of Oversight, in addition to information reported into ORFS and the Department's Computerized Accident / Incident Reporting System (CAIRS), have highlighted deficiencies in oversigggaetor activities. Therefore, the additional risks posed as more H&R tasks involving cranes are performed by subcontractor personnel heightens the concem ever H&R safety and the need for effective oversight of subcontractor performance. Information contained in ORPS does not explicitly and formally identify whether an H&R incident is associated with a contractor or subcontractor activity. While it wa! --* M Sis review to make this determination for some of the 131 incidents analyzeOM nod possible to resolve this issue for the entire sample.

The following standards describe the guidelines currently in use based on requirements of the U.S. Occupational Safety and Health Administration (OSHA) and the American National Standards Institute (ANSI):

DOE-STD-1090-99, " Hoisting and Rigging (Formerly Hoistirig and Rigging Manual),"

U.S. Depar' ment of Energy, March 1999.

. MIL-HDBK-1038, " Weight Handling Equipment," Department of De'ense Handbook, March 06,1998.

- NAVFAC P-307 U.S. Navy, June 1998.

The DOE standard occasionally goes beyond the minimum general industry standards totablished by OSHA and ANSl; and also delineates the me:e stringent requirements necessary to accomplish the extremely complex, diversified, critical, and oftentimes hazardous hoisting and rigging work found within the DOE complex, in doing so, it addresses the following items which are not covered in detail in the general industry standards:

t Management responsibility and accountability

2. Operator /inspwtor training and qualification requirements

3. Definition of critical lifts and the additional requirements for making them

4. The need and responsibilities of a persn-in-charge for critical lifts AS-7 l

f

5. The need and responsibilities of a designated leader for ordinary lifts I

6. ~ The definition and special requirements for pre engineered production lifts l l

7. Special requirements for the testing, inspection, and maintenance of hoisting equipment j

in hostile environments

8. Nondestructive testing /nondestruMMnItion requirements for such items as hooks, welds, and spreader bars i

g. Special requirements for inspection and load-testing of hoisting and rigging equipment /accessones DRAFT Hook latch requirements for cranes, slings, and rigging accessories

10. i 1

11. Design standards for such equipment as cranes, forklifts, and hooks

12. Operating practices for hoisting ardsreiar ~= rations

.) TR ~1

13. Rigging information and load tables

14. Good and bad rigging practices.

. J A5-8 J

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l l Appendix 6 Aircraft Crashes into Spent Fuel Pools The staff used the generic data provided in DOE-STD-3014-96, "Acculent Analysis for Aircraft Crash into Hazardous Facilities," U.S. Department of Energy (DOE), October 1996, to assess the likelihood of an aircraft crash into or near a Mhi%ed spent fuel pool. Aircraft damage can l affect the structural integrity of the spent fuel pool or affect the availability of neart>y support systems, such as power supplies, heat exchangers and water makeup sources, and may also affect recovery actions.

The staff obtained DOE-STD-3014-96, the and will frequency refer to this as t 1eof e an aircra@ffective aircraft target area model:s F = [ N, P, f,(x, y). A ii Equation 1 i.a where: DRAFT N, = estimated annual number of site-specific aircraft operations (nolyr) l P, = aircraft c ash rates (per takeoff and landing for near-airport phases) and per {

flight for in-flight (non airport) phase of operation '

f,(x, y) = aircraft crash location probability (per square mile)

Ag = site-specific effective area for the facility of interest including skid and fly-in !

effective areas (square miles) i i = (index for a flight phase): i=1,2, and 3 (takeoff, in-flight, landing)

J = (index for aircraft category, or subcategory)  !

k = (index for flight source): there could be multiple runways and non airport operations  :  ;

The site-specific area is further defined as: i eff

• f s Equation 2 where:

A = (WS + R)-(H cotq)+ 2 L W.WS +L W f R A s= (WS + R) S I

where:

A, = total effective target area H = height of facility A, = effective fly-in area L = length of facility A, = effective skid area W = width of facility WS = wing span S = aircraft skid distance cote = mean of cotangent of aircraft R = length of facility diagonal impact angle Alternatively, a point target area model can be defined as just the area (length times width) of the facility in question, in Table A6-1 the staff has summarized the generic aircraft data and crash '

frequency values for five aircraft types from Tables B-14 through B-18 of DOE-STD-3014-96.

