1 Of ENGINEERING DATA TRANSMITTAL 616269

s P.PS 1 Of 1

ENGINEERING DATA TRANSMITTAL 616269

2. To: (Receiving Organization) 3. from: (Originat ing Organization) 4. Related EDT No.: S o l i d i f i e d HLW I n t e r i m Pro jec ts SAR Engineer ing None Storage P r o j e c t (8M200) 5. Proj./Prog./Dept./Div.: 6 . Design Authority/ Design Agent/Cog. 7. Purchase order No.:

Engr.: TDP/TWR R. J. Kidder NA 8. Or ig inator Remarks:

Th is e v a l u a t i o n i s issued f o r approval and re lease. NA 9. Ec#ip./Cmnponent No.:

10. SystemfBldg.lFacility:

HLW/CSB 11. Receiver Remarks: 11A. Design Baseline Dacunent? [ ] Yes [ x ] No 12. Major Assm. Dug. No.:

13. PermitfPermit Appl icat ion No.: t-f?- 1 I&. Rewi red ResDonse Date:

COB 09/13/96 15. DATA TRANSMIllED ( f ) (6) ( H ) ( 1 )

Approval Reason Origi- Recelv. Des,g~ for nalor er

IAI IC) ID1

nator Trans- Dlspo- D l ~ p o ltam N O NO NO

mittal sition sition

1%) Title or Description of Data Transmitted I81 DocurnentlOrawing No Sheet

1 WHC-SD-WM-TI-781 0 Canis ter Storage - 2 1 N A --

B u i l d i n g Evaluat ion o f Nuclear Sa fe ty f o r NA Sol i d i f i ed High-Level Waste T rans fe r and Storage

16 KEY Approval Designator IF1 Reason far Transmmal IGI Disposition IHI & 11)

1 Approved 4 Reviewed ndcomment lree WHC-CM-3-5. 2 Release 5 Post-Review 2 Approved wlcommenr 5. Reviewed wlcomrnent Sec.12.71 3 Information 6 Dist. IReceipt Acknow Requiredl 3 Disapproved Wlcomment 6. Receipt acknowledged

E. S. Q. D or NiA 1 A P P ~ O Y ~ I 4 Review

17 SIGNATUREIDISTRIBUTION Wee Approval Designator for required signatured

BD-7hOO-172-2 (05/96) GEf097

BD-7-172-1

WHC-SD-WM-TI-781, Rev. 0

Canister Storage Building Evaluation of Nuclear Safety for Solidified High-Level Waste Transfer and Storage

R. J . Kidder Westinghouse Hanford Company, Richland, WA 99352 U . S . Department o f Energy Cont rac t DE-AC06-87RL10930

EDT/ECN: 616269 UC: UC230 Org Code: EM200 Charge Code: D4N87 B&R Code: EW3130010 T o t a l Pages: 55

Key Words: C a n i s t e r Storage, High-Level Waste, Safe ty E v a l u a t i o n

A b s t r a c t : T h i s document i s i ssued t o eva lua te t h e s a f e t y impacts t o t h e C a n i s t e r Storage B u i l d i n g f rom t r a n s f e r and s to rage o f s o l i d i f i e d h igh- l e v e l waste.

TRADEMARK DISCLAIMER. Reference herein t o any speci f ic comnercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily const i tu te o r imply i t s endorsement, recmendation, o r favoring by the United States Government o r any agency thereof or i t s contractors or subcontractors.

Pr in ted i n the United States of America. To obtain copies of t h i s document, contact: WHC/BCS Document Control Services, P.O. Box 1970, Mailstop H6-08, Richland UA 99352, Phone (509) 372-2420: Fax (509) 3 i b - h ~ a 9 .

=*-_--* L )& 7 - - )7 y b R ease Approval U Date Release stamp

Approved for Public Release

A-6400-073 (10/95) GEF321


CANISTER STORAGE BUILDING EVALUATION OF NUCLEAR SAFETY FOR SOLIDIFIED HIGH-LEVEL WASTE TRANSFER AND STORAGE

SEPTEMBER 1996

Prepared by: R. J . Kidder


EXECUTIVE SUMMARY

This report provides support for a Westinghouse Hanford Company (WHC) scoping evaluation to identify significant issues for use of the Canister Storage Building (CSB) for interim storage of solidified high-level waste (HLW) produced during the Phase I privatization effort. evaluated were glass canisters (large and small) and containers of loaded cesium ion exchange resin.

evaluations and some draft operating manuals for spent nuclear fuel (SNF) handing in the CSB were reviewed. Potential abnormal occurrences which could cause a release of radioactive material from either or both of the container types that could challenge the CSB safety systems was developed. abnormal occurrences was classified by types (accident phenomena) and were combined to form accident analyses for the bounding releases from each type o f accident. From these bounding analyses, preliminary lists of safety class and significant systems for the CSB were developed. Detailed structural or thermal analyses were not performed when analysis information was not available, or loss of confinement was assumed for containers and/or CSB components when subjected to abnormal thermal or mechanical stresses. The preliminary conclusion was that no additional engineered safety features would be required for interim storage of SNF.

the CSB was to continue to classify the building as Hazard Category 2. single HLW container (glass canister or cesium container per the inventory information provided by WHC) has sufficient radionuclide inventory to result in a Hazard Category 2 per DOE-STD-1027-92. facility is a decision that could be made by the Program Secretarial Officer (PSO).

container should be double or single contained. information and assumptions, the improvement in residual risk provided by the secondary cesium container would be relatively small.

This evaluation did not find any additional Nuclear Regulatory Commission (NRC) requirements beyond the present SNF Project requirements but a recommended future detailed NRC equivalency review may provide additional requ i remen t s .

It is recommended that for future safety analysis, during conceptual design, a formal hazards identification/evaluation should be performed of the entire HLW container handling and storage process. validation of the inventory assumptions (radionuclide content and physical form) used for accident analysis should be performed. This i s particularly important for the physical form of the cesium ion exchange resin. If the resin is grounded or somehow reduced to small particles during the drying process, this will potentially change the accident analysis results. The safety equipment list should be verified, to make sure that the appropriate

The waste forms

To identify bounding accident scenarios, several safety analyses and

The

The effect of the storage of HLW in the CSB on the hazard category of A

Designating the CSB a Category 1

A qualitative evaluation was performed to determine if the cesium Based on current design

Verification and

i i


CSB systems a r e i n c o r p o r a t e d . A s t r u c t u r a l a n a l y s i s shou ld be performed t o determine t h e p o t e n t i a l e f f e c t s o f d ropp ing a new m u l t i c a n i s t e r overpack h a n d l i n g machine (MHM) cask on t h e s a f e t y - c l a s s v a u l t r o o f , e i t h e r d u r i n g c o n s t r u c t i o n o r a se ismic event. A p r e l i m i n a r y hazard c a t e g o r i z a t i o n must be documented. Because t h e des ign o f t h e f a c i l i t y and process i s i n a r e l a t i v e l y e a r l y stage, no a t tempt was made t o eva lua te a c c i d e n t f requenc ies . e v a l u a t e r i s k , some s o r t o f f requency d e t e r m i n a t i o n w i l l need t o be performed. A s e t o f a d m i n i s t r a t i v e c o n t r o l s and Technical Safe ty Requirements/Operational Safe ty Requirements (TSR/OSR) must be developed t o assure t h a t t h e assumptions i n t h e s a f e t y a n a l y s i s a re implemented i n t h e des ign and o p e r a t i o n o f t h e f a c i l i t y . I n p a r t i c u l a r , t h e a p p r o p r i a t e s u r v e i l l a n c e program f o r c o n t a i n e r s i n i n t e r i m s to rage must be developed. T h i s r e p o r t assumes t h a t some p e r i o d i c m o n i t o r i n g o f t h e s to rage tubes w i l l be performed by a s u r v e i l l a n c e program.

To

iii

WHC.SD.WM.TI.781, Rev . 0

CONTENTS

1.0

2.0

3.0

4.0

5.0

6.0

7.0

8.0

9.0

SCOPE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

FACILITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.1 FACILITY/PROCESS DESCRIPTION . . . . . . . . . . . . . . . . 1

INVENTORY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3 .1 RADIONUCLIDE INVENTORIES . . . . . . . . . . . . . . . . . . 6

3.1.1 Glass C a n i s t e r s . . . . . . . . . . . . . . . . . . . . 6 3.1.2 Cesium I o n Exchange Resin Conta iners (Cesium

Containers) . . . . . . . . . . . . . . . . . . . . . . 6

ACCIDENT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . 7 4.1 CSB STORAGE HAZARDS . . . . . . . . . . . . . . . . . . . . . 7

4.1.1 Abnormal Occurrences Dur ing CSB Storage . . . . . . . . 7 4.1.2 RESULTS/DISCUSSION OF CSB STORAGE . . . . . . . . . . . 10 4.1.3 Acc idents for F u r t h e r A n a l y s i s . . . . . . . . . . . . 11

4.2 CONSEQUENCES OF ACCIDENTS . . . . . . . . . . . . . . . . . . 12 4.2.1 Consequence A n a l y s i s . . . . . . . . . . . . . . . . . 12 4.2.2 Rad ionuc l ide I n v e n t o r i e s . . . . . . . . . . . . . . . 14 4.2.3 Acc ident Scenar ios . . . . . . . . . . . . . . . . . . 14

4.3 SAFETY CLASSIFICATION SUMMARY . . . . . . . . . . . . . . . . 31

HAZARD CLASSIFICATION EVALUATION . . . . . . . . . . . . . . . . . . 34

CESIUM CONTAINER EVALUATION . . . . . . . . . . . . . . . . . . . . 37 6.1 CONTAINER TRANSPORTATION . . . . . . . . . . . . . . . . . . 37 6.2 CONTAINER STORAGE . . . . . . . . . . . . . . . . . . . . . . 38

6.2.1 Mechanical Damage . . . . . . . . . . . . . . . . . . . 38 6.2.2 Cor ros ion . . . . . . . . . . . . . . . . . . . . . . . 39 6.2.3 P r e s s u r i z a t i o n . . . . . . . . . . . . . . . . . . . . 39 6.2.4 Overheat ing . . . . . . . . . . . . . . . . . . . . . . 40 6.2.5 Exp los ion . . . . . . . . . . . . . . . . . . . . . . . 40 6.2.6 Summary . . . . . . . . . . . . . . . . . . . . . . . . 41

NRC EQUIVALENCY REVIEW . . . . . . . . . . . . . . . . . . . . . . . 42 7.1 DESIGN AND CONSTRUCTION MEASURES . . . . . . . . . . . . . . 42

7.1.1 R e s u l t s . . . . . . . . . . . . . . . . . . . . . . . . 43 7.2 CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . . . 43

SAFETY ANALYSIS REQUIREMENTS . . . . . . . . . . . . . . . . . . . . 44

REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46

i v


LIST OF TABLES

3-1. 3-2. 3-3. 3-4.

4-1.

4-2.

4-3.

4-4.

4-5.

4-6.

4-7 .

4-8.

4-9.

Glass Properties. . . . . . . . . . . . . . . . . . . . . . . . . . Radionuclide Composition of Glass . . . . . . . . . . . . . . . . . Cessium-Loaded Ion Exchange Resin Properties . . . . . . . . . . . Radionuclide Compositions in Undiluted Supernate IX Column Feed. . . . . . . . . . . . . . . . . . . . . . Postulated Abnormal Occurrences During CSB Receiving and Storage (2 Sheets). . . . . . . . . . . . . . . . Atmospheric Dispersion Coefficients Used in Accident Analyses for the CSB . . . . . . . . . . . . . . . . . . . Unmitigated Dose Consequences for a Dropped Glass Canister Onto Another Canister. . . . . . . . . . . . . Mitigated Dose Consequences for a Dropped Glass Canister Onto Another Canister. . . . . . . . . Unmitigated Dose Consequences for a Dropped Glass Canister. . . . . . . . . . . . . . . . . . . , . . Mitigated Dose Consequences for a Dropped Glass Canister. . . . . . . . . . . . . . . . . . . . . . . . . . . Unmitigated Dose Consequences for Glass Canister Over-Pressuri zati on. . . . . . . . . . . . . . . . . . . . Unmitigated Dose Consequences for a Dropped Cesium Container. . . . . . . . . . . . . . . . . . . . . . . . . . Mitigated Dose Consequences for a Dropped Glass Canister. . . . . . . . . . . . . . . . . . . . . . . . . . .

4-10. Unmitigated Dose Consequences for a Pressurized Release from a Cesium Container. . . . . . . . . . . . . . . . . .

4-11. Mitigated Dose Consequences for a Pressurized Release from a Canister. . . . . . . . . . . . . , . . . . . . .

4-12. Safety Systems, Structures, and Components Criteria. (2 sheets) . . . . . . . . . . . . . I . . . . . . .

4-13. Preliminary List of CSB Safety-Class and Safety-Significant SSCs. . . . . . . . . . . . . . . . . . . . . .

5-1. Preliminary Hazard Categorization For One Large Glass Canister. (2 sheets) . . . . . . . . . . . . . . . . . . . .

6-1. Summary of Double Containment Evaluation. . . . . . . . . . . . . . 8-1. Resources and Schedule. . . . . . . . . . . . . . . . . . . . . . .

3 3 5

5

8

13

16

17

20

21

23

26

27

30

31

32

34

35 42 45

V


LIST OF ACRONYMS

ARF CSB DOE EDE HLW HWVP MCO MHM NRC PSAR PSO RF RFP SNF ssc TQ TSR/OSR WHC

a i rbo rne re lease f r a c t i o n Can is te r Storage B u i l d i n g U.S. Department o f Energy e f f e c t i v e dose equ iva len t h i g h l e v e l waste Hanford Waste V i t r i f i c a t i o n P l a n t mu1 t i - c a n i s t e r overpacks m u l t i - c a n i s t e r overpack hand1 i n g machine Nuclear Regulatory Commission P r e l i m i n a r y Safety Ana lys i s Report Program S e c r e t a r i a l O f f i c e r r e s p i r a b l e f r a c t i o n request f o r proposal spent nuc lea r f u e l systems, s t ruc tu res , and components t h r e s h o l d q u a n t i t y Technica l Safety Requirement/Operational Safety Ana lys i s Report Westinghouse Hanford Company

v i

WHC-SO-WM-TI-781, Rev. 0

T h i s page i n t e n t i o n a l l y l e f t b lank.

v i i


1.0 SCOPE

Th is e v a l u a t i o n i d e n t i f i e s nuc lea r s a f e t y issues and formulates p r e l i m i n a r y f i n d i n g s . B u i l d i n g (CSB), as designed f o r SNF storage, can p rov ide f o r t h e sa fe s torage o f s o l i d i f i e d h igh - leve l waste (HLW) i n accordance w i t h a l l a p p l i c a b l e U.S. Department o f Energy (DOE) guidance documents.

Th is r e p o r t evaluates whether t h e Can is te r Storage

2.0 FACILITY

The pa th forward f o r s o l i d i f i e d HLW i n t e r i m storage i s a f f e c t e d by whether t h e SNF CSB, as designed f o r storage, can p rov ide f o r t h e sa fe s torage o f s o l i d i f i e d HLW (WHC 1996). Th i s e n t a i l s t h e use o f t h e CSB f o r i n t e r i m storage o f s o l i d i f i e d HLW produced du r ing t h e Phase I p r i v a t i z a t i o n e f f o r t . An i n i t i a l engineer ing assessment i n d i c a t e d t h a t excess CSB v a u l t s (two southern most) could, w i t h some m o d i f i c a t i o n , be used f o r immobi l ized HLW i n t e r i m storage (Jacobs 1996). However, t h i s engineer ing e v a l u a t i o n i nc luded o n l y a cu rso ry assessment o f s a f e t y impacts on CSB design and opera t i on .

and con ta ine rs c o n t a i n i n g cesium separated from waste by i o n exchange i n t o r e s i n d u r i n g l ow- leve l waste pret reatment . e i t h e r standard s i z e o r elongated. Standard can is te rs , l a r g e c a n i s t e r s , and cesium con ta ine rs phys i ca l c h a r a c t e r i s t i c s are presented i n Sect ion 3.1.

