ML20206C209

' O l

H. B. ROBINSON STEAM ELECTRIC PLANT INDEPENDENT SPENT FUEL STORAGE' INSTALLATION SAFETY ANALYSIS REPORT AMENDMENT NO.1

~

O

~

D b

j e

e

o l

l Amendment No. 1 l

ISFSI SAR DOCUMENT AMENDMENT RECORD SHEET O

Keep this page in the GeneralInformation Section of the SAR. Record the entry of amendments on this sheet as they are inserted. This will then serve as a record of the completeness of this SAR.

Amendment No./Date Issued Date Amendment Entered Initials No. 1/

No. 2 /

No. 3 /

No. 4 /

No. S/

No.6/

No. 7 /

No. 8 /

No. 9 /

No.10 /

No.11/

No.12 /

No.13 /

No.14 /

No.15 /

j No.16 /

No.17 /

~

i No.18 /

No.19 /

No. 20 /

No. 21/

No. 22 /

No. 23 /

0 Amendment No. 1

ISFSI SAR 4

CAROLINA POWER & LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT INDEPENDENT SPENT FUEL STORAGE INSTALLATION MATERIALS LICENSE NO. SNM-2302 SAFETY ANALYSIS REPORT - AMENDMENT NO.1 INSTRUCTION SHEET This amendment contains additional or revised technica1 information and i

editoria1 changes in the form of replacement pages for the ISFSI SAR. Each revised page bears the notation " Amendment No.1" at the page bottom. Vertical bars in the margins of revised pages to indicate the locations of technical revisions on the page.

The following removals and insertions should be made to incorporate Amendment No.,1 into the SAR:

i REMOVE INSERT EXISTING PAGES REPLACEMENT PAGES CHAPTER 1.

CHAPTER 1

]

1.1-1 through 1.14 1.1-1 through 1.14' 4

l 1.2-1 through 1.2-6 1.2-1 through 1.2-6 1.3-1 through 1.3-6 1.3-1 through 1.3-6 T

l 1.3-1 1.3-1 1.R-1 1.R-1 Figure 1.1-1 Figure 1.1-1 Figure 1.2-1 Figure 1.2-1 Figure 1.3-1 Figure 1.3-1 l

Figure 1.3-2 Figure 1.3-2 Figure 1.3-3 Figure 1.3-3 f

Figure 1.34 Figure 1.34 Figure 1.3-3 Figure 1.3-3 CHAPTER 2 CHAPTER 2 2.1-1 through 2.1-3 2.1-1 through 2.14

[

l 2.2-1 through 2.2-2 2.2-1 through 2.2-2 2.6-1 through 2.64 2.6-1 through 2.64 i Amendment No. 1 l

ISFSI SAR REMOVE INSERT EXISTING PAGES REPLACEMENT PAGES CHAPTER 3 CHAPTER 3 3.1-2 through 3.1-5 3.1-2 through 3.1-6 3.2-1 through 3.2-3 3.2-1 through 3.2-4 3.3-1 through 3.3-4 3.3-1 through 3.3-4 CHAPTER 4 CHAPTER 4 4.2-1 through 4.2-2 4.2-1 through 4.2-2 4.3-1 4.3-1 4.6-1 4.6-1 4.7-1 through 4.7-2 4.7-1 through 4.7-2

~

4.R-1 4.R-1 Figure 4.2-1 Figure 4.2-1, Sh. I a

Figure 4.2-1, Sh. 2 Figure 4.2-2 Figure 4.2-2 CHAPTER 5 CHAPTER 5

-1 through 5.1-5 5.1-1 through 5.1-5 5.1 5.2-1 through 5.2-4 5.2-1 through 5.2-4 5.3-1 5.3-1 5.4-1 5.4-1

~

5.R-1 5.R-1 Figure 5.2-1 Figure 5.2-1 Figure 5.2-2 Figure 5.4-1 j

O Amendment No. 1

ISFSI SAR REMOVE INSERT O

EXISTING PAGES REPLACEMENT PAGES CHAPTER 7 CHAPTER 7 7.1-2 7.1-2 7.3-1 through 7.3-4 7.3-1 through 7.3-7 7.4-1 through 7.4-3 7.4-1 through 7.4-4 7.6-1 7.6-1 7.R-1 7.R-1 Figure 7.3-1 Figure 7.4-1 Figure 7.4-1 Figure 7.4-2 Figure 7.4-2 r

Figure 7.4-3

)

Figure 7.6-1 Figure 7.6-1 CHAPTER 8 CHAPTER 8 l

I 8.0-1 8.0-1 through 8.0-2 8.1-1 through 8.1-2 8.1-1 through 8.1-6 8.2-1 through 8.2-3 8.2-1 through 8.2-11 8.3-1 through 8.3-2 8.3-1 through 8.3-7 l

8.4-1 through 8.4-2 8.R-1 8.R-1 Figure 8.2-1

, Figure 8.2-2 l

Figure 8.2-3 Figure 8.2-4 Figure 8.3-1 l

Figure 8.3-2 Figure 8.4-1 1 hd-nt No. 1

ISFSI SAR REA40VE INSERT EXISTING PAGES REPLACEMENT PAGES CHAPTER 10 CHAPTER 10 i

10.0-1 10.0-1 10.2-1 10.2-1 10.3-1 10.3-1 i

10.5-1 CHAPTER 11 CHAPTER 11 11.1-1 through 11.1-2 11.1-1 through 11.1-2 l

11.R-1 11.R-1 i

4 1

7 t

l i

i A

I I

1 i

i 1

l l Amendment No. 1 I

I l

-- - - - - -, -, - - - - - - - -, - -, - - -.. - -,.. -, -. - -, - ~ - - - - - - - - -, - -,,.,.,,

r---,---.

=

f T

iO ISFSI SAR AMENDHENT NO. 1 JUSTIFICATION TABLE SECTION JUSTIFICATION i

i 1.1 A revision was made to include reference to Revision 1 of j

the NUTECH Topical Report, provide clarity in wording, and j

add an acronym.

1.2.3 Revisions were made in response to NRC Question No. 2 i

concerning the principal design criteria for the fuel, and to include the eight-foot drop accident analysis addressed in NRC Question No. 3 (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

Table 1.2-1 Revisions were made to clarify DSC dimensions, specify a three-module unit instead of an eight-module unit and add a reference for design life.

4 Table 1.2-2 Revisions were ande to accurately,repres'ent ISFSI fuel handling operations.

Table 1.2-3 Revisions were made in response to NRC Question No. 4 j

concerning cask movement and transport (NRC letter dated l

7/12/85, CP&L response dated 10/30/85).

l 1.3.1 Revisions were made in response to NRC Question Nos. 8, 9,.

10, 12, 14 and 15,concerning canister design, seismic analysis, civil work, and modifications to the 17-300 system operation (NRC letter dated 7/12/85, CP&L responses dated 10/30/85 and 2/17/86).

1.5, 1.R Revisions were made to include references to Revision 1 of the NUTECH Topical Report and to RBR2 plant documents.

Figure 1.1-1 Revisions were made to more clearly show the system.

2.1.1 Revisions were made in response to NRC Question No. 21 concerning the correct latitude of the site, and to correct the. location of lots sold around the lake (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

2.1.2 Revisions were made to include discussion of the impact of l

planned and existing chemical waste storage facilities, add j

further discussion of the railroad track usage, and include a statement defining the controlled area (CP&L letter dated 4/22/86).

)

2.2.1 A revision was made in response to NRC Question No. 24 to i

correct the location of the railroad track (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

4 l

(1043NEL/vy)

SECTION JUSTIFICATION l

1 l

O 2.2.2 A revision was made in response to NRC Question No. 25 to reflect the correct location of the highway (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

l 2.6 A revision was made in response'to NRC Question No. 27 concerning a reference to the United States Coast and Geodetic Survey Map, to reflect information identical to that in the NUTECR Topical Report, and to add a reference for the foundation analysis (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

3.1.2 Revisions were made in response to NRC Question Nos. 32 and 33 to reflect instrumentation of two of the units, changes made to the IF-300 cask, and to clarify the description of the. positioning skid (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

l l

3.2 Revisions were made in response to NRC Question Nos. 20, j

29. and 30 concerning material incorporated by reference and thermal loading (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

i 3.2.3 Revisions were made in response to NRC Question No. 37 in j

conjunction with changes to Section 8.2 concerning site-l specific seismic analysis for the DSC, DSC support assembly.

O and the HSM (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

?

3.3.2 Revisions were made in response to NRC Question No. 20 concerning material incorporated by reference (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

i l

3.3.3 Revisions were made in response to NRC Question No. 32 I

concerning instrumentation (NRC letter dated 7/12/85, CP&L 1

response dated 10/30/85).

l

)

4.2.1 Revisions were made in response to NRC Question No. 42 to include specific informatics about staudards used in design, fabrication, and construction (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

]

4.2.3 Revisions were made in response to NRC Question No. 44 concerning. consistency between the NUTECR Topical Report and this SAR and to include information about instrumentation of two of the DSCs (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

4.3.2 Revisions were made in response to NRC Question No. 45 concerning instrundatation of the first two DSCs (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

4.6 Revisions were made to provide more detail about corrosion protection.

(1043NRL/vp)

i SECTION JUSTIFICATION 4.7.1 Revisions were made to provide additional information about handling and transport equipment.

4.1 Reference 4.1 was revised to reflect Revision 1 of the j

NUTECH Topical Report.

i 5.1.1.1 Revisions were made in response to NRC Question No. 49 concerning spanner references (NRC letter dated 7/12/85, I

CP&L response dated 10/30/85).

]

5.1-2 Changes for clarity.

5.1.1.6 Revisions were made to clarify canister loading procedures.

5.1.1.7 Section 5.1.1.7 has been rewritten to include more specific information.

5.1.1.8 Revisions were made in response to NRC Question No. 57 to reflect possible conditions for returning the cask to the fuel pool (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

5.1.3.1 Revisions were made for clarity.

5.1.3.4 Revisions irere made in response to NRC Question No. 59 to O

include references to details of the instrumented DSC (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

5.2.1.1 References to the cask liner were deleted.

l 5.2.1.2 Revisions were made in response to NRC Question Nos. 62 and 78 concerning the maximum force which may be exerted by the ram (NRC letter dated 7/12/85, CP&L responses dated 10/30/85 and 2/17/86).

5.2.2 Revisions were made to reflect a visual inspection interval of once per day.

Table 5.2 Table 5.2-1 was revised in response to NRC Question No. 60 concerning a typographical error (NRC 1stter dated 7/12/85, CP&L response dated 10/30/85).

5.3.2 Revisions were made in response to NRC Question Nos. 63 and 80 concerning spare air outlet shielding blocks (NRC letter i

dated 7/12/85, CP&L response dated 10/30/85).

I 5.4 Revisions were made in response to NRC Question No. 64 concerning instrumentation of t;he DSCs (NRC letter dated 7/12/85, CP&L response dated 10/30/85).

5.R Reference 5.1 was changed to reflect Revision 1 of the t

NUTECH Topical Report.

O 3

.(1043NRL/vy) i

i' JUSTIFICATION SECTION 7.1.2 Revisions were made to reflect development of procedures consistent with ALARA concepts.

7.3.2 Revisions were made in response to NRC Question Nos. 34, l

67, and 70'concerning the shielding analysis (NRC letter dated 7/12/85, CP&L responses dated 10/30/85 and 2/17/86).

Table 7.3-1 Table 7.3-1 was revised to correctly reflect design drawings.

Table 7.3-2 Table 7.3-2 was revised in response to NRC Question Nos. 13, 34, 35, 68, and 69 concerning the shielding analysis (NRC letter dated 7/12/85, CP&L responses dated 10/30/85 and 2/17/86).

l l

7.4.1 Revisions were made in response to NRC Question Nos. 70 and 72 concerning dose rates (NRC letter dated 7/12/85, CP&L responses dated 10/30/85 and 2/17/86).

7.4.2 Revisions were made in response to NRC Question No. 74 concerning Figure 7.4-2 (NRC letter dated 7/12/85, CP&L responses dated 10/3.0/85 and 2/17/86)'.

I Table 7.4-1 Table 7.4-1 was revised in response to NRC Question Nos. 70, i

71, and 72 concerning dose rates (NRC letter dated 7/12/85, CP&L responses dated 10/30/85 and 2/17/86).

