Severe Accidents in Spent Fuel Poolsin Support of Generic Safety Issue 82

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i

INUREG/CR-4882

BNL-NUREG-52093

AUG 1 7 1987

Severe Accidents in Spent Fuel Pools

in Support of Generic Safety

Issue 82

< ; >

Prepared by V . L. Sailor K. R. Perkins J . R. W ee ks H. R. Connell

Brookhaven Nat ional Laboratory

Prepared for

U.S. Nuclear Regulatory

Commission

DISTRIBUTION QF T iiiS DO CUMENT IS UNLIMIT


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DISCLAIMER

This report was prepared as an account of work sponsored by anagency of the United States Government. Neither the United StatesGovernment nor any agency Thereof, nor any of their employees,makes any warranty, express or implied, or assumes any legalliability or responsibility for the accuracy, completeness, orusefulness of any information, apparatus, product, or processdisclosed, or represents that its use would not infringe privatelyowned rights. Reference herein to any specific commercial product,process, or service by trade name, trademark, manufacturer, orotherwise does not necessarily constitute or imply its endorsement,recommendation, or favoring by the United States Government or anyagency thereof. The views and opinions of authors expressed hereindo not necessarily state or reflect those of the United StatesGovernment or any agency thereof.


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NUREG/CR-4982

BNL-NUREG-52093

Severe Accidents in Spent Fuel Pools

in Support of Generic Safety

Issue 82

NUREG/CR4982

TI87 013442

Manuscript Completed: June 1987

Date Published: July 1987

Prepared by

V. L. Sailor, K. R. Perkins, J. R. Weeks, H. R. Connell

Department of Nuclear Energy

Brookhaven National Laboratory

Upton, New York 11973

Prepared for

Division of Reactor and Plant Systems

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Wa shingto n, DC 20555

NRC FIN A3786

MASTER

DISTRIBUTION OF [HIS liCCUMERT IS UNLIMITED


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ABSTRACT

This investigation provides an assessment of the likelihood and conse

quences of a severe accident in a spent fuel storage pool - the comple te

draining of the

pool .

Potential mechanisms and conditions for failure of the

spent fuel , and the subsequent release of the fission p rod ucts , are identi

fie d. Two older PWR and BWR spent fuel storage pool desig ns are consi dered

based on a preliminary screening study which tried to identify vulner abili

ties . Internal and external events and accid ents are asses sed. Condi tions

which could lead to failure of the spent fuel Zircal oy claddi ng as a result of

cladding rupture or as a result of a self-sustaining oxidation reaction are

presented. Propagation of a cladding fire to older stored fuel assemblies is

eval uate d. Spent fuel pool fission product inventory is estima ted and the

releases and consequences for the various cladding failure scenarios are pro

vided. Possible preventive or mitigative measures are qualitatively evalu

ated. The uncertainties in the risk estimate are la rge , and areas where ad

ditional evaluations are needed to reduce uncertainty are identifi ed.

iii


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TABLE OF CONTENTS

Paj e

ABSTRACT iii

LIST OF TABLES viii

LIST OF FIGURES xi

PREFACE xiii

ACKNOWLEDGEMENTS*.

'.*. *.'.'.'.*..'.*.'. ..*.*. '..1'.'....*.

xv

EXECUTIVE SUMMA RY xvi i

1. INTROD UCTION 1

1.1 Prev ious Inves tiga tion s 1

1.2 Rel ated Even ts 2

1.3 Risk Pot ent ial 3

1.4 Disc uss ion of Spen t Fuel Stor age Pool Desi gns and Feat ure s 3

1.5 More Detaile d Stud ies 4

1.6 Repor t Conten t 4

1.7 Referen ces of Secti on 1 5

2.

ACCIDENT INITIATING EVENTS AND PROB AB ILITY ESTIMA TES 15

2.1 Loss of Wate r Circulat ing Capab ilit y 15

2.2 Stru ctur al Fai lu re of Pool 16

2.2.1 Structural Fa ilur e of Pool Resul ting from Seis mic

Events 16

2.2.1.1

A Review of Seismi c Hazard Data 17

2.2.1.2

Seis mic Hazard Estim ates for Eastern United

State s Sites 20

2.2.1.3

Seis mic Fragili ty of Pool Stru ctur es 20

2.2.1.4

Seismic ally-Induced Failure Probabi lities 22

2.2.1.5

Sensitivity Studies 23

2.2.1.6

Conclus ion s on Seismi c Risk 23

2.2.2 Structura l Failu res of Pool Due to Miss ile s 23

2.3 Partial Drai ndow n of Pool Due to Refu elin g Cavity Seal

Failures 23

2.4 Pool Stru ctur al Fai lur e Due to Heavy Load Drop 25

2.5 Summary of Acciden t Probabi lities 28

2.6 Refer enc es for Sec tio n 2 28

3. EVA LUATION OF FUEL CLAD DING FA ILURE 49

3.1 Sum mar y of SFUEL Resul ts 49

3.1.1 Model Desc rip ti on 49

3.1.2 Clad Fir e Initia tio n Resul ts 50

3.1.3 Clad Fire Pro pag ati on 51

v


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Paje

3.1.3.1 Perfect Ventilation 53

3.1.3.2 Inadequate Ventilation 55

3.2 Validation of the SFUEL Compute r Code 56

3.3 Conclusi ons Regarding SFUEL Analyses 57

3.4 References for Section 3 58

4.

CONSEQUENCE EVALUATION 63

4.1 Radionucli de Inventories 63

4.2 Release Estim ates 63

4.2.1 Estimated Release s for Self-Sust aining Cladding Oxida

tion Cases (Cases 1 and 2) 64

4.2.2 Estimated Release for Low-Temperature Cladding Failure

(Cases 3 and 4) 65

4.3 Off-Site Radiological Conseque nces 65

4.3.1 Scenarios for Conseq uence Calcula tions 65

4.3.2 Conseq uence Result s 66

4.4 Refere nces for Section 4 67

5. RISK PROFILE 75

5.1 Failure Frequency Estimates 75

5.1.1 Spent Fuel Pool Failure Probability 75

5.1.2 Spent Fuel Failure Likelihood 75

5.2 Conclusion s Regarding Risk 76

5.3 Referen ces for Section 5 76

6. CONSIDERATION OF RISK REDUCTION MEAS URES 79

6.1 Risk Preven tion 79

6.2 Accident Mitigation 80

6.3 Conclusions Regarding Preventive and Mitigative Measures 80

6.4 Reference s for Section 6 81

AP PENDIX A - RADIOACTIVE INVENTORIES 83

A. 1 INTRODUCTION 83

A. 2 SIMULATION OF OPERATING HISTORIES 83

A.2.1 Thermal Energy Productio n vs Time 83

A.2.2 Fuel Burnup Calcula tions 83

A.2.3 Calculation of Radioactive Inventories 84

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Page

A. 3 DATA FOR MILLSTONE 1 85

A.3.1 Reactor and Fuel Cycle Paramet ers 85

A.3.2 History of Operations 85

A.3.3 BWR Fuel Assembly Model Used in 0RIGEN2 Calculati ons 86

A.3.4 Calculated Radioactive Inventories 86

A.3.5 Decay Heat 86

A. 4 DATA FOR GINNA 87

A.4.1 Reactor and Fuel Cycle Parameters 87

A.4.2 History of Operations 87

A.4.3 PWR Fuel Assembly Model Used in 0RIGEN2 Calcul ation s 87

A. 4.4 Calculated Radioactive Inventories 88

A.4.5 Decay Heat 88

A. 5 REFERENCES FOR AP PENDIX A 89

AP PENDIX B - IMPACT OF REVISED REACTION ON THE LIKELIHOOD OF ZIRCONIUM

FIRES IN A DRAINED SPENT FUEL POOL 106

REFERENCES FOR APP ENDIX B 112

AP PENDIX C - EXAMPLE INPUT FILES FOR SFUEL AND CRAC2 127

C.