A6-1 I

Table A6-1 Generic Aircraft Data cote Crashes per mir-yr Notes:

Aircraft Wingspan Skid distance DRAFT Min Ave Max d

1440 10.2 'lx10-7 ~2x10 3x104 l General 50 aviation 4 4x10-7 2x10*

Air carrier 98 60 m rf/ m7x10 60 'Y 8 ' 4x10 7 1x10* 8x10*

Air taxi 58 780 7.4 6x10* 2x10-7 7x10-7 takeoff 1.arge 223 military 4x104 ex10* landing Small 100 447 DRAFT 4x104 military

.The working group used the data presented in Table A6-1 to determine the frequency of aircraft hits per year for various building sizes (length, width, and height) for the average and maximum crash rates. The resulting frequencies are presented in Table A6-2. The product N,*P,*f,(x,y) for Equation 1 is taken from the crashes per mi7-yr and A,is obtained from Equation 2 based on the aircraft characteristics. Two sets of data were generated: one including the wing and skid lengths using the effective aircraft target area model and a second case which considered only the area (length times width) of the site using the point target area model.

The working group chose the building or facility characteristics to cover a range typical of a spent fuel pool to that of the PWR auxiliary building or the BWR secondary containment structure.

4 A6-2

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Table A6 Aircraft Hits per Year Building (L x Wx H)(ft) Enoctive Area (mi) Average Hits / year Maximum Hits / year with the doe enective DRAF T aircraft target area model 100 x 50 x 30 1.0x108 2.1x104 3.1x104 200 y 100 x 30 3.7x104 5.5x104 1.8x14 A r- T 400 x 200 x 30 3.5x1I2 ' 7.0x104 1.0x10d 200 x 100 x 100 2.5x104 5.0x104 7.6x104 400 x 200 x 100 4.7x104 ,,, 9.5x104 1.4x10d l ,

I 80 x 40 x 30 9.0x1M 1.8x104 2.7x104 l

With the point target area model 100 x 50 x 0 1.8x10d 3.7x104 5.4x104 200 x 100 x 0 7.2x10d 1.5x104 2.2x104 l 400 x 200 x 0 2.9x104 5.9x104 8.6x10*

1 i 80 x 40 x 0 1.1x10d 2.4x104 3.5x104 l The working group compared the DOE effective aircraft target area model, using the generic

, data in Table A6-1, to the results of two evaluations reported in "Probabilistic Safety Assessment and Management," A. Mosleh and R.A. Bari(editors), PSAM 4, Volume 3, Proceedings of the 4-th Intemational Conference on Probabilistic Safety Assessment and Management,13-18 September 1998, New York City, USA.

The first evaluation of aircraft crash hits was summarized by C.T. Kimura, et al., in " Aircraft Crash Hit Analysis of the Decontamination and Waste Treatment Facility (DWTF) at the <

Lawrence Livermore National Laboratory (LLNL)." DWTF Building 696 was assessed, it is a 254 feet long by 80 feet wide,1 story,39 feet high structure. The results of Kimura's study are i shown in Table A6-3.

l 2

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a Table A6-3 DWTF Aircraft Crash Hit Frequency (peryear)

Military Aviation TotaP Period Air Carners Air Taxies~ n gifviation 1995 1.72x104 2.47x104 235E105 5.03x103 2.76x108 1993-1995 1'.60x104 2.64x104 2.82x10 4 6.47x104 3.16x104 1.57x10# 2.58x104 n jlpgxg_ 7.23x104 3.23x10 4 1991-1995 1986-1995 1.52x10 4

2.41x104 N.'89x'10" 8.96x10 4 3.23x105 Note (1): Venous penods were studied to assess variations in air Geld operatsons.

4 Applying the DOE generic data to the Dh rNebn a frequency range of 4.4x10 hits per year to 6.6x10-5 hits per year for the effective aircraft target area model. For the point target area model, the range was 1.5x104 to 2.2x104 per year.