For purposes o f t h e eva lua t i on bas is , t h e c a n i s t e r / c o n t a i n e r should be assumed t o p rov ide pr imary containment. The c a n i s t e r / c o n t a i n e r are s t a i n l e s s s t e e l w i t h a welded l e a k - t i g h t c losu re . The f r a c t i o n o f waste m a t r i x (HLW g lass o r cesium) t h a t cou ld y i e l d r e s p i r a b l e f i n e s under acc iden t scenar ios i s unknown. Standard assumptions rega rd ing r e s p i r a b l e f r a c t i o n (RF) were adopted f o r t h e e v a l u a t i o n bas i s .

S o l i d i f i e d HLW encompasses c a n i s t e r s c o n t a i n i n g v i t r i f i e d (g lass ) HLW

The v i t r i f i e d HLW c a n i s t e r s may be

2.1 FACILITY/PROCESS DESCRIPTION

The CSB i n t e r i m storage f a c i l i t y was designed and p a r t i a l l y cons t ruc ted f o r v i t r i f i e d waste i n t e r i m storage t o immobi l ize HLW a t t h e Hanford Waste V i t r i f i c a t i o n P lan t (HWVP). Cons t ruc t i on o f t h e CSB was h a l t e d i n 1993 a f t e r c a n c e l l a t i o n o f t h e HWVP P ro jec t , w i t h o n l y the concrete pad and a w a l l s e c t i o n b u i l t . Removal o f SNF from the K Basins and o t h e r Hanford SNF storage l o c a t i o n s became a S i t e p r i o r i t y , and an i n t e r i m storage f a c i l i t y was needed. The CSB was subsequently re-scoped t o accommodate Hanford S i t e SNF i n t e r i m storage. Cons t ruc t i on o f t h e CSB was r e s t a r t e d i n A p r i l , 1996. Although o n l y one o f t h e t h r e e v a u l t s i s r e q u i r e d f o r SNF i n t e r i m storage, t h e SNF p r o j e c t p lans t o c o n s t r u c t a l l t h r e e v a u l t s , i n c l u d i n g s t r u c t u r a l components and decking. excess v a u l t s , bu t w i l l be prov ided i f necessary f o r waste immob i l i za t i on .

Storage tubes and exhaust s tacks w i l l n o t be prov ided f o r t he two

1


The CSB des ign i nc ludes t h r e e below-grade v a u l t s approx imate ly 50 m wide by 55 m l o n g by 14 m deep, w i t h each v a u l t c o n s i s t i n g o f a m a t r i x o f 22 rows by 10 columns f o r a t o t a l o f 220 s t e e l s torage tubes. S i x ove rs i zed s torage tubes are a l s o prov ided i n each v a u l t t o accommodate overpacked c a n i s t e r s . Each tube can h o l d t h r e e can is te rs , equat ing t o 678 c a n i s t e r s p e r v a u l t .

Decay heat from t h e g lass c a n i s t e r s and cesium con ta ine rs w i l l be removed by n a t u r a l convect ion. Cool ing a i r i s drawn through an i n l e t duc t i n t o a plenum t h a t d i s t r i b u t e s i t i n t o the v a u l t , f l ows across t h e tubes, and e x i t s through an e leva ted exhaust s tack. a i r through t h e v a u l t i s n a t u r a l convect ion caused by a d e n s i t y d i f f e r e n c e between h o t a i r i n s i d e t h e v a u l t and stack, r e l a t i v e t o cool i n t a k e a i r .

The g l a s s c a n i s t e r s and cesium con ta ine rs w i l l be t ranspor ted t o t h e CSB i n an o n s i t e t r a n s p o r t a t i o n cask v i a a diesel-powered t r a c t o r / t r a i l e r . The t r u c k v e s t i b u l e w i l l be an extens ion o f t h e CSB t h a t p rov ides an a i r - l o c k e d t r a n s i t i o n area f o r t he t r a n s p o r t s e m i - t r a i l e r and t h e r e c e i v i n g cask crane t o unload t h e g lass c a n i s t e r o r cesium con ta ine r cask i n t o t h e r e c e i v i n g p i t . A f t e r t h e t r a n s p o r t t r a i l e r has been p o s i t i o n e d i n t h e t r u c k ves t i bu le , t h e t r a c t o r w i l l be disconnected and removed from t h e f a c i l i t y . The r e c e i v i n g crane w i l l be used t o l i f t t h e t r a n s p o r t cask c o n t a i n i n g t h e g lass c a n i s t e r o r cesium con ta ine r o f f t he t r a i l e r and lower i t i n t o t h e s e r v i c e p i t .

The g l a s s c a n i s t e r o r cesium con ta ine r w i l l be t r a n s f e r r e d t o the s torage tube. CSB w i l l be accomplished by the m u l t i - c a n i s t e r overpacks (MCO) hand l i ng machine, and t h e MHM, whose f u n c t i o n i s t o p i c k up g l a s s c a n i s t e r s o r cesium con ta ine rs from t h e s e r v i c e p i t i n t he r e c e i v i n g area o f t h e CSB and t r a n s f e r them t o s torage tubes i n one o f t h e v a u l t s o f t h e CS8. As r e q u i r e d f o r f i n a l d i s p o s i t i o n , t h e MHM w i l l be capable o f reve rs ing t h e process and r e t r i e v i n g t h e g lass c a n i s t e r s o r cesium con ta ine rs .

The mot ive f o r c e f o r drawing c o o l i n g

Movement o f t h e g lass c a n i s t e r s o r cesium con ta ine rs w i t h i n t h e

3 .0 INVENTORY

The exact CSB inven to ry o f HLW c a n i s t e r s and cesium con ta ine rs i s Three cases are, t he re fo re , prov ided t o bound t h e p o s s i b l e unknown.

i nven to ry . These cases are descr ibed below.

Case A (s tandard c a n i s t e r maximum) - To ta l p e r v a u l t l o a d i n g o f 678 s tandard c a n i s t e r s (3 standard c a n i s t e r s p e r s torage tube) .

Case B ( l a r g e c a n i s t e r maximum) - To ta l pe r v a u l t l o a d i n g o f 452 l a r g e c a n i s t e r s (2 l a r g e c a n i s t e r s pe r s torage tube) .

Case C (cesium con ta ine r maximum) - Per v a u l t l o a d i n g o f cesium con ta ine rs cou ld range from 226 (1 cesium con ta ine r pe r s torage tube) t o 1356 (6 cesium con ta ine rs pe r s torage tube) . Hopever,57the aggregate pe r s torage tube inven to ry w i l l n o t exceed 3.18 x 10 C i Cs, 3.18 x lo5 C i

2


137mBa, and 2.1 C i 135Cs r e g a r d l e s s o f t h e number o f c o n t a i n e r s p e r s to rage tube.

S o l i d i f i e d HLW encompasses c a n i s t e r s c o n t a i n i n g v i t r i f i e d ( g l a s s ) HLW (Tables 3-1 and 3-2) and c o n t a i n e r s o f cesium-loaded i o n exchange r e s i n generated d u r i n g l o w - l e v e l waste p re t rea tment (Tables 3-3 and 3-4).

Notes: Values based on Manuel, A. F . , J. D. Galbraith, S. L . Lambert, and G. E. Stegen, 1996, Phase I High-Level Uaste Pretreatment and Feed Staging Plan, UHC-SD-ES-370, Rev. 0, Uestinghouse Hanford Conpany, Richland, Washington.

Glass Density - 2,640 kgfd

Table 3-2. Rad ionuc l ide Composition o f Glass. (3 sheets) I s o t o p e I c i / g I DCF (rem/Ci) I Rem/g

res . i r a b l e 4.29E-02 3.04E-04 1.05E+01 8.06E-02 6.69E-05 1.20Et04 4.22E+02 4.09E-02 7.38E-01 7.45E-02 1.53E+00 6.82E-04 4 8.32E-04

3


4


Oxide

I so tope

8.1E-08 6.OE-07 1.5E-05

Inc luded i n To ta l ‘None prov ided i n EPA (1988)

kg Oxide/kg Z e o l i t e

T iO,

H.0

Others

I I S O , 0.14

0.36

0.10

0.24

Table 3-4. Radionucl ide Compositions i n Und i l u ted Supernate I X Col umn Feed .*

*This source term i s used o n l y f o r t h e abnormal event t h a t l i q u i d i s r e t a i n e d i n the i o n exchange r e s i n .

5


3.1 RADIONUCLIDE INVENTORIES

The f o l l o w i n g descr ibes t h e r a d i o n u c l i d e i n v e n t o r i e s o f t h e s o l i d i f i e d HLW g lass c a n i s t e r s and cesium i o n exchange r e s i n con ta ine rs (cesium con ta ine rs ) .

3.1.1 Glass Can is te rs

There are two s i zes o f g l a s s can is te rs , one c o n t a i n i n g 1,650 kg o f g l a s s and the o t h e r w i t h a capac i t y o f 2,735 kg o f g lass . The g lass c a n i s t e r s are s t a i n l e s s s t e e l w i t h a welded l e a k - t i g h t c losure. The d e n s i t y o f t h e g lass i s 2,640 kg/m3. The r a d i o n u c l i d e composi t ion o f t h e g lass i n cur ie /gram (Ci /g) i s shown i n Table 3-2. A lso shown are t h e dose convers ion f a c t o r s (EPA 1988) f o r each o f t h e isotopes, w h i l e t h e f i n a l column o f Table 3-2 shows t h e e f f e c t i v e dose equ iva len t (EDE) c o n t r i b u t i o n i n rem/g o f g lass, assuming 1 g o f g l a s s as r e s p i r a b l e (s10 micron) . The Pu isotopes a re a l l assumed t o be i n the ox ide form. The t o t a l f o r a l l i so topes ' DCFs i s 3.30 x IO5 rem p e r g o f g lass re leased as r e s p i r a b l e . A d d i t i o n a l p e r t i n e n t i n f o r m a t i o n on t h e two g l a s s c a n i s t e r s are as f o l l o u s :

Standard Can is te r

Nominal l e n g t h - 3.0 in Nominal d iameter - 60.96 cm Wall t h i ckness - 0.95 cm M a t e r i a l o f c o n s t r u c t i o n - 304L s t a i n l e s s s t e e l Can is te r weight (empty) - 454 kg Glass content - 1,650 kg To ta l c a n i s t e r weight ( f u l l ) - 2,100 kg I n t e r n a l volume - 0.74 m3 Maximum a l l owab le c e n t e r l i n e temperature - 400 O C

Large Can is te r

Nominal l e n g t h - 4.5 m Nominal d iameter - 60.96 cm Wall t h i ckness - 0.95 cm M a t e r i a l o f c o n s t r u c t i o n - 304L s t a i n l e s s s t e e l Can is te r weight (empty) - 667 kg Glass con ten t - 2,735 kg To ta l c a n i s t e r weight ( f u l l ) - 3,402 kg I n t e r n a l Volume - 1.15 m3 Maximum a l l owab le c e n t e r l i n e temperature - 400 O C

3.1.2 Cesium I o n Exchange Resin Conta iners (Cesium Conta iners)

have up t o 3.2 x l o 5 C i o f 137Cs/137mBa i n each con ta ine r . The cesium con ta ine rs w i l l con ta in i no rgan ic i o n exchange r e s i n and can

However, t he

6


r e a t e p e r s torage tube inven to ry w i l l n o t exceed 3.2 x lo5 C i o f Ys/'"Ba and 2.1 C i o f 13%s regard less o f t h e number o f con ta ine rs p e r s torage tube. l e a k - t i g h t c losu re . chemical composi t ion o f t h e m a t e r i a l as rece ived from t h e vendor i s shown i n Table 3-3. s i z e o f 30 t o 60 mesh (595 t o 250 pm). t h e percentage o f f i n e s i n the r e s i n , i t i s assumed f o r acc iden t ana lys i s purposes t h a t 0.1% o f t h e r e s i n w i l l have been reduced t o p a r t i c l e s w i t h a d iameter o f < 10 pm. The i o n exchange r e s i n con ta ins t e n weight percent water as rece ived f rom t h e vendor. Although t h e product s p e c i f i c a t i o n s r e q u i r e t h a t t h e s to red product be d r ied , i t i s assumed t h a t a volume o f i o n exchange feed equal t o t h e volume o f water from the vendor could be r e t a i n e d i n t h e cesium con ta ine r f o r one o f t h e acc ident scenar ios. The r a d i o n u c l i d e concen t ra t i on o f t h e u n d i l u t e d supernate used i n Cesium Demonstrat ion U n i t P re l im ina ry Safety Eva lua t i on (Ebasco/BNFL, 1994) i s assumed (Table 3-4). A d d i t i o n a l p e r t i n e n t i n f o r m a t i o n f o r t h e cesium con ta ine rs i s as f o l l o w s :

The cesium con ta ine rs w i l l be s t a i n l e s s s t e e l w i t h a welded

The i o n exchange media as rece ived from t h e vendor has a p a r t i c l e Because no i n f o r m a t i o n i s prov ided on

The b u l k d e n s i t y o f t h e r e s i n i s 1,000 kg/m3. The

Cesium Conta iner

Maximum l e n g t h - 1.37 m Maximum diameter - 33 cm Wall t h i ckness - 0.95 cm M a t e r i a l o f c o n s t r u c t i o n - 304L s t a i n l e s s s t e e l Conta iner weight (empty) - 212 kg Cesium-loaded i o n exchange r e s i n weight - 84 kg To ta l weight - 296 kg

4.0 ACCIDENT ANALYSIS

The acc iden t ana lys i s was conducted t o eva lua te t h e hazards, develop acc ident scenar ios, and d e f i n e systems, s t ruc tu res , and components (SSCs) impor tan t t o s a f e t y f o r t h e t r a n s f e r s and s torage o f HLW i n the CSB f a c i l i t y .

4.1 CSB STORAGE HAZARDS

Th is e v a l u a t i o n i d e n t i f i e d t h e abnormal occurrences d u r i n g CSB storage t o determine t h e hazardous events f o r t h e acc ident ana lys i s .

4.1.1 Abnormal Occurrences Dur ing CSB Storage

rad ionuc l i des t o be breached are l i s t e d i n Table 4-1. The p o s t u l a t e d abnormal occurrences t h a t cou ld cause t h e confinement o f

7


CAUSE

Seismic event

Lightning

Airplane crash Hunan error

Hoist f a i l u r e

Tab le 4-1.

SAFEGUARDS

Emergency procedures

Training and procedures

CSB I ICCURRENCE

~

Hydrogen bui ldup i n cesiun container, Leakage of hydrogen from the container and spark source

ix ternal phenomena

Design of container. adnin is t ra t ive controls on uater removal

lamage t o canister or container 3ecause of dropping heavy object (sh ie ld r i n g or cask l i d ) on :anister or container during receiving lamage t o cesiun container because ,f hydrogen explosion

rydrogen explosion i n storage tube

External leakage o f uater i n t o storage tube. rad io lys is o f the mater. and spark Source

Hydrogen buildup i n cesiun container, leakage of hydrogen frm the Container and w a r k source

Hunan error

Hydrogen explosion i n MHN

Design of storage tubes; survei l lance

Adnin is t ra t ive controls on mater removal

Training and procedures

Maintenance and inspection of c w o n e n t s

lamage t o glass canister or cesiun zontainer during receiving sequence

rruck accident i n receiving bay Hunan error

Truck mechanical f a i l u r e

S a m as above Hunan error

Receiving crane f a i l u r e

Fai lure of l i f t i n g voke

Buildup of hydrogen because of inadequate uater removal

Buildup of steam because of inadequate mater removal and overheating during transport Hunan error

Fai lure of control system

rruck f i r e i n receiving bay )rapping cask/canister or cesium :mtainer i n t o receiving area when l i f t i n g cask from transport t r a i l e r

T r e i ni ng and procedures- -dr i ver q u a l i f i c a t i o n

Maintenance and inspection of truck carponents S a m as above Training and procedures

Maintenance and inspection of l i f t i n g carponents

Controls t o assure conplete dryin5

Control system i n h i b i t s t ravel unless locking pins are engaged

lverpressurization of the cesium :ontainer during receiving

IHM t rave ls W o locking pins tngaged causing glass canister or :esiun container t o drop if ieismic event occurs

Hydrogen buildup i n cesiun container from i n c q l e t e drying causing rad io lys is and i g n i t i o n

Controls t o assure conplete dry in i

8


Tab le 4-1. CSB

OCCURRENCE Drop glass canister or cesium container because of damaged hois i

Damage t o glass canister or cesi ln container because of ho is t snagging

Dropping glass canister or cesium container i n t r a n s i t t o storege locat i on

Damage t o glass canister or cesiun container during lowering i n t o storage tube

Damage t o glass canister or cesiun container during ra is ing from storage tube Glass canister o r cesium container dropped i n t o service p i t onto neuly received glass canister o r cesium container prepared f o r storage Melt ing of glass canister or cesium container because o f overheating i n storage tube Over-pressurization of the cesium container during storage

Corrosion of glass canister or cesiun container

lamage t o tu0 containers during storage tube loading

lamage t o lower container because 3f placing second unl ike container 3n top of lover container

l e l t i n g o f contents of glass :mis ter o r cesium container =cause of overheating i n MHM

Postu la ted Abnormal Occur s c e i v i n g and Storage ( 2 SI

CAUSE Overraising the hois t u h i l e carrying the glass canister or cesiun container Overlouering the canister ho is t u h i l e carrying the container

Grapple prematurely disengaged o r engaged uhen load i s inproperly posit ioned

Lateral mavement o f MUM u h i l e container i s being lowered or stuck part way because of too many impact absorbers, urong vault, or

poor c lea rance Same as above

Misrouting of an MHU already containing a cesium container o r glass Canister t o service s t a t i o n

Decay heat and loss o f storage tube cool ing system

Hydrogen a c c m l a t i o n because of f a i l u r e t o cwnpletely dry ion exchange res in

Steam pressurization because of heating of cesium container because of decay heat and loss of cool ing

~~

Materials i n c o n p a t i b i l i t y of uaste u i t h container (excess chlorides, e. G.)