1

INTRODUCTION 127

C.2 SFUEL INPUT 127

C.3 CRAC2 INPUT 127

vii


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LIST OF TABLES

Table Page

S.l Estima ted Risk for the Two Spent Fuel Pools from the Two

Dominant Contr ibuto rs xxi

i i

1.1 Data on Spent Fuel Basins (as of December 31 , 198 4) 7

2.1 Typical Spent Fuel Pool Dimen sions and Water Inventories 32

2.2 Decay Heat as a Function of Time Since Last Refueling (Data

from Appendi x A) 32

2.3 Examples of Thermal-Hydra ulic Transient Paramet ers, Assuming

Complete Loss of Pool Coolant Circul ation 32

2.4 Fragility Paramet ers Assumed in This Study for Spent Fuel

Storage Pools 33

2.5 Weighting Factors Assigned to the Various Hazard and Fragility

Curves for the Millstone Case 33

2.6 Summary of Convolu tions of Seismic Hazard Curves with Fragility

Curves 34

2.7 Events in Which Inflated Seals Have Failed 35

2.8 Estimated Distrib ution of Human Error in Heavy Crane Ope rat ion s.. . 36

2.9 Assumpt ions Used in Calcul ating the Hazard of Catas troph ic Struc

tural Damage to Pool Res ulting from the Drop of a Shipping Cas k... 37

2.10 Summary of Estimated Prob abiliti es for Beyond Design Basis Acc i

dents in Spent Fuel Pools Due to Comple te Loss of Water Inventor y. 38

3.1 Summary of Critical Cond ition s Necess ary to Initiate Self-

Sustai

ni

ng

Oxi dati

on 59

3.2 Summary of Radial Oxidation Propagation Results for a High

Densitiy PWR Spent Fuel Rack with a 10 Inch Diameter Inlet and

Perfect Venti

1

ation 59

3.3 Summary of Radial Oxidation Propagation Results for a Cyli n

drical PWR Spent Fuel Rack with a 3 Inch Diamet er Hole and

Perfect Venti

1 ati

on 60

3.4 Summary of Radial Oxidation Propagation Results for a Cyli n

drical PWR Spent Fuel Rack with a 1.5 Inch Diame ter Hol e and

Perfect Venti

1

ation 60

3.5 Summary of Radial Oxidation Propagation Results for Various PWR

Spent Fuel Racks with No Ventila tion 61

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Table

Paae

3.6 Comparison of SNL Small Scale Oxidation Tests to Calcul ation s

with CLAD 61

4.1 Comparison of Radioac tive Inventories of Equilibr ium Core with

Spent Fuel Assemblies for Selected Isotopes (Millstone 1) 68

4.2 Estimated Radionu clide Release Fraction During a Spent Fuel

Pool Accident Resulting in Complete De struction of Cladding

(Cases 1 and 2) 69

4.3 Estimated Releases of Radio nucli des for Case 1 in Which a

Zirconium Fire Propagates Throughout the Entire Pool Inventory

(Worst Case ) 70

4.4 Estimated Releases of Radion uclide s for Case 2 in Which Only the

Last Discharged Fuel Batch Suffers a Zirconium Fire 71

4.5 Estimated Releases of Radio nucli des for Cases 3 and 4 in Which

Low-Temp erature Cladding Failures Occur 72

4.6 Comparison of Radioa ctive Inventories of Equilibri um Core with

Spent Fuel Assemblies for Selected Isotopes (Ginna) 73

4.7 CRAC2 Results for Various Releases Correspo nding to Postulated

Spent Fuel Pool Acciden ts with Total Loss of Pool Water 74

5.1 Estimated Risk for the Two Spent Fuel Pools from the Two Domi

nant Contri butors 77

A.l Reactor and Fuel Cycle Parameters for Mills tone 1 90

A. 2

Summary of Operational Milesto nes for Millst one 1 91

A.3 Summary of Spent Fuel Batches in Millst one 1 Storage Basin

(With Projections to 198 7) 92

A . 4

Comparison of Cumula tive Gross Thermal Energy Production with

Calculated Fuel Burnup from Start of Opera tions in 1970 to

April 1, 1987 (Millst one 1) 93

A.5 Comparison of Radioac tive Inventories of Reactor Core and Spent

Fuel Basin (Millstone 1) 94

A.6 Comparison of Radioact ive Inventories of Most Recently

D i s

charged Fuel Batch (Batch 11) with Longer Aged Discharged

Batches (Batches 1-10) (Millstone 1) 95

A.7 Decay Heat Released from Spent Fuel Inventory for Vario us D i s

charged Fuel Batches (Millstone 1) 96

ix


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Table Page

A. 8 Radionuclide Contributions to Decay Heat for Various Spent Fuel

Batche s (Millstone 1) 97

A.9 Reactor and Fuel Cycle Parameters for Ginna 98

A. 10

Summary of Operational Milestone for Ginna 99

A.11 Summary of Spent Fuel Batches in Ginna Storage Basin (With

Projec tions to 1987) 100

A.12 Comparison of Radioactive Inventories in Reactor Core and Spent

Fuel Basin (Ginn a) 101

A.13 Comparison of Radioactive Inventories in Most Recently Discharged

Fuel Batch with Longer Aged Fuel Batch es (Ginna) 102

A. 14 Decay Heat Released from Spent Fuel Inventory for Various

Dis

charged Fuel Batche s (Ginna ) 103

A.15 Radionuclide Contributions to Decay Heat for Various Spent Fuel

Batc hes (Ginna) 104

x


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PREFACE

This study is an initial atte mpt by Broo khav en National La borat ory to

characterize the radiological risks posed by storage of spent reactor fuel at

commercial reactor sites in the United Sta tes. This work was done at the

request of the U.S. Nuclear Regulatory Commission in support of their tech ni

cal analy sis related to Generic Safety Issue 82, Beyond Design Basis A c c i

dents in Spent Fuel Pools . The method of analysis used in this study was to

a) survey the spent fuel pool c onfigura tions at commercial reactor sites in

terms of the characteris tics that are important to risk and b) perform d e

tailed ana lyses of those spent fuel c onfigurati ons for which the risk appeared

to be potentially s ignifi cant. The detailed analyses were performed by using

the methodology of probabilistic risk assessment that has been used extensi ve

ly in the assessment of power plant risks during normal op erat ion.

T h u s ,

t h i s

initial stud y, while limited in resour ces , required the integration of several

t e c h n o l o g i c a l l y d i s t i n c t d i s c i p l i n e s (e . g . , s ei s m i c a n a l y s i s , f ue l d e g r a d a t i o n

a n a l y s i s ,

offsite conseq uence

a n a l y s i s).

Although these disciplin es have been

integrated before in the normal operation risk ass ess men ts, the application t o

the spent fuel problem posed novel and uncertain condition s not encountered in

the normal operation risk asse ssme nts. The present study did not add res s:

the potential for recritical ity; the fuel damage process during a slow pool

dra ina ge; and the fuel reconfiguration after a clad f i r e . The results of this

study have additional u ncert ainty , beyond those character istic of traditional

r i sk a s s e s s m e n t s t u d i es f o r r e a c t or o p e r a t i o n s , w h i c h i s as so c i a t e d w i t h t h e

novel aspects of the phenomenology and the limitations of the data b a s e .

xiii


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ACKNOWLEDGEMENTS

This work was performed for the Reactor and Plant Safety Issues Branch of

the Division of Reactor and Plant Sy st em s, NRC/RES. The NRC Manag ers for the

program we re Mr. E. Throm and Dr. M. Wohl wh o provided consider able input and

technical direction to the program. Mr. E. Throm also assisted by coordinat

ing a thorough NRC review of the initial draft of this r eport.

As with most integrated programs technical contributio ns were provided by

many people within and external to BNL. In part icul ar, the authors are in

debted to Dr s. A. Benjam in (SNL) and F. Best (Texas ASM) who provided consi d

erable assistance in implementing and understanding the SFUEL

code .

The

authors are also grateful for several technical contributions from the DNE

staff at BNL. Dr. K. Shiu provided considerab le assistanc e in evaluating th e

seismic hazard. Dr. T. Teichman assisted in several statistical eva luat ion s.

Dr.M. Reich and Dr. J. Pires were was especially helpful in the interpr eta

tion of pool structural fragility results and Dr. L. Teuton ico provided an

evaluation of the oxidation rate dat a. Dr. A. Tingle helped set up and inter

pret the consequence calculations with the CRAC2

code.

Mr. A. Aronson imple

mented the 0RIGEN2 code and provided the calculations for spent fuel pool

fis

sion product inventories for the actual disc harge hist orie s. Dr s. W. Pratt

and R. Bari provided administra tive assistance and were very helpful in pro

viding a thorough technical review of the final report.

The authors are especially grateful t o Ms. S. Flippen for her excellent

typing of this report and for cheerfully accepting the numerous additions and

revisions to this manuscript.

xv


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building in about 20 mi nu te s. Gates to the transfer tube and the fuel stora ge

p oo l w e r e i n th e cl o s e d p o s i t i o n , s o no wa t e r d r a i n e d f r o m t h e p o o l .

6

The second pneumatic seal failur e incident occurre d in the Hatch spent

s t o r a g e p o o l/t r a n s f e r c a n a l , (th e se al f a i l u r e a t Ha t c h w a s n o t in th e

refueling cav ity ) which released appro ximate ly 141, 000 gallons of water and

resulted in a drop in wate r level in the pool of about five f ee t.

7

Ho w e v e r , t h e BNL re v i e w of th e s e e v e n t s i n d i c a t e s t h a t t h e y a r e u n i q u e t o

t h e p l a n t s i n vo l v e d a n d s uc h e v e n t s a r e u n l i k e l y t o ca u s e a su b s t a n t i al l o s s

o f p oo l i n v e n t o r y f o r o t h e r p l a n t s . Ho w e v e r , p n e u m a t i c s ea l f a i l u r e s m a y

expos e individual fuel bundles during refueli ng and these events are being

investigat ed as part of Generic Issue 137, Refueling Cavity Seal Failu re.

S.2 OBJECTIVE

The obje c t i v e o f th i s i n v e s t i g a t i o n i s to pr o v i d e a n as s e s s m e n t o f th e

p o t e nt i a l r is k f r o m p o s s i b l e a c c i d e n t s i n sp e n t fu el p o o l s . The ri s k s a r e d e

fined in terms of:

- th e pr o b a b i l i t i e s o f va r i o u s i n i t i a t i n g e v e n t s t h a t m i g h t c o m p r o m i s e

t h e s t r u c t ur a l i n t e g r i t y o f th e poo l o r it s co o l i n g c a p a b i l i t y ,

- t h e p r o b a b i l i ty o f a sy s t e m f a i l u r e , g i v e n an in i ti a t i n g e v e n t ,

- f ue l f a i l u r e m e c h a n i s m s , g i v e n a sy s t e m f a i l u r e ,

- p o t e n t i al r a d i o n u c l i d e r e l e a s e s , a n d

- c o n s e q u e n c e s o f a sp e c i f i e d r e l e a s e .

Th i s s t u d y g e n e r a l l y f o l l o w s t h e l o g i c o f a ty p i c a l p r o b a b i l i s t i c r i sk

a n a l y s i s

(PRA);

h o w e v e r , b e c a u s e o f th e re l a t i v e l y l i m i t e d n u m b e r o f po t e nt i a l

a c c i d e n t s e q u e n c e s w h i c h c o u l d r e s u l t i n th e dr a i n i n g o f th e

p o o l ,

t h e a n a l y

s e s h a v e be e n g r e a t l y s i m p l i f i e d .

Th e co n f i g u r a t i o n s o f sp e n t f u el s t o r a g e p o o l s v a r y f r o m p l a n t t o pl a n t .