4 Table 7.4-2 Table 7.4-2 was revised to correspond to Figure 7.4-2, which was revised in response to NRC Question No. 74 7'

j concerning dose rates (NRC letter dated 7/12/85, CP&L responses dated 10/30/85 sad 2/17/86).

7.6.2 Revisions were made in response to NRC Question No. 75 concerning assinua dose (NRC letter dated 7/12/85, CP&L responses dated 10/30/85 and 2/17/86).

7.R Pertinent references relating to shielding were added, and reference was made to Revision 1 of the NUTECH Topical Report.

4 l

8.0 Revisions were made in response to NRC Question No. 76 concerning design events (NRC letter dated 7/12/85, CP&L 4

responses dated 10/30/85 and 2/17/86).

8.1 Revisions were made in response to NRC Question Nos. 76 and 77 concerning design events sad accident pressure (NRC

~

letter dated 7/12/85, CP&L responses dated 10/30/85 and 2/17/86).

Revisions were made in response to NRC Question Nos. 3, 12, 8.2 37, 38, 42, 76, and 84 concerning features of the DSC and cask drop accidents, seismic analysis, site-specific analysis, and design' events (NRC letter dated 7/12/86, CP&L l

responses dated 10/30/85, 2/17/86, and 4/22/86).

(1043NEL/vp)

4 SECTION JUSTIFICATION 8.3 Revisions were made in response to NRC Question Nos. 38, 76, and 86 concerning foundation design and design events (NRC letter dated 7/12/85, CP&L responses dated 10/30/85 l

and 2/17/86).

8.4 Added to provide description of instrumentation.

l 8.R Additional references were added concerning foundation design, and Reference 8.1 was revised to reflect Revision 1 of the NUTECH Topical Report.

i i

10.0 A revision was made to reference new Section 10.5.

10.1 Revisions were made to add fuel specifications.

I 10.2 A revision was made to reflect correct dose rates.

10.3 Revisions were made to reflect proper actfons to be taken 4

if air inlets and/or outlets become bloeied.

j e

l 10.5 Section 10.5 was added to describe surveillance of the RSM inside cavity.

i i

11.1 Revisions were made in response to a reviewer's co m ats (CP&L letted dated 4/22/86).

1 l

e 1

e O

t (1043NRL/vp)

i l

l

\\

E. B. ROBINSON INDEFENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT l

TABLE OF CONTENTS

}

SECTION TITLE PAGE

1.0 INTRODUCTION

AND GENERAL DESCRIFFION OF INSTALLATION 1.1-1 4

1.1 INTRODUCTION

1.1-1 i

l 1.2 GENERAL DESCRIPTION OF INSTALLATION 1.2-1 i

1.2.1 GENIEAL DESCRIFTION 1.2-1 1

1.2.2 FRINCIPAL SITE CHARACTERISTICS 1.2-1 1.2.3 FRINCIPAL DESIGN CRITERIA 1.2-1 i.

1.2.3.1 Structural Features 1.2-1 1.2.3.2 g East Dissipation 1.2-2 i

1.2.4 OPERATING AND FUEL BANDLING SYSTEMS 1.2-2 1.2.5 SAFETY FEATURES 1.2-2 1.2.6 RADI0 ACTIVE WASTE AND AUEILIARY SYSTEMS 1.2-3,

1.3 GENERAL' SYSTEMS DESCRIFFIONS 1.3-1 1

1.3.1 SYSTEMS DESCRIPTIONS 1.3-1 i

1.3.1.1 Canister Desian 1.3-1 i

1.3.1.2 Borisontal Storane Module 1.3-1 1.3.1.3 Transfer Cask 1.3-1 1.3.1.4 Transporter 1.3-1 1.3.1.5 Skid 1.3-2 1

l 1.3.1.6 Borisontal Eydraulie Ras 1.3-2 1

1.3.1.7 System operation 1.3-2 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.5 MATERIAL INCORPORATED BY REFERENCE 1.5-1

{

i Amendment No. 1

f I

i O B. B. ROBINSON INDEFENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT l

TABLE F CONTENTS (Cont'd) i SECTION TITLE PAGE i

2.0 SITE CHARAcrERISTICS 2.1-1 2.1 GEOGRAFNY AND DENDGRAPHY 2.1-1 l

l 2.1.1 SITE LOCATION 2.1-1 l

i 2.1.2 SITE DESCRIPTION 2.1-1 i

l 2.1-1 2.1.2.1 Other Activities within the site Boundary 2.1.2.2 5_oundaries for Establishing Effluent Release Limits 2.1-3 I

2.1.3 F0FULATION DISTRIBUTION AND TRENDS 2.1-3 1

l 2.1.3.1 Posulation Within 10 Miles 2.1-3 1

)

2.1.3.2 Population Between 10 and 30 Miles 2.1-4 2.1.3.3 Transient Population 2.1-4 j

2.1.4 USES OF NEARBY LAND AND WATERS 2.1-4 i

i 2.2 NEARBY INDUSTRIAL. TRANSPORTATION. AND MILITARY l

FACILITIES 2.2-1 2.2.1 LOCATIONS AND ROUTES 2.2-1 2.

2.2 DESCRIPTION

S 2.2-1 j

2.3 NETE030 LOGY 2.3-1 j

2.3.1 REGIONAL CLIMAT014GY 2.3-1 1

3 2.3.2 LOCAL METEOROLOGY 2.3-2 2.3.2.1 Data Sources 2.3-2 i

}

2.3.2.2 Topography 2.3-3 1

l 2.3.3 ONSITE MTEOR0 LOGICAL MASUREMNTS PROGRAMS 2.3-3 l

2.3.3.1 Onette coerational Prosrae 2.3-3 i

j 2.3.3.2 Onsite Data 2.3-4 J

l 2.3.4 DIFFUSION ESTIMATES 2.3-4 11 Anandment No. 1

I E. B. ROBINSON INDEFENDElfr SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REFORT t

i TABLE OF CONTENTS (Cont'd)

SECTION TITLE PAGE 2.4 SUEFAG EYDROLOGY 2.4-1 l

2.4.1 EYDROLOGIC DESCRIPTION 2.4-1 2.4.1.1 Site and Facilities 2.4-1

)

2.4.1.2 Eydrosphere 2.4-1 2.4.2 FLOODS 2.4-1 2.4.3 FROBABLE MARIMUM FLOOD ON STREAMS AND RIVERS 2.4-1 l

2.4.4 FOTBrrIAL DAM FAILURES 2.4-2 l

2.4.5 FROBABLE MARIMUM SRGE AND SEICE FLOODING 2.4-2 i

l 2.4.6 PROBABLE MARIMM TSUNAMI FLOODING 2.4-2 l

jO 2.4.7 IG FLOODING 2.4-2

.g 1

2.4.8 FLOODING FROTECTION REQUIRD err 8 2.4-2 i

l 2.4.9 ENVIROBerrAL ACCEPTANG OF EFFLUBrrs 2.4-2 i

1 i

2.5 SUBSURFACE EYDROLOGY

2. 5-1.

t

)

2.5.1 GROUNDWATER USAGE 2.5-1 2.5.2 SITE CEARAcrERISTICS 2.5-1 i

j 2.5.3 CONTAIlnerr TRANSFort ANALYSIS 2.5-1 2.6 450LOGT AND SEISMOLOGY 2.b1 i

l 2.6.1 BASIC GE0 LOGIC AND SEISMIC INFORMATION 2.6-1

]

2.6.2 VIBRAT0tY GROUND ICTION 2.6-2

{

2.6.2.1 Eartheuake Eistory 2.6-2 i

2.6.2.2 Earthauske Probabilities 2.6-2 2.6.2.3 Desian Earthauske 2.6-2

{

2.6 3 SURFAG FAULTING 2.6-3 1

4 d

111 Amendesat No. 1 L

O H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Cont'd)

SECTION TITLE PAGE 2.6.4 STABILITY OF SUBSURFACE MATERIALS 2.6-3 2.6.4.1 Geologie Features 2.6-3

2. 6.4. 2 Properties of Subsurface Materials 2.6-3 2.6.4.3 ISFSI Foundation 2.6-4 2.6.5 SLOPE STABILITY 2.6-4 t

2.7

SUMMARY

OF SITE CONDITIONS AFFECTING CONSTRUCTION AND OPERATING REQUIREMENTS 2.7-1 3.0 PRINCIPAL DESIGN CRITERIA 3.1-1 i

j 3.1 FURPOSE OF THE INSTALLATION 3.1-1 l

}

3.1.1 NATERIAL TO BE STORED 3.1-1 7

3.1.1.1 Physical Charseteristics 3.1-1 j

3.1.1.2 Theras1 Characteristics 3.1-1 3.1.1.3 Radiological Characteristics 3.1-1 3.1.2 GENERAL OPERATING FUNCTIONS 3.1-2 i

3.1.2.1 overall Functions of the Facility 3.1-2

'3.1.2.2 Handling and Transfer Equipment 3.1-3 j

3.2 STRUCTURAL AND MECHANICAL SAFETY CRITERIA 3.2-1 3.2.1 TORNADO AND WIND LOADINGS 3.2-2 3.2.1.1 Applicable Design Farsasters 3.2-2 1

3.2.1.2 Determination of Forces on the Structures 3.2-2 3.2.1.3 Ability of Structures to Perfore 3.2-2 3.2.2 WATER LEVEL (FLOOD) DESIGN 3.2-2 3.2.3 SEISMIC DESIGN 3.2-3 iv Amendment No. 1

s i

i l

E. B. ROBINSON INDEPENDENT SPElff FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Cont'd)

SECTION TITLE PAGE l

3.2.3.1 Input Criteria 3.2-3 3.2.3.2 Seisaie-System Analysis 3.2-3 1

3.2.4 SNOW AND ICI LOADS 3.2-3 3.2.5 C0tSINED LOAD CRITERIA 3.2-3

.I 3.3 SAFETY PROTECTION SYSTEM 3.3-1 i

3.3.1 GENERAL 3.3-1 1

PROTECTION BY MULTIPLE CONFINEMENT BARRIERS AND SYSTEMS 3.3-1 3.3.2 j

)

l 3.3.2.1 Confinement Barriers and Systems 3.3-1 l

3.3.2.2 Ventilation - Offsas 3.3-1 3.3.3 FROTECTION BY EQUIPMENT AND INSTRUMElfrATION SELECTION 3.3-2 j

l 3.3.3.1 Equisesnt 3.3-2 3.3.3.2 Instrumentation 3.3-2 3.3.4 NUCLRAR CRITICALITY SAFETY 3.3-2 3.3.5 RADIOLOGICAL PROTECTION 3.3-2 1

3.3.5.1 Access Control 3.3-2 3.3.5.2 Shielding 3.3-2 l

3.3.5.3 Radiological Alarm System 3.3-3 3.3.6 FIRE AND IIPLOSION PROTECTION 3.3-3 3.3.6.1 Fire Protection 3.3-3 4

)

3.3.6.2 Explosion Protection 3.3-3 1

3.3.7 MATERIALS HANDLING AND STORAGE 3.3-3 3.3. 7.1 Irradiated Fuel Handling and Storage 3.3-3 3.3. 7.2 Radioactive Weste Treatment 3.3-4 i

v Amendment No. I 1

- - _. - _ _. _ _... _. _, _ ~. _ _ _ _ _. _ _ _..