Contamination of uelds

Drop of top container on top of second container

~

Hunen error i n container placement

Decay heat and loss of MHM cooling system and loss of a b i l i t y t o Rove Rhm t o location where container can be removed

!nces Dur ing bets).

SAFEGUARDS Limi t sui tches

L imi t suitches and resolver

L imi t suitches

Ueight sensing system

Control system Mechanical in ter lock on MHM

Same as above

Adnin is t ra t ive control and a c c m t a b i I i t y

Weight ind icat ion on mhm grapple

Emergency procedures

Controls t o assure cwnplete drying of i o n exchange res in

Q u a l i t y control; design o f containers

Inpact absorbers a t bottom of storage tube and between :ontainers :esium container designed t o bear f u l l Weight o f glass canister and inpact absorber (not yet p e c i f i ed)

Inergency procedures

lackup pouer t o MHM v e n t i l a t i o n system

9


4.1.2 RESULTS/DISCUSSION OF CSB STORAGE

The CSB storage activities, as evaluated in this report, include a number of uncertainties. of the MHM internal transport cask. performed, safety-class structures could be at risk. This activity will be included in the evaluation performed during conceptual design. Preliminary Hazard Analysis, which have been performed for the CSB and the safety analyses for the HWVP, a number of draft operating manuals were reviewed, and a number of assumptions made about the applicability of the process description to the glass canisters and cesium containers. are as follows:

One activity that was not evaluated was replacement Depending upon where the upgrade work is

In addition to

Assumptions

. .

.

.

. .

.

.

.

.

.

Glass canister sizes will be those shown in Section 3.1.1

Cesium container sizes are those shown in Section 3.1.2

The radionuclide content of the glass is as shown in Table 3-2

The maximum radionuclide content of a single cesium container or a single storage well holding up to six containers is 3.2 x lo5 Ci 13'Cs

The maximum normal fines fraction (based on fracturing during cooling) of glass is the same as that used in the HWVP Preliminary Safety Analysis Report (PSAR) (see Section 4.2.3.1, Source Term Analysis)

The cesium resin used will be inorganic

Cesium ion exchange resin will be dried so that there are no free liquids and insufficient remaining water to generate sufficient hydrogen to overpressure the container or create an explosion hazard

The drying and handling process for cesium ion exchange resin will not create greater than 0.1 wt% "fine" (less than 10 micron) particles

Containers (cesium containers and glass canisters) are not vented

Storage tube depth is 12.8 m (per the HWVP PSAR; the SNF Preliminary Safety Evaluation assumed a storage tube depth of slightly over 13 m)

Criteria for safety-class and safety-significant equipment remain as stated in WHC-CM-4-46

The use of impact absorbers at the bottom of the storage tubes and in between containers is assumed, although no further credit is taken in the unmitigated analysis

It is assumed that the MHM grapple will be designed to not allow release of the grapple with a load.

10


A scenar io t h a t may n o t have been examined i s t h e p o s s i b i l i t y o f m i s r o u t i n g a f u l l MHM t o p i c k up a f u l l con ta ine r (g lass c a n i s t e r o r cesium con ta ine r ) t h a t has j u s t a r r i v e d a t t h e CSB. The MHM grapp le does have weight i n d i c a t i o n , b u t t h a t seems t o be t h e o n l y hardware method o f a s c e r t a i n i n g whether the MHM has a payload. An i n i t i a l s tep i n t h e process o f p i c k i n g up a con ta ine r i s r e l e a s i n g the grapple so t h a t i t can be used t o grasp t h e newly- a r r i v e d con ta ine r . I f t h e weight i n d i c a t i o n d i d n o t show t h a t t h e MHM had a payload, i t migh t be dropped onto t h e newly-arr ived con ta ine r . Confinement would probably be maintained, b u t recovery a c t i o n s m igh t be extens ive.

Yet another scenar io t h a t was pos tu la ted i n p rev ious hazards ana lys i s was dropping a con ta ine r onto another con ta ine r a l ready i n a s torage tube. The hazard a n a l y s i s assumption was t h a t each tube would c o n t a i n two impact absorbers t o prevent the con ta ine rs from being damaged by a drop acc ident , one t o prevent t h e bottom con ta ine r from being damaged i f dropped, t h e o t h e r t o prevent t h e t o p con ta ine r and bottom con ta ine r from being damaged i f t h e t o p one was dropped on t h e bottom one. Th is ana lys i s does n o t t ake c r e d i t f o r t h e impact absorbers f o r s a f e t y c lass determinat ion.

noted i n the opera t i ng manuals t h a t t h e r e c e i v i n g crane i s r a d i o c o n t r o l l e d . I f random r a d i o s i g n a l s cou ld l e a d t o a c o n d i t i o n t h a t might r e s u l t i n dropping t h e cask and/or t h e con ta ine r , i t should be assured t h a t t h e crane s i g n a l i s unique.

t o mix con ta ine rs (a g lass c a n i s t e r and a cesium con ta ine r ) i n the s torage tube. I t i s f u r t h e r assumed t h a t t h i s event would n o t be planned b u t might be p o s s i b l e i f bo th k inds o f con ta ine rs were being p laced i n t o s torage d u r i n g the same t ime pe r iod . (damage t o a cesium con ta ine r ) would be des ign ing the ex te rna l frame f o r t he cesium con ta ine r t o bear t h e weight o f an impact absorber and a g lass c a n i s t e r . f e a t u r e w i l l be implemented.

Dropping a f u l l cask/conta iner i s an undes i rab le occurrence. It was

It i s assumed f o r t he purposes o f t h i s r e p o r t t h a t i t migh t be poss ib le

A s a f e t y f e a t u r e t o prevent undes i rab le consequences

It i s assumed i n t h e abnormal occurrence t a b l e t h a t t h i s s a f e t y

It i s unknown whether o r n o t t h e s torage tubes w i l l be p e r i o d i c a l l y moni tored as t h e s torage tubes f o r spent f u e l are planned t o be. assumed i n t h i s r e p o r t t h a t p e r i o d i c mon i to r i ng o f t h e s torage tubes f o r re lease o f r a d i o a c t i v e m a t e r i a l from a con ta ine r w i l l be performed. Because i t i s assumed t h a t t he i o n exchange r e s i n w i l l be d r i e d be fo re t h e column i s sealed, i t i s assumed t h a t t h e s torage tubes w i l l n o t be i ne r ted , no r w i l l t hey be moni tored f o r hydrogen.

It i s

4.1.3 Acc idents for Fur the r Ana lys i s

c r e d i b l e t h a t might cause damage t o the con ta ine rs (g lass c a n i s t e r s o r cesium con ta ine rs ) . The types o f abnormal events are:

Conta iner c o r r o s i o n

F i ve types o f abnormal events o t h e r than e x t e r n a l events are deemed

11


Overheating (because of loss of cooling in either the MHM or the storage tube)

Drop accidents of various types

hydrogen evolution, or hydrogen burn) of the cesium containers because of failure to completely dry the ion exchange resin

Catastrophic failure of the container because of shearing during handling or external pressure from hydrogen detonation in the storage tube.

These accidents were further grouped into bounding scenarios for

s

Pressurization (caused by steam from the residual water overheating,

mechanical damage and internal pressurization for each of the container types.

4.2 CONSEQUENCES OF ACCIDENTS

4.2.1 Consequence Analysis

Radiological inhalation dose consequences for each accident analyzed are based on the following.

Mass of material available for release Airborne release fraction (ARF) and (RF) Dose conversion factors

Duration of exposure Breathing rates.

Atmospheric dispersion of airborne particles

The radiological dose to a maximum receptor of interest is typically determined by using the following equation.

X D = M x - x R x C Q

where D = Dose (rem) M = Mass of material released (9)

R = Breathing rate (m3/s) C = Dose per unit respirable radioactive material inhaled (rem/g)

The quantity of respirable material released (M) is determined by the

x / Q = Time-integrated atmospheric dispersion coefficient (s/m3)

specific accident scenario. The atmospheric dispersion coefficient ( x / Q ) is based on specific release conditions (e.g., ground level or elevated, long or short duration) and the receptor's distance from the release. dispersion coefficient is the time-integrated normalized air concentration at the receptor's location.

12

The atmospheric

The coefficient represents the dilution of an


~

Onsi t e (100 rn E)

Hanfard S i t e boundary (18.3 krn E)

Near r i v e r bank (15.0 kn E )

a i r b o r n e contaminant caused by atmospheric m ix ing and tu rbu lence . values f o r these analyses have been generated us ing t h e GXQ computer program (Hey 1993a and Hey 1993b). determine o n s i t e and o f f s i t e consequences.

t h e f u l l range o f compass d i r e c t i o n s . For these analyses, t h e worst-case meteorology (windspeed and d i r e c t i o n ) f o r t he Hanford S i t e was used t o determine t h e l o c a t i o n a t t h e S i t e boundary where the maximum doses would occur. doses a t 100 m (328 ft).

A l l x/Q Table 4-2 con ta ins t h e x/Q values used t o

Values o f x/Q were c a l c u l a t e d f o r va r ious d i s tances f rom t h e re lease i n

The same approach was used t o determine t h e d i r e c t i o n o f t h e maximum

4.18 E-05 (230 m Y)

1.32 E-05 1.11 E-05 7.14 E-06

1.96 E-05 1.55 E-05 8.31 E-06

3.41 E-02 1.13 E-02

Table 4-2. Atmospheric D ispe rs ion C o e f f i c i e n t s Used i n Acc ident Analyses f o r t h e CSB.

X / Q Point source (s/m’)

Y i thout p l m Receptor Location Descr ipt ion meander, without I mee%!,’LZout I f a n a t 5.66 m’ls) (One

stack stack

For c a l c u l a t i o n o f o f f s i t e re leases, values o f x/Q were c a l c u l a t e d f o r a recep to r i n t h e h ighes t dose d i r e c t i o n a t t he Hanford S i t e boundary and f o r a maximum dose r e c e p t o r a t t h e Columbia R ive r t o t h e east and a proposed boundary a t Highway 240 t o the west. S i t e boundary was found t o be a t a l o c a t i o n 18.3 km (11.3 m i ) east o f t h e CSB. The maximum o f f s i t e recep to r a t t h e near r i v e r boundary 15 km (9.3 m i ) east o f t he f a c i l i t y was found t o be more l i m i t i n g than one a t Highway 240. t he case even though Highway 240 i s c l o s e r t o t h e f a c i l i t y . o n s i t e (100 m [328 ft]) recep to r was east o f t he f a c i l i t y f o r t h e ground- level re1 ease.

The maximum o f f s i t e recep to r f o r t he

Th is i s The maximum

Cor rec t i on t o t h e x/Q may be taken f o r b u i l d i n g wake e f f e c t s i f t h e acc ident does n o t change the b u i l d i n g c o n f i g u r a t i o n . i n t h e ana lys i s , none o f t he acc idents analyzed i n t h i s document used b u i l d i n g wake e f f e c t c o r r e c t i o n s . re leases assumed t o occur d u r i n g n e u t r a l o r s t a b l e atmospheric c o n d i t i o n s w i t h windspeeds l e s s than 6 m/s (19.7 f t / s ) and w i t h re lease t imes g r e a t e r than 1 hour.

To p rov ide conservat ism

Cor rec t i on f o r plume meander i s c r e d i t e d f o r ground

The b rea th ing r a t e ( R ) depends on a c t i v i t y f a c t o r s and exposure du ra t i on . A l l acc idents analyzed assume 8-hour exposure ( t h e d u r a t i o n o f a t y p i c a l work ing s h i f t ) t o t h e o n s i t e recep to r and 24-hour exposure t o t h e o f f s i t e recep to r . It i s assumed t h a t w i t h i n one day ’s t ime, emergency

13


measures will have caused evacuation of offsite areas affected by the accident. rate given in Reference Man: Characteristics (ICRP 1975). Its value is 3.3 x m3/s. The dose per unit material inhaled (C) is the value for the total committed effective dose equivalent.

inhalation of radioactive material. Dose contributions from the submersion pathway were not calculated and are assumed to be negligible with respect to the total dose for the radionuclides of interest. also expected to be negligible, therefore, the dose from groundshine and submersion are not included in the radiological dose calculations.

DOE, state, and federal emergency preparedness plans limit ingestion of contaminated food in the event of an accident. Emergency Response PI an (DOE-RL 1995), governs emergency response for a1 1 Hanford Site facilities. The primary determinant of exposure from the ingestion pathway is the effectiveness of public health measures (i.e., interdiction) rather than the severity of the accident itself. pathway, if it occurs, is a relatively slow-to-develop pathway and is not considered an immediate threat to an exposed population in the same sense as the inhalation pathway.

Therefore, the breathing rate used is the acute, light activity Anatomical, Physiological, and Metabolic

The major radiation exposure pathway for the identified accidents is

Doses from groundshine are

Potential doses from the ingestion pathway are not considered because

DOE/RL-94-02, Hanford

The ingestion

4.2.2 Radionuclide Inventories

cesium ion exchange resin containers (cesium containers) are described in Section 3.0.

The radionuclide inventories of the solidified HLW glass canisters and

4.2.3 Accident Scenarios

4.2.3.1 Multiple Glass Canister Failure during Handling-HHM Drop. Dropping a glass canister from the MHM into a tube that already contains one or more glass canisters is an operational accident caused by human error or equipment failure. catastrophic failure o f a glass canister would be shearing the container because of inadvertently closing the MHM lower shielding doors during the process of raising or lowering the canister. The accident is classified as a containment failure.

Other initiating events that could potentially result in

Scenario Development. A MHM failure or human error during raising or lowering operations results in dropping a glass canister into a tube that already contains one or more canisters. The storage tube is not actively ventilated, although it will be vented via a cartridge filter to the operating area. vault. This natural convection flow i s around the exterior of the tube, and is then exhausted through the CSB stack. The MHM will be in place over the tube when the drop occurs and the MHM tube seal is designed to maintain a seal

14

Cooling is provided by natural convection that occurs through the


for the tube. flat, essentia radionuclides.

The canister is designed to withstand a drop of 7 m onto a

Since a tube can hold two large canisters or three small lly unyielding surface without breaching or dispersing

canisters and since the material at risk is larger for two large canisters, it is assumed that one large canister is already in the tube. Each large canister is 4.5 m in height and with one already in the tube, the height of the drop is approximately 12.8 m - 4.5 m = 8.3 in. The canister is designed to not fail catastrophically if it is dropped with an impact absorber in place at the bottom of the tube and with an impact absorber above the existing canister in the tube. For this scenario, it is assumed the impact absorbers are not in place, so the dropped canister is assumed to rupture along with the one already in the tube, releasing airborne particulate from the shattered glass within. The tube is initially not assumed to maintain its containment function so that releases from the canisters enter the vault area or the CSB area. the building through breaches/unfiltered leak paths in the system.

