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.
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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
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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
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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
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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
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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 .
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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.
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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
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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.
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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.
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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.)
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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
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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
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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
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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
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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
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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.
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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
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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
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
S t o r e d I n v e n t o r y
( 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
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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.
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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
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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
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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
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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
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 )
65
130
352
N/A
52
9
6 0 8
-
208
0
312
445
430
N/A
250
S t o r e d I n v e n t o r y
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 a d i o a c t i v i t y
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 )
R a d i o a c t i v i t y
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
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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