3 i

i H. B. ROBINSON i

INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Cont'd)

SECTION TITLE PAGE l

3.3.7.3 Waste Storage Facilities 3.3-4 3.3.8 INDUSTRIAL AND CHEMICAL SAFETY 3.3-4 1

3.4' CLASSIFICATION OF STRUCTURES COMPONE!rtS. AND SYSTEMS 3.4-1 1

l 3.5 DECOMMISSIONING CONSIDERATIONS 3.5-1

)

4.0 INSTALLATION DESIGN 4.1.1 4.1

SUMMARY

DESCRIPTION 4.1-1 4.1.1 LOCATION AND LAY 0tTF OF INSTALLATI5N 4.1-1 1

4.1.2 FRINCIPLE FEATURES 4.1-1 O

4.1.2.1 Site Boundary 4.1-1 i

4.1.2.2 controlled Area 5.1-1 4.1.2.3 Emergency Planning Zone 4.1-1 l

l 4.1.2.4 Sito Utility Supplies and Systems 4.1-1 4.1.2.5 Storage Facilities 4.1-1 1

4.1.2.6 Stack 4.1-1 1

4.2 STORAGE STRUCTURES 4.2-1 4.2.1 STRUCTURAL SPECIFICATIONS 4.2-1 4.2.1.1 Design Basis 4.2-1 4.2.1.2 Construction. Fabrication, and Inspection 4.2-1 4.2.2 INSTALLATION LAY 0tTI 4.2-1 i

i 4.2.2.1 Building Plans 4.2-1 4

4.2.2.2 Confinement Features 4.2-2

)

4.2.3 INDIVIDUAL UNIT DESCRIFFION 4.2-2 s

i e

vi Amendment No. 1

= - - - - -

_ - - ~

l

]

H. B. ROBINSON INDEPENDENT SPENT PUEL STORAGE INSTALLATION l

SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Cont'd)

SECTION TITLE PAGE 4.3 AUXILIARY SYSTEMS 4.3-1 4.3.1 VENTILATION AND OFFGAS SYSTEM 4.3-1 4.3.1.1 Ventilation System 4.3-1 4.3.1.2 offsas Systes 4.3-1 4.3.2 ELECTRICAL SYSTEM 4.3-1 4.3.3 AIR SUPPLY SYSTEM 4.3-1 4.3.3.1 Compressed Air 4.3-1 4.3.3.2 Breathing Air 4.3-1 4.3.4 STEAM SUPPLY AND DISTRIBUTION SYSTEM 4.3-2 O

4.3.5 WATER SUPPLY SYSTEM 4.3-2 i

.g 4.3.5.1 Major Components and operating Characteristics 4.3-2 4.3.5.2 Safety donsideration and Controls 4.3-2 4.3.6 SEWAGE TREATEMENT SYSTEM 4.3-2 4.3.6.1 Sanitary Sewage 4.3-2 4.3.6.2 Chemical Sewase 4.3-2 4.3.7 COBOEINICATIONS AND ALARM SYSTEM 4.3-2 4.3.8 FIRE PROTECTION SYSTEM 4.3-2 4.3.9 MAINTENANCE SYSTEMS 4.3-3 4.3.10 COLD CHEMICAL SYSTEMS 4.3-3 4

4.3.11 AIR SAMPLING SYSTEM 4.3-3 4.4 DECONTAMINATION SYSTEMS 4.4-1 4.5 SHIPPING CASE REPAIR AND MAINTENANCE 4.5-1 4.6 CATHODIC PROTECTION 4.6-1 vii Amendment No. 1

O E. B. ROBINSON INDEFENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTDrfS (Cont'd)

SECTION TITLE PAGE 4.7 FUEL HANDLING OPERATION SYSTEM

4. 7-1 4.7.1 STRUCTURAL SPECIFICATIONS

4. 7-1 4.7.2 INSTALLATION LAY 0tFr

4. 7-1

4. 7.2.1 Building Plans

4. 7-1

4. 7.2. 2 Confinement Features

4. 7-1 4.7.3 INDIVILUAL UNIT DESCRIPTION

4. 7-1

4. 7.3.1 shipping Cask Freparation

4. 7-1 4.7.3.2 Spent Fuel Loading

4. 7-2

4. 7.3.3 DSC Drying. Backfilling, and Sealing

4. 7-2 0

5.0 0FIEATION SYSTEMS 5.1-1 5.1 OPERATION DESCRIPTION 5.1-1 5.1.1 NARRATIVE DESCRIPTION 5.1-1 5.1.1.1 Freparat' ion of the Transfer Cask and Canister 5.1-1 5.1.1.2 Fuel Loading 5.1-1 5.1.1.3 Cask Drying Process 5.1-2 5.1.1.4 DSC Sealing Operations 5.1-2 5.1.1.5 Transport of the Cask to the Horisontal Storage Module (ESM) 5.1-2 5.1.1.6 Loading of the Canister into the BSM 5.1-3 5.1.1.7 Monitoring operations 5.1-3 5.1.1.8 Uniosding the DSC from the HSM 5.1-4 5.1.2 FLOW SEErr.

5.1-5 5.1.3 IDElffIFICATION OF SUBJECTS FOR SAFETY ANALYSIS 5.1-5 viii Assadaant No. 1

4 l

J j

E. B. ROBINSON l

INDEFENDENT SPENT FUEL STORAGE INSTALLATION l

SAFETY ANALYSIS REPORT

.i i

TABLE OF CONTENTS (Cont'd)

SECfION TITLE PAGE l

6.3 LIQUID WASTE TREATMENT AND RETENTION 6.3-1

)

6.3.1 EBR2 LIQUID WASTE MANAGEMNr SYSTEM 6.3-1 l'

6.3.1.1 Desian Objectives 6.3-1 l

6.3.1.2

System Description

6.3-2 6.4 SOLID WASTES 6.4-1 i

{

6.5 RADIOLOGICAL IMPACf 0F NORMAL OPERATIONS-StDGIARY 6.5-1 7.0 RADIATION FR&rECTION 7.1-1 i

7.1 ENSURING THAT OCCUPATIONAL RADIATION EIF08URES ARE ALARA 7.1-1 l

7.1.1 POLICY CONSIDERATIONS 7.1-1 1

T 7.1.2 DESIGN CONSIDERATIONS 7.1-1 i

I 7.1.3 OPERATIONAL CONSIDERATIONS 7.1-2 4

7.2

, RADIATION SOURGS 7.2-1 7'. 2.1 CEARACTERIZdTION OF SOURGS 7.2-1 i

7.2.2 AIRBORNE SOURCES 7.2-1 J

)

7.3 RADIATION FSOYECTION DESIGN FEATURES 7.3-1 7.3.1 INS"ALLATION DESIGN FEATURES 7.3-1 1

7.3.2 SEIELDING 7.3-1 4

I 7.3.2.1 Radiation Shielding Design Features 7.3-1 t

Shiel' ins' Analyste 7.3-1 7.3.2.2 d

7.3.3 VENTILATION 7.3-4 l

l 7.3.4 RADIATION MONITORING INSTRUMENTATION 7.3-4 E

Amendment No. 1

...... ?

O E. R. ROBINSON INDEFENDENT SPElff FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT TABLE OF CONTElffS (Cont'd)

SECTION TITLE PAGE 7.4 ESTTMATED ONSITE COLLECTIVE DOSE ASSESSMENT 7.4-1 7.4.1 OPERATIONAL DOSE ASSESS E NT 7.4-1 7.4.2 STORAGE TERN DOSE ASSESSMElff 7.4-1 7.5 EEALTE PETSICS PROGRAN 7.5-1

^

7.5.1 ORGANIZATION 7.5-1 7.5.2 EQUIPM NT, INSTRUMENTATION, AND FACILITIES 7.5-1 7.5.3 FROCEDURES 7.5-1 7.6

' ESTIMATED OFFSITE COLLECf1VE DOSE ASSESSENT 7.b1 7.6.1 EFFLUIlff AND ENVIRONMElrTAL NDNITORING FROGRAN 7.6-1 7.6.2 ANALYSIS OF MLTIFLE ColfrRIBUTION 7.6-1 "

7.6.3 ESTIMATED DOSE IQUIVALElffS 7.b1 8.0 AllALYSIS OF DESIGIf IVElrFS 8.0-1 8.1 NORMAi, AllD OFF-IIORMAL OPERATIONS 8.1-1 8.1.1 NORMAL OPERATION AllALYSIS 8.1-1 8.1.2 0FF-Il0RMAL OPERATION ANALYSIS 8.1-3 8.1.2.1 Transport 8.1-3 8.1.2.2 Air Flow Blockase 8.1-4 8.1.3 BADI0thG1 CAL IMPACF FROM 0FF-IIORMAL OPERATIONS 2.1-4 8.2 ACCIDElff ANALYSIS 8.2-1 8.2.1 1 ASS OF AIR OUTLET SEIELDIl10 8.2-1 8.2.2 TORNAD0/ TORNADO GENERATED MISSILE 8.2-1 O

\\

l mi Amendasat No. 1

O R. S. ROSINSON INDEFENDENT SPElff FUEL STORAGE INSTALLATION SAFETY ANALYSIS BEFORT TABLE OF ColffElffS (Cont'd)

SEcf10N TITLE 3

8.2.3 EARTEQUAEI 8.2-2 8.2.3.1 Accident Analyste 8.2-2 4.2.3.2 Accident Dose Calculacion 4.2-3 8.2.4 DROPECCIDENT 8.2-3 8.2.4.1 Foetulated Cause of Evente 8.2-3 8.2.4.2 Dros Accident Analyste 8.2-4 8.2.5 LIGETWilIG 8.2-7 8.2.5.1 Foetulated Cause of Evente 8.2-7 8.2.5.2 Analyste of Effects and Conseeuences 8.2-7 8.2.6 SLOCEAGE OF' AIR INLETS AND OtFfLETS 8.2-4

.g 8.2.7 ACCIDElff PRES $URIEATION OF DSC 4.2-4 8.2.8 FIRE 8.2-4 4.2.9 DET, STORAGE CANISTER LEAEAGE 8.2-9 8.2.10 LOAD C00SINAT10W 8.2-9 8.3 FOUNDATION 0881GW 8.3-1 4.4 DSC INFTRUMNTATION F8WFTRATION D8810N 8.4-1 9.0 conduct 0F OPERATIONS

. 9.1-1 9.1 080AW13AT10NAL FFRUCTUM 9.1-1 9.1 1 CDEFORATE ORGANIEATION 9.1-1 9.1.1.1 Cornorate Functions. Reessasibilities and Authorities 9.1-1 9.1.1.2 Apolicant's In-Ecuse Orasatsstion 9.1-1 9.1.1.3 Interrelationshise with Contractors and Suss11ere 9.1-1 l

9.1.1.4 Aeolicant's Technical Staff 9.1-2 i

sit Amendment No. 1

O E. B. ROSIE00t INDEFENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT i

' TABLE OF CONTENTS (Cont'd) 1 g

PM g

9.1.2 OPERATING ORGANIZATION, MANAGBENT, AND ADNINISTRATI'rt COWrROLS STSTEN 9.1-4 9.1.2.1 Onsite 0rassisation 9.1-4 9.1.2.2 Forsonnel Functions. Reasonsibilities. Authorities 9.1-5 9.1.3 PERSONNEL QUALIFICATION REQUIRBerf8 9.1-5 9.1.3.1 Minians Qualification Reevireuente 9.1-5 9.1.3.2 Qualifications of Personnel 9.1-6 i

e 9.1.4 LIAIS0N NITE OFFSIDE ORGANIZATIONS 9.1-6 1

9.2 FRE-0PERATIONAL TESTING AND OPERAT10N 9.2-1 9.2.1 ADMINISTRATIVE PROCEDURES FOR CONDUCTING TEST PROGRAN 9.'2-1 9.2.2 CF6L TEST FROGRAN DESCRIPTION 9.2-1 7

9.2.3 TEST DISCUSSION 9.2-1 9.2.3.1 Phreical Facil:, ties feettaa (Thermal Teotian of ESN and DSC?

9.2-1 i

9.2.3.2 Oserations Testina (Bandlina Teste) 9.2-2 9.3 TRAINING FROGRAN 9.3-1 9.3.1 FLANr NfAFF TRAINING FROGRAN 9.3-1 9.3.2 REFLACDerr AND RETRAININC PROGRAM 9.3-1 9.4 NORMAL OPERATIONS 9.4-1 i

9.4.1 PROCEDURES 9.4-1 9.4.2 RE00RDS 9.4-1 9.5 DERGENCY PLANNING 9.5-1 9.6 DEC000GSSIONING FLAN 9.6-1 Bill Agnadasat No. 1 I

= - - -

O N. B. ROBINSON l

INDEFENDENT SPENT FUEL STORACE INSTALLATION SAFETT ANALYSIS REPORT TABLE OF CONTENTS (Cont'd)

SECTION TITLE FACE I

10.0 OPERATING ColtrROLS AND LIMITS 10.0 ;

10.1 FUEL SPECIFICATIONS 10.1-1 j

10.2 LIMITS FOR THE SURFACE DOSE RATE OF THE ESM WEILE TM OSC IS IN STORACE 10.2-1 10.3 LIMITS FOR TM MAXIWM AIR TEMPERATURE RISE i