Particles suspended in the MHM-tube confinement are assumed to enter

Source Term Analysis. The inventory at risk is the total inventory of two large canisters, or 5,470 kg of glass. exceeds the 7 m the canisters are designed. Farnsworth et al. (1988) performed drop tests o f simulated HWVP glass from a 9 m (30 ft) height. disassembling the canister Farnsworth et al. determined that 0.10 wt% of the waste form had fractured into fines t74 pm, and 0.014 wt% had fractured into fines 4 0 pm. 0.021 wt% and 0.004 wt% respectively. Farnsworth et al. concluded that the fines measured in the canister dropped 0.3 m (1 ft) were primarily created before impact, because of thermal stresses induced during normal cooling of the glass. Although the drop height in this case is slightly less than the 9 m tests, it is assumed the ARF times the RF (ARF x RF) for this scenario is 1.4 x With 5,470 kg of %lass for the !YO large canisters and with the canisters breaching, 5.47 x 10 g x 1.4 x 10 = 766 g are released as respirable within the tube. No credit is taken for any source term reduction because of agglomeration, gravitational settling, and plateout on the surfaces.

The drop height is 9.5 m, which

Upon

A drop of the canister from 0.3 m (1 ft), the values were

Unmitigated Consequence Analysis. The released source term is used to determine the radiological doses to the onsite and offsite receptors. following assumptions are used to determine the radiological consequences to human receptors.

The

Only short-term exposure effects resulting from inhalation are evaluated. .

. The values of x/Q for the selected receptor locations are those given in Table 4-2 for releases without plume meander.

The dose per unit of respirable radioactive material inhaled is the total given in Table 3-2 (3.3 x lo5 rem/g).

Using the relationship of: The unmitigated onsite dose is calculated as:

15


P o i n t Source Without Stack o r P1 ume Meander

( s / m v5 1 mSv (rem)

Receptor Loca t ion EDE

Ons i te 2.8Et04

Hanford S i t e boundary .32E-05 l . l E t 0 1

Near r i v e r bank 1.6E+01

(100 m E) 3'41E-02 (2.8E+03 rem)

(18.3 km E) ( l . l E t 0 0 rem)

(15.0 km E ) 1*96E-05 (1.6Et00 rem)


Guide1 i n e mSv (rem)

5.OEt01 (5.OEt00 rem)

5.OE+OO (5.OE-01 rem)

5.OE+OO (5.OE-01 rem)

Doseonsite = (766 g)(3.41 x s/m3)(3.3 x m3/s)(3.3 x l o 5 rem/g) =

2,845 rem

The unmi t i ga ted doses a re summarized i n Table 4-3.

Table 4-3. Unmi t igated Dose Consequences f o r a Dropped Glass Can is te r Onto Another Can is te r .

EDE = e f fec t i ve dose equivalent.

M i t i g a t e d Consequence Analys is . With t h e tube remain ing i n t a c t and t h e MHM tube seal i n p lace a t t h e t o p o f t h e tube, o n l y a small $mount o f leakage i s poss ib le . The i n t e r i o r volume y f an empty tube i s 4.73 m and t h e i n t e r n a l volume o f t h e MHM c a v i t y i s 1.50 m . occupies approx imate ly 1.15 m3. r u p t u r e d c a n i s t e r s i s approx imate ly 0.12 m3. Therefore, t h e t o t a l v o i d volume w i t h i n t h e tube assuming no impact l i m i t e r s i n p lace i s approx imate ly 4.73 m3 t 1.50 m3 - (2)(1.15 m3) - (2)(0.12 m3) = 3.7 in3. The breached c a n i s t e r s and broken g l a s s are a t t he bottom o f t h e tube a f t e r t h e drop. p a r t i c l e s l e a k i n g from t h e c a n i s t e r i n t o the tube would m ig ra te ve ry s l o w l y t o l e a k paths above t h e opera t i ng deck l e v e l . It i s reasonable t o assume t h a t o n l y a f r a c t i o n o f t h e p a r t i c l e s w i l l remain a i rbo rne l o n g enough t o t r a v e l t h e d i s tance t o t h e a v a i l a b l e l e a k paths, as t h e r e i s no a i r f l o w w i t h i n t h e tube. P a r t i c l e s w i l l be removed through agglomeration, g r a v i t a t i o n a l s e t t l i n g , and p l a t e o u t on sur faces. A maximum quas i - s tab le concen t ra t i on f o r r e s p i r a b l e p a r t i c l e s i n a i r f o l l o w i n g mechanical d i s tu rbance and t ime f o r s e t t l i n g i s g i ven as 100 mg r e s p i r a b l e pa r t i c l es /m3 (NUREG/CR-2651, pp 2.58 and 2.66 [Nuclear Regulatory Commission [NRC] 19821). The e n t i r e gas v o i d volume o f t h e tube i s assumed t o l eave t h e confinement and e n t e r t h e

The g lass i n each l a r g e g l a s s c a n i s t e r The volume occupied by each o f t h e l a r g e

Therefore,

16


environment. this volume is therefore 3.7 m x 100 mg/m = 370 mg. mitigated respirable source term released to the operating area. mitigated dose is calculated as:

With a void voluye of 3.7 m3j the respirable particulate mass in This is taken to be the

The onsite

Doseonsite = (0.370 g)(3.41 x s/m3)(3.3 x m3/s)(3.3 x lo5 rem/g) =

1.37 rem

The mitigated doses are summarized in Table 4-4.

Table 4-4. Mitigated Dose Consequences for a Dropped Glass Canister Onto Another Canister.

Point Source Without Stack or P1 ume Meander Guideline

Receptor Location mSv (rem)

mSv (rem)

Onsite 3.41E-02 (100 m E)

I I

Hanford Site boundary 5.3E-03 5.OE+OO .32E-05 (18.3 km E) (5.3E-04 rem) (5.OE-01 rem) I Near river bank 7.9E-03 5.OE+OO (15.0 km E) I I (7.9E-04 rem) I (5.OE-01 rem)

~~

Note:

EDE = effective dose equivalent

Guidelines are from UHC-CM-4-46, Nonreactor FaciLitv Safety Analysis Manual, Vestinghouse Hanford Conpany.

Comparison to Guidelines. If this accident is not mitigated or prevented by designed safety features, the radiological dose consequences exceed the onsite and offsite guidelines, assuming a probability of one for the accident. an early stage, no attempt is made to determine initiating event frequencies. Regardless of the frequency of the event, if still found to be credible in subsequent analyses, a structure, system of component would have to be designated as a least safety-significant because of the onsite consequences. (See Section 4.3 for safety-class and safety-significant criteria). probability of one for the event, a structure, system or component would need to be designated as safety-class. The mitigated consequences taking credit for the tube integrity and MHM tube seal, are within the onsite and offsite guide1 ines. class, as in case of a drop of a canister, the impact absorbers would prevent a breach. As the tube and MHM tube seal contribute significantly to defense- in-depth, they should be designated as safety-significant. To prevent an unfiltered release because of premature movement of the MHM while the glass canister has been partially lowered into the tube, the MHM interlocks and MHM/ valve turntable interlocks should be designated as safety-class. Also, the

As the design of the CSB and the operations envisioned are at

At a

Therefore, the impact absorbers should be designated as safety-

17


MHM should be seismically restrained to prevent canister shear during a seismic event when insertion/removal is taking place.

4.2.3.2 similar accident as above, but one in which the tube is initially empty. Dropping a glass canister from the MHM into a tube is an operational accident caused by human error or equipment failure. containment failure.

A MHM failure or human error during raising or lowering operations results in dropping a glass canister into an empty tube. The storage tube is not actively ventilated, although it will be vented via a cartridge filter to the operating area. convection that occurs through the vault. around the exterior of the tube, and is then exhausted through the CSB stack. The MHM will be in place over the tube when the drop occurs and the MHM tube seal is designed to maintain a seal for the tube. The glass canister could be sheared during MHM handling because of the lower shielding doors if the MHM inadvertently closed on the canister when it is partially in the MHM or by a bending moment applied to the canister because of differential movement between the MHM and the storage tube when the canister is partially in the tube and partially in the MHM. portion of the glass in the canister, the bounding scenario with the greater release of fines is the drop of the canister.

The canister is designed to withstand a drop of 7 m onto a flat, essentially unyielding surface without breaching or dispersing radionuclides. The tube is designed to hold two large canisters, or three small canisters. Since a large canister contains a larger quantity o f glass, the accident scenario assumes a large glass canister is dropped. The large canister is 4.5 m in height and the height of the drop is approximately 12.8 m. The impact absorber at the bottom of the tube is designed to prevent a catastrophic canister failure. For this scenario, it is assumed the impact absorber is not in place, so the dropped canister is assumed to rupture and release airborne particulate from the shattered glass. not assumed to maintain its containment function so that releases from the canister enters the vault area or the CSB area. Particles suspended in the MHM-tube confinement are assumed to enter the building through breaches/unfiltered leak paths in the system.

The inventory at risk is the total inventory of one large canister, or 2,735 kg of glass. considerably exceeds the 7 m the canisters are designed. Since this drop height exceeds the Farnsworth et al. (1988) drop tests of simulated HWVP glass from a 9 m (30 ft) height, a higher ARF could be anticipated for a 12.8 m drop. To estimate the amount of respirable particles produced in a drop of 12.8 m, the methodology of NUREG-1320, Nuclear Fuel Cycle Facility Accident Analysis Handbook (NRC 1988) is used. Figure 4.11 of NUREG-1320 provides a graph in which the mass fraction of glass and ceramics produced as a function of effective impact energy density from crush-impact. A large glass canister holds approximately 2,735 kg of glass. Including the weight of the canister,

Single Glass Canister Failure during Handling-MHM Drop. This is a

The accident is classified as a

Scenario Development.

Cooling is provided by natural This natural convection flow is

As the shear would be distributed over only a

The tube is initially

Source Term Analysis. The drop height is 12.8 m, which

18


t h e t o t a l weight i s 3,402 kg. c a n i s t e r i s c a l c u l a t e d as:

The crush-impact energy f o r t h e dropped

E = mgh = (3,402 kg)(9.8 m/s2)(12.8 m) = 4.27 x l o 5 J.

NUREG-1320 s t a t e s t h a t o n l y one-hal f o f t h e impact energy should be used t o es t ima te t h e re lease f r a c t i o n because o f rebound and t h e f a c t t h a t t h e subs t ra te w i l l absorb some o f t h e impact energy. conta ined i n a l a r g e c a n i s t e r i s 2,735 kg/ 2,640 kg/m3 = 1.03 m3, o r 1.03 x l o 6 cm3.

The volume o f g lass

Therefore, t h e e f f e c t i v e energy d e n s i t y i s c a l c u l a t e d as:

E / V = (4.27 x l o 5 J /2) (1.0 x l o 7 ergs/J) / (1 .03 x l o 6 cm3) =

2.1 x l o 6 ergs/cm3

F igu re 4.11 o f NUREG-1320 ass igns RFs o n l y down t o an e f f e c t i v e energy d e n s i t y o f 1.0 x l o 7 ergs/cm3, w i t h 0.025 w t % f o r t h i s energy dens i t y . While one cou ld e x t r a p o l a t e t h e s t r a i g h t l i n e s o f F igu re 4.11 down t o 2.1 x l o 6 ergs/cm3, no at tempt i s done so, and the more conserva t i ve re lease f r a c t i o n o f 2.5 x Apply ing t h i s re lease f r a c t i o n t o 2,735 kg o f g lass, 684 g are re leased from t h e rup tu red c a n i s t e r as r e s p i r a b l e . No c r e d i t i s taken f o r any source term r e d u c t i o n because o f agglomeration, g r a v i t a t i o n a l s e t t l i n g , and p l a t e o u t on the surfaces.

Unmi t igated Consequence Analys is . determine t h e r a d i o l o g i c a l doses t o t h e o n s i t e and o f f s i t e recep to rs . t h e r e l a t i o n s h i p o f :

i s used.

The re leased source term i s used t o Using

X D = M x - x R x C 9

The unmi t i ga ted o n s i t e dose i s c a l c u l a t e d as:

Doseonsite = (684 g)(3.41 x lo-‘ s/m3)(3.3 x m3/s)(3.3 x l o 5 rem/g) =

2,540 rem

The unmi t i ga ted doses are summarized i n Table 4-5.

19


Hanford S i t e boundary (18.3 km E )

Near r i v e r bank (15.0 km E )

Table 4-5. Unmi t igated Dose Consequences f o r a Dropped G1 ass Can is te r .

.32E-05 9.8+00 5.OE+OO (9.8E-01 rem) (5.OE-01 rem)

1.5E+01 5.OE+OO 1'96E-05 (1.5Et00 rem) (5.OE-01 rem)

P o i n t Source Wi thout Stack o r P1 ume Meander Guide1 i n e

Receptor Loca t ion mSv (rem)

mSv (rem) I I . . . . .

Ons i te 2.5E+04 5.OE+01 I100 m E) I 3'41E-02 I (2.5Et03 rem) I (5.OEt00 rem)

M i t i g a t e d Consequence Analys is . With t h e tube remain ing i n t a c t and t h e MHM tube seal i n p lace a t t h e t o p o f t h e tube, o n l y a small amount o f leakage i s poss ib le . The i n t e r i o r volume o f an empty tube i s 6.23 m3 and t h e volume o f g lass i s 1.15 m3. approx imate ly 0.12 m3. approx imate ly 6.23 m3 - 1.15 m3 - 0.12 m3 = 4.96 m3. The breached c a n i s t e r and broken g l a s s i s a t t h e bottom o f t h e tube a f t e r t h e drop. p a r t i c l e s l e a k i n g from t h e c a n i s t e r i n t o t h e tube would m ig ra te ve ry s l o w l y t o l e a k paths above t h e opera t i ng deck l e v e l . It i s reasonable t o assume t h a t o n l y a f r a c t i o n o f t h e p a r t i c l e s w i l l remain a i rbo rne l o n g enough t o t r a v e l t he d i s tance from t o the a v a i l a b l e l e a k paths, as t h e r e i s no a i r f l o w w i t h i n t h e tube. P a r t i c l e s w i l l be removed through agglomeration, g r a v i t a t i o n a l s e t t l i n g , and p l a t e o u t on sur faces. A maximum quas i - s tab le concen t ra t i on f o r r e s p i r a b l e p a r t i c l e s i n a i r f o l l o w i n g mechanical d i s tu rbance and t ime f o r s e t t l i n g i s g i ven as 100 mg r e s p i r a b l e pa r t i c l es /m3 (NUREG/CR-2651, pp 2.58 and 2.66 [NRC 19821). The e n t i r e gas v o i d volume o f t h e tube i s assumed t o leave t h e confinement and e n t e r t h e environment. Wi th a v o i d volume o f 4.96 m3, t h e r e s p i r a b l e p a r t i c u l a t e mass i n t h i s volume i s t h e r e f o r e 4.96 m3 x 100 mg/m3 = 496 mg. Th is i s taken t o be t h e m i t i g a t e d r e s p i r a b l e source term re leased t o t h e opera t i ng area.

Doseonsite = (0.496 g ) (3 .41 x

The volume occupied by t h e l a r g e r u p t u r e d c a n i s t e r i s Therefore, t h e t o t a l v o i d volume w i t h i n t h e tube i s

Therefore,

The o n s i t e m i t i g a t e d dose i s c a l c u l a t e d as:

s/m3)(3.3 x

1.84 rem

m3/s)(3.3 x IO5 rem/g) =

The m i t i g a t e d doses are summarized i n Table 4-6.

20


Hanford Site boundary (18.3 km E)

Near river bank (15.0 km E)

Table 4-6. Mitigated Dose Consequences for a Dropped Glass Canister.

1.32E-05 7.1E-03 5.OE+OO (7.1E-04 rem) (5.OE-01 rem)

l.lE-02 5.OE+00 1'96E-05 (1.lE-03 rem) (5.OE-01 rem)

Point Source Without Stack or P1 ume Meander Guideline


mSv (rem)

I Onsite 1.8Et01 5.OE+01 (100 m E) I 3'41E-02 I (1.8E-00 rem) I (5.OE+00 rem)

21


Scenar io Development. It i s n o t a n t i c i p a t e d t h a t a g l a s s c a n i s t e r w i l l be breached due t o over-pressure caused by l o s s o f c o o l i n g f o r t h e g l a s s c a n i s t e r . However, i f c o r r o s i o n were t o occur, o r an undetected weld d e f e c t e x i s t s , t h e c a n i s t e r cou ld be breached because o f heat bu i l d -up i n s i d e t h e c a n i s t e r caused by t h e r a d i o a c t i v e decay hea t ing o f t h e g lass. For t h i s scenario, i t i s assumed t h a t one had occurred ( c o r r o s i o n and weld d e f e c t ) a f t e r severa l years o f s torage and t h e c a n i s t e r d i d over-pressur ize, and a re lease o f f i n e s w i t h i n t h e c a n i s t e r resu l ted . would no rma l l y be i n p lace, which i s designed t o ma in ta in a seal f o r t h e tube. The tube i s i n i t i a l l y n o t assumed t o ma in ta in i t s containment f u n c t i o n so t h a t t h e re lease f rom t h e c a n i s t e r en te rs t h e v a u l t area o r t h e CSB opera t i ng area. P a r t i c l e s suspended i n the tube confinement area are assumed t o e n t e r t h e b u i l d i n g through b reaches /un f i l t e red l e a k paths i n t h e system.