In

B WR ' s ,

t h e p o o l s a r e l o c a t e d w i t h i n t h e r e a c t o r b u i l d i n g w i t h t h e b o t t o m o f

the pool at about the same elevation as the upper portion of the r eactor

p r e s

s u r e v e s s e l . Du r i ng r e f u e l i n g t h e c a v i t y a b o v e t h e t o p of th e pr e s s u r e v e s s el

i s f l o o d e d t o th e sa m e e l e v a t i o n a s th e st o r a g e

p o o l ,

so that fuel assemb lies

c a n b e tr a ns f e r r e d d i r e c t l y f r o m t h e r e a c t o r t o th e po ol v i a a ga t e w h i c h s e p

arates the pool from the cav ity . In PWR pl an ts , the storage pool is located

in an auxilia ry build ing. In some cases the pool sur face is at about grade

l e v e l , i n ot h e r s t h e po ol b o t t o m is at gr a d e . The re f u e l i n g c a v i t i e s a r e

u s u a l l y c o n n e c t e d t o th e st o r a g e p oo l b y a tr a n s f e r t u b e . Du r i n g r e f u e l i n g

the spent assembly is removed from the reactor vessel and placed in a cont ain

e r w h i c h t h e n t u r n s o n it s s i d e , m o v e s t h r o u g h t h e t r a n s f e r t u b e t o th e st o r

age p o o l , i s se t up r i g h t a g a i n a n d r e m o v e d f r o m t h e t r a n s f e r c o n t a i n e r t o a

s t o r a g e r a c k . Va r i o u s g a t e s a nd w e i r s s e p a r a t e d i f f e r e n t s e c t i o n s o f th e

t r a n s f e r a n d s t o r a ge s y s t e m s . A sc r e e n i n g s t u dy w a s p e r f o r m e d t o id e n t i f y

p o t e n t i a l l y r i sk s i g n i f i c a n t s e q u e n c e s i n v o l v i n g s p e n t f u el p o o l s , t h e p o ol

d e s i g n f e a t u r e s o f th e co m m e r ci a l p o w e r p l a n t s w e r e r e v i e w ed a n d s u m m a r i ze d .

In or d e r t o pr i o r i t i ze t h e pr e s e n t r i sk a n a l y s i s , a pr e l i m i n a r y r i s k

a s s e s s m e n t w a s p e r f o r m e d f o r s p e n t f u el p o o l s u s i n g t h e RSS me t h o d o l o g y

2

and

t h e r e s u l t s o f th e ab o v e s c r e e n i n g s t u d y . Thi s pr e l i m i n a r y s t ud y i n d i c a t ed

that a seismic i nitiated failur e of the pool was the domin ant risk

xviii


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computer code , based on the operating histor ies of each of the plants (Appen

dix A ) . The calculated data included the 1987 inventories for each fuel batch

discharged at each refueling over the operating histo ry.

Fractional releases for various groups of radionucli des were estimated

based on the physical p arameters charact erizing the SFUEL failure sce nari o.

T h u s , source terms were estimated correspon ding to the seven accident s cenar

i o s :

five involving cladding fire for various amounts of fue l, and two in

volving cladding rupture (without a

f i r e ) .

Off-sit e radiological consequ ences were calculated using the CRAC2 com

p u t e r c o d e .

9

Because of several features in the health physics modeli ng in

the CRAC2 cod e, the population dose res ults are not very sensitive to the

estimated fission product relea se. A more sensitive measu re of the accident

severity appears to be the interdiction area (contaminated land ar ea) which in

the worst cases was about two hundred square mi le s. While the long-term

health effects (i.e., per son- rem) are potentially l arg e, it is important to

note that no "prompt fa tali ties" were predicted and the risk of injury was

also negligible.

5.6 RISK PROFILE

The likelihood and cons eque nces of variou s spent fuel pool acciden ts have

been combined to obtain the risks which are summarized in Table S.l. The

population dose results are insensit ive to the fission product release be

cause they are driven by decontamination levels assigned within the CRAC2

cod e. The health physics models in CRAC2 assign a maximum allow able dose for

each individual before the contaminat ed area is reoccupi ed. This allowable

dose for the returning population is the dominant contrib utor to total exp o

sure and limits the utility of the dose calc ulati on. Thus the land interd ic

tion area is included in Table S.l as a more sens itive re presentation of the

severity of the postulated a ccide nt.

The unique character of fuel pool acciden ts (potentially large releases

of long lived isot opes) makes it difficult to compare directly to reactor core

melt acci dent s. There are no early health effe cts . The long-term exposure

calculati ons are driven by assumptions in the CRAC modeling and the results

are not very sensitive to the severity of the accid ent. There is substantial

uncertainty in the fission product release esti mate s. These uncertai nties are

due to both uncertainty in the accident p rogression (fuel tempe rature after

clad oxidation and fuel relocation occurs) and the uncertainty in fission

product decontamination.

5.7 CONSIDERATION OF MEASURE S WHICH MIGHT REDUCE CONSE QUENCE S

A number of potential preventi ve and mitig ative measur es were identi fied,

but because of the large uncertainty ranges in Table S.l, the potential bene

fits of such measures are also uncertain and plant spec ific. A cost benefit

analysis has not been perfo rmed. Rath er, the phenomenological insi ghts , de

veloped during the investigation, have been used to generate a list of

p o s

sible risk reduction me asu re s. Calculati ons with the SFUEL code indicate

that, for those plants that use a high density s torage rack co nfig urati on, a

factor of five reduction in the fire probability (given loss of pool inven

tory ) can be achieved by improved air circulation ca pabil ity. This reduction

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f a c t o r is ba s e d up o n th e ti m e pe r i o d af t e r di s c h a r g e fo r wh i c h SFUEL pr e d i c t e d

t h a t th e de c a y he a t is su f f i c i e n t t o in i t i a t e a cl a d

f i r e .

Co n s i d e r i n g t h e

l a r g e u n c e r t a i n t y i n r i s k , a p l a n t s p e c i f i c c o s t /b e n e f i t a n a l y s e s s h o u l d b e

p e r f o r m e d b e f o r e s u c h r is k r e d u c t i o n m e a s u r e s a r e i m p l e m e n t e d .

5.8 CONCLUSIONS

Thi s li m i t e d ri s k as s e s s m e n t , w h i c h wa s pe r f o r m e d fo r tw o ol d e r sp e n t

fuel

p o o l s ,

i n d i c a t e s th a t th e ri s k es t i m a t e s ar e qu i t e un c e r t a i n an d co u l d

p o t e n t i a l l y (u n d e r w o r s t c a s e a s s u m p t i o n s ) b e s i g n i f i c a n t . Th e u n c e r t a i n t y i n

r i s k is do m i n a t e d by th e es t i m a t e d un c e r t a i n t y in th e li k e l i h o o d of th e lo s s

o f po ol in t e g r i t y du e to be y o n d de s i g n ba s i s se i s m i c ev e n t s . Thi s u n c e r t a i n t y

i s , i n tu r n , d r i v e n by th e un c e r t a i n t y in th e se i s m i c haz a r d an d th e sp e n t

f u el po ol fr a g i l i t y . The s e ri s k ra n g e s ar e co n s i s t e n t wi t h th e cu r r e n t me d i u m

p r i o r i t y as s i g n e d to th e is s u e by th e NRC.

1

It is no t cl e a r th a t th e s e un c e r

t a i n t y r a n g e s a r e d i r e c t l y a p p l i c a b l e t o o t h e r p l a n t s b e c a u s e t h e p l a n t s

s e l e c t e d f o r d e t a i l e d s t u d y w e r e c h o s e n s p e c i f i c a l l y f o r t h e i r p e r c e i v e d v u l

n e r a b i l i t y t o s e i s m i c e v e n t s a f t e r a n e x t e n s i v e s c r e e n i n g p r o c e s s (r e f e r t o

S e c t i o n S . 2 ) . F or e x a m p l e , if t h e f r a g i l i t y e s t i m a t e s f o r p l a n t s , w h i c h m e e t

t h e n e w s e i s m i c d e s i g n c r i t e r i a , w e r e u s e d , a s i g n i f i c a n t r e d u c t i o n i n t h e

p r e d i c t e d li k e l i h o o d of se i s m i c a l l y in i t i a t e d po ol fa i l u r e wo u l d re s u l t . In

a d d i t i o n m a n y o f t h e n e w p l a n t s h a v e po ol c o n f i g u r a t i o n s a n d a d m i n i s t r a t i v e

p r o c e d u r e s w h i c h w o u l d p r e c l u d e c a s k d r o p a c c i d e n t s . T h e r e f o r e , i n o r d e r t o

d e t e r m i n e w h e t h e r o t h e r p l a n t s h a v e a s i g n i f i c a n t r i sk p r o f i l e , a pl a n t

s p e

c i f i c ev a l u a t i o n wo u l d be re q u i r e d . A ke y pa r t of su c h an ev a l u a t i o n wo u l d be

t o ob t a i n a re a l i s t i c se i s m i c fr a g i l i t y es t i m a t e fo r th e sp e c i f i c sp e n t fu e l

p o o l .

5.9 REFERENCES FOR SUM MA RY

1. A Pr i o r i t i z a t i o n of Gen e r i c Sa f e t y Iss u e s , D i v i s i o n of Sa f e t y Tec h n o l o

g y ,

O f f i c e of Nuc l e a r Rea c t o r Reg u l a t i o n , U. S . Nuc l e a r Reg u l a t o r y Com m i s

s i o n ,

NUREG-0933, Dec emb er 1983, pp. 3.82-1t h r o u g h 6 .

2 . Rea c t o r Sa f e t y St u d y , A n As s e s s m e n t of Ac c i d e n t Ris k s in U.S . Com m e r c i a l

Nuc l e a r Po w e r Pl a n t s , U. S . Nuc l e a r Reg u l a t o r y Com m i s s i o n , NUREG-75/014

( W A S H - 1 4 0 0 ) ,

Octob er 1975, Ap p. I, Sec tio n 5.