ArrER STORAGE 10.3-1 10.4

$URVEILLANCE OF THE ESM AIR INLITS 10.4-1 10.5 SURVEILLANCE OF THE HSM INSIDE CAVITY 10.5-1 l

1 11.0 QUALITY ASSURANCE 11.0-1 11.1 CORPORATE QUALITY ASSURANCE 11.1-1 11.2

'E. 3. ROBINSON QUALITY ASSURANCE PROGRAM 11.2-1 t

l 11.3 NUTECW QUALITY ASSURANCE 11.3-1

)

i i

t l

l i

\\

l I

i i

l i

miv Ameadsent No. 1

)

i 1

ll g

i E. B. ROBINSON i

INDEFENDENT SPENT FUEL STORAGE INSTALLATION l

SAFETT ANALYSIS REPORT i

LIST OF TABLES l

TABLE TITLE PAGE 4

1.1-1 ACRONYMS 1.1-3 1.1-2 ABBREVIATIONS 1.1-5 1.2-1 DESIGN PARAMETERS FOR TER EBR ISFSI 1.2-4 1.2-2 SUteE H OF ISFSI FUEL BANDLING OPERATIONS 1.2-5 j

1.2-3 FRIMARY DESIGN PARAMETE23 FOR THE ISFSI OPERATING SYSTDis 1.2-6 l

1.3-1 MMOR SYSTERS, SUBSYSTEMS, AND COMPONENTS OF THE l

E. B. ROBINSON ISFSI 1.3-6 3.1-1 FRYSICAL CEARACTERISTICS OF FWR FUEL ASSEW LIES l

BASED ON NOMINAL DESIGN 3.1-5 3.1-2 ACC5FTABLERADIOLOGICALCRITERIAFORSTORAGEOF MAftETAI IN T H EBR ISFSI 3.1-6 3.3-1 E. B. ROBINSON ISFSI IMPORTANT TO SAFETY l

(SAFETY RELATED).FIATURES 3.3-5 1

3.3-2 RADIOACTIVITY CONFINEMENT BARRIERS AND SYSTEMS l

OF TE ISFSI 3.3-6 5.2-1 TRANSFR SYSTEM CONFONENT DESCRIPTION 5.2-4 7.3-1 DSC MD SHIELDING MATERIAL THICENESSES 7.3-5 l

7.3-2 SEIELDING ANALYSIS RESULTS 7.$6 7.4-1 SUDGERY OF ESTIMATED ONSITE. DOSES DURING FUIL BANDLING OPERATIONS 7.4-2 i

7.4-2 ESTIMATE AlatUAL ONSITE DOSES DURING STORAGE FEASE 7.4-4 8.1-1 DRY SEIELDED CANIST R AND EDRIZOIFFAL STORAGE 8.1-5 4

ICDULE CONFONElff NEIGEf8 8.1-2 MAEDEIN DRY STORAGE CANISTER SELL STUSSES 8.1-6 FOR IIORMAL OPERATING LOADS 3

8.2-1 MAEIMUM DSC STRESSES FOR 8-F00T BOTTOM IND 8.2-10 DROP ACCIDElff l

xv Amendment No, 1

~...

'O H. B. ROBINSON INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT LIST OF TABLES (Cont'd) l TABLE TITLE PAGE 8.2-2 DSC ENVELOPING LOAD COMBINATION 8.2-11 9.1-1 H.1. BANKS, MANAGER - CORPORATE QUALITY ASSURANCE DEPARTMElff 9.1-7 9.1-2 G. P. BEATTY, JR., MANAGER - ROBINSON NUCLEAR PROJECT DEPARTMENT 9.1-9 l

9.1-3 J. M. CURLEY, MANAGER - TECHNICAL ~ SUPPORT 9.1-11 9.1-4 A. B. CITITER, VICE PRESIDENT - NUCLEAR ENGINEERING &

LICENSING DEPARTMENT 9.1-12

'9 1-5 D.M.ROSS,'PRINCIPA5 ENGINEER 9.1-14 9.1-6 F. L. LOWERY, MANAGER - OPERATIONS 9.1-15 i

9.1-7 L. E. MARTIN, MANAGER - CIUCLEAR FUEL SECTION 9.1-16 9.1-8 R. E. W RGAN, GENERAL MANAGER - ROBINSON PLANT 9.1-17 7

9.1-9 M. J. REID, MANAGER - PROJECT CONSTRUCTION 9.1-19 9.1-10 R. M. SMITE, MANAGER - ENVIRONMENTAL & RADIATION ColffROL 9.1-20 1

10.1-1 ACCEPTABLE RADIOLOGICAL CRITERIA FOR STORAGE OF MATERIAL IN THE HER ISFSI 10.1-2 l

1 i

i xvi Amendment No. 1

-.,--y-

. _ _. - - - ~ - - - - - - -, - - - - - -,.., -.,, - -, _.., -.,

,,,,-,n

i O.

H. B. ROBINSON INDEPENDENT SPENT FUEL STORACE INSTALLATION SAFETY ANALYSIS REPORT F

LIST OF FIGURES FICURE TITLE 1.1-1 Primary Components of the ISFSI 1.1-2 Plot Plan 1.2-1 Borizontal Storage Module 1.3-1 Dry Shielded Canistfer and Internal Basket

~

l 1.3-2 ESM Air Flow Diagram l

1.3-3 Skid Features i

j 1.3-4 Hydraulic Ram l

1.3-5 Primary Canister Handling Operations 2.1-1 General Site Location May 2.1-2 H. B. Robinson Site Boundary and Exclusion Zone 4.2-1 Dry Shielded Canister (sheet 1 of 2)

I 4.2-1 Dry Shielded Canister l

(sheet 2 of 2) 4.2-2 Horizontal Storage Module 4.5-1 General Arrangement - HB12 Fuel Handling, Building 5.1-1 Cask Extension for CE IF-300 cask 5.1-2 Cask Liner for CE IF-300 Cask 5.1-3 Handling Operations Flow Sheet

]

(sheet 1 i

of 4) 5.1-3 Handling operations Flow Sheet (sheet 2 of 4) xvii I

e

l V

H. B. ROBINSON i

INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT LIST OF FIGURES (Cont'd) i FIGURE TITLE 5 1-3 Handling Operations Flow Sheet (sheet 3 of 4) 5 1-3 Handling Operations Flow Sheet (sheet 4 3

.of 4) 5 2-1 Hydraulic Ram System 5.2-2 DSC Grappling Systes j

5.4-1

,DSC Instrumest Locations 7.3-1 Location of Reported Dose Rates (Table 7 3-2) 74-l' Annual Dose (ares /yr) from 3 ESMs ( Assuming 2080 hours0.0241 days 0.578 hours 0.00344 weeks 7.9144e-4 months /yr) l 7 4-2 Dose Rate vs. Distance from Surface of HSM (Assuming 3 Wodules) 7.'4-3 Radiation Zone Map of Module Surface Dose Rates i

7.6-1 Anom *.Off aite Dose' (area per year) from 3 HSMs (Based on 1W hrs / day, 365 days /yr) 8.2-1 Cask Drop Height Criteria 8.2-2 Cask Deceleration vs. Time, 8-foot drop 8 2-3 DSC Botton Region ANSYS Model 8 2-4 DSC Top Region ANSYS Model.

8.3-1 Mat Foundation STARDYNE Model s

l 8 3-2 Foundation Uplift Model i

8.4-1 Penetration Model for Horisontal Drop Analysis i

9.1-1 CP&L Operations and Engineering & Construction Organization 9 1-2 Robiuson Nuclear Project Organisation O

I

~

xviii Amendment No. 1

i

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF INSTALLATION

1.1 INTRODUCTION

Carolina Power & Light Company (CP&L) has entered into an agreement with the U. S. Department of Energy (DOE) to conduct a licensed at-reactor dry storage demonstration program for spent nuclear fuel to be located at the H. B. Robinson Steam Electric Plant Unit No. 2 (HBR2). This document provides the safety analysis report (SAR) required as part of the license application under 10 CFR Part 72, " Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation (ISFSI)." This SAR is organised in accordance with the guidelines contained in Regulatory Guide 3.48.

Table 1.1-1 lists the acronyms and Table 1.1-2. lists the abbreviations used throughout this document.

The Nuclaar Waste Policy Act of 1982.(MWPA) estab11shed 1998 as an operational i

date for a spent fuel /high level radioactive waste repository. To assist utilities in providing for spent fuel storage until a repository is operational, the Nuclear Waste Policy Act requires DOE to "... establish a demonstration program in cooperation with the private sector, for the dry storage of spent nuclear fuel at civilian nuclear power reactor sites, with the objective of establishing one or more technologies that the Commission (NRC] may by rule approve for use at the sites of civilian nuclear power reactors without, to the maximum extent practicable, the need for additional, si'te-specific approvals by the Commission." Accordingly, on May 9, 1983, the DOE issued its Solicitation for Cooperation Agreement Proposal O

' Program.

(#DE-SC06-83RL10432) for'a Licensed, At-Reactor, Dry Storage Demonstration Carolina Power & Light Company, in response to the DOE solicitation,

~

submitted a proposal on, August 23, 1983 to DOE to demonstrate the NUTECH Engineers, Inc. (NUTECH) Horizontal Modular Storage (NUHOMS) System at the sits of the H. B. Robinson Steam Electric Plant Unit No. 2.

DOE accepted CP&L's proposal in October 1983, and the contract was signed'in March 1984 (Reference 1.1).

The Electric Power Research Institute-(EPRI) has also entered into the program as a participant.

EPRI, in particular, is involved in the research and development aspects of the program.

The NUHOMS system is the dry' storage design used for the H. B. Robinson (HBR)

ISFSI. In addition to this SAR, Revision 1 of the generic Topical Report for the NUHOMS system, submitted by NUTECH in November 1985 (Reference 1.2),

provides the details of the system to be utilized at HER2. Figure 1.1-1 shows the primary components of the HBR ISFSI. The location of the ISFSI on the

' Robinson site is shown on Figure 1.1-2.

The NUROMS system provides long-term interim storage for irradiated fuel assemblies. The fuel assemblies are confined in a helium atmosphere by a stainless steel canister. The canister is protected and shielded by a massive i

concrete module. Decay heat is removed by thermal radiation, conduction and convection from the canister to an air plenum inside the concrete module. Air flows through this internal plenum by natural draft convection.

The canister containing seven irradiated fuel assemblies is transferred from 8

the reactor fuel pool to the concrete module in a transfer cask. The cask is W

precisely aligned and the canister is inserted into tne module ey means of a hydraulic ram.

1.1-1

%t No. 1

ISFSI SAR S

The NUHOMS system is a totally passive installation that is designed by l

analysis to provide shielding and safe confinement of irradiated fuel. The dry storage canister and horizontal storage module have been designed to function during and withstand certain accidents as described in this SAR.

The fuel assemblies to be stored in the ISFSI are currently located in the HBR2 spent fuel pool and were irradiated in the HBR2 reactor. Seven fuel assemblies are stored in each dry shielded canister. One dry shielded canister is stored in each concrete module. The license application by CP&L requests a license to construct and operate a total of eight modules. CP&L initially intends to construct three modules. Construction of the ISFSI will take approximately one year.

In accordance with the DOE agreement, the first year of operation of the three module ISFSI will be part of a test program.

Normal operation of the facility will continue past the first year for up to 20 years under the initial license until a permanent federal repository is available to store the spent fuel. CP&L may expand the. facility to eight modules in the future if more storage becomes necessary.

O O

e l

l O 1.1-2 Amendment No. 1 l

t

ISFSI SAR TABLE 1.1-1 ACRONYM 3 ACI American Concrete Institute A/E architect / engineer AIF Atomic Industrial Forum AISC American Institute of Steel Construction l

ALARA as low as reasonably achievable ANSI American National Standards Institute ASTM American Society of Testing and Materials l

CP&L Carolina Power & Light Company CQAD Corporate Quality Assurance Department CVCS Chemical and Volume Control System DBT Design Basis Tornado DCE U. S. Department of Energy DSC dry shielded canister E

East EEI Edison Electric Institute ENE East Northeast EOC Emergency Operations Center EPRI Electric Power Research Institute FD Fuel Department j

FHB Fuel Handling Building 4

FSAR Final Safety Analysis Report CE Ceneral Electric Company O

H8E' H. 8. Robinson Steam Electric Plant HBR2 H. 8. Robinson Steam Electric Plant Unit No. 2 HSM horizontal storage module IC internal combustion IFA irradiated fuel assembly ISFSI independent spent fuel storage installation i

N North NC North Carolina NE Northeast NELD Nuclear Engineering & Licensing Department NME North Northeast NNW North Northwest NRC Nuclear Regulatory Coemission NUTECH Horizontal Modular Storage NUROMS NUTECH NUTECH Engineers, Inc.