Source Term Analys is . As discussed i n Sect ion 4.2.3.1, Source Term Analys is , t h e r e e x i s t t h e p o s s i b i l i t y t h a t d u r i n g cool-down o f t h e g lass, f i n e s can be produced due t o thermal s t resses induced d u r i n g normal coo l i ng . Farnsworth e t a l . a t t r i b u t e d 0.004 w t % o f p a r t i c l e s l e s s than 10 pm diameter as p r i m a r i l y c rea ted be fo re t h e impact, because o f t h e coo l i ng . o f $ lass f o r a l a r g e c a n i s t e r , t h i s would rep resen t 2.735 x IO6 g x 4.0 x 10- = 109.4 g o f g lass p a r t i c l e s ( i n powder form) l e s s than 10 pm t h a t cou ld be a v a i l a b l e f o r t h e p ressu r i zed re lease. It i s assumed t h a t these p a r t i c l e s t h a t are l e s s than 10 pm r e s i d e on t h e o u t e r sur faces o f t h e g lass . mass o f 2,735 kg o f g lass and d e n s i t y o f 2,640 kg/rn3, t h e h e i g h t o f t h e g lass i n s i d e t h e g lass c a n i s t e r i s 378.2 cm. g lass c y l i n d e r , t h e r a t i o o f t h e t o p sur face area t o t h e t o t a l su r face area o f t h e g lass c y l i n d e r i s 0.036. Therefore, 0.036 x 109.4 g = 3.94 g o f p a r t i c l e s l e s s than 10 pm would r e s i d e on t h e t o p su r face o f t h e g lass .

NUREG/CR-3093, Aerosols Generated by Release o f P ressu r i zed Powders and S o l u t i o n s i n S t a t i c A i r ( S u t t e r 1983) i n v e s t i g a t e d t h e p ressu r i zed re leasee o f DUO and TiO, powders from pressures from 50 p s i g t o 1000 ps ig . The average l a r g e s t f r a c t i o n made a i rbo rne from t h e experiments was f o r T i 0 powder a t 500 p s i g and r e s u l t e d i n 24.1% being made a i rbo rne . Th is ARF o f 0.$41 i s used and w i t h 3.94 g o f g lass p a r t i c l e s a t r i s k on t h e upper su r face o f t h e g lass c y l i n d e r , 0.95 g o f g lass t h a t i s l e s s than 10 pm i s expected t o be re leased and made a i rbo rne .

The re leased source te rm i s used t o determine t h e r a d i o l o g i c a l doses t o t h e o n s i t e and o f f s i t e recep to rs . c r e d i t i s taken f o r any source term r e d u c t i o n because o f agglomeration, g r a v i t a t i o n a l s e t t l i n g , and p l a t e o u t on t h e sur faces. Using t h e r e l a t i o n s h i p

The s torage tube s h i e l d p l u g

Wi th 2,735 kg

Wi th a

Wi th a r a d i u s o f 29.53 cm f o r t h e

Unmit igated Consequence Analys is . No

X D = M x - x R x C 9


Doseonsite = (0.95 g)(3.41 x s/m3)(3.3 x m3/s)(3.3 x lo5 rem/g) =

22


Point Source Without Stack or P1 ume Meander

Receptor Location I

EDE mSv (rem)

Guideline mSv (rem)

Onsite 3.5E+01 5.OE+01 I 3'41E-02 I (3.5E+00 rem) I (5.OE+00 rem)

Hanford Site boundary (18.3 km E)

Near river bank (15.0 km E )

1.32E-05 1.4E-02 5.OE+OO (1.4E-03 rem) (5.OE-01 rem)

2.OE-02 5.OE+OO 1*96E-05 (2.OE-03 rem) (5.OE-01 rem)

~

Note: Guidelines are f r m YHC-CU-4-46. Nonreactor Facilitv Safetv Analvsis Manual, Uestinghouse Hanford Company.


Mitigated Consequence Analysis. With the tube remaining intact and the storage tube shield plug in place at the top of the tube, only a small amount of leakage is possible. Consequence Analysis, for the dropped glass canister applies for this scenario. The tube is assumed to Sontain a single large glass canister with the lower impact absorber (4.21 m volume) in place. With a maximum quasi- stable concentration of 100 mg/m3, the mitigated consequences are slightly less than those for the dropped large glass canister, due to the reduced void volume (4.73 m3) with the shield plug in place and the absence of the MHM cavity. m3 and the mitigated doses are slightly less than those of Table 4-4.

for the onsite and offsite receptors are within the risk acceptance guidelines. guidelines and due to the uncertainties in the design and early analysis presented, it is prudent to designate the tube and storage tube shield plug as safety-signi ficant.

4.2.3.4 Cesium Container Failure. Dropping a cesium container during unloading or during placement into the storage tube is an operational accident potentially caused by human error or equipment failure which could result in a failure o f the cesium container. potentially result in catastrophic failure of a cesium container would be shearing the container because of inadvertently closing the MHM lower shielding doors during the process of raising or lowering the MCO or failure

The same discussion of Section 4.2.3.1, Mitigated

The total void volume is 4.73 m3 - 1.15 m3 - 0.12 m3 - 0.21 m3 = -3.25

Comparison to Guide1 ines. The mi tigated and unmitigated consequences

However, as the unmitigated dose is significantly close to the

Other initiating events that could

23


of the container because of external pressure caused by a hydrogen detonation in the storage tube.

materials handling could result in dropping a cesium container into the transport cask (during unloading) or into the storage tube. could cause the greatest damage to the cesium container would be into the storage tube (approximately 12.8 m). normally placed at the bottom o f the storage tube will prevent catastrophic failure of the cesium container. For this scenario it is assumed that, in addition to the drop of the container, the impact absorber has not been placed in the bottom of the storage tube. It is also conservatively assumed that, if the impact absorber is not in place, the storage tube would be breached by the drop of a cesium container. proposed is much less than the weight o f glass canisters, it might not breach the tube if dropped.)

The storage tube is not actively ventilated, although it will be vented via a cartridge filter to the operating area. actively ventilated via a cartridge filter. For the bounding consequences of the accident, it is assumed that the storage tube and/or the MHM storage tube seal provides an unfiltered leak path for the respirable particles released from the cesium container.

lower shielding doors o f the MHM being inadvertently closed on the container when it is partially in the MHM or by a bending moment applied to the container because of differential movement between the MHM and the storage tube when the container is partially in the tube and partially in the MHM. The release from such an accident is assumed to be the same as the unmitigated release from the container drop with a direct leak path into the CSB operating area.

The final scenario with similar unmitigated and mitigated consequences to the container drop is the failure o f the cesium container because of external pressure caused by a hydrogen detonation in the storage tube. assumed that this accident would also fail the storage tube.) The hydrogen buildup could occur because o f two extremely unlikely scenarios. the radiolysis of water that had leaked into the storage tube. assumed that the hydrogen could build up to a detonatable concentration, detonate, and cause the cesium container and the storage tube to breach. scenario is extremely unlikely, because no internal sources of water are provided in the vault design. The second, even more unlikely scenario, is that the ion exchange resin would contain water, which would be subject to radiolysis. and be detonated in the storage tube.

Source Term Analysis. inventory of a cesium container, 3.2 x 10 Ci o f 137Cs. is assumed that the resin has been dried; therefore it is assumed that no ion exchange liquid feed will be released.) It is assumed that 0.1% of the resin (84 kg x 0.001 = 84 g) is in respirable particles.

Scenario Development. A MHM hoist failure or human error during

The drop that

It is assumed that the impact absorber

(Because the weight of the cesium container as

The MHM is assumed to be

The cesium container could be sheared during MHM handling because of the

(It is

One would be It is then

This

The hydrogen generated would then leak from the cesium container

The inventory at risk is the maximum radionuclide (For this accident, it

For the bounding case, all

24


of t h i s m a t e r i a l i s assumed t o be re leased as r e s p i r a b l e . unmi t i ga ted acc iden t ana lys i s envelopes a l l acc iden t consequences i n v o l v i n g cesium con ta ine rs i n a s i n g l e s torage tube, i n c l u d i n g a l l drop acc idents and a l l acc iden ts t h a t i n v o l v e ca tas t roph ic f a i l u r e o f a s i n g l e c a n i s t e r . t h e i nven to ry o f cesium pos tu la ted f o r t h i s acc ident i s t h e maximum a l l owab le amount f o r a s i n g l e s torage tube, i t i s t h e r e f o r e t h e maximum amount o f m a t e r i a l t h a t cou ld be conta ined i n a s i n g l e cesium con ta ine r . Because up t o s i x cesium con ta ine rs would f i t i n t o a s torage tube, t h e amount o f cesium i n any combinat ion o f two con ta ine rs i nvo l ved i n a drop acc iden t would no rma l l y be expected t o be l e s s than t h e pos tu la ted amount.

r a d i o l o g i c a l doses t o t h e o n s i t e and o f f s i t e receptors . assumptions are used t o determine t h e r a d i o l o g i c a l consequences t o human recep to rs .

Note t h a t t h e

Because

Consequence Analys is . The re leased source term i s used t o determine the The f o l l o w i n g

.

. Only shor t - term exposure e f f e c t s r e s u l t i n g from i n h a l a t i o n are evaluated.

The values o f x / Q f o r t h e se lec ted r e c e p t o r l o c a t i o n s are those g i ven i n Table 4-2 f o r re leases w i t h o u t plume meander.

The dose e r u n i t o f r e s p i r a b l e r a d i o a c t i v e m a t e r i a l i n h a l e d i s 1.22 x 10 rem/g (3.2 x lo5 Ci / (84 kg)(3.19 x l o 4 rem/Ci). D

Using t h e r e l a t i o n s h i p



Doseongite = (84 g ) (3 .41 x s/m3)(3.3 x IO-& m3/s)( l .22 x IO5 rem/g) =

116 rem

The unmi t i ga ted doses are summarized i n Table 4-8.

25


P o i n t Source Without Stack o r P1 ume Meander

(slm x/Q3 ) mSv (rem)

Receptor Loca t ion EDE

Gu ide l i ne mSv (rem)

Ons i te

Note: Guidelines are from UHC-CM-4-46, Nonreactor F a c i l i t y Safety Analysis Manual, Vestinghouse

EDE = e f f e c t i v e dose equivalent

Hanford Company.

M i t i g a t e d Consequence Ana lys i s . I f t h e tube remained i n t a c t and t h e s torage tube s h i e l d p l u g o r MHM tube seal were i n p lace a t t h e t o p o f t h e tube, o n l y a small amount o f leakage woulg be poss ib le . The i n t e r i o r volume of an empty tube and MHM c a v i t y i s 6.23 m and t h a t o f t h e i o n exchange r e s i n con ta ine r i s 0.117 m3. approx imate ly 6.11 m3. a f t e r t h e drop. Therefore, p a r t i c l e s l e a k i n g from t h e con ta ine r i n t o t h e tube would m ig ra te ve ry s l o w l y t o l e a k paths above t h e opera t i ng deck l e v e l . I t i s reasonable t o assume t h a t on l y a f r a c t i o n o f t h e p a r t i c l e s w i l l remain a i rbo rne l o n g enough t o t r a v e l t h e 9.5 - 12.8 m d i s tance from t h e bottom o f t h e tube t o t h e a v a i l a b l e l e a k paths. agglomeration, g r a v i t a t i o n a l s e t t l i n g , and p l a t e o u t on sur faces. A maximum quas i - s tab le concen t ra t i on f o r r e s p i r a b l e p a r t i c l e s i n a i r f o l l o w i n g mechanical d i s tu rbance and t ime f o r s e t t l i n g i s g i ven as 100 mg r e s p i r a b l e pa r t i c l es /m3 (NUREG/CR-2651, pp 2.58 and 2.66 [NRC 19821). The e n t i r e gas v o i d volume o f t h e tube i s assumed t o leave t h e confinement and e n t e r t h e environment. i n t h i s volume i s t h e r e f o r e 6.11 m3 x 100 mg/m3 = 611 mg. t h e m i t i g a t e d r e s p i r a b l e source term re leased t o t h e opera t i ng area. o n s i t e m i t i g a t e d dose i s c a l c u l a t e d as:

Doseonsite = (0.611 g)(3.41 x

Therefore, t he t o t a l v o i d volume w i t h i n t h e tube i s The breached con ta ine r i s a t t h e bottom o f t h e tube

P a r t i c l e s w i l l be removed through

Wi th a v o i d volume o f 3.25 m3, t h e r e s p i r a b l e p a r t i c u l a t e mass Th is i s taken t o be

The

m3/s)( l .22 x l o 5 rem/g) = s/m3)(3.3 x

0.84 rem


26


Near river bank (15.0 km E)

Table 4-9. Mitigated Dose Consequences for a Dropped Glass Canister.

Point Source Without Stack or P1 ume Meander Guide1 ine


(slm 1 mSv (rem)

(100 m E) 3*41E-02 (8.4E-01 rem) (5.OEt00 rem) Onsite 8.4Et00 5.OE+01

4.8E-03 5. OEtOO 1*96E-05 (4.8E-04 rem) (5.OE-01 rem)

Hanford Site boundary 3.2E-03 5.OE+OO I (18.3 km E) I 1'32E-05 I (3.2E-04 rem) I (5.OE-01 rem)

Comparison to Guidelines. If this accident is not mitigated or prevented by designed safety features, the radiological dose consequence exceed the onsite guidelines, assuming a probability of one for the accident. As the design of the CSB and the operations envisioned are at an early stage, no attempt is made to determine initiating event frequencies. Regardless of the frequency of the event, if still found to be credible in subsequent analyses, a structure, system of component would have to be designated as safety significant because of the onsite consequences. consequences, taking credit for the tube integrity and storage tube shield plug, are within the onsite and offsite guidelines. tube seal and MHM tube seal need to be designated as safety-significant for this accident. Furthermore, the impact absorber should be designated as safety-significant, because the impact absorber would prevent a breach of the storage tube (note that this equipment would be designated safety-class for accident scenarios involving the glass canisters). For accidents involving shearing a container, the preventive equipment is the set of interlocks on the MHM that prevent the lower shield doors from being shut while the container is partway out of the MHM and the set of interlocks that prevent MHM lateral movement during the lowering of or raising of cesium containers. equipment would be designated safety significant on the basis of this accident scenario. For the hydrogen detonation accident, administrative controls which prevent the buildup of hydrogen in the storage tube to a detonatable concentration should be developed. These include a surveillance routine that would reliably detect the presence of hydrogen or moisture in the storage tube (because of the in-leakage of water) and controls to assure that cesium ion exchange resin stored in the cesium containers is completely dried.

4.2.3.5 Cesium Container Pressurized Release. It is specified in the request for proposal (RFP) that the ion exchange resin in the cesium container will be dried to eliminate the possibility of gas generation or the presence of free liquids.

The mitigated

Therefore, the tube and

This

This section assumes that, because o f a failure of administrative

27


c o n t r o l s , a cesium con ta ine r con ta in ing a volume o f i o n exchange process feed equal t o t h e volume o f water t h a t i s present w i t h t h e i o n exchange r e s i n as rece ived f rom t h e vendor i s i n a d v e r t e n t l y sent t o t h e CSB f o r s torage. Th is cou ld r e s u l t i n e i t h e r an i n t e r n a l hydrogen de tona t ion o r a p ressu r i zed re lease because o f overheat ing t h e con ta ine r and genera t i ng steam. The consequences o f t h e two scenar ios are assumed t o be s i m i l a r .

c o n t r o l s , a cesium con ta ine r i s sent t o t h e CSB f o r s torage c o n t a i n i n g f u l l y loaded i o n exchange r e s i n t h a t has n o t been d r ied . I n a d d i t i o n t o t h e f u l l y loaded r e s i n , t h e con ta ine r has a volume o f i o n exchange feed (concen t ra t i ons of r a d i o n u c l i d e s as pe r Table 3-4), equal t o the volume o f t h e 10 weight percent water (8.4 kg) t h a t t he r e s i n would con ta in as rece ived from t h e vendor. It i s assumed t h a t , because o f a l o s s o f c o o l i n g i n e i t h e r t h e MHM o r the s torage tube, t h e con ta ine r f a i l s and steam w i t h en t ra ined rad ionuc l i des and r e s i n p a r t i c l e s i s d ischarged from t h e a breach i n t h e con ta ine r . The m a t e r i a l would then be d ischarged t o e i t h e r t h e s torage tube o r t h e MHM.