3. A . S . B e n j a m i n , D . J. M c Cl o s k s y , D .A . P o w e r s , a n d S . A. D u p r e e , S p e n t Fu el

H e a t u p F o l l o w i n g L o s s o f W a t e r D u r i n g S t o r a g e , p r e p a r e d f o r t h e U. S .

N uc l e a r Reg u l a t o r y Com m i s s i o n by Sa n d i a L a b o r a t o r i e s , NUREG/CR-0649

( S A N D 7 7 - 1 3 7 1 ) ,

May 1979.

4 . N. A . Pi s a n o , F . Be s t , A . S . B e nj a m i n and K .T . St a l k e r , The Po t e n t i a l fo r

P r o p a g a t i o n o f a S e l f - S u s t a i n i n g Z i r c o n i u m O x i d a t i o n F o l l o w i n g L o s s o f

Wa ter in a Spent Fuel St ora ge

P o o l ,

p r e p a r e d fo r th e U.S . Nuc l e a r Reg u

l a t o r y Co m m i s s i o n b y S a n d i a L a b o r a t o r i e s , (D r af t M a n u s c r i p t , J a n u a r y

1984) (Note: th e pro ject ran out of fun ds befo re th e report was p u b

l i s h e d .)

5. IE Bul let in No.

8 4 - 0 3 :

Ref u e l i n g Cav i t y Wa t e r

S e a l ,

U.S . Nuc l e a r Reg u

l a t o r y Com m i s s i o n , O f f i c e of Ins p e c t i o n an d Enf o r c e m e n t , A u g u s t 24, 1984.

xx i


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Licensee Event Repo rt , LER No. 84-013-00, Haddam Neck, Docket No. 50-213,

Failure of Refuel ing Pool S eal , 09/21/84.

Nucleonics Wee k, December 11, 1986, pg. 3-4.

A. G. Crof f, 0RIGEN2: A Versat ile Compu ter Code for Calc ulat ing the

Nuclide Composition and Characte ristics of Nuclear Materia ls, Nuclear

Technolo gy, Vol. 6 2, pp . 335-352, September 1983.

L.T. Ritchi e, J.D. Johnson and R.M. Blo nd, Calculatio ns of Reactor Acci

dent Conseque nces Version 2, CRAC2: Computer Code User's Gui de , prepared

by Sandia National Laboratories for the U.S . Nuclear Regulatory Commis

si on , NUREG/CR-2326 (SA ND81-1994), Febr uar y 1983.

xxii


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Table S . l Estimated Risk f o r t h e T w o Spent Fuel Pools from

the T w o Dominant Contrib utors

Acc iden t

I n i t i a t o r

Seismic induced

PWR pool f a i l u r e

Seismic induced

BWR pool f a i l u r e

Cask drop

2

induced

PWR pool f a i l u r e

Cask drop

2

induced

BWR pool f a i l u r e

Spent Fuel

Pool Fi re

P r o b a b i l i t y / R y

2 . 6 x l 0 "

l t

- 1 . 6 x l 0 -

1 0

6 . 5 x l 0 -

5

- 4 x l 0

- 1 1

3 x l 0 -

5

- 3 x l 0 ~

1 2

8 x l 0 "

6

- 8 x l 0 -

1 3

Hea l th R isk

1

(Man-rem/Ry)

600-Neg.*

156-Neg.

70-Neg.

20-Neg.

I n t e r d i c t i o n

1

Risk

(Sq.

M i . / Ry )

.011-Neg.

.003-Neg.

.001-Neg.

4 x l 0 - '

t

- Ne g .

* N e g . - Negligible.

The upper

e n d o f t h e

risk ranges assu mes

n o

fire p ropagation from

t h e

last fuel discharge

t o

older fuel. However,

t h e

fission products

in

the last fuel discharge were assumed

t o b e

released during

t h e

fire

with

n o

fission product deconta minati on

o n

s t r u c t u r e s .

2

After removal o f accumulated inventory resu mes. Presently, most plants

are accu mula ting spent fuel i n t h e pool witho ut shipp ing t o permanent

stor age. (Note that many n e w plants have pool confi gurati ons a n d admin

istrative procedures which would preclude this failure mode.)

xxiii


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1. INTRODUCTION

Generic Safety Issue 82, Beyond Design Basis Accident s in Spent Fuel

Pools, was assigned a MEDIUM priority in November 1983.

l

In this prio riti za

ti on , the NRC staff considered t hree fac tors that had not been included in

e a r l ie r r is k a s s e s s m e n t s :

2

1. Spent fuel is currently being stored rather than shipped for repro

c e s s in g o r re p o s i t o ry d i s p o s a l , r e s u l ti n g in mu c h l a r g e r i n v e n t o r i e s

of spent assemb lies in reactor fuel basins than had previously been

a n t i c i p a t e d;

2. In or d e r t o ac c o m m o d a t e t h e l a r g e r i n v e n t o r y , h i g h d e ns i t y r a c k i n g i s

n e c e s s a r y , a nd

3. A th e o r e ti c a l m o d e l

3

s u g g e s t ed t h e p o s s i b i l i t y o f Zi r c a l o y f i r e ,

propagating from assembly to assembly in the event of complete drain

a g e o f wa t e r f r o m t h e p o o l .

1.1 Previo us Investigat ions

The Reactor Safety Study,

2

commonly referred to as WASH-1400, concluded

that the risks associated with spent fuel stor age were extremely small in com

p a r i s on w i t h a c c i de n t s a s s o c i a te d w i t h t h e r e a ct o r c o r e . Tha t co n c l u s i o n w a s

based on design and operational feature s of the stor age pools which made the

l o s s o f wa t e r i n v e n t o r y h i g h l y u n l i k e l y , e . g . ,

The pool structures were designed to withstan d safe shutdown e art h

q u a k e s ,

The fuel racks were designed to preclude cr itic ali ty,

Pool design and instrument ation precluded inadvertent and undetected

l o s s of wa t e r i n v e n t o r y ,

Procedur es and interlocks prevente d the drop of heavy loads on stored

a s s e m b l i e s , a nd

The storage struc tures were designed to accommoda te the forces and

m i s s i l e s g e n e r a t e d b y vi o l e nt s t o r m s .

Probabilitie s of pool fai lures due to external ev ents (earthq uake s,

m i s

sile s) or heavy load drops were estimated to be in the range of 10~

6

/ y e a r .

Radioac tive release estimates were based on meltin g of 1/3 of a core f or var

i o u s de c a y p e r i o d s , w i t h a n d w i t h o u t f i l t r a t i o n o f th e bu i l d i n g a t m o s p h e r e

(see Ref . 2, Table I 5- 2).

S u b s e q u e n t t o th e Rea c t o r S a f e t y S t u d y , A . S . B e nja m i n e t a l .

3

i n v e s t i g a t

e d t h e h e a t u p o f spe n t f ue l f o l l o w i n g d r a i n a g e o f th e po o l . A co m p u t e r c o d e ,

S FUEL, wa s de v e l op e d t o an a l y ze t h e r m a l - h y d r a u l i c p h e n o m e n a o c c u r r i n g w h e n

storage racks and spent assemblies bec ome exposed to air. The compute r model

t a k e s i n t o a c c ou n t d e c a y t i m e , f ue l a s s e m b l y d e s i g n , s t o r a ge r a c ks d e s i g n ,

p a c k i n g d e n s i t y , r o o m v e n t i l a t i o n a n d o t h e r v a r i a b l e s t h a t a f f e c t t h e h e a t u p

o f t he f u e l .

Calculat ions with SFUEL indicated that , for some storage c onfig urat ions

a nd d e c a y t i m e s , t h e Z i r c a l o y c l a d d i n g c o u ld r e ac h t e m p e r a t u r e s a t wh i c h t h e

e x o t h e r m i c o x i d a t i o n w o u ld b e c o m e s e l f - s u s t a i n i n g w i t h r e s u l t a nt d e s t r u c t i o n

1


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of the cladding and fission product rel ease. The possibility of propagatio n

to adjacent assemblies (i.e., the cladding would catch fire and burn at a hot

enough temperature to heat neighboring fuel assemblies to the ignition point)

was also identif ied. Under certain condi tio ns, the entire inventory of stored

fuel could become invo lved. Cladding fires of this type could occur at tem

peratures well below the melting point of the U 0

2

fuel* The cladding ignition

point is about 900C compared to the fuel melt ing point of 2880C.

Uncer tainti es in the SFUEL calcula tions were primarily att ributed to un

certain ties in the zirconium oxidation rat es.

Further work was done to refine the SFUEL com puter model and to compare

calculated results with experimental data.

1

* These more recent results have

generally confirmed the earlier concepts of a Zircaloy fire which, given the

right conditions, will propagate to neighboring assembl ies. However, compari

sons to out-of -pile heat-up data have not shown good agreement with the co de .

As discussed in Section 3, the SNL author s noted that more work in several

areas was needed to define more precisely the condi tions and confi gurat ions

which allow or prevent pr opaga tion.

Several studies have been conducted on alternative spent fuel st orage

conc ept s. Among these is a report published by the Electric Power Research

Institute

( E P R I ) ,

which applies probabil istic risk assessment technique s to

several storage con cep ts.

5

While this study does not directly address Generic

Safety Issue 82, it does provide useful insight on approp riate analytical

method ology as well as useful data on an in-ground (on-site) storage po ol.

1.2 Related Events

There is no case on record of a significant loss of water inventory from

a dome stic , commercial spent fuel storage pool . Howe ver, two recent incidents

have raised concern about the possib ility of a partial draindow n of a storage

pool as a result of pneu mati c seal fa ilu re s.

The first incident occurred at the Haddam Neck reactor during prepara

tions for refueling.

6

An inflatable seal bridging the annulus between the

reactor vessel flange and the reactor cavity bearing plate extruded into the

gap, allowing 200,000 gallons of borated w ater to drain out of the refueling

cavity into the lower levels of the containm ent building in about 20 min ute s.