NW Northwest NWPA Nuclear Waste Policy Act of 1982 ONRR Office of Nuclear Reactor Regulation PWR pressurized water reactor QA quality assurance QC quality control RC Regulatory Guide SAR Safety Analysis Report SC South Carolina SE Southeast SSE safe shutdown earthquake SSE South Southeast 1.1-3 y,,

g

ISFSI SAR TABLE 1.1-1 (Cont'd)

SSW South Southwest SW Southwest US United States UTM Universal Transverse Mercator WDS Weste Disposal System WNW West Northwest 10CFR Code of Federal Regulations, Title 10 i

T 3

i i

1.1-4

,__4__

l

i L

ISFSI SAR 1.2 CENERAL DESCRIPTION OF INSTALLATION O

1.2.1 CENERAL DESCRIPTION The ISFSI provides for the horizontal, dry storage of irradiated fuel assemblies (IFAs) in a concrete module. The principal components are a concrete horizontal storage module (HSM) and a steel dry shielded canister i

(DSC) with an internal basket which holds the IFAs.

Each HSH contains one DSC and each DSC contains seven fuel assemblies. The modules are to be constructed on a common foundation but are not interconnected. The outer, exposed walls are 3 1/2 feet thick concrete to l

provide the'necessary shielding. The initial phase of construction includes 1

three modules. Up to eight modules. may be built and operated at the Robinson site. The second foundation for the additional five modules would be 1

constructed nearby, but separate from the initial three. It has been determined that construction of the additional five modules would not have any impact on continued operation of the initial three modules. Figure 1.2-1 shows the configuration of the inital phase of the HBR ISFSI.

In addition to these primary components, the HBR ISFSI also requires transfer equipment to move.the DSCs from the irradiated fuel pool *(where they are~

loaded with the IFAs) to the HSMs where they are stored. This transfer system consists of a transfer cask, a hydraulic ram, a truck, a trailer and a cask skid. This transfer system will interface with.the existing HBR2 irradiated fuel pool, the cask crane, the site layout (i.e., roads and topography) and other procedural requirements.

1.2.2 PRINCIPAL SITE CHARACTERISTICS The ISFSI is located on the H. B. Robinson Steam Electric Plant site near Hartsville, South Carolina. Carolina Power & Light Company owns and operates a 2300 MWt nuclear generating unit (Unit 2) and a 185 MWe fossil-fueled generating unit (Unit 1) on the Robinson site. The ISFSI is located within the Unit 2 protected area approximately 600 ft. west of the containment building.

1.2.3 PRINCIPAL DESIGN CRITERIA The principal design criteria and parameters for the HBR ISFSI are shown in Table 1.2-1.

The radiation sources are for the maximum burnup fuel. For the fuel to be stored, the radiation sources shall be less than or equal to the sources described in Table 1.2-1.

1.2.3.1 Structural Features The HSM is a low profile reinforced concrete structure designed to withstand normel operating loads, the abnormal loads created by seismic activity, tornados and other natural events and the postulated accidental loads which j

may occur during operation.

)

The structural features of the DSC are defined, to a large extent, by the cask i

drop accident. The operational procedures for the transfer of the cask from fuel pool to the module site are such that the maximum height at which a l

\\

I*2~1

.j hw m 1 o

,--,_.-,,.,.,,,----..-.,..,..-.e---.---.-n-.

- -...w-

l ISFSI SAR l

credible cask drop can occur is limited to 2.44 m (8 ft). The detail description of the cask handling procedure and the transfer operation is presented in Sections 1.3.1.7 and 8.2.4 of this report. The canister body, the double contair. ment welds on each end, and the DSC internals are designed to provide their intended safety functions after a 2.44 m (8 ft) drop.

In fact, the original design of the DSC and its internals, as presented in the NUHOMS topical report, has been modified to withstand decelerations associated with drop heights significantly higher than 2.44 m (8 ft) limit. However, the 2.44 m (8 ft) drop is the minimum requirement used as a design criteria, since it envelopes any actual drop accident that could occur at the Robinson site.

l The details of the cask drop accident are contained in Section 8.2.4 of this report.

1.2.3.2 Decay Heat Dissipation The decay heat of the IFAs is removed.from the DSC by natural draft convection.. Air enters the lower part of the HSM, rises around the DSC and exits through the top shielding slab. The flow cross-sectional area is designed to provide adequate air flow from the draft height of the HSM and the inlet and outlet air temperature differences for the hottest day conditions 8

0 (i.e., 51.7 C (125*F) inlet and 98.9'C (210 F) outlet).

1.2.4 OPERATING AND FUEL HANDLING SYSTEMS The major operating systems of the ISFSI are those required for fuel handling and transport of the fuel from the spent fuel pool to the ISFSI. General O

operations are outlined in Table 1.2-2 and the primary design parameters of the required systems are listed in Table 1.2-3.

The mejority of the fuel handling operations involving the cask (i.e., fuel loading, drying, trailer loading, etc.) utilize standard procedures at HBR2 for spent fuel shipment.

The remaining operations (cask-HSM~ alignment and DSC transfer) are unique to the ISFSI.

1.2.5 SAFETY FEATURES The principal safety feature of the ISFSI is the containment provided by the DSC and the concrete shielding of the HSM. This shielding reduces the gamma and neutron flux emanating from the IFAs inside a DSC so that the average outside surface dose rate on the HSM is less than 20 meen/hr. Additional ISFSI safety features include:

a)

Filling the DSC and cask with domineralized water prior to lowering them into the spent fuel pool - Prevents contamination of the DSC exterior by pool water.

b)

Internal shield blocks inside the HSM - Reduces scatter dose out of the air inlet.

c)

External shield blocks on the HSM - Reduces scatter dose out of the air outlet.

O d)

Shield plugs on the DSC - Reduces dose during DSC drying, helium filling and seal welding.

/

~

1.1-2 a-.nA=.ng go, 1

i ISFSI SAR e)

Double containment closure welds on each end of the DSC - Prevent leakage of radioactive gases or particulates if the fuel rods should fail.

1.2.6 RADIOACTIVE WASTE AND AUXILIARY SYSTEMS Because of the passive nature of the ISFSI, there are no radioactive waste or auxiliary systems required during normal storage operations. There are, however, some waste and auxiliary systems required during DSC loading, drying and transfer into the module. The HBR2 waste systems handle the fuel pool water and air and inert gas which are vented from the DSC and cask during drying. Auxiliary handling systems (such as hydraulic pressure control, alignment, crane, etc.)'are also required during the loading and transfer operation.

I*2-3

%c No. 1

ISFSI SAR TABLE 1.2-1 DESIGN PARAMETERS FOR THE HBR ISFSI Category Criterion or Parameter Value Fuel Acceptance Fissile Content 3.5% Fissile Criteria (U-235 Equivalent)

Radiation Source 15 l

camma 5.73 x 10 photons /sec/ assembly 8

Neutron 1.67 x 10 neutron /sec/ assembly Heat Load 1 KW/ Assembly Dry Shie.1ded Capacity 7 PWR Fuel Assemblies Canister per Canister Size Length (typical) 4.56m (179.5 in)

Diameter 0.94m (36.9 in)

Temperature (max. fuel rod clad) 380 C (716 F)

Cooling Natural Convection 2

Design Life 50 Years O

Material 304 Stainless Steel with Lead End-Shields Internal Helium 0.981 bar (1 sta)

Pressure Horizontal Capacity 1 Dry Shielded Canister Storage Module per Module Unit Size 3 modulr.s per Unit Length 6.71m (22.00 ft)

Height 3.81m (12.50 ft)

Width 7.54m (24.75 ft)

Surface Radiation Dose 20 ares /hr Rate (average on contact)

Material Reinforced Concrete 2

Design Life 50 years j

Actualdesigglimitsareforsevenassemblieg0in the DSC with source rates 1

. of 1.17 x 10 neutrons /se'c/DSC<and 4.01 x 10 photons /sec/DSC.

O 2 Expected life is much longer -(hundreds of years); however, initial license application is for 20 years only. Future amendments may seek to extend the life.

/

i 1

1.2-4 Amendment No. 1

._-,.~-.-_,.--___-----.m

.m

,,_.,e

l ISFSI SAR l

TABLE 1.2-2

SUMMARY

OF ISFSI FUEL HANDLINC OPERATIONS 1.

Clean the DSC and Load it into the Transfer Cask 2.

Fill the DSC and Cask with Domineralized Water 3.

Lift the Cask Containing the DSC into the Spent Fuel Pool j

4.

Load the Fuel into the DSC 5.

Place the Top Lead Plug on the DSC 1

6.

Place the Lid on the Cask 7.

Lift the Cask Containing'the Filled DSC out of the Spent Fuel Pool and Place it on the Drying Pad 8.

Remove the cask Lid 9.

Drain the Water from the Cask and DSC to a Level Two Inches Below the Top Surface of the Lead Plug 10.

Seal Weld the Upper Steel Cover of the Top Lead Plug onto the DSC Body j

11. Hydrotest the sealed DSC l

12. Drain, Evacuate and Dry the DSC

13. Backfill the DSC with Helium

14. Seal Weld Plugs in the Drain and Vent Line of the DSC

15. Place and Seal Weld the Top Cover Plate

16. Drain the Water from the cask l

17. Replace the Cask Lid

18. Lift the Cask onto the Trailer and Lower it into the Horizontal Position O

20.

19. Tow the T' railer to the HSM Remove the HSM Front Access Cover

21. Remove the Cask Lid

22. Align the cask and the HSM

23. Insert the Hydraulic Ram

24. Pull the DSC into the HSM

25. Replace and Tack Weld the HSM Front Access Cover

26. Install Seismic Retainer and Rear Cover Plate

O O

1.2-5 Amendment No. 1.

l --

ISFSI SAR TABLE 1.2-3 PRIMARY DESICN PARAMETERS FOR THE ISFSI OPERATING SYSTEMS j

System Parameters Value Cask Cavity Diameter O.953m (37.5 in.)

Cavity Length 4.572m (180 in.)

Payload Capacity 9524 kg (21,000 lb)

Heat Rating

> 7Kw Shielding (Surface Dose) 200 ares /hr Cask Movement Liftable by Crane N/A Rotatable by Crane from N/A Vertical to Horizontal Cask Lid Removable in Horizontal Position N/A Trailer and

. Truck Transportable N/A Skid Cask Lid Must Protrude Past 15.25cm (6 in.)

End of Trailer and Skid Capacity (Trailer) 109,000kg (120 tons)

(Skid) 100,000kg (110 tons) 3 Positioning Capability t7.62cm (3 in.)

Vertically i

15.08cm (2 in.)

Towards Module 15.08cm (2 in.)

Parallel to Module l

O 1,2-6 A==ad=aat No. 1

,.,.... _ _ _ - -. _ _ _ _ - _., _ _.. _ _ _ -,._._ ~.._. _.-,,

1.3 CENERAL SYSTEMS DESCRIPTIONS The major systems, subsystems, and components of the HBR ISFSI are shown in Table 1.3-1.

The following subsections briefly describe the principal systems and components and their operation.

1.3.1 SYSTEMS DESCRIPTIONS 1.3.1.1 Canister Design Figure 1.3-1 shows the dry shielded canister. The DSC is sized to hold seven irradiated pressurized water reactor (PWR) fuel assemblies. The main component of construction is a stainless steel cylinder with a 0.5 inch wall and 0.94m (36.9 in.) outside diameter. The overall length is 179.5 in.

The general system description of the canister design is available in the NUTECH Topical Report.

1.3.1.2 Horizontal Storage Module An isometric view of a unit of three HSMs is shown in Figure 1.2-1.

The HSM provides a unitized modular storage location for irradiated fuel. The HSM is constructed.from reinforced concrete and structural steel. The modules.will be constructed in place at the storage location. The thick concrete top and front of the HSM provide adequate neutron and gamma shielding. The general systems description of the HSM is provided in the NUTECH Topical Report.