Scenar io Development. Because o f a breakdown o f a d m i n i s t r a t i v e

The s torage tube i s n o t a c t i v e l y v e n t i l a t e d , a l though i t w i l l be vented v i a a c a r t r i d g e f i l t e r t o t h e opera t i ng area. The MHM i s assumed t o be a c t i v e l y v e n t i l a t e d v i a a c a r t r i d g e f i l t e r . For t h e bounding consequences o f t h e acc ident , i t i s assumed t h a t t h e s torage tube and/or t h e MHM storage tube seal p rov ides an u n f i l t e r e d l e a k pa th f o r t h e r e s p i r a b l e p a r t i c l e s re leased from t h e cesium con ta ine r .

Source Term Analys is . The inven to ry a t r i s k i s t h e maximum r a d i o n u c l i d e i nven to ry o f a cesium con ta ine r , 3.2 x 10 C i o f 13'Cs and 8.4 l i t e r s o f i o n exchange feed. It i s assumed t h a t 0.1% o f t h e r e s i n (84 kg x 0.001 = 84 g) i s i n l e s s than t e n micron p a r t i c l e s . Per Su t te r , 1983, a p ressu r i zed re lease o f t i t a n i u m ox ide a t 500 p s i has an ARF o f 4.5 percent . Thus, t h e amount o f i o n exchange r e s i n assumed t o be re leased i n r e s p i r a b l e form i s 4.5 percent o f t h e r e s p i r a b l e amount (0.045 X 84 g = 3.78 9) . feed, a bounding ARF o f 1.0 x l o - ' and RF o f 0.7 f o r h i g h l y superheated l i q u i d are used, p e r DOE-HDBK-3010-94 (Mishima 1994), Sec t i on 3. The t o t a l amount o f i o n exchange feed re leased i s then 0.588 L (0.1 x 0.7 x 8.4 l i t e r ) .

The re leased source term i s used t o determine the r a d i o l o g i c a l doses t o the o n s i t e and o f f s i t e recep to rs . assumptions a re used t o determine t h e r a d i o l o g i c a l consequences t o human recep to rs .

For t h e re lease o f i o n exchange

Consequence Analys is . The f o l l o w i n g

Only shor t - term exposure e f f e c t s r e s u l t i n g f rom i n h a l a t i o n are evaluated.

The values o f x / Q f o r t h e se lec ted r e c e p t o r l o c a t i o n s are those g i ven i n Table 4-2 f o r re leases w i t h o u t plume meander.

The dose p e r u n i t o f r e s p i r a b l e r a d i o a c t i v e m a t e r i a l i nha led i s 1.22 x 10 rem/g (3.2 x l o 5 C i / (84 kg)(3.19 x l o 4 rem/Ci) f o r i o n exchange res ign . exchange feed i s 1.8 x l o 4 rem/L [6.8 x 10 rem/ga1/(3.79 L / g a l ) ] .

The dose pe r u n i t o f r e s p i r a b l e aerosol f rom i o n

28

Using t h e r e l a t i o n s h i p


X D = M x - x R x C 9

The unmi t i ga ted o n s i t e dose from t h e p ressu r i zed re lease o f r e s i n i s c a l c u l a t e d as:

Doseonsite = (3.78 g)(3.41 x l o - * s/m3)(3.3 x

5.19 rem

The unmi t i ga ted o n s i t e dose from t h e p ressu r i zed re lease o f i o n exchange feed i s c a l c u l a t e d as:

Doseonsite = (0.588 L)(3.41 x s/m3)(3.3 x

0.12 rem

The t o t a l unmi t i ga ted dose i s then 5.31 rem. summarized i n Table 4-10.

m3/s)( l .Z2 x l o 5 rem/g) =

m3/s)(1.79 x l o 4 rem/L) =

The unmi t i ga ted doses are

29


Po in t Source Without Stack o r P1 ume Meander

Y / O EDE Receptor Loca t ion

Gu ide l i ne mSv (rem)

I Ons i te 5.3Et01 5.OEt01 I 3*41E-02 I (5.3E+00 rem) I (5.OE+00 rem)

I Hanford S i t e boundary 2.OE-02 5. O E t O O I (18.3 km E) I 1*32E-05 I (2.OE-03 rem) I (5.OE-01 rem)

I I Near r i v e r bank 3.OE-02 5.OE+OO (15.0 km E) I 1'96E-05 I (3.OE-03 rem) I (5.OE-01 rem) Note: Guidelines are fram WHC-04-4-46. Nonreactor Facility Safetv Analysis Manual, Westinghouse

Hanford Company.


M i t i g a t e d Consequence Analys is . I f the re lease occurred f rom a c a n i s t e r i n a s torage tube, an even sma l le r re lease would occur. The i n t e r i o r volume o f an empty tube i s 4.73 m3 and t h a t o f t h e i o n exchanqe r e s i n con ta ine r i s 0.117 in3. The lower impact absorber volume i s -0.21 m . Therefore, t h e t o t a l v o i d volume w i t h i n the tube i s approx imate ly 4.40 m . P a r t i c l e s and aerosol l e a k i n g f rom t h e con ta ine r i n t o t h e tube would m ig ra te ve ry s l o w l y t o l e a k paths above t h e opera t i ng deck l e v e l . It i s reasonable t o assume t h a t o n l y a f r a c t i o n o f t h e p a r t i c l e s and aerosol w i l l remain a i rbo rne l o n g enough t o escape f rom t h e tube v i a t h e a v a i l a b l e leakpaths. P a r t i c l e s w i l l be removed through agglomeration, g r a v i t a t i o n a l s e t t l i n g , and p l a t e o u t on sur faces. A maximum quas i - s tab le concen t ra t i on f o r r e s p i r a b l e p a r t i c l e s i n a i r f o l l o w i n g mechanical d i s tu rbance and t ime f o r s e t t l i n g i s g i ven as 100 mg r e s p i r a b l e pa r t i c l es /m3 (NUREG/CR-2651, pp 2.58 and 2.66 [NRC 19821). The e n t i r e gas v o i d volume o f t h e tube i s assumed t o leavf t h e confinement and e n t e r t h e environment. i n t h i s volume i s t h e r e f o r e 4.40 m3 x 100 mg/m3 = 440 mg. Th is i s taken t o be t h e m i t i g a t e d r e s p i r a b l e source term re leased t o t h e opera t i ng area. The 440 mg cou ld be composed o f a combination o f i o n exchange r e s i n f i n e s w i t h a u n i t dose o f 1.22 x lo5 rem/g and t h e i o n exchange feed s o l u t i o n w i t h a u n i t dose o f 1.8 x 10' rem/g [ conse rva t i ve l y assuming a d e n s i t y o f 1,000 g / l i t e r , t he u n i t dose i s 1.8 x l o 4 r e m / l i t e r / (1,000 g / l i t e r ) = 1.8 rem/g]. I f it i s conserva t i ve l y assumed t h a t a l l t h e re leased m a t e r i a l i s i o n exchange r e s i n , t he o n s i t e m i t i g a t e d dose i s c a l c u l a t e d as:

Doseonsite = (0.440 g)(3.41 x s/m3)(3.3 x

0.60 rem

With a v o i d volume o f 4.40 m , t h e r e s p i r a b l e p a r t i c u l a t e mass

m3/s)(l.22 x l o 5 rem/g) =


30


Point Source Without Stack or P1 ume Meander

(s/m xfQ3 1 mSv (rem)

Receptor Location E D E

Guideline mSv (rem)

Onsite 6.OE+OO

(15.0 km E ) Note: Hanford Copany.


Guidelines are from UHC-CM-4-46, Nomeactor Facility Safety Analysis Manual, Vestinghouse

The doses from a pressurized release into an MHM would be less because of the lower volume of the MHM. Furthermore, the MHM will be equipped with a high-efficiency particulate air-filtered ventilation system. efficiency particulate air filters could withstand the pressurization, the release would be further mitigated.

If this accident is not mitigated or prevented by designed safety features, the radiological dose consequence could exceed the onsite guidelines, assuming a probability of one for the accident. When the design of the CSB and the operations envisioned are at an early stage, no attempt is made to determine initiating event frequencies. mitigated consequences, taking credit for the tube integrity and storage tube shield plug, are within the onsite and offsite guidelines. Therefore, the tube and tube seal need to be designated as safety-significant for this accident. Furthermore, the MHM confinement should be designated as safety- significant (note that this equipment would be designated safety class for accident scenarios involving the glass canisters). In addition, the preventive features for this scenario are the administrative control requiring removal of water from the resin before it is stored in the CSB and the cooling of the storage tubes and the MHM (for the overpressurization because of steam formation if water is present).

If the high

Comparison to Guidelines.

The

4.3 SAFETY CLASSIFICATION SUMMARY

Safety class determinations in this document are in accordance with the criteria and requirements of Section 9.0 of WHC-CM-4-46, Safe ty Analysis Manual. The two safety designations are: 1) safety-class, and 2) safety- significant. The selection of safety-class and safety-significant SSCs is based primarily on their particular importance to defense-in-depth. class S S C s prevent or mitigate releases to the public that would otherwise

Safety-

31


Safety Structures, Systems, and CDnpOnents (SSCs):

Prevent or mi t igate o f f s i t e publ ic exposure i n excess of 500 mrem (5 mSv) EDE. See Notes 1 and 2.

Place or maintain an operating process i n a safe condit ion that prevents or mit igates consequences t o the public i n excess 500 mrem EDE. See

exceed the offsite radiological risk guide1 ine, or prevent accidental nuclear criticality. radiological materials to onsite workers and toxic chemicals to the offsite public and onsite workers.

as compared to the current safety-class and safety-significant. comparisons are reproduced in Table 4-12.

Safety-significant SSCs prevent or mitigate releases of

Table 1 in Chapter 9 of WHC-CM-4-46 shows the previous SSC designations These

Previous SSC Neu SSC Designation. Designation

SC-1 safety class

SC-1 Safety class

sc-2

N/A

sc-3

Monitor the release of radioactive materials t o the e n v i r o m n t during and af ter accidents uhere the monitor 's output i n i t i a t e s Emergency Response Plan actions or operator actions t o place the operating process i n a safe condit ion per Cr i ter ion 2.

Maintain operating parameters u i t h i n the TSRs or OSRs that protect the publ ic per C r i t e r i a 1 or 2.

Maintain double contingency protect ion against an accidental nuclear c r i t i c a l i t y as defined i n UHC-CM-4-29. Nuclear C r i t i c a l i t y Safety.

Prevent or mi t igate onsite exposure t o rad io log ica l materials i n excess of 5 rem (50 mSv) EDE. See Notes 1 and 2.

Prevent or mi t igate tox ic chemical exposure t o u i t h i n the r i s k guidel ines of UHC-CM-4-46, Chapter 7.0. See Notes 2 and 4.

Place or maintain an operating process i n a safe condit ion that prevents or mit igates consequences that exceed C r i t e r i a 6 or 7.

Prevent or mit igate exposure i n excess of 5 rem EDE or an airborne concentration of tox ic material i n excess of the applicable chemical ERPC-2 l i m i t t o f a c i l i t y operators uho are r e l i e d on t o achieve the safe condit ion of C r i t e r i a 2 and 8.

Monitor the release of radioactive and/or hazardous materials t o the e n v i r o w n t during and a f t e r accidents h e r e the monitor's o u t p l t i n i t i a t e s Emergency Response Plan actions or operator actions t o place the operating process i n a safe condit ion per Cr i ter ion 8.

Maintain operating parameters u i t h i n the TSRs or OSRs that protect the ansite uorker per C r i t e r i o n 6.

Provide defense-in-depth prevention o r mi t igat ion of an uncontrol led release of radioactive and/or hazardous material deemed s ign i f icant . See Note 4 .

Prevent or mi t igate an acute f a t a l i t y t o a f a c i l i t y uorker or serious i n j u r y t o a group of uorkers, except uhere the SSCs are contro l led through an implemented i n s t i t u t i o n a l safety or rad iat ion protect ion

Safety s ign i f icant



Safety class

Safety class

Support the safety function of a safety class SSC. This includes contro l and monitoring functions (operating a i r , e l e c t r i c a l power, instrunentation, etc.). See Note 3.

sc-1 I safety class I

sc-1 Safety class

s ign i f icant

s ign i f icant


Safety Structures, Systems, and COmponents (SSCs):

Support the safety function of a safety s ign i f icant SSC. cont ro l and m n i t o r i n g functions (operating a i r , e l e c t r i c a l pouer, i n s t r w n t a t i o n , etc.).

This includes

See Notes 3 and 4.

Previous SSC Neu SSC Designation* Designation

SC-1 or SC-2 Safety s ign i f icant

As a result of the postulated accidents analyzed, Table 4-13 summarizes the structures, systems, and components (SSCs) that are candidates for safety- class or safety-significant designation. There are no new designations required because of the handling and storage of glass or cesium containers, beyond those envisioned for SNF storage. However, if a container were to rupture while in a tube, special recovery procedures may be required to stabilized the situation, and to recover and process the large quantity of fines, to allow final disposal of the material.

33


confinewent of releases

Table 4-13. P re l im ina ry L i s t o f CSB Safety-Class and S a f e t y - S i g n i f i c a n t sscs.

Over-pressurization during handling. Single canister f a i l u r e during handling. Mul t ip le f a i l u r e during handling.

Structure, System or c m o n e n t

Storage tubes

HHH

HHH in ter locks

HHHlvalve turntable in ter lock

Impact absorbers

Safety Function I Referenced Accident Analysis I clas~;:;::tion I Safety-signif icant

canister i s i n forces. intermediate w s i t i o n

Prevent catastrophic f a i l u r e of canister i f dropped to the b o t t m of the tube; protect confinement function

Prevent f a i l u r e of both canisters i f one i s d r o w d on another

Single canister f a i l u r e during handling.

Mul t ip le canister f a i l u r e during handling.

Safety-class

CBS = Canister Storage Building HHH = HCO handling machine SSC = structures, systems, and ccqmnents

5.0 HAZARD CLASSIFICATION EVALUATION

The hazard c l a s s i f i c a t i o n f o r t he CSB f o r SNF opera t i ons has been determined t o be a Hazard Category 2 f a c i l i t y us ing t h e guidance o f DOE-STD-1027-92, Hazard Ca tegor i za t i on and Acc ident Ana lys i s Techniques f o r Compliance with DOE Order 5480.23, Nuclear S a f e t y Ana lys i s Repor ts (DOE 1992). The f i n a l hazard category f o r t h e CSB was de r i ved i n WHC-SD-SNF-HC-007, Hazard Category A n a l y s i s f o r t he Can is te r Storage B u i l d i n g (Kummerer 1995). document was prepared i n accordance w i t h WHC-CM-4-46, Sec t i on 4.0, "Determin ing and Documenting F a c i l i t y Hazard Category", t h e WHC gu ide f o r implementing and documenting the requirements o f DOE Order 5480.23 and DOE-STD-1027-92. Th is hazard c l a s s i f i c a t i o n assignment f o r t h e CSB has been approved by DOE ( S e l l e r s 1996). The f i n a l hazard c a t e g o r i z a t i o n f o r t h e SNF was based upon a beyond des ign bas i s earthquake i n which t h e CSB opera t i ng area f l o o r co l l apsed and a l l 390 MCOs were breached. A t o t a l o f 2.56 x l o 6 g o f f u e l i n f i n e s form a t r i s k f o r re lease (no t a l l re leased) , and t h e sum of t he r a t i o s was determined t o be 19.0 when compared t o t h e Hazard Category 2 t h r e s h o l d q u a n t i t y (TQ) value.