Gates to the transfer tube and the fuel storage pool w ere in the closed posi

tio n, so no water drained from the po ol .

7

Had thes e gates been open at the

time of the leak, and had they not been closed within 10 to 15 min ute s, the

pool would have drained to a depth of about 8.5 feet, exposing the upper 3

feet of the active fuel region in the spent fuel as se mbl ie s.

7

Als o, had the

transfer of spent fuel been in progress with an assembly on the refueling

mac hin e, immediate action would have been necessary to place the assembly in a

safe location under water to limit exposure to per sonn el. The NRC has identi

fied this aspect of a seal failure accident as potential Generic Issue 137,

"Refueling Cavity Seal Failure."

8

The current schedule for evaluation of the

issue is December 1987.

The NRC Office of Inspection and Enforce ment required all lic ensees to

promptly evaluate the potential for refueling cavity seal fa ilu re s.

6

Re

sponses indicated that the refueling cavity configuration at Haddam Neck is

2


31/166

u n i q u e in t ha t t h e a n n u l u s b e t w e e n t h e r e a c t o r f l a n g e a n d t h e c a v i t y b e a r i n g

p l a t e is m o r e t h a n 2 f e e t w i d e . In m o s t p l a n t s t h i s g a p i s o n l y a b o u t 2

i n c he s w i d e .

9

A b o u t 4 0 o p e r a t i n g ( or s o o n t o o p e r a t e ) r e a c t o r s u s e i n f l a t a b l e

s e a l s in t h e r ef u e l i ng c a v i t y . H o w e v e r , b e ca u s e o f d e s i g n d i f f e r e n c e s , t h e

H a d d a m N e c k f a i l u r e d o e s n ot a p p e a r t o b e d i r e c t l y a p p l i c a b l e t o th e o t h e r

p l a n t s . It i s n o t e d t h a t m o s t B W R p l a n t s h a v e p e r m a n e n t s t e e l b e l l o w s s e a l s

t o fi ll t h e g ap b e tw e e n t h e r e a c t o r f l a n g e an d t h e c a v i t y b e a r i n g p l a t e . T h i s

i s s u e i s d i s c u s s e d m o r e f u l l y in S e c t i o n 2 . 3 .

T h e s e co n d p n e u m a t i c s e a l f a i l u r e i n c i d e n t o c c u r r e d i n t h e H a t c h s p e n t

s t o r a g e p o o l / t r a n s f e r c a na l i n D e c e m b e r 1 9 8 6 .

1 0

In t h i s i n c i d e n t , a p a i r o f

p n e u m a t i c s e a l s d e f l a t e d w h e n t h e c o m p r e s s e d a i r s u p p l y w a s i n a d v e r t e n t l y s h u t

o f f . T h e s e a l s i n v o lv e d w e r e in t h e t r a n s f e r c an a l f l e x i b l e s e i s m i c j o i n t .

T h e l e a k d e t e c t i o n a n n u n c i a t o r f a i l e d t o a l ar m a n d t h e l e a k w a s no t d i s c o v e r e d

f o r a b o ut 7 - 1 /2 h o u r s . A p p r o x i m a t e l y 1 4 1 , 0 0 0 g a l l o n s o f w a t e r l e a k e d f r o m t h e

s t o r a g e f u el a n d t h e w a t e r l ev e l d r o p p e d a b o u t 5 - 1 / 2 f e e t .

1 .3 R i s k P o t e n t i a l

T h i s s t u d y a d d r e s s e s b e y o n d d e s i g n b a s i s a c c i d e n t s i n s p e nt f u el p o o l s

t h a t m i g h t r e s u l t i n t h e c o m p l e t e l o s s of p o ol w a t e r d u e t o s t r u c t u r a l f a i l

u r e , m a s s i v e l e a k s or b o i l - o f f o f i n v e n t o r y d u e t o p r o l o n g e d f a i l u r e o f

c o o l i n g s y s t e m s . T h e r i s k p o t e n t i a l s a r e d e f i n e d i n t e r m s o f

- t h e p r o b a b i l i t i e s o f v a r i o u s i n i t i a t i n g e v e n t s t h a t m i g h t c o m p r o m i s e

t h e s t r u c t u r al i n t e g r i t y o f t h e p oo l o r i t s c o o l i n g c a p a b i l i t y ,

- t h e p r o b a b i l i t y o f a s y s t e m f a i l u r e , g i v e n a n i n i t i a t i n g e v e n t ,

- f ue l f a i l u r e m e c h a n i s m s , g i v e n a s y s t e m f a i l u r e ,

- p o t e n t i a l r a d i o n u c l i d e r e l e a s e s , a nd

- c o n s e q u e n c e s o f a s p e c i f i e d r e l e a s e .

T h e a n a l y s e s g e n e r a l l y f o l l o w t h e l o g i c o f t y p i c a l p r o b a b i l i s t i c r i s k

a n a l y s e s

( P R A ) ;

h o w e v e r , b e c a u s e o f t h e r e l a t i v e l y l i m i t e d n u m b e r o f p o t e n t i a l

a c c i d e n t s e q u e n c e s , w h i c h c o u l d r e s u l t i n t h e d r a i n i n g o f t h e p o o l , t h e a n a l y

s e s ha v e be e n g r e a t l y s i m p l i f i e d .

1 .4 D i s c u s s i o n o f S p e n t F u el S t o r a g e P o ol D e s i g n s a n d F e a t u r e s

T h e g e n e r al d e s i g n c r i t e r i a f o r s p e n t fu el s t o r a g e f a c i l i t i e s a r e s t a t e d

in A p p en d i x A of 1 0 C F R 5 0 ,

n

a n d a r e d i s c u s s e d m o r e f u l l y i n R e g u l a t o r y G u i d e

1 . 1 3 .

1 2

T h e p oo l s t r u c t u r e s , s p e n t f u el r a c k s a n d o v e r h e a d c r a n e s m u s t b e d e s i g n

e d t o S e i s m i c C a t e g o r y I s t a n d a r d s . It i s r e q u i r e d t h a t t h e s y s t e m s b e d e

s i g n e d ( 1 ) w i t h c a p a b i l i t y t o p e r m it a p p r o p r i a t e p e r i o d i c i n s p e c t i o n a nd t e s t

i ng o f c o m p o n e n t s i m p o r t a n t t o s a f e t y , ( 2 ) w i t h s u i t a b l e s h i e l d i n g f o r r a d i a

t i o n p r o t e c t i o n , ( 3 ) w i t h a p p r o p r i a t e c o n t a i n m e n t , c o n f i n e m e n t , a n d f i l t e r i n g

s y s t e m s ,

( 4 ) w i t h a re s i d ua l h e a t r e m ov a l c a p a b i l i t y h a v i n g r e l i a b i l i t y a n d

t e s t a b i l i t y t h a t r e f l e c t s t h e i m p o r t a n c e t o s a f e t y o f d e c a y h e a t a nd o t h e r

r e s i du a l h e a t r e m o v a l , a n d ( 5 ) t o p r e v e n t s i g n i f i c a n t r e d u c t i o n i n f u el s t o r

a g e c o o l an t i n v e n t o r y u n de r a cc i d e n t c o n d i t i o n s .

1 1

A s p a rt o f t h e p r e l i m i n a r y s c r e e n i n g s t u d y f o r a c c i d e n t v u l n e r a b i l i t i e s ,

t h e d e s i g n f e a t u r e s o f t h e s p e n t fu el p o o l s f o r t h e c o m m e r c i al p o w e r p l a n t s

3


32/166

were reviewed and assembl ed. The configu ration s of spent fuel storag e pools

vary from plant to plan t. Table 1.1 summarizes thi s information for each of

the p o o l s .

In BWRs the pools are located within the reactor building with the bottom

of the pool at about the same elevation as the upper portion of the reactor

pressure v ess el. (For exa mpl e, at Oyster Creek the bottom of the pool is at

e l e v a t i o n

80'6 ,

and the top at

119'3 .

The water depth is 38 feet .) During

refu elin g, the cavity above t he top of the pressure vessel is flooded to the

s a m e e l e v a t i o n a s th e st o r a g ep o o l , so that fuel ass emblies can be transferre d

directly from the reactor to the pool via a gate which separates the pool from

t h e c a v i t y .

In PWR plants the stor age pool is located in an auxili ary b uil din g. In

some cases the pool surface is at about grade le vel , in others the pool bottom

is at gra de. The refueling cavities are usually connected to the storage pool

b y a tr a n s f e r

t u b e .

During refueling the spent assembly is removed from the

reactor vessel and placed in a container which then turns on its s i d e , m o v e s

t h r o u g h t r a n s f e r t u b e t o st o r a g e p o o l , set upright again and removed from the

transfer container to a storage rack. Various gates and weirs separate d i f

f e r e n t s ec t i o n s of th e tr a n s f e r a n d s t o r ag e s y s t e m s . Mo r e d e t a i l s c o n c e r n i n g

various configurations are given in Section 2.3.

1.5 More Detail ed Studi es

The overall objective of the present investigation was to determ ine

whethe r possible severe accidents involving spent fuel pools posed a signifi

cant risk to the publ ic. In order to prioritize the invest igation a prelim

inary risk assessment was performed using

RS S

2

m e t h o d o l o g y t o id e n t i f y t h e

potentially important accident sequences and the characteristics of specific

fuel pools which could lead to unusually high vulnerability to accid ents .

This preliminary risk assessme nt indicated that seismically induced structural

failure of the pool appeared to dominate the spent fuel pool risk. This

appeared to be particularly true for older plants in the eastern states where

recent studies have indicated an increase in the estimated seismic ha zard.