The HSMs are placed in service on a load bearing foundation. Certain civil work is required to prepare the storage site for a level foundation and access 9

area..Emis work includes the relocation of any existing underground utilities, excavation, backfill, compaction and leveling. Also, a 4 inch mud slab will be placed on the leveled subgrade to provide smooth working surface for the placement of the foundation.

1.3.l.3 Transfer Cask The' transfer cask used with the ISFSI provides shielding during the DSC drying operation and during the transfer to the HSM. For the HBR ISFSI, the IF-300 cask (which CP&L owns) licensed under 10CFR71 as a transportation cask will be used (Reference 1.3).

In order to meet the cask cavity minimum length requirement and the criteria for cask lid removal in the horizontal position, the IF-300 cask requires an addition. The addition includes an extension collar (11 inches long and 6 inches thick) with the inside diameter same as that of the cask, and a'2 inch thick cask lid.

In this modified configuration the energ'y absorbing properties of the cask is significantly reduced.

However, as described in detail in Section 8.2.4 of this report there is no credible condition during the cask handling and transfer operation in which the cask could be dropped on its head. Hence, the removal of the cask head with its radial impact limiting fins does not affect the safety features of i

the ISFSI transfer operation.

l 1.3.1.4 Transporter The transporter consists of a trailer with a capacity of 100 tons. The trailer carries the cask skid and the loaded transfer cask. The trailer is 1.3-1 Amendment No. 1

ISFSI SAR designed to ride as low to the ground as possible to minimize the HSM height. Four hydraulic jacks are placed under the trailer to provide vertical movement for alignment of the cask and HSM. The trailer is pulled by a conventional tractor.

1.3.1.5 Skid The cask positioning skid is a we,1ded widefiange frame assembly, which houses the cask cradle support and the cask saddle assemblies. These components are welded to the skid frame. The cask cradle and its support provide the rotational capabilities required to orient the cask from vertical to horizon-tal position. Once in the horizontal position the cask will be seated on the saddle and will be prevented from further movement. The skid is seated on 100 i

ton capacity guided Hilman Rollers at each corner. There,are four hydraulic positioner cylinders mounted on the skid frame and the trailer bed. The rollers and the cylinders will be used for the final alignment of the cask and the HSM. The entire skid assembly is s,eated on a trailer bed. During towing of the trailer, the skid is tied'down to the trailer bed by means of tie down brackets.

The skid and its various. components are designed to withstand the inertia forces associated with transportation shocks. The fes.tures of.the skid described above are shown in Figure 1.'3-3.

1.3.1.6 Horizontal Hydraulic Ram The horizontal hydraulic ram is a telescopic, hydraulic boom with a capacity of 97,860 N (22,000 lb ) and a reach of 7.62m (25 ft). The ram will be g

mounted on the concrete foundation and wall of the HSM on the opposite side from the loading position. Figure 1.3-4 shows the hydraulic ram.

~

1.3.1.7 System operation 1

~

The primary operations (in sequence of occurrence) for the HBR system era shown schematically in Figure 1.3-5 and are described below:

a)

Cask Preparation - Cask preparation includes exterior washdown and interior decontamination. These operations are done on the HBE2 decontamination pad outside the spent fuel pool area. The operations are standard cask operations and have been previously performed by CP&L personnel. Detailed procedures for these operations are described in Chapter 5.

1 b)

Canister Preparation - The internals and externals of the canister are thoroughly washed. This ensures that the newly fabricated canister will meet existing HBE2-specific criteria for placement in the spent fuel pool.

c)

Cask-Canister Loadina - The empty canister is inserted into the cask.

Proper alignment is assured by visual inspection.

i d)

Cask Lifting and Placement in the Pool - The cask and DSC inside the cask are filled with domineralized water. This prevents an inrush of pool l

water when they are placed in the spent fuel pool. This will also reduce (if l.3-2 Amendment No. 1

ISFSI SAR f

O not prevent) contamination of the DSC outer surface by the pool water.

The I

water-filled cask with the DSC inside is then lifted into the fuel pool.

e)

Canister-Assembly Loading - Seven assemblies are placed into the canister basket. This operation is identical to the existing HBR2 cask loading operation.

f)

Cask Lid Placement - The cask lid placement operation consists of placing the DSC upper end-shield plug inside the DSC using the overhead The cask lid is then placed on the cask and it is raised to the crane.

surface where the cask lid bolts are attached. The cask lid firmly holds the j

DSC in place while in the cask.

1 4

3)

Cask Lifting out of the Pool - The filled and closed cask is lifted out i

of the spent fuel pool and placed (in the vertical position) on the drying pad inside the decontamination area. This operation is identical to existing HBR2 cask lifting operations. During this operation the overhead crane is equipped with a redundant yoke and as such is operating in a single failure proof mode. The use of the redundant yok'e eliminates the possibility of any drop accident at this stage of operation.

h)

Canister Sealing - The cask drain is opened and enough of the domineralised water in the cask-canister gap is removed to lower the water level to two inches below the top surface of the upper lead shield plug.

Using a pump, the water level in the canister is reduced to approximately two O_

applied to the outer surface of the upper shield plug.

inches below' the bottom surface of the upper shield plug. A fillet weld is This provides the primary seal for the DSC. After welding, water is pumped back into the DSC to

?

pressurise the internals to 1.5 atmospheres for a hydrostatic test.

i i)

Cask-Canister Drying - Air lines are connected to the DSC and the water inside the canister is forced out by air pressure. The water which is removed from the e,ask and the DSC is routed to the HER2. radioactive waste processing equipment.

In addition to forcing air into the DSC, the air line is used to draw a vacuum to facilitate drying until the water content meets the design I

criteria.

j)

Belius Filling - In order to ensure that no fuel and/or cladding oxidation occurs during storage, the canister is filled with helium (Be). To accomplish this, the. air line (which was used for drying) is removed, and a portable Be gas bottle is connected.

i The canister is filled with He gas to a pressure of 1.47 bar (1.5. atmospheres) and a helium leak sniff test is performed. The pressure is then reduced to 0 981 bar (1 sta). After the canister is filled with the inert gas, the l

filling line is removed and the DSC line connector is plugged acid welded closed. This plug and weld is at the surface of the upper steel plate of the and plug. When the steel cover plate is then welded in place, the integrity j

of the penetrations is assured.

i k)

Final Canister Sealing - Af ter the filling, the steel cover plate is positioned and seal welded. This provides a redundant seal at the upper end I

of the DSC. The lower and also has redundant seal welds, which were made and

(

tested during fabrication. This operation provides the double seal integrity of the DSC.

1 3-3 Amendment No. 1

l ISFSI SAR j

I 1)

Transporter-Skid Loadina - The cask transport tid is positioned and l

bolted in place to close the cask for transport to the HSM. The water in the cask-canister gap is then drained. After draining, clean domineralized water is flushed through the canister-cask gap to remove any contamination left on the canister exterior. The gap is then drained and the cask lid removed, i

allowing access to the DSC. A swipe is taken over the canister exterior at the DSC upper surface and one foot below the DSC head. The cask is then lifted from the decontamination area by the overhead crane with its redundant yoke attached, and is placed on a concrete pad adjacent to the skid / trailer.

Once the cask is on the concrete pad the redundant yoke is removed to allow the cask to fit into the cask cradle. The cask is then raised just above the 1

cradle stop the skid assembly which is held in the horizontal position. At this juncture the crane is operating trithout the redundant yoke. The maximus j

height required to raise the cask above the cradle is 2.44 m (8 ft). The cask j

is then lowered into the cradle until it is firmly seated. Next, the cask is j

tilted from the vertical to the horizontal position until the top region of the cask is firmly positioned on the saddle stop the skid. The lifting and tilting procedures used are identic'ai to those used for loading the IF-300

j.

shipping cask onto its rail car skid (Raference 1.5). The crane yoke is then removed and the cask is properly tied dcwn to the skid frame. At no time i

during the tilting operation is the cask more than 2.44 m- (8 f t) off the ground. Furthermore, since the cask is always lifted from the cask trunnions I

located on the upper region of the cask, the failure of the yoke will not I

cause the cask to drop on its head; hence, no possibility of a cask top end drop. If the yoke fails during tilting operation, the cask would land on its steel ring fins located near the top of the cask's outer shell. The cask drop T

l accident analysis and further discussion on the postulated drop heights sad orientations are presented in Section 8 2.4.

3 m)

Transfer - once loaded and secured, the transte'r trailer is towed to j

the RSM. This movement is completely within the HER2 plant site and protected area. The skid assembly is designed such that none of the cask redundant tie downs and support mechanism can fail due to the inertia forces associated with the transportation shocks and vibrations. Additionally, the skid is tied down to the trailer bed by means of tie down brackets, designed to withstand the same forces. The possibility of cask dropping from the skid, or the cask / skid / trailer tipping or rolling ovsr is extremely remote. The impact decelerations generated by such unlikely events are enveloped by the 2.44 m (8 l

f t) horisontal and vertical drop criteria due to the fact that the cent.or of gravity of the cask is less than 2.44 m (8 f t) from the ground level during i

the transport operation.

n)

Cask-Module Preparation - At the HSM storage area, the cask transport j

lid is removed. Next, the transfer trailer is backed into position and the i

RSM front access cover is raised and removed. The rear access cover' plate is I

l also removed. An optical alignment system and the hydraulic skid positiccers l

are used for the final alignment of the canister and module.

o)

Module Loadina - After final alignment, the canister is pulled into the RSM by the hydraulic ram.(located at the rear of the HSM).

i 1.3-4 Amendment No. 1 l

i O

ISFSI SAR p)

Storane - After the DSC is inside the HSH, the hydraulic ram is released from the DSC. The transfer trailer is pulled away and the HSH front access cover (steel plate) is closed. The rear access steel cover plate is l

also installed and secured in place. The DSC is now in storage within the HSM.

q)

Retrieval-For retrieval, the cask is positioned as previously described and the hydraulic ram is used to push the DSC into the cask. All coupling, attachment, alignment, and closure operations are done in the same manner as previously described, but in reverse order..Once back in the cask, the DSC and its cargo of canistered irradiated fuel assemblies are ready for shipment to a permanent repository or other storage location. During the one year demonstration phase, provisions will be made'to return the canister to the HBR2' spent fuel pool if necessary.

e j

I l

7 4

4 9

l 9

9 l

4 e

i 0

1 e

j 1.3-5 l

a==ad==at No. 1 i

ISFSI SAR TABLE 1.3-1 MAJOR SYSTEMS, SUBSYSTEMS AND COMPONENTS OF THE H. B. ROBINSON ISFSI Dry Shielded Canister Canister Basket Square Cells Spacer Disk Support Rods Canister Body Shie'ided End Plugs Top Cover Steel Plate l

Horizontal Storage Module Concrete Module Precast Outlet Shielding Blocks Dry Shielded Canister Support Assembly O-DSC Seismic Retaining Assembly Alignment System y

Front Access Cover (Ste,1 Plate)

Rear Access Cover (Steel Plate)

Air Flow Penetrations Trailer Cask Positioning Skid I

Skid Positioning System (Vertical and Horizontal)

Transfer Cask Cask Body Cask Lids Cask Drains 1

Cask Extension Collar l

j See Figures 1.1-1, 1.2-1, and 1.3-1.

O l

1.3-6 Amendment No. 1

I ISFSI SAR i

1.5 MATERIAL INCORPORATED BY REFERENCE i

O i

The Topical Report for the NUTECH Horizontal Modular Storage (NUHOMS) System for Irradiated Nuclear Fuel (NUH-001, Revision 1, ADV001.0100) submitted to i

the Nuclear Regulatory Commission by NUTECH Engineers, Inc. in November 1985 is hereby incorporated into this SAR by reference. The NUHOMS topical is i

referenced in Chapters 1, 3, 4, 5, 7, and 17.

The Ceneral Electric Company Safety Analysis Report for the IF-300 shipping Container (NEDO-10048-2) is also incorporated by reference. The IF-300 SAR is referenced in Chapters 3, 5, and 8.