That

34


For a preliminary hazard categorization of the glass and cesium canisters, the three cases o f Section 3.0 were examined. standard canisters, 1.1 x lo6 kg of glass would be the inventory. with 452 large canisters, 1.2 x lo6 \s of glass would be the inventory. there is a limit of 3.2 x lo5 Ci o f per vault, Case B provides a higher inventory than Case C. the sum of the ratios for a single large glass canister (Case B), using the methodology of DOE-STD-1027-92 for a preliminary hazard categorization and comparing the sum of the ratios to the Hazard Category 2 TQ value. For those isotopes not listed in Table A.l of the Standard, a determination was made whether the isotope was a beta-gamma emitter, mixed fission product, or alpha emitter, and the recommended TQ 2 value used from Table A.l for the particular isotope.

For Case A, with 678 For Case B,

Since

Table 5-1 provides Cs per storage tube and with 226 tubes

Table 5-1. Preliminary Hazard Categorization

35


sun of Ratios 4.391+01

CAT = category TQ = Threshold Quantity XX = ISOtope's TQ included uith parent.

36


From Table 5-1, one l a r g e c a n i s t e r represents 43.9 t imes t h e TQ va lue f o r a Hazard Category 2 f a c i l i t y , o r 87.8 t imes t h e TQ p e r tube. 452 l a r g e c a n i s t e r s i n a v a u l t , t h i s would rep resen t 19,843 t imes t h e Hazard Category TQ 2 va lue.

i n v e n t o r y f rom t h e s torage o f t h e g l a s s c a n i s t e r s and cesium con ta ine rs i n the CSB, compared t o t h a t o f SNF (15,700 t imes t h e TQ 2 va lue [WHC-SD-SNF-HC-OOI]), a1 though t h e f i n a l hazard c a t e g o r i z a t i o n o f t h e CSB may n o t change. To be a Hazard Category 1 f a c i l i t y r e q u i r e s t h e f a c i l i t y t o be a Category A r e a c t o r o r be des ignated as Hazard Category 1 by t h e PSO. Therefore, t h e f i n a l hazard category document f o r t h e CSB would need t o be up-dated t o i n c l u d e t h e g lass c a n i s t e r s and cesium con ta ine rs and submi t ted t o DOE f o r rev iew and approval .

Wi th

Th is rep resen ts a s u b s t a n t i a l increase i n q u a n t i t i e s o f r a d i o n u c l i d e

6.0 CESIUR CONTAINER EVALUATION

The cesium con ta ine r was evaluated f o r b e n e f i t s o f double-containment d u r i n g t r a n s f e r and s torage o f t he HLW.

6.1 CONTAINER TRANSPORTATION

The requi rements f o r s a f e l y packaging i o n exchange r e s i n contaminated w i t h cesium-137 f o r t r a n s p o r t are s i m i l a r t o s torage requirements. Packaging des ign f o r t r a n s p o r t i nvo l ves cons ide ra t i on o f t h e p o t e n t i a l dynamic s t resses t h a t w i l l a f f e c t t h e package. Furthermore, t r a n s p o r t ope ra t i ons i n v o l v e t h e hand l i ng o f t h e package i n the environment w i t h o u t t h e p r o t e c t i o n o f a f a c i l i t y . Because t r a n s p o r t a t i o n i s a t r a n s i e n t operat ion, however, many s torage requi rements f o r packaging w i l l be more r e s t r i c t i v e than t h e t r a n s p o r t requirements, i n c l u d i n g p r o v i s i o n s f o r p r e s s u r i z a t i o n .

The t r a n s p o r t s a f e t y o f t h e cesium con ta ine rs depends on t h e s h i e l d i n g and containment c a p a b i l i t i e s o f t h e t r a n s p o r t cask. The t r a n s p o r t cask w i l l p rov ide t h e pr imary containment boundary, so t h e b e n e f i t o f hav ing a doubly encapsulated cesium con ta ine r is minimal f o r t ranspor t . Because n e i t h e r o f t h e two cesium con ta ine rs w i l l be designed t o w i ths tand t r a n s p o r t acc ident cond i t i ons , t h e s a f e t y ana lys i s f o r packaging must r e l y on t h e assumption t h a t i f one con ta ine r undergoes ca tas t roph ic f a i l u r e , then t h e second con ta ine r w i l l a l s o f a i l . Furthermore, t h e consequences o f a l e a k i n t h e cesium con ta ine r d u r i n g t r a n s p o r t are i n s i g n i f i c a n t compared t o such an event i n the CSB, s i nce t h e t r a n s p o r t cask i s designed t o w i ths tand t h e e n t i r e spectrum o f t r a n s p o r t a t i o n acc idents . i n a d v e r t e n t l y contaminated, t h e cask can be r e l o c a t e d t o a s u i t a b l e f a c i l i t y f o r decontamination.

I f the i n t e r i o r o f t h e t r a n s p o r t cask i s

The s h i e l d i n g prov ided by the t r a n s p o r t cask must be s u f f i c i e n t t o ma in ta in t h e dose r a t e below 100 mr/hr on t h e cask su r face and 10 mr/hr a t 2 meters. Th is s h i e l d i n g must be s u f f i c i e n t t o prevent t h e dose r a t e a t 1 meter

37


from exceeding 1 rem/hr when the transport cask is subjected to credible transportation accident conditions.

The credible accidents were identified based on the extent of the shipping campaign and the design o f the transport cask. accidents include a collision or drop during transport.

Typical credible

The design o f the transport cask must be adequate to maintain the containment boundary when subjected to normal transport conditions and credible transportation accident conditions.

protected against the effects of decay heat and pressurization.

for the transport of the cesium container include the imposition of speed 1 imits for the transport vehicle and restrictions on transport during bad weather.

The requirements governing the packaging and transportation o f

The containment and shielding integrity of the transport cask must be

The additional operational controls that may be determined appropriate

radioactive material onsite are provided in WHC-CM-2-14, Hazardous Material Packaging and Shipping.

6.2 CONTAINER STORAGE

The following is an evaluation of the safety benefits and tradeoffs that may be realized by the double containment of the cesium ion exchange resin containers (cesium containers). greater safety for the handling o f the ion exchange resin during loading, transport, unloading, and storage; however, some important considerations must be addressed before a decision is rendered. It is assumed that a double containment design can be developed that will maintain the same outer dimensions and meet the hand1 ing requirements for the CSB.

A double-containment container may provide

The principal safety considerations of single verses double containment are related to the effects on the potential accidents that have been postulated in evaluating the safety o f the CSB handling and storage of the containers. There are five possible means of container breach postulated in the evaluation that could lead to a release of radioactive cesium from the cesi um containers: mechani cal damage, corrosion, pressurization, overheating and explosion. The impact o f double containment on each of these accident initiators is discussed below.

6.2.1 Mechanical Damage

Most of the postulated accidents that contribute to the overall risk associated with the handling and storage of the cesium containers are mechanical damage events. Damage events have been considered that result from dropping a container, mishandling a container, and collisions involving a container. There is also a postulated event that involves the interaction and

38


r u p t u r e o f two con ta ine rs . parameter t h a t w i l l e f f e c t t h e i r res i s tance t o damage, and t h e p r o v i s i o n o f double containment i s expected t o c o n t r i b u t e some amount o f a d d i t i o n a l s t r e n g t h t o t h e package. The double containment w i l l a l s o add weight t o the package and may compl icate hand1 i n g d u r i n g t r a n s p o r t and s torage operat ions. The se ismic performance o f a heav ie r con ta ine r may a l s o be d i f f e r e n t than a s i n g l e con ta ine r . gross weight near 300 kg (650 l b s ) , and a double containment package cou ld be double t h a t weight o r 600 kg (1,300 l b s ) . The mechanical f o rces associated w i t h most drop scenar ios o f con ta ine rs are expected t o r u p t u r e bo th confinements. Design o f a double containment package cou ld address each o f these cons ide ra t i ons , and perhaps a robus t secondary confinement cou ld be p rov ided t h a t would increase t h e p r o b a b i l i t y o f t h e con ta ine r s u r v i v i n g a drop acc iden t o r se ismic event. However, a few r i s k dominant acc iden ts would s t i l l be pos tu la ted f o r t h e double containment package and r e q u i r e cons ide ra t i on . The e v a l u a t i o n o f these acc idents w i l l l i k e l y show o n l y a s l i g h t r e d u c t i o n i n t h e o v e r a l l r i s k as a r e s u l t o f a l ower p r o b a b i l i t y o f damage and/or s l i g h t r e d u c t i o n i n t h e amount o f m a t e r i a l re leased.

The mechanical s t r e n g t h o f t h e con ta ine rs i s a key

The s i n g l e containment packages as s p e c i f i e d have a maximum

6.2.2 Cor ros ion

package contents and i t s m a t e r i a l s o f c o n s t r u c t i o n cou ld l e a d t o c o r r o s i o n and a breach o f t h e con ta ine r . s t a i n l e s s s t e e l and the package contents i s expected t o always be cesium i o n exchange r e s i n . Abnormal events i n c l u d e t h e r e t e n t i o n o f up t o 10% water t o t h e con ta ine rs , which cou ld l e a d t o o t h e r s a f e t y concerns b u t probably n o t co r ros ion . The a d d i t i o n o f ac ids o r o t h e r unan t i c ipa ted and incompat ib le m a t e r i a l s i s ext remely u n l i k e l y , b u t cou ld l ead t o a c o r r o s i o n f a i l u r e o f t h e con ta ine r .

It has been pos tu la ted t h a t m a t e r i a l i n c o m p a t i b i l i t i e s between t h e

The con ta ine r m a t e r i a l has been s p e c i f i e d t o be

The a p p l i c a t i o n o f a double confinement con ta ine r would a f f e c t t h e c o r r o s i o n performance o f t h e package, b u t probably o n l y t o de lay the f a i l u r e and re lease; t h a t i s , t o increase t h e t ime t o f a i l u r e by i nc reas ing t h e amount o f m a t e r i a l t h a t must corrode be fo re a re lease occurs. f a i l u r e i s n o t expected t o r e s u l t i n a ca tas t roph ic re lease o f m a t e r i a l , t h i s event i s l e s s severe than most pos tu la ted mechanical f a i l u r e s , and i s t h e r e f o r e n o t a r i s k dominant event. package f o r cesium i o n exchange r e s i n does n o t seem t o p rov ide much s a f e t y b e n e f i t f o r c o r r o s i o n res i s tance .

S ince a c o r r o s i o n

Development o f a double containment

6.2.3 P r e s s u r i z a t i o n

The presence o f water o r o t h e r unexpected and incompat ib le m a t e r i a l s w i t h i n a con ta ine r has been pos tu la ted as a means t o p ressu r i ze t h e package. The genera t i on o f gas o r steam cou ld r e s u l t i n t h e s t r e s s i n g o f t h e con ta ine r and eventual r u p t u r e and re lease o f t he r a d i o a c t i v e m a t e r i a l contents . The a p p l i c a t i o n o f double containment i s again expected t o p rov ide o n l y a small b e n e f i t .

39


The presence of a secondary containment during a pressurization failure of the primary containment may mitigate or prevent the release of material from the package, but the cause of the pressurization may still be present and, without detection and correction, would continue and also pressurize and fail the secondary containment. detection features or rupture disks, it is not likely that double containment will offer any significant reduction in the safety risk from pressurization of cesium containers.

Unless the package is designed with special

6.2.4 Overheating

The loss of convective cooling or the presence of an unexpectedly large heat source within a container has been postulated as a means to overheat a cesium container and lead to melting or a pressurized rupture. of double containment may help to delay the release from such failures and may even prevent container rupture; but if not specifically designed for such conditions, double containment may exacerbate the overheating problem by degrading heat transfer. Overheating accidents have been considered for the CSB and are a concern for the HLW glass canisters as well. contribution to the overall safety risk from overheating of cesium containers is small, the benefit of double containment for the prevention or mitigation of this event is deemed small.

The presence

Since the

6.2.5 Explosion

The generation of hydrogen by the radiolysis of water has been postulated as an initiator of either an internal or an external explosion leading to the release of radioactive materials. Water inside a cesium container is an abnormal event. Since the ion exchange resin is packaged in air and since the radiolysis of water produces both hydrogen and oxygen, the generation of hydrogen might lead to the development of a detonable mixture. The free volume inside the container is small and but no detailed analysis of the explosive power of the worst case mixture has been completed. Similarly, the presence of water in a storage tube has been postulated as a source of radiolytic hydrogen generation for both cesium containers and glass canisters. An explosion under these conditions might implode and rupture a container.

The application of a double container to these scenarios seems to offer some limited benefit. The detonation of a hydrogen mixture within the primary containment of a double containment package may result in a reduction in the amount of radioactive material that is postulated to be released. If strong enough, the secondary containment may even sustain the explosion and rupture of the inner primary containment. Conversely, the secondary containment may offer an added gas volume that would increase the available oxygen and lead to an even greater volume of detonable gas and a more severe accident. For externally initiated explosions, the secondary containment seems to offer two effects. and the water may decrease the rate of hydrogen generation by radiolysis, which may also decrease the probability of a detonation.

The increased shielding and distance between the radiation source

The secondary

40


containment may also offer greater resistance to damage of the inner containment by an external explosion.

6.2.6 Sumnary

radioactive material from the ion exchange containers was considered and the effects of both single and double containment packaging was evaluated to estimate the safety benefit. disadvantages of double containment for each accident type and assigns a subjective score to aid quantitative judgement.

cesium containers may be improved by the application of double containment, but the benefit may be small as compared to the already low risk of the container storage operation. Several safety-classlsafety-significant structures, systems and components are included in the design of the CSB to assure the low risk of SNF and HLW glass canister handling and storage. potential consequences from SNF accidents is roughly 4 times greater than that of cesium container accidents for the same source term mass (note that the unit gram dose for SNF is 4.3 x lo5 rem and for the cesium it is 1.1 x lo5 rem). The CSB safety risk is dominated by SNF and HLW glass canister handling. containers appears to be o f small benefit to reduce risk. A rigorous cost- benefit analysis and risk assessment may be necessary to provide definitive evidence, but the subjective evaluation provided here suggests that the added safety benefit of double containment is small and does not justify the required effort.

Each of the postulated accident initiators leading to the release of

Table 6-1 compares the expected benefits and

The overall safety risk associated with the handling and storage of

The

The addition of a double confinement container for cesium

4 1


TYPE ADVANTAGES DISADVANTAGES SCORE

Mechanical impact I + 2 ' - 2 = o I Mechanical strength, Increased ueight and

increased resistance handling d i f f i c u l t i e s t o sore impacts, might hold up release i f inner container f a i l e d I first I

Corrosion Increased t i n e t o Uithout detection, second +1,-1 = 0 f a i l , hold up of b n r r i e r u i l L l i k e l y also I release f a i l

Pressurization Larger t o t a l volune t o Uithout detection, second +1,-1 = 0 pressurize b a r r i e r u i l l l i k e l y also

f a i l

Overheating

Explosion

General

Increased heat Decreased thermal +1,-1 = 0

of release

explosions

detection before release

+1,-2 = -1

7.0 NRC EQUIVALENCY REVIEW

The purposes o f t h i s NRC eva lua t i on are: 1) a p r e l i m i n a r y rev iew o f t h e SNF P r o j e c t nuc lea r s a f e t y equiva lency NRC comparison t o t h e proposed HLW storage i n CSB, and 2) as a r e s u l t o f t h i s rev iew and e v a l u a t i o n and t h e conclus ions drawn, i d e n t i f y any ac t i ons f o r cons ide ra t i on t h a t may be necessary t o demonstrate nuc lea r s a f e t y equiva lence f o r t h e p r e l i m i n a r y design.

P r o j e c t Pa th Forward, Nuclear S a f e t y Equiva lency t o Comparable NRC-Licensed F a c i l i t i e s . The rev iew bas is was presented i n f o u r p r i n c i p a l subgroups: 1) pa r t s , sec t i ons and appendices determined t o be n o t app l i cab le ; 2) NRC r e g u l a t i o n s accommodated by WHC s i te -w ide programs; 3) NRC r e g u l a t i o n s p o t e n t i a l l y a p p l i c a b l e t o new SNF P r o j e c t f a c i l i t i e s ; and 4) NRC guidance p o t e n t i a l l y a p p l i c a b l e t o new SNF P r o j e c t f a c i l i t i e s .