Based on this preliminary stud y, two older BWR and PWR plants were selected

for more detailed studies because of their perceived vulnerabilit y to seismic

e v e n t s . S p e c i f i c a l l y , Mi l l s t o n e 1 an d Gin n a , w e r e s e l e c te d b e c a u s e o f av a i l a

bility of dat a, fuel pool inve ntor y, and the relative familiarity of the BNL

staff with the various candidate s i t e s . The operating histo ries of the two

p l a n t s w e r e m o d e l e d t o ob t a i n a re a l i s t i c r a d i o a c t i v e i n v e n t o r y i n th e va r i o u s

spent fuel batc hes . Details of the modeli ng procedure s and a listing of the

calculated radionuclide content are presented in Appendix A.

It should be noted that both plants have relatively large inventories of

spent fuel assemblies in their spent fuel p o o l s .

1.6 Report Content

Accident initiating e vents and their proba biliti es are covered in Section

2. Fuel cla dding fa ilur e scen ario s based on the SFUEL1W Compu ter Code are

evaluated in Section 3. Included are sensitivity analyses of the failure

sce

narios arising from uncertaint ies in Zircaloy oxidation reaction rate da ta ,

and hardware configuration as sump tion s. Section 4 presents data on the

4


33/166

potential for releases of radionuclides under various cladding failure scenar

ios and compares the projected releases with releases associated with severe

core accident seque nces . In Section 5, risk profiles are developed in terms

of person-rem population doses for several accident se quen ces. Section 6 con

siders measures that might mitigate pool draining and/or Zircaloy fire propa

g a t i o n .

1.7 Referenc es for Sectio n 1

1. A Prioritization of Generic Safety Issues, Division of Safety Technol o

g y ,

Office of Nuclear Reactor Regula tion, U.S. Nuclear Regulatory Commis

s i o n ,NUREG-0933, Decemb er 1983, pp. 3.82-1through 6.

2. Reactor Safety Stud y, An Assessment of Accident Risks in U.S. Commercial

Nuclear Power Plan ts, U.S. Nuclea r Regula tory Comm iss io n, NUREG-75/014

(WASH-1400),October 1975, Ap p. I, Section 5.

3. A. S. Benja min, D.J. McClosksy, D.A. P owe rs, and S.A. Dupr ee, Spent Fuel

Heatup Following Loss of Water During Storage, prepared for the U.S.

Nuclear Regulatory Commis sion by Sandia Labo rat ori es, NUREG/CR-0649

(SAND77-1371),

May 1979.

4.

N.A. Pisano , F. Be st , A. S. Benjamin and K.T. Stal ker, The Potential for

Propagation of a Self-Sustaining Zirconium Oxidation Following Loss of

Wat er in a Spent Fuel St orage P o o l , prepared for the U.S. Nuclear Regu

latory Commission by Sandia Laboratories, (Draft Manuscript, January

1984) (Note: the project ran out of funds befo re the report was pub

lished.)

5. D.D.

O r v i s ,

C. J ohns on, and R.

J o n e s ,

Review of Proposed Dry-Storage

Concepts Using Probabilistic Risk Assessm ent, prepared for the Electric

Power Research Institute by the NUS Corpo rat ion , EPRI NP-3365, Februar y

1984.

6. IE Bul let in No. 84-03: Refueling Cavity Water S e a l , U.S. Nuclear Regu

latory Commiss ion, Office of Inspection and Enforc ement , August 24, 1984.

7. Licensee Event Repo rt , LER No. 84-013-00, Haddam Nec k, Docket No. 50- 213,

Failure of Refueling Pool S e a l ,

09/21/84.

8. Generic Issue Manage ment Control System - First Quarter FY-87 Update s,

Memorandum from T.P.

S p e i s ,

Directo r, Division of Safety Review and Over

s i g h t ,

to H.R. Dent on, Directo r, Office of Nuclear Reactor Regulat ion,

U.S.Nuclear Regulator y Com mis si on, February 13, 1987.

9. Licen see Resp ons es to NRC IE Bul let in No.

84-03.

10. U.S. Nuclear Regulatory Commi ssio n, Morning Report - Region II, Decem

ber 5, 1986.

11. Code of Federal Regulat ion s, Title 10, Part 50, Domestic Licensing of

Production and Utilization Faci liti es, Appendix A, 'General Design Cri

teri a for Nuclear Pow er Plan ts,' General Design Criterion 6 1, 'Fuel Stor

age and Handling and Radioactivity Cont rol

1

.

5


34/166

1 2 .

U.S. Nuclea r Regul ator y Com mis sio n, Regula tory Guide 1.13, Spent Fuel

Storage Facility Design Basis, December 1981.


35/166

Table 1.1 BWR' s: DATA ON SPENT FUEL STORAGE BASINS. Included are spent fuel storage Inventories as of December 1984,

fraction s of core in stora ge, comparisons with the "reference case " of radionuclide inv entory, locations of

spent fuel basins, and seismic design bases of pools.

Radioactivity

Thermal Number of Spent Fuel Relative to Seismic

Power Fuel Assemblies Stored Invento ry

3

Stored Inventory Reference Case

c

Storage Pool Design

Plant (MWt) in Cor e

3

(No. of Assemblies) Fractions of Cor e

0

(per cent) Locationd Bas is

e

Big Rock Point

Browns Ferry-1

Browns Ferry-2

Browns Ferry-3

Brunswick-1

Brunswick-2

Cooper

Dresden-l

Dresden-2

Dresden-3

Duane Arnold

Fitzpatrick

Grand Gulf-1

Hatch-1

240

3293

3293

3293

2436

2436

2381

700

2527

2527

1658

2436

3833

2436

84

764

764

764

560

560

548

464

724

724

368

560

i

N/A

560

172

1068

889

1768

f

1056

9

924

985

221

h

2014

-

576

816

0

140

2.05

1.40

1.16

2.31

1.89

1.65

1.80

0.48

h

2.78

-

1.57

1.46

0.00

0.25

4.9

46.1

38.2

76.1

46.0

40.2

42.9

3.36

h

70.3

-

26.0

35.6

0.0

6.1

AB,

RB,

RB,

RB,

RB,

RB,

RB,

RB,

RB,

RB,

RB,

grd

ele

ele

ele

ele

ele

ele

ele

ele

ele

ele

N/A

RB, ele

DBE^O.OSy

DBE=0.20g

DBE=0.20g

DBE=0.20g

DBE=0.16g

DBE=0.16g

DBE=0.2g

DBE=0.20g

DBE=0.2g

DBE=0.2g

DBE=0.12g

DBE=0.15g

DBE=0.15g


36/166

T a b l e 1. 1 ( C o n t ' d )

T h e r m a l N u m b e r o f

P o w e r F u e l A s s e m b l i e s

P l a n t

( M W t ) i n

C o r e

3

Hatch-2

Humboldt Bay

LaCrosse

LaSalle-1

LaSalle-2

Limerick-1

Mi l ls tone-1

Monticello

Nine Mile Point-1

Oyster Creek

Peach Bottom-2

Peach Bottom-3

Pi lgr im-1

quad Cities-1

Quad Cities-2

2436

220

165

3323

3323

3293

2011

1670

1850

1930

3293

3293

1998

2511

2511

560

172

72

N/A

N/A

N/A

580

484

532

560

764

764

580

724

724

Spent Fuel

S t o r e d I n v e n t o r y

3


( N o .

o f

A s s e m b l i e s ) F r a c t i o n s

o f

Core'

1284

251

207

0

0

0

1346

1137

1244

1375

1361

1212

1128

1730

412

2.29

1.46

2.88

0.00

0.00

0.00

2.32

2.35

2.34

2.46

1.78

1.59

1.94

2.39

0.57

R a d i o a c t i v i t y

R e l a t i v e

t o

S e i s m i c

R e f e r e n c e

C a s e

c

S t o r a g e P o o l D e s i g n

( p e r c e n t ) L o c a t i o n ^ B a s i s

e

55.8

3.2

4.8

0.0

0.0

0.0

46.7

39.2

43 .3

47.5

58.6

52.4

38.8

60.0

14.3

RB, e le

N/A

AB,

RB,

RB,

RB,

RB,

RB,

RB,

RB,

RB,

RB,

RB,

RB,

RB,

grd

e le

e le

e le

e le

e le

e le

e le

e le

e le

e le

e le

e le

DBE=0.15g

DBE=0.50g

DBE=0.12g

SSE=0.20g

SSE=0.20g

SSE=0.13g

DBE=0.17g

DBE=0.12g

DBE=0.11g

DBE=0.22g

DBE=0.12g

DBE=0.12g

DBE=0.15g

0BE=0.24g

DBE=0.24g


37/166

Table 1.1 (Cont'd)

Plant

Susquehanna-1

Susquehanna-2

Vermont Yankee

Wash. Nucl.-2

Footnotes

Thermal

Power

(MWt)

3293

3293

1593

3323

Number of

Fuel Assemblies

in Core

3

764

764

368

N/A

Spent Fuel

Stored Inventory

3

( N o . of Assemblies)

0

0

1174

0

Stored Inventory

Fractions of Core

0

0.00

0.00

3.19

0.00

Radioactivity

Relative to

Reference Case

c

(per cent)

0.0

0.0

50.8

0.0

Storage Pool

Location

0

"

R B , ele

R B , ele

R B , ele

N/A

Seismic

Design

B a s i s

e

SSE=0.1g

SSE=0.1g

DBE=0.14g

SSE=0.32g

a) Source: U. S. Nuclear Regulatory Commission, Licensed Operating Reactors, NUREG-0020, Vol. 9, No. 1, January 1985.

b) (Stored Assembl 1es)/(Ass emblies 1n

C o r e ) .

c) "Reference Source Terra" assumes a thermal power of 3000 MWt, stored Inventory from ten annual di schar ges, last discharge six months ag o,

total invent ory 1750 assemb lies. Source term relative to "Reference Source Term" has not been corrected for age of fuel in storag e.

d) Locati on: RB = reactor buildi ng, AB = auxiliary buil ding, grd = pool at grade level, ele = pool at high elevation in building.

e) Seismic design basis as a function of the gravitational acceleratio n ( g) : DBE = design basis earth quake , or equivalent as used for older

vintage pla nts; SSE = safe shutdown ear thquake as defined in 10 CFR 100, App. A. Entry shown is the horizontal componen t.

f) Brunswick- 1 has in storage 160 PWR + 656 BWR assemblie s, equivalent to 1056 BWR assembl ies.

g) Brunswick-2 has in storage 144 PWR + 564 BWR assemblies, equivalent to 924 BWR assemblies.

h) Dresden Units 2 and 3 have two pools in one stru ctur e. The data cited are total of the two.

i) N/A = data not available.