I 6

i i

l I

l l

9 A

1.5-1 Amendment No. 1

(

REFERENCES:

CHAPTER 1 1

(

l.1 CP&L/ DOE Licensed At-Reactor Dry Storage Demonstration Program, Cooperative Agreement No. DE-FC06-84RL10532, Amendment No. A000, March 1984.

I 1.2 NUTECH Engineers, Inc., " Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel," NUH-001, Revision 1, November 1985.

J l.3 Dockat Number 71-9001, Certificate of Compliance Number 9001 for General Electric Model No. IF,-300 Shipping Container, Package l

Identification No. USA /9001/B( )F.

j 1.4 Carolina Power and Light Company, "H. B. Robinson Steam Electric j

Plant Unit No. 2 Updated Final Safety Analysis Report," Docket No.

50-261, License.No. DPR-23.

1.5 Carolina Power and Light Company, H. B. Robinson Steam Electric Plant, Plant Operating Manual, " Refueling Instruction Sp.ent Fuel Cask Handling.Instrugtions for Loading and Shipping of Power Fuel,"

FHP-034.

1

(

l l

7 l

i

)

i 1

i l

1 i

I l

l 1.R-1 hdment No.1

\\

)

_ ~_

ISFSI SAR 2.0 SITE CHARACTERISTICS 2.1 CEOGRAPHY AND DEMOCRAPHY 2.1.1 SITE LOCATION The Independent Spent Fuel Storage Installation (ISFSI) is located on the H. B. Robinson (HBR) Plant site. The site is located in northwest Darlington County, South Carolina, approximately 3 miles west-northwest of Hartsville, South Carolina; 24 miles northwest of Florence, South Carolina; 34 miles north-northeast of Sumter, South Carolina; and 54 miles east-northeast of Columbia, South Carolina. The North Carolina border is 28 miles north of the site and the Atlantic Ocean is about 88 miles southeast (Figure 2.1-1).

The site is on the southwest shore of Lake Robinson, a cooling impoundment of Black Creek. Coordinates are 34' 24' 12" north latitude and 80* 09' 30" sest l

longitude. Universal Transverse Mercator (UTH) coordinates are 3,806,800 north and 577,500 east.

The site is located in the Coastal Plain physiographic, province, approximately 15 miles southeast of the Piedmont province. Topography 6f the region.

(Figure 2.1.1-2 of Reference 2.1) is characterized by r'olling sand hil.ls interspersed with water courses.

i 2.1.2 SITE DESCRIPTION I

A map of the site showing property lines (site boundary) is provided in Figure 2.1-2.

The site currently covers approximately 2,500 acres of land and surrounds Lake Robinson, a 2280-acre impoundment. Carolina Power & Light Company (CP&L) owns all land below the 230 ft. con.cour surrounding the lake.

Since the original purchase of the land, CP&L sold lots on the lake's east l

shore. In addition to these lots, CP&L sold 4.4 acres of land to,the New Market United Methodist Church for expansion of church facilities. In all cases, CP&L retained ownership of land below the 230 ft. contour. However, property owners were allowed to lease land between the 230 ft. contour and the lake (220 ft. contour normal operating level) and to construct piers, boathouses, and ramps. Provisions were also made for means of access to the lake by the general public. As a result, three privately owned recreational areas,.one private sailboat club, and numerous access points throughout the lake allow for the use of the lake by the local population.

1 j

2.1.2.1 Other Activities Within the Site Boundary

~i Carolina Power & Light Company owns and operates a 2300-Mwt nuclear generating plant (Unit 2) on the Robinson site. The ISFSI is located within the protected area for the nuclear unit. Unit 2 received an operating license from the U. S. Atomic Energy Commission in 1970 (Docket No. 50-261/ License No. DPR-23).

Carolina Power & Light Company also owns and operates a 185 Mwe fossil-fueled generating plant (Unit 1) adjacent to the nuclear unit (Unit 2).

Unit I was placed in service in 1960, prior to the construction of Unit 2.

Additionally, CP&L leases the Darlington IC Plant, a 572 Mwe internal combustion plant, from-i 2.1-1 Amendment No. 1 I

l-ISFSI SAR O

Westinghouse, Inc. The Darlington Plant is located approximately 1 1/3 miles NNW of the plant on land originally included as part of the Robinson site i

property.

Approximately 1000 ft. from the Unit 2 reactor, CP&L operates a nuclear visitors' center. As part of the center's facilities, there is a covered picnic pavilion, located on the southwest shore of Lake Robinson.

A spur track of the Seaboard Coastline Railroad branches from a mainline at i

McBee, South Carolina, and passes 1600 ft.' west of the plant. An extension of this spur enters the immediate plant area north of the plant and allows for delivery of coal to Unit 1.

The maximum speed limit for these tracks is observed as 10 mph. As a result of having extended the security fence, paralleling the siding tracks, to the we s,t, the siding track closest to the new security fence is not used for storage of empty cars. This track is used only as a "run through" track for railroad cars.

l Figure 1.1-2 shows two supporting facilities recently construc,ted on plant property. The Chemical Barrel Storage Building will be located southwest'of the HSMs and the oil storage facility will be located near the path to be used to transport the DSCs from the fuel handling area to the HSMs.

The Weste 011 Storage System will provide a two-year temporary storage facility of four classes of fluids: low-level radioactively contaminated lube 4

]

oil, contaminated solvents, noncontaminated lube oil, and noncontaminated solvents. The facility is located on the area north of the site of the future Y

Operations and Maintenance Building and west of the Shredder-Compactor Building.

Because of the classes of fluide handled, the system is broken down further into two separate systems. The, Contaminated Weste Oil System is designed to store radioactively contaminated waste liquids. The Noncontaminated Waste Oil System is designed to store noncontaminated liquids. Neither will tie into i

the existing waste disposal systems.

i Each (contaminated or noncontaminated) system consists of ones 75-gallon oil fill tank, strainer, 20 gym fill / recirculation pump,150 gym transfer /

recirculation pump, 10,000-gallon storage tank for oil, 400-gallon storage tank for solvent, and 40 gym transfer / recirculation pump for solvents.

The Waste Oil Storage System is designed for future disposal and is sized to store fluids until a permanent disposal system is implemented.

2 A 2'-6" dike (774 f t ) surrounds each contaminated and noncontaminated.

system. The total capacity of the dike is 14,474 gallons which is adequate to i

prevent the release of radioactive liquids in the event of a fire or a major spill.

1 O

The Chemical / Barrel Storage Facility is a 2,250 square foot, pre-engineered building located west of the Unit 2 protected area boundary and north of the Chemical / Bulk Receiving Warehouse. The facility was designed for storage of paints, lab chemicals to support the on-site chemistry lab, and provide storage of bottled gas.

2.1-2 h adment No. I

ISFSI SAR The building has no floor drains. It does have a 4" high perimeter curb to contain any spills. The spills will be dry treated and cleaned up rather than drained to the Stora Sewer System.

Ventilation is accomplished by the power roof ventilators and filtered wall louvers.

Fire protection is provided by dry pipe sprinklers and portable fire extinguishers which are located in each area. Full-height, fire-rated concrete masonry unit (CMU) walls divide the building. All equipment and fixtures are explosion proof suitable for Class I, Isivision 2, Group D atmospheres, per NFPA 70, NEC Article 500. Bottled gases are separated into groups of compatible gases by CHU partitions.

2.1.2.2 Boundaries for Establishing Effluent Release Limits The exclusion zone is defined as the 1400 ft. radial area surrounding the 4

plant. ' There are no residences or agricultural activities inside of the exclusion zone.

The 1400 ft. exclusion zone encompasses land and a portion "of Lake Robinson.

All land included in this area is owned in fee simple, without reservations, by CP&L. As such, all mineral rights are owned by CP&L. The public is allowed access to that part of Lake Robinson which is included in the exclusion zone.

The controlled area for the ISFSI is defined as being contained within the HBR

?

Unit 2 exclusion area. The protective area is within the Plant's ' exclusion sone.

A spur track of the Seaboard Coastline Railroad branches from a mainline in McBee, South Carolina, and passes 1600 ft. west,of the plant. An extension of i

this spur is operated by CP&L and intersects the exclusion zone north of the plant. Approximately three coal deliveries are made to Unit 1 each week.

There is no passenger traffic on either the spur extension or the spur track.

An easement has been granted to Southern Bell Telephone and Telegraph Company for maintenance of telephone lines within the exclusion zone.

2.1.3 POPULATION DISTRIBUTION AND TRENDS 2.1.3.1 Population Within 10 Miles The 1980 estimated resident population between sero and ten miles of the l

Robinson site is presented in Section 21 of Reference 2.1 as well as estientes of future population.

The area between zero and ten miles includes parts of four counties:

Darlington, Chesterfield, Kershaw, and Lee, and the total 1980 resident population was approximately 31,000. The majority of these residents live in or around the city of Hartsville, 3 miles SSE (7631 city /11,529 suburban).

One other small concentration of resident population was indicated for the city of McBee, 7 miles NW (774 city). Other population within the area is generally considered to be rural.

2.1-3 Amendment No. 1

I ISFSI SAR 2.1.3.2 Population Between 10 and 50 Miles The 1980 estimated resident population between ten and fifty miles is presented in Section 2.1 of Reference 2.1 along with estimates of future population.

The area is generally rural and is characterized by population concentrations in and around Florence, SC, 24 miles SE (30,062 city /9951 suburban); and Sumter, SC, 34 miles.SSW (24,890 city /10,485 suburban). Cities with area populations over 10,000 include Laurenburg, NC, 44 miles ENE (11,480 city /536 suburban); Monroe, NC, 44 miles NNW (12,639 city /none suburban); Rockingham, NC, 42 miles NNE (8300 city /7203 suburban)) and Lancaster, SC, 40 miles WNW (9603 city /6082 suburban).

2.1.3.3 Transient Population

'The transient population within 10 miles of the Robinson site is composed of i

four major components: the industrial labor force, seasonal population variation, school population, and hospital / nursing home populations. These specific effects are discussed in Section 2.1 of Reference 2.1.

2.1.4 USES OF NEARBY LAND AND WATERS Uses of land and water and respective populations in the 10-mile area surrounding the site is discussed in Section 2.1 of Reference 2.1 and O

Section 2.2 of this report.

?

i O

2.1-4 Amendment No. 1

l i

2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES j

2.2.1 LOCATIONS AND ROUTES i

The Independent Spent Fuel Storage Installation is located on the Robinson Plant site. The Robinson Plant is located in an area which is generally rural i

and undeveloped. Isssediately west of the site and along the eastern shore of Lake Robinson, residential development has occurred, as well as the establishment of various public and private recreational areas. Other i

residential development within the 5-mile area surrounding the plant is confined to the Hartsville area (3 miles SSE).

The ISFSI is located within the controlled ar'es of the 2300 MWt nuclear unit. A coal fired electric generating plant is located west and adjacent to the nuclear unit. The Darlington IC Plant is located approximately 1 1/3 1

miles NNW of the nuclear unit. Isssediately north of the Darlington IC Plant is a gas pipeline. Other industrial development within 5 miles is limited to j

the areas in and surrounding Hartsville (3 miles SSE).

Agricultural development has occurred within the five-mile area, especially in areas north and dest of the plant.

Principal transportation routes or facilities include highways, a railroad line (1600 ft. W), and a small airport (2 1/2 miles E).

1 There are no military bases within the five mile area.

2.

2.2 DESCRIPTION

S V-Residential development along the shores of Lake Robinson is confined to the eastern and northern shore of the lake. Since 1960, numerous permanent and vacation homes have been built above the 230 ft. contour. Below the 230 ft.

contour, property owners.have constructed small private piers, boat docks, and i

ramps) access is provided by lease agreements between landowners and Carolina Power & Light Company.

Public recreational areas include Easterling's Landing, 1.7 miles NNE (a beach, picnic and paved boat launch area); Atkinson's Landing, 1.2 miles NME (a beach and boat launch area); and J & M Marina, 4300 ft. E (a beach, paved boat launch, and boat gasoline facility). A small private sailboat club is l

also located on the lake, 2 miles NNE. All facilities are on the eastern lake l

shore. Several other areas provide recreational access to the lake, but this use is limited compared to,that of other facilities.

i East and adjacent to the nuclear unit (Unit 2), CP&L owns and operates a 185 Mwe coal fired electric generating plant (Unit 1). Unit I was placed in service in 1960.

i The Darlington IC Plant (1 1/3 miles NNW) is a 572 MWe internal combustion electric generating plant. The plant is owned b'y Westinghouse, Inc., but is leased and operated by CP&L.

Carolina Pipeline transports natural gas via an underground pipeline (2 miles i

N).

The pipeline transects Lake Robinson in an east / west direction. That part of the pipeline which crosses the discharge canal esteads above ground.

2.2-1 Amendment No. 1

ISFSI SAR I

i Other industrial development within b miles of.the plant is not extensive, and includes eleven firms which employ more than 100 people.

Principal products are paper products, textiles, fertilizer, eeds, and bearings (Reference 2.2).

All of these firms are located in or near Hartsville (3 miles SSE).

Agricultural development has occurred within the five-mile area especially in areas north and west of the plant. Acreage to the north includes numerous peach orchards. Associated with the peach orchards is a fruit processing firm which processes and distributes local peaches, as well as other non-local produce.

Principal transportation routes include SC 151 (1/2 mile W), a two-lane highway running north and south connecting McBee and Hartsville; and numerous state maintained secondary roads.

A small private airport is located 2 1/2 miles east of the plant. Only small aircraft use the runway.

4 i

?

I l

j 1

i l

\\

1 2.2-2 Amendment No. 1 i~

- =.

ISFSI SAR 2.6 CEOLOGY AND SEISMOLOGY Specific soil testing has been performed at the designated location for the ISFSI. The data obtained from this testing is' utilized in the foundation design for the ISFSI. As part of the foundation analysis / design, the subject of soil liquefaction is addressed. The following sections discuss the Robinson site geology and seismology. The foundation analysis is presented in Section 8.3 of this report.

2.6.1 BASIC CE0 LOGIC AND SEISMIC INEORMATION The Robinson site is locate,i in the Coastal Plain physiographic province approximately 15 miles southeast of the Piedmont. province. In South Carolina, g

the Coastal Plain is composed of largely unconsolidated sediments which overlie a slightly sloping surface of crystalline rock. The Coastal Plain sediments in the area of the situ were formed at the same time as the Tuscaloosa Formation, but locally are known as the Middendorf Formation.

The surficial materials at the Robinson site are recent sands or soils developed from the Middendorf.

Because of the high quarts content of the sands and the climatic environment, the surficial soils may not weather sufficiently to differ considerably from the parent material. Thus, it is nearly impossible to distinguish the recent alluvial soils from the parent Middendorf sand since both the alluvial and weathered soils are derived from the Middendorf. Only their manner of placement would be different. From an engineering standpoint, the difference is minor.

p The subsurfaca materials encountered in the test holes drilled at the site are completely consistent with recent alluvium and Middendorf Formations encountered throughout the vicinity. Discontinuities within the strata are sedimentary and no structural deformation is apparent 1n the Middendorf Formation in the site area.

The Middendorf is about 400 ft. thick and overlies an eroded, slightly sloping surface of Piedmont crystallines that may be somewhat weathered near the surface.

Triassic basins are known in the area; however, it is believed that the likelihood of a Triassic basin at the site is quite small. The basement rock at the site is considered to be Piedmont crystalline since the results of the seismic surveys indicate a high velocity material at a depth consistent with the depth of Piedmont crystallines encountered in wells in the area.

In general., the upper alluvial sands and gravels are moderately compact.

Layers of compressible material occur in the upper 30 to 50 ft.

Because of the quantity of fines in the sand and gravel, it could not be considered free-draining material. The underlying Middendorf contains generally compact l

relatively incompressible sands and firm to hard clayey soils. Several strata l

of cemented sandstone were encountered in the borings at depths of roughly 90 i

to 100 ft.

/

2.6-1 Amendment No. 1

ISFSI'SAR O

unconsolidated formation.,

From a geological standpoint, the Middendorf is considered to be an From an engineering point of view, however, the materials are firm and compact and would provide good foundation support for 1

the proposed construction. The materials range in texture from a hard or l

compact soil to a soft rock.

I Further details of the regional geology are discussed in Section 2.5.1 of i

Reference 2.1.

2.6.2 VIBRATORY GROUND MOTION A seismological study for the Robinson site has been performed to determine the design and hypothetical earthquakes for the site and the ground motion spectra associated with them. Details are discussed in Section 2.5.2 of Reference 2.1.

2.6.2.1 Earthquake History The largest earthquake in this region occurred at Charleston, South Carolina l

in August 1886. This shock had an intensity of about Modified Mercalli IX at l

the epicenter and it is estimated that this shock had a magnitude of 6 1/2 to-7'with epicentral acceleration of 0.253 to 0.30g.

However, damage was t

confined to a relatively small area.

Only one earthquake of intensity V or greater has ever been recorded within 50 miles of the Robinson site. This shock occurred on October 26, 1959, near 1

McBee, Chesterfield County, South Carolina, with an intensity of Modified Mercalli VI.

The epicenter was located about 15 miles from the site. The

?

estimated intensity at the site was about V.

]

Except for a trend of epicenters paralleling the Blue Ridge, there is no apparent trend of other epicenters in the region. Most of the smaller historical shocks were reported in scattered population centers. Further details of area seismicity are described in Section 2.5.2 of Reference 2.1.

The site is located in Zone 1 of the U. S. Coast and Geodetic Survey Map and Equal Seismic Probability. Zone 1 is characterized as a zone of light earthquake activity which would result in minor damage. Therefore, on an historical basis, it would appear that the site will not experience damaging earthquake motion during the life of the planned facility.

2.6.2.2 Earthquake Probabilities, l

On the basis of historical data, it is expected that the site area could esperience a shock on the order of the 1959 McBee shock once during the life of the facility. This shock could be as far distant as in 1959 (15 miles), or perhaps ' closer.

2.6.2.3 Design Earthquake Comparison between the Robinson site and certain areas in California indicc e I

a' similarity in the depth and type of overburden material. For this reason, Dr. C. W. Housner of the California Institute of Technology recommended the use of his average California response spectra to-define the earthquakes (see 2.6-2 Amendment No. 1 l

Appendix 2.5E of Reference 2.1).

Dr. Housner's specific recommendations aret design for maximum horizontal ground acceleration of 0.lg with a vertical component of 2/3 of the horizontal acceleration, and hypothetical earthquake maximum horizontal ground acceleration of 0.20s.

The safe shutdown earthquake and operating basis earthquake for HBR2 are 0.2g and 0.1, respectively. It is important to note that even if an earthquake 3

comparable to the Charleston shock were to occur 35 miles from the site, the j

ground acceleration would not exceed 0.2. These values are identical to 3

1 those reported in Reference 2.1.

2.6.3 SURFACE FAULTING A study of the possibility of the existence of faults in the ares indicated that no active faulting was apparent. The sediments underlying the site are quite thick and apparently undisturbed. The surface of the buried crystallines is an ancient eroded one, and active faults are unknown in the vicinity of the site.

No faulting is apparent in the unconsolidated sediments of the Coastal Plain. The underlying basement rocks are effectively masked by more than 400 ft. of sediments at the site,and cannot be directly observed below the Fall Zone. However, faulting in the basement complex is known from exposures above the Fall Zone and cores from scatterd borings drilled through the Coastal Plain sediments.

2.6.4 '

STABILITY OF SUBSURFACE MATERIALS 2.6.4.1 Geologic Features i

The test boring program, refraction surveys, and laboratory tests, when combined, present the following picture of the subsurface and geologic site conditions: The piedmont crystalline basement rock at the site is overlain with approximately 460 ft. of unconsolidated coastal plain sediment. These sediments are comprised of about 30 fe' of surface alluvium over 430 f t. of i

the Middendorf formation. The Middendorf is made up of sands, silty and sandy clay, sandstone, and siltstone. Compressional wave velocities are 17,500 fps in the basement rock, 7,200 fps in the Middendorf, and 1,500 fps in the top 30 ft. of alluvium.

2.6.4.2 -

Properties of Subsurface Materials In order to evaluate changes in the properties of the soils which underly EBR2, a series of static and dynamic triaxial compression tests and confined compression tests were performed at the time HBR2 was. designed.

a)

Dynamic Confined Compression Tests: The. test results indicated that the compressibility characteristics of the silty and clayey soils encountered at the site are not appreciably affected by dynamic loading. The settlement of the sandy soils increased somewhat when the sample was subjected to an oscillating load of short duration.

b)

Dynamic Triaxial Compression Tests: The test results indicated that the available shearing strength of the soils at the site are generally 2.6-3

,,,,,,,,t,,, l

i ISFSI SAR l

slightly reduced when subjected to dynamic influence. The upper moderately O

firm clayey silt is not appreciably affected, but the stiff, silty clay, and dense sandy soils apparently experience some minor strength reduction.

2.6.4.3 ISFSI Foundation A specific soil testing and foundation evaluation has been performed at the designated location for the ISFSI. The data obtained from this testing is being utilized in the foundation design for the'ISFSI. The subject of soil liquefaction has been addressed as follows:

The analysis of the subsurface profile at.the site of the ISFSI confirms previous findings made from other subsurface investigations carried out on the BBR2 site. Th1 analysis of the data ebrained from the ISFSI site borings used j

the procedures recommended by Seed, Idrissi and Arango for estimating liquefaction susceptibility using standard penetration resistances (N-values).

c 1

Data collected from the borings indicat'es potential liquefaction; hoeever, i

this sample represents the soil conditions at a depth of approximately 100 feet. This loose soil zone is less than 5 feet thick, is located beneath approximately 23 feet of the very hard silty clay of the Middendorf Formation and is surrounded by approximately 25 feet of firm to dense sand. Hard silty clay continues below the sand stratum at a depth of approximately 112 feet.'

Based on a review of the previous explorations performed at the H. 8. Robinson site, there is no indication of a similar zone,of loose sand located at the depths encountered by this boring. the 5-foot zone of loose sand is believed to be an isolated pocket or lens not representative of an area-wide layer in the Middendorf Formation.

Based on this analysis, the ISFSI site profile is considered to have a very low to no likelihood of liquefaction during a design basis earthquake.

Because liquefaction of the loose sand pocket at 100 feet cannot be completely discounted, the effects of liquefr.ction were examined. In the case of a confined layer of potentially liqusfiable saturated sands at large depths, as in this boring for the ISFSI, the paramount problem is'not one of bearing capacity or landslide susceptibility as in surficial sands, but one of surface settlement following complete liquefaction.

It can be expected that the hard 1

silty clay and dense sands surrounding the loose sand layer would bridge any effects from the liquefaction of the loose sands and no' surface settlements would occur.

2.6.5 SLOPE STABILITY The failure of any slopes at the Robinson site will not adversely effect the ISFSI. A discus,sion of slope stability of the earth dam and appurtenances at j

the Robinson ' site is provided in Section 2.5.5 of Reference 2,1,.

4 9

d

+

f O

]

2.6-4 6"and==at No. 1

~

3.1.2 GENERAL OPERATING FUNCTIONS s

3.1.2.1 Overall Functions of the Fac.ility l

The ISFSI is designed to be totally passive, requiring no utilities or waste processing system and to utilize HBR2's existing cask handling, fuel handling and associated auxiliary equipment in preparing the IFAs for dry storage. The j

. cask to be used for the onsite transfer operation is the GE IF-300 shipping cask which is fully documented in Reference 3.5.

i The DSC will be placed into the cavity of the GE IF-300 shipping cask. The shipping cask will then be lowered into the existing H8R2 spent fuel pool where seven fuel assemblies will be placed into the DSC. Once the top lead shield plug is placed onto the loaded DSC, the cask containing the loaded DSC will be raised out of the pool and the water drained partially from the DSC cavity. At this juncture, the top lead shield plug is welded into place. The DSC cavity will then be drained and vacuum dried of all water and backfilled with helium through the vent and siphon tube penetrations. After backfilling 1

the DSC with helium, the penetrations will be seal welded and the top cover i

plate will be welded into place. With the sealed DSC still within the confines of the shipping cask, the shipping cask will be transported to the ESM and aligud with the front access of the HSM. A hydraulic ram will then j

extend from the rear access of the HSH through the HSM and attach onto the DSC j

grappling plate. The hydraulic ran will be retracted through the HSM, pulling O

the DSC into the HSM. Once the DSC is properly positioned within the HSM, the front and rear accesses of the HSM will be closed.

t The HSMs which house the DSCs are located on a level, reinforced concrete