The rev iew b a s i s was prov ided from WHC-SO-SNF-DB-002, Spent Nuc lea r Fue l

7.1 DESIGN AND CONSTRUCTION MEASURES

I n a d d i t i o n , NRC guidance t h a t may have d i r e c t a p p l i c a t i o n t o CSB P r o j e c t des ign and c o n s t r u c t i o n a c t i v i t i e s , was reviewed as a prudent s tep i n implementing t h e P o l i c y ' s ob jec t i ves . scope o f p o t e n t i a l l y a p p l i c a b l e NRC t e c h n i c a l requi rements are t h e des ign and

I n accordance w i t h t h e Po l i cy , t h e

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c o n s t r u c t i o n measures (as opposed t o a l s o i n c l u d i n g p reopera t i ona l o r ope ra t i ona l measures) mandated by t h e NRC regu la t i ons , f o r t h e purposes o f t h i s eva lua t i on , T i t l e 10 Code o f Federal Regulat ions, Pa r t s 0 through 199, r e v i s e d as o f January 1, 1995. These requirements which are i d e n t i f i e d i n WHC-SD-SNF-DB-003, Spent Nuc lea r Fue l P r o j e c t Path Forward, A d d i t i o n a l NRC Requirements, apply t o t h e CSB.

7.1.1 Resul ts

t o the CSB HLW t r a n s f e r and s torage. w i t h t h e except ion o f t h e des ign earthquake, which i s be ing implemented i n a manner t h a t es tab l i shes equivalence i n sa fe ty , as opposed t o d i r e c t equiva lence t o t h e r e g u l a t i o n . separate WHC document, WHC-SD-SNF-DB-004, Spent Nuc lea r Fue l P r o j e c t Seismic Design C r i t e r i a , Nuclear Regulatory Commission Equiva lency E v a l u a t i o n Report, which i s based upon: 1) t h e c u r r e n t CSB des ign meeting bo th t h e DOE and NRC s a f e t y goals , and 2) t h e o v e r a l l seismic r i s k i s lower than t h e r i s k associated w i t h l i censed nuc lea r power p lan ts .

Table 1, which i d e n t i f i e s t h e respons ib le d i s c i p l i n e ( s ) f o r implementat ion o f t h e requi rements (e.g., c i v i l s t r u c t u r a l and mechanical) t o h e l p f a c i l i t a t e va lue t o t h e end user . Where the t a b l e i nc ludes a reference(s) , t h i s does n o t imp ly t h a t t h e i d e n t i f i e d a d d i t i o n a l NRC requi rement(s) and r e l a t e d DOE requi rements n e c e s s a r i l y s a t i s f y the requirements o f t h e re fe rence (s ) , o r even t h a t t h e SNF P r o j e c t Path Forward must s a t i s f y t h e requi rements o f t h e re fe rence (s ) . and where t h e reader may r e f e r t o i n WHC-SD-SNF-DB-002, and NRC r e g u l a t i o n s and guidance f o r r e l a t e d i n fo rma t ion .

The NRC requi rements conso l i da ted i n t o 29 i tems i n WHC-SD-SNF-003, apply A l l i tems comply w i t h NRC r e g u l a t i o n s

The se ismic i ssue was presented i n d e t a i l i n a

The rev iew b a s i s i n f o r m a t i o n was prov ided from WHC-SD-SNF-DB-003,

I t i s s imply a n o t a t i o n as t o where t h e bas i c i ssue i s ra i sed ,

7.2 CONCLUSIONS

I n t h e comparison o f DOE and NRC requirements r e l e v a n t t o t h e des ign and c o n s t r u c t i o n o f CSB f a c i l i t y HLW t r a n s f e r and storage, a number o f d i f f e r e n c e s were i d e n t i f i e d i n WHC-SD-SNF-DB-002. Where d i f f e r e n c e s e x i s t , t h e NRC requi rements f o r t h e most p a r t were e i t h e r more conserva t i ve o r r e q u i r e t h e p repara t i on o f a d d i t i o n a l documentation (e.g., as p a r t o f a s a f e t y ana lys i s r e p o r t ) .

Considerat ion were i d e n t i f i e d f o r meaningful d i f f e r e n c e s where e x i s t i n g SNF P r o j e c t Path Forward requi rements need t o be enhanced. i d e n t i f i c a t i o n , a l l d i f f e rences , where Ac t i ons f o r Considerat ion were i d e n t i f i e d , were conso l i da ted i n t o WHC-SD-SNF-DB-002, Table 2. The more impor tan t o f these i tems, i .e., may have subs tan t i ve p r o j e c t impacts, are i d e n t i f i e d below w i t h t h e i r corresponding r e g u l a t i o n o r guidance number and a b r i e f cap t i on o f t he i tem. p r e l i m i n a r y s a f e t y c l a s s equipment r e q u i r e d f o r t h e CSB HLW t r a n s f e r s and

I n regards t o e s t a b l i s h i n g nuc lea r s a f e t y equivalency, Ac t i ons f o r

For ease o f

Th is l i s t was compiled based on t h e c u r r e n t

43


storage, and need for HLW transfer and storage to comply with the SNF Project requ i rement s .

Section 72.3, Definitions (definition of structures, systems and components [SSCs] important to safety embraces additional Safety Class 1 SSCS).

Section 72.92, Design basis external natural events, and 872.102, Geologic and seismological characteristics (more conservative seismic design criteria).

utilities, tornado and tornado missiles).

SRP 9.1.2, Spent Fuel Storage (use of a keff criticality safety limit of 0.95).

A detailed NRC equivalency review for the HLW storage should be

Section 72.122, Overall requirements (inclusion of sharing of common

conducted to verify these conclusions and establish the baseline NRC equivalency requirements for the addition of HLW storage in the CS8.

Also, equivalency was established for the most part by DOE orders and in some instances by WHC procedures and instructions in WHC-SD-SNF-DE-002. If the SNF Project takes steps to acquire waivers for some of the DOE orders used to establish nuclear safety equivalency, or if relevant DOE orders or WHC procedures and instructions are revised or canceled, a number of the conclusions regarding nuclear safety equivalency in this evaluation could be nul 1 ified.

Future detailed NRC equivalency review may provide additional requirements for handling and storage of HLW in the CSB.

8.0 SAFETY ANALYSIS REQUIREMENTS

Safety analysis requirements for the safe storage of HLW in the CSB will be determined by a safety review which will address similar questions as defined in WHC-IP-0842, TWRS Administration, Unreviewed Safety Questions, and the possible revision of the CSB facility hazard classification, WHC-SD-SNF-HC-007. Deviations from the safety basis will require additional safety analysis and revision to WHC-SD-SNF-RPT-004, Canister Storage Building Safety Analysis Report - Phase 2 Safety Analysis Documentation Supporting Canister Storage Building Subsurface Construction and Fabrication o f Tubes, Multi-Canister Overpack Handling Machine and Receiving Crane. This forms the basis for the extent of safety analysis for the storage of HLW in the CSB.

Based on the preliminary accident analysis results from Sections 4.0 and 5.0, only following items would require additional analysis for storage of HLW: 1) the consequences and probabilities of the accidents previously evaluated in WHC-SD-SNF-RPT-004 would increase, 2) a new accident for shearing the canisters would be added to the accidents in WHC-SD-SNF-RPT-004, and the

44


ACTIVITY

a d d i t i o n a l hazardous m a t e r i a l may r e v i s e t h e hazard category. These i tems w i l l r e q u i r e a d d i t i o n a l s a f e t y analyses and r e v i s i o n o f t h e WHC-SD-SNF-RPT-004 and WHC-SD-SNF-HC-007 (Kummerer 1995).

FTE DURATION

It i s recommended t h a t f o r f u t u r e s a f e t y ana lys i s , d u r i n g conceptual design, a formal hazards identification/evaluation should be performed o f t h e e n t i r e HLW con ta ine r hand l i ng and s torage process. v a l i d a t i o n o f t h e i nven to ry assumptions ( rad ionuc l i de con ten t and phys i ca l form) used f o r acc iden t ana lys i s should be performed. impor tan t f o r t h e phys i ca l form o f t h e cesium i o n exchange r e s i n . Th is e v a l u a t i o n assumed, f o r normal cond i t i ons , a d r y product w i t h a r e l a t i v e l y smal l percentage o f r e s p i r a b l e p a r t i c l e s (0.1%). somehow reduced t o small p a r t i c l e s d u r i n g t h e d r y i n g process, t h i s w i l l p o t e n t i a l l y change t h e acc ident ana lys i s r e s u l t s . should be v e r i f i e d , t o make sure t h a t t h e approp r ia te CSB systems are i nco rpo ra ted . Furthermore, a d e t a i l e d rev iew o f f a c i l i t y m o d i f i c a t i o n s should be performed, p a r t i c u l a r l y t h e p o t e n t i a l need f o r replacement o f t h e MHM casks. A s t r u c t u r a l ana lys i s may need t o be performed t o determine t h e p o t e n t i a l e f f e c t s o f dropping t h e new MHM cask on the sa fe ty -c lass v a u l t r o o f , e i t h e r d u r i n g c o n s t r u c t i o n o r a se ismic event. c a t e g o r i z a t i o n must be documented. process i s i n a r e l a t i v e l y e a r l y stage, no at tempt was made t o eva lua te acc iden t f requencies. I n o rde r t o evaluate r i s k , some s o r t o f frequency de te rm ina t ion w i l l need t o be performed. TSR/OSR must be developed t o assure t h a t t h e assumptions i n t h e s a f e t y ana lys i s a re implemented i n t h e des ign and opera t i on o f t h e f a c i l i t y . p a r t i c u l a r , t h e approp r ia te s u r v e i l l a n c e program f o r con ta ine rs i n i n t e r i m storage must be developed. o f t h e s torage w e l l s w i l l be performed by a s u r v e i l l a n c e program.

V e r i f i c a t i o n and

Th is i s p a r t i c u l a r l y

I f t h e r e s i n i s ground up o r

The s a f e t y equipment l i s t

A p r e l i m i n a r y hazard Because t h e des ign o f t h e f a c i l i t y and

A se t o f a d m i n i s t r a t i v e c o n t r o l s and

I n

Th is r e p o r t assumes t h a t some p e r i o d i c mon i to r i ng

The c o s t and schedule f o r t h e proposed s a f e t y a n a l y s i s a c t i v i t i e s are prov ided i n Table 8-1.

4 Months

DOE rev iew 3 Months FSAR r e v i s i o n

DOE = U.S. Department o f Energy FSAR = F i n a l Sa fe ty Ana lys i s Report FTE = F u l l - t i m e Equ iva len t

45


9.0 REFERENCES

DOE-RL, 1995, Hanford Emergency Response P lan , DOE/RL-94-02, U . S . Department o f

DOE, 1992, Hazard C a t e g o r i z a t i o n and Acc ident A n a l y s i s Techniques f o r

Energy, Richland Operat ions O f f i c e , R ich land, Washington.

Compliance w i t h DOE Order 5480.23, Nuc lear S a f e t y A n a l y s i s Repor ts , DOE Standard 1027-92, U . S . Department o f Energy, Washington, D.C.

P r e l i m i n a r y S a f e t y E v a l u a t i o n , February 28, 1994.

Dose Conversion Fac tors For I n h a l a t i o n , Submersion, And I n g e s t i o n , EPA-520/1-88-020, U . S . Environmental P r o t e c t i o n Agency, Washington, D . C .

Farnsworth, R . K . , and J . Mishima, 1988, DWPF C a n i s t e r Impact T e s t i n g and Analyses f o r t h e T r a n s p o r t a t i o n Center , PNL-6379 (UC-70), P a c i f i c Northwest Labora tory , R ich land, Washington.

West i nghouse Hanford Company, Ri ch l and, Washington.

WHC-SD-GN-SWD-30003, Rev. 0 , Westinghouse Hanford Company, R ich land, Washington.

C h a r a c t e r i s t i c s , P u b l i c a t i o n 23, I n t e r n a t i o n a l Commission on Rad io log ica l P r o t e c t i o n , Elmsford, New York .

Jacobs, E . R . , 1996, HLW I n t e r i m Storage F a c i l i t y F e a s i b i l i t y S tudy , (Ex terna l L e t t e r FRF-016 t o R . B . Calmus, WHC) F luor D a n i e l , I n c . , R ich land, Washington.

B u i l d i n g , WHC-SD-SNF-HC-007, Rev. 0 , Westinghouse Hanford Company, R ich land, Washington.

Mishima, J . , 1994, A i r b o r n e Release F r a c t i o n s l R a t e s and R e s p i r a b l e F r a c t i o n s l R a t e s f o r Nonreactor Nuc lear F a c i l i t i e s , DOE-HDBK-3010-94, U . S . Department o f Energy, Washington, D.C .

N R C , 1982, Acc ident Generated P a r t i c u l a t e M a t e r i a l s and T h e i r C h a r a c t e r i s t i c s -- A Review o f Background I n f o r m a t i o n , NUREG/CR-2651, U . S . Nuclear Regulatory Commission, Washington, D.C .

U . S . Nuclear Regulatory Commission, Washington, D . C .

EBASCO/BNFL, 1994, I n i t i a l Pretreatment Module - Cesium Demonstrat ion U n i t

EPA, 1988, L i m i t i n g Values o f Rad ionuc l ide I n t a k e And A i r Concent ra t ion and

Hey, 8 . E . , 1993a, GXQ Program Users’ Guide, WHC-SD-GN-SWD-30002, Rev. 1,

Hey, 8 . E . , 1993b, GXQ Program V e r i f i c a t i o n and V a l i d a t i o n ,

I C R P , 1975, Reference Man: Anatomical , P h y s i o l o g i c a l , and M e t a b o l i c

Kummerer, M . , 1995, Hazard Category A n a l y s i s f o r t h e C a n i s t e r S torage

N R C , 1988, Nuc lear Fuel Cyc le F a c i l i t y Acc ident A n a l y s i s Handbook, NUREG-1320,

46


S e l l e r s , E . D . , 1996, Hazard Category f o r t h e Spent Nuc lea r Fuel P r o j e c t C a n i s t e r Storage B u i l d i n g (CSB) ( L e t t e r 95-SFD-250 t o P res iden t , Westinghouse Hanford Company, January l o ) , U . S . Department o f Energy, Rich land Operat ions O f f i c e , Rich land, Washington.

S o l u t i o n s i n S t a t i c A i r , NUREG/CR-3093, U . S . Nuclear Regulatory Commission, Washington, D . C .

Forward, WHC-SD-WM-SP-011, Rev. 0 , Westinghouse Hanford Company, R ich land , Washington.

Company, Rich land, Washington.

Washington.

Hanford Company, Ri ch l and, Washington.

Equiva lency t o Comparable NRC-Licensed F a c i l i t i e s , Rev. 2 , Westinghouse Hanford Company, Rich land, Washington.

Requirements, Rev. 2 , Westinghouse Hanford Company, Rich land, Washington.

WHC-SD-SNF-DB-004, Spent Nuclear Fuel P r o j e c t Seismic Design C r i t e r i a , Nuclear Regu la to ry Commission Equiva lency Eva lua t i on Repor t , Westinghouse Hanford Company, R ich land , Washington.

S a f e t y A n a l y s i s Documentation Suppor t i ng C a n i s t e r Storage B u i l d i n g Subsur face Cons t ruc t i on and F a b r i c a t i o n o f Tubes, M u l t i - C a n i s t e r Overpack Handl ing Machine and Receiv ing Crane, Rev.2, Westinghouse Hanford Company, Richland Washington.

S u t t e r , 1983, Aerosols Generated by Releases o f P ressu r i zed Powders and

WHC, 1996, Immobi l ized High-Level Waste I n t e r i m Storage P r e l i m i n a r y Path

WHC-CM-2-14, Hazardous M a t e r i a l Packaging and Sh ipp ing , Westinghouse Hanford

WHC-CM-4-46, S a f e t y A n a l y s i s Manual, Westinghouse Hanford Company, Rich land,

WHC-IP-0842, TWRS A d m i n i s t r a t i o n , Unreviewed S a f e t y Ques t ions , Westinghouse

WHC-SD-SNF-DB-002, Spent Nuclear Fuel P r o j e c t Path Forward, Nuc lea r S a f e t y

WHC-SD-SNF-DB-003, Spent Nuclear Fuel P r o j e c t Path Forward, A d d i t i o n a l NRC

WHC-SD-SNF-RPT-004, C a n i s t e r Storage B u i l d i n g S a f e t y A n a l y s i s Report - Phase 2

47

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Project TitleNVork Order C a n i s t e r Storage B u i l d i n g Evaluat ion o f Nuclear S a f e t y f o r S o l i d i f i e d High-Level Waste T r a n s f e r and Storage, WHC-SD-WM-TI-781, Rev 0

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