38/166

Table 1.1 ( C o n t ' d ) P W R ' s : D A T A O N SPENT FUEL STORAGE BASINS . Included a r e spent fuel storage inventories a s of December 1 9 8 4 ,

f r a c t i o n s of core in s t o r a g e , c o m p a r i s o n s w i t h t he " r e f e r e n c e c a s e " of radionuclide inventor y, locationso f

s p e n t f u e l b a s i n s , a nd seismic design bases of p o o l s .

Plant

Arkansas-1

Arkansas-2

Beaver Vail

Byron-1

Callaway-1

Calvert

CI i

Calvert

CI i

Catawba-1

Cook-1

Cook-2

ey-1

f f s - 1

f fs-2

Crystal River-3

Davis Besse

Diablo Cany

Farley-1

- 1

on-1

Thermal

Power

(MWt)

2568

2815

2660

f

N/A

3411

2700

2700

N/A

3250

3411

2544

2772

3338

2652

Number

o f

Fuel Assemblies

in Core

3

177

177

157

N/A

N/A

217

217

N/A

193

193

177

177

N/A

157

Spent Fuel

Stored Inventory

Stored Inventory

(No.

o f A s s e m b l i e s ) F r a c t i o n s of C o r e

0

Radioactivity

Relative

t o

Reference Case

c

(per

cent)

56.3

26.7

17.6

0.0

N/A

9

108.0

-

N/A

9

93.1

-

24.6

31.2

N/A

19.3

Storage Pool

Locat ion**

AB,

AB,

FB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

Seismic

Design

B a s i s

e

DBE=0.2g

DBE=0.2g

SSE=0.125g

SSE=0.2g

SSE=0.2g

DBE=0.15g

DBE=0.15g

SSE=0.12g

SSE=0.20g

SSE=0.20g

SSE=0.10g

DBE=0.15g

J

DDE=0.4g

SSE=0.10g

388

168

104

0

N/A

g

8 6 8

N/A

9

553

171

199

N/A

114

2.19

0.95

0.66

0 . 0 0

N/A

9

'4.00

N/A

9

2.87

0.97

1.12

N/A

0.73


39/166

Table 1.1 (Cont'd)

Plant

Fa.rley-2

Fort Calhoun

Ginna

Haddam Neck

Indian Point-1

Indian Point-2

Indian Point-3

Ke w a u n e e

Maine Yankee

M c G u i r e - 1

M c G u i r e - 2

M i l l s t o n e - 2

North Anna-1

North Anna-2

Oconee-1

Thermal

Power

(MWt)

2652

1500

1520

1825

h

2758

3025

1650

2630

3411

3411

2700

2775

2775

2568

Number of

Fuel Assemblies

in Core

3

157

133

121

157

h

0

193

193

121

217

193

N/A

217

157

157

177

Spent Fuel

Stored Inventory

3

(No.

o f

Assemblies)

62

305

340

545

160

332

140

268

577

91

N/A

376

9

220

-

g

1037

Stored Inventory

Fractions

o f

C o r e

0

0.39

2.29

2.81

3.47

h'

1.72

0.73'

2.21

2.66

0.47

N/A

1.73

g

1.40

-

9

5.86

Radioactivity

Relative

t o

Reference

C a s e

c

(per cent)

10.5

34.4

42.7

63.4

h

47.4

21.9

36.5

69.9

16.1

N/A

46.8

9

38.9

-

g

150.5

Storage Pool

Location

0

"

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

Seismic

Design

B a s i s

e

SSE=0.10g

DBE=0.17g

DBE=0.20g

DBE=0.17g

DBE=0.10g

DBE=0.15g

DBE=0.1'5g

DBE=0.12g

DBE=0.1'0g

SSE=0.15g

SSE=0.15g

DBE=0.17g

SSE=0.12g

SSE=0.12g

DBE=0.10g


40/166

Table 1.1 (Cont'd)

Radioactivity

Thermal Number of Spent Fuel Relative to Seismic

Power Fuel Assembli es Stored Inven tory

3

Stored Inventory Reference Case

c

Storage Pool Design

Plant (MWt) in Cor e

3

(No. of Assemblies) Fractions of Core

0

(per cent) Location

0

B a s i s

e

Oconee-2

Oconee-3

Palisades

Palo Verde-1

Point Beach-1

Point Beach-2

Prairie Island-1

Prairie Island-2

Rancho Seco-1

Robinson-2

Salem-1

Sal em-2

San Onofre-1

San Onofre-2

San Onofre-3

2568

2568

2530

N/A

1518

1518

1650

1650

2772

2300

3338

3411

1347

3410

3390

177

177

204

N/A

121

121

121

121

177

157

193

193

157

217

217

-

218

480

N/A

g

524

-

g

601

-

260

152

296

265

94

217

0

-

1.23

2.35

N/A

g

4.33

-

g

4.97

-

1.47

0.97

1.53

1.37

0.60

1.00

0.00

-

31.6

59.5

N/A

9

65.7

-

g

82.0

-

40.7

22.3

51.2

46.8

8.1

34.1

0.0

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

AB,

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

grd

DBE=0.1g

DBE=0.1g

DBE=0.20g

SSE=0.20g

DBE=0.18g

DBE=0.18g

SSE=0.12g

SSE=0.12g

SSE=0.25g

DBE=0.20g

DBE=0.20g

DBE=0.20g

DBE=0.50g

SSE=0.67g

SSE=0.67g


41/166

T a b l e 1 .1 ( C o n t ' d )

P l a n t

S e , q u o y a h - l

S e q u o y a h - 2

S t . L u c i e - 1

S t . L u c i e - 2

S u m m e r - 1

S u r r y - 1

S u r r y - 2

T h r e e M i l e

I s l a n d - 1

T h r e e M i l e

I s l a n d - 2

T r o j a n

T u r k e y P o i n t - 3

T u r k e y P o i n t - 4

W a t e r f o r d - 3

Y a n k e e R o w e

T h e r m a l

P o w e r

( M W t )

3 4 1 1

3 4 1 1

2 7 0 0

2 5 6 0

2 7 7 5

2 4 4 1

2 4 4 1

2 5 3 5

i

3 4 1 1

2 2 0 0

2 2 0 0

N/A

600

N u m b e r o f

F u e l A s s e m b l i e s

i n C o r e

3

193

193

217

N/A

157

157

157

177

177

193

157

157

N/A

76

S p e n t F u e l


3

( N o . o f A s s e m b l i e s )

65

130

352

N/A

52

9

6 0 8

-

208

0

312

445

430

N/A

250


F r a c t i o n s o f C o r e

0

0 . 3 4

0 . 6 7

1 . 6 2

N/A

0 . 3 3

g

3 . 8 7

-

1 . 1 8

0 . 0 0

1 . 6 2

2 . 8 3

2 . 7 4

N/A

3 . 2 9


R e l a t i v e t o

R e f e r e n c e C a s e

c

( p e r c e n t )

S t o r a g e P o o l

L o c a t i o n * *

S e i s m i c

D e s i g n

B a s i s

e

1 1 . 5

2 3 . 0

4 3 . 8

N / A

9 . 2

9

9 4 . 5

2 9 . 8

0 . 0

A B , g r d

A B , g r d

A B , g r d

A B , g r d

A B , g r d

A B , g r d

A B , g r d

A B , g r d

A B , g r d

S S E = 0 . 1 8 g

S S E = 0 . 1 8 g

D B E = 0 . 1 0 g

S S E = 0 . 1 0 g

S S E = 0 . 1 5 g

S S E = 0 . 1 5 g

S S E = 0 . 1 5 g

D B E = 0 . 1 2 g

S S E = 0 . 1 2 g

5 5 . 1

6 2 . 4

6 0 . 3

N/A

1 9 . 7

A B , g r d

A B , g r d

A B , g r d

A B , g r d

A B , g r d

D B E = 0 . 2 5 g

D B E = 0 . 1 5 g

D B E = 0 . 1 5 g

S S E = 0 . 1 0 g

N o n e


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T a b l e 1 .1 ( C o n t ' d )


T h e r m a l N u m b e r o f - S p e n t F u e l R e l a t i v e t o S e i s m i c

P o w e r F u e l A s s e m b l i e s S t o r e d I n v e n t o r y

3

S t o r e d I n v e n t o r y R e f e r e n c e C a s e

c

S t o r a g e P o o l D e s i g n

P l a n t ( M W t ) i n C o r e

3

( N o . o f A s s e m b l i e s ) F r a c t i o n s o f C o r e' ' ( p e r c e n t ) L o c at i o n '' B a s i s

e

g g g

Z i o n - 1 3 2 5 0 1 9 3 8 6 3 4 . 4 7 1 4 5 . 3 A B , g r d S S E = 0 . 1 7 g

Z i o n - 2 3 2 5 0 1 9 3 - - A B , g r d S S E = 0 . 1 7 g

F o o t n o t e s

a ) S o u r c e : U . S . N u c l e a r R e g u l a t o r y C o m m i s s i o n , L i c e ns e d O p e r a t i n g R e a c t o r s , N U R E G - 0 0 2 0 , V o l . 9 , N o . 1 , J a n u a r y 1 9 8 5 .

b ) ( S t or e d A s s e m b l i e s ) / ( A s s e m b l i e s i n C o r e ) .