The second evaluation was presented in a paper by K. Jamali, et al., " Application of Aircraft Crash Hazard Assessment Methods to Various Facilities in the Nuclear Industry,"in which additional facility evaluations were summarized. For the Seabrook Nuclear Power Station, Jamali's application of the DOE effective aircraft target area model to the Final Safety Analysis 4

Report (FSAR) data resulted in an impact frequency 2.38x10 per year. The Millstone 3 plant area was reported as 9.5x104 square miles and the FSAR aircraft crash frequency was reported .

to be 1.6x10* per year. Jamali applied the DOE effective aircraft target area model to information found in the Millstone 3 FSAR. Jamali reported an impact frequency of 2.74x10* per year using the areas published in the FSAR and 2.31x104 per year using the effective area calculated by the effective aircraft target ares model.

When the working group used the generic DOE data in Table A6-1, it resulted in an estimated impact frequency range of 2x104 to 3x105 per year for the point target area model, and 1.6x104 {

to 2.4x10d per year for the effective aircraft target area model.

The staff documented a site-specific evaluation for Three-Mile Island Units 1 and 2 in f NUREGICR-5042, " Evaluation of Extemal Hazards to Nuclear Power Plants in the United States," Lawrence Livermore National Laboratory, December 1987. The NUREG estimated the aircraft crash frequency to be 2.3x10" accidents per year, about the same value as would be predicted with the DOE data set for the maximum crash rate for a site area of 0.01 square miles, in NUREG/CR 5042, the staff summarized a study of a power plant response to aviation accidents. The results are presented in Table A6-4.

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Table A64 Probability of Penetrabon as a Function of Location and Concrete Thickness DRAFT Thickness of reinfor'ced concrete Plant locahon Aircraft type 1 foot 1.5 feet 2 feet 6 feet s 5 miles Small s 12,000 lbs. m .p g 0 0 0 Large > 12,000 lbs. '0996 O.52 0.28 0

> 5 miles Small s 12,000 lbs. 0.28 0.06 0.01 0 from airport Large : 12,000 lbs. 1.0 1.0 0.83 0.32 UIW I The working group has reasonable assurance that the DOE model and generic data provide a range of aircraft crash hit frequencies that would be consistent with plant-spo ,ific evaluations.

At this level of effort, the resulting damage from an aircraft crash cannot be fully evaluated on a plant-specific bases.

. A detailed structural evaluation is beyond the scope of this effort. In general, PWR spent fuel pools are located on, or below grade, and BWR spent fuel pools, while generally elevated about 100 feet above grade, are located inside a secondary containment structure. The vulnerability of support systems (power supplies, heat exchanges and makeup water supplies) requires a knowledge of the size and location of these systems, information not readily available.

Calculated values for risk-informed assessment of spent fuel pool Significant pool damage PWR The working group calculated a value for significant PWR spent fuel pool damage resulting from a direct hit. It is based on the point target area model for a (100 x 50) foot pool with a conditional probability of 0.3 (large aircraft per.etrating 6-ft of reinforced concrete) that the crash results in significant damage. If 1-of-2 aircraft are large and 1-of-2 crashes result in spent fuel uncovery, then the estimated range is 2.7x10* to 4.0.x104 per year.

BWR The working group calculated a value for significant BWR spent fuel pool damage resulting from a direct hit. It is the same as that for the PWR,2.7x104to 4.0x104 per year. Mark-l and Mark-Il secondary containments do not appear to offer any significant structures to reduce the likelihood of penetration, although on one side there may be.a reduced likelihood due to other structures. Mark-Ill secondary containments may reduce the likelihood of penetration as the spent fuel pool may be considered to be protected by additional structures.

A6-5 iu t

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,o These values for significant damage to spent fuel pools are significantly less than estimates for otherinitiators.

. Loss of support system The working group estimated the value fo@/(Fs}pport system (power supply, heat exchanger or makeup water supply) based on the DOE model including wing and skid area for a (400 x 200 x 30) foot area with d a conditional probability of 0.01 that one of these systems i The estimated value range is 7x10 to 1.0x104 per year.

The working group estimated the value fo Ks3pport system (power supply, heat DOE model including wing and skid area for a exchanger or makeup water supply) base on 4 4 (10 x 10 x 10) foot structure. The estimated 4

value range is 7.3x10 ,to 1.1x10 per year with the wing and skid area, and 7.4x10"* to 1.1 x10 per year for the point model.

These expected values of support systemggg than those for otherinitiators.

A6-6