c ) R e f e r e n c e S o u r c e T e r m a s s u m e s a t h e r m a l p o w e r of 3 0 0 0 M W t , s t o r e d i n v e n t o r y f r o m t e n an n u a l d i s c h a r g e s , l a s t d i s c h a r g e s i x m o n t h s a g o ,

t o t a l i n v e n t o r y 7 0 0 a s s e m b l i e s . S o u r c e t e r m r e l a t i v e t o R e f e r e n c e S o u r c e T e r m h a s n o t b e e n c o r r e c t e d f o r a g e o f f u el i n s t o r a g e .

d ) L o c a t i o n : R B = r e a c t o r b u i l d i n g , A B = a u x i l i a r y b u i l d i n g , F B = fu e l b u i l d i n g , g = p o o l a t g r a d e l e v e l , e = p o o l a t h i g h e l e v a t i o n i n

b u i l d i n g .

e ) S e i s m i c d e s i g n b a s i s a s a f r a c t i o n o f t h e g r a v i t a t i o n a l a c c e l e r a t i o n ( g ) : D B E = d e s i g n b a s i s e a r t h q u a k e , or e q u i v a l e n t a s u s e d f o r o l d e r

v i n t a g e p l a n t s ; S S E = s a f e s h u t d o w n e a r t h q u a k e a s d e f i n e d i n 1 0 C F R 1 0 0 , A p p . A . E n t r y s h o w n i s t h e h o r i z o n t a l c o m p o n e n t .

f ) N / A = d a t a n o t a v a i l a b l e .

g ) S p e n t f ue l b a s i n s h a r e d b y t w o u n i t s . E n t r i e s s h o w n a r e t o t a l s ,

h ) I n d i a n P o i n t - 1 is p e r m a n e n t l y s h u t d o w n .

1 ) T M I - 2 i s i n d e f i n i t e l y s h u t d o w n .

j ) D i a b l o C a n y o n o r i g i na l l y u s e d t h e D o u b l e D e s i g n E a r t h q u a k e , D D E a c c e l e r a t i o n = 2 D B E . L a t e r , m o r e e l a b o r a t e an a l y s i s w a s d o n e to

p o s t u l a t e a n e a r t h q u a k e o f 0. 5 g a s s o c i a t e d w i t h t h e H o s g r i F a u l t .


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2.

ACCIDENT INITIATING EVENTS AND PRO BA BILITY ESTIMA TES

2.1 Loss of Wate r Circul ating Capability

The spent fuel basins of U.S . nuclear power stations contain a large in

ventory of wat er, primarily to provide ample radiation shielding o ver the top

of the stored spent fuel. Some typical pool dimensions and water i nventories

are shown in Table 2.1. The heat load from decay hea t of spent fuel depe nds

on decay time since the last refueling . Heat loads for the entire spent fuel

inventory of the two older plant s are shown in Table 2.2 (data extrapola ted to

the 1987 scheduled

r e f u e l i n g s).

The coolin g systems provided for spent fuel

pools typically have a capacity in the range of 15 to 20xl 0

6

Btu /hr (4.4 to

5 .9x l0

3

k w ) .

In the event that normal circulation of the cooling water is disru pted ,

e . g . , d u e to s t at i o n b l a c k o u t , p u m p f a i l u r e , p i p e r u p t u r e , e t c . , t h e w a t e r

tempera ture of the pool would steadily increase until bulk boiling occurr ed.

(Note: In a situation where the stored inventory was sm all , an equil ibrium

tem per atu re, below the boiling p oin t, would be reached at which surface evap

oration balanced th e decay heat

load).

Thermal-hyd raulic analyses of the consequences of partial or co mplete

loss of pool cooling capability are a routine part of the safety analysis re

ports required for licensing and amendments there to. Genera lly, these analy

ses consider several sc enarios ranging from typical to extremely conserv ative

cond itio ns. A sampling of conservativ e results for several plants is given in

Table 2.3. The data clearly demonstrate that the time interval from loss of

circulation until exposure of fuel to air is quite long. Even in the most

pessi misti c c ase cited in Table 2.3 (Docket No.

50-247),

the wate r level in

the pool would drop only about 6 inches per ho ur. T h u s , t h e r e i s c o n s i d e r a b l e

time availab le to restore normal cooling or to implement one of several alter

native backup options for cooling.

For licensing pur pos es, it has been accepted that the time interval for

restoring cooling manuall y from availabl e water sources is adequate withou t

requiring active (automatic) redundant cooling syste ms.

Howev er, in considering the prioritization of Generic Issue 82, Beyond

Design Basis Accidents in Spent Fuel Pools , the NRC staff recognized that

there is a finite probability that cooling could not be restored in a timely

m a n n e r .

2

The case treated in Ref. 2 was for a BWR. The estimated frequency

for the loss of one (of two ] cooling train s was taken to be 0.1/Ry (the

value assumed in WASH-1400).

3

This combined with the conditional probabi li

ties of failure/non-availability of the second train yielded a combined f r e

quency of a pool heatup event of 3.7xlO

2

/Ry. (This esti mate appear s to be

somewhat conservative since no pool heatup event s are on record after ~10

3

reactor years of accumulated e x p e r i e n c e ) .

To escalate from a pool heat up event to an event which results in fuel

damage requires the failure of several alternat ive syste ms that are capable of

supplying makeup w ater (the RHR and conde nsate t ransfer sys tem s, or , as a last

r e s o r t , a f i r e

hose).

Estimated frequencies of failure for each of the alter

nat ive s, combined with th e frequency of a pool heatup ev en t, resulted in an

15


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The three steps and the treatment of the uncertai nties have been summar

ized by Reed,

6

who notes that the largest uncertai nties are associated with

s t e p 1 ), i . e . , t h e p r o b a b i l it i e s o f o c c u r r e n c e o f s e v e r e e a r t h q u a k e s h a v i n g

correspondingly very large ground accel erati ons. Reed makes the assertion

that due to the large uncertainties in the ground shaking hazard, it is un

productive to refine the structure and equipment capacity calculations to

accuracies which are inconsistent with the hazard uncertainty.

6

The specific

applicability to spent fuel pools of Reed's assertion is discussed in Section

2.2.1.3.

2.2.1.1

A Review of Seismic Hazard Data

The primary difficulty in characterizing the seismic hazard at specific

sites in the Eastern United S tates

(EUS),

i.e. , sites to the east of the Rocky

Mountains is that severe earthquakes are rare events in the EUS. A systematic

analysis of recorded earthquakes and their relationship to geological features

has yielded seismic zonation maps of theEUS.

7

Howev er, such information can

not readily be translated into the type of seismic hazard functions needed as

input for PRA. Conseq uentl y, available historical data alone are insufficient

for obtaining meaningful site specific estimates of the frequency of severe

e v e n t s .

During the past 6 or 7 ye ar s, the methodologie s for seismic hazard analy

ses have been under intensive deve lopme nt. Henc e, the analyses presented in

this report must be considered provisional and subject to future refinem ent.

At the present tim e, an intensive effort to refine the methodology is in prog

ress under the auspices of the Electric Power Research Institute

(EPRI).

8

The

m e t h o d s , i n p ut p a r a m e t e r s , c om p u t e r p r o g r a m m in g a nd u s e r s ' m a n u a l s a r e p r e

sented in a ten volume report which is currently in the process of distri bu

t i o n .

8

This is referred to as the Seism icity Ow ners Group (SOG) seis mic h a z

ard methodology devel opment program, or SOG Method ology . Unfortunately the

SOG Methodology was not available for the calculations carried out in this

r e p o r t .

The SOG Methodology is a refinement and elaboration of the met hodol ogies

develo ped earli er at Lawrence Livermore National Laboratory (LLNL) by D.L.

Bernreuter and his colleagues under NRC sponso rship . The initial study was a

part of the NRC's System atic Evaluation Program (SEP).

9

The methodolo gy has

been expanded and modified in a subsequent study , EUS Seismic Hazard Charac

t e r i z at i o n P r o j e ct (S HC P ).

1 0

n

Since the SHCP results are used for the seismic hazard e sti mat es, some

further discussion of the Bernreuter methodolog y is appro priat e. Three basic

steps are involved:

1. Expert opinion was elicited to delineate and characterize se ismically

active zones in the EUS, and to define earthquak e ground motion

m o d e l s .

The experts also provided estimates of uncertainti es assoc i

ated with their assumptions.

2.

Seismic zonatio n, seismicity and ground motion inputs are integrated

into hazard functions at specified site s.

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3. Modeling and parameter uncertaint ies are reflected in the form of

best estimat es and 15th, 50th and 85th percentile seismic hazard

c u r v e s .

The various steps are carried out in a highly disciplined and systematic

mann er. Provision is made at various stages for peer review of the methods

and input op inio n, feedback to the experts and critical evaluation of the re

s u l t s .

In step 1, each expert prepares a best estima te map which del ineates

the seismic z o n e s . Each zone is characterized by a set of parameters that

give the maximum earthqua ke intensity to be expected for that zone (upper mag

nitude

c u t - o f f ),

the expected frequency of eart hqua kes , and the magni tude re

currence relatio n. For each input (zone bou nda rie s, seismic p a r a m e t e r s ) , t h e

expert provides a measure of his degree of confi dence . Also each expert is

given the option of submitting alt ernative map s of differing zonations and

chara cter izatio ns (up to as many as 30m a p s ) . The data from each expert are

evaluated separately through step 2.

In step 2, the cont ribution at a given site from each zone is integrated

over the zone area and then over all

z o n e s .

This requires the use of ground

motion models for which a range of alternative m odels are employed to yield a

set of alternative hazard cur ves . A Ground Motion Panel of experts have

selected several alternative mode ls to be used , each having a weighting factor

(see Ref. 10, App. C) . Also each ground motion model incorporates a site spe

cific correction to account for local geology.

In step 3, the results of the individual

Severe Accidents in Spent Fuel Poolsin Support of Generic Safety Issue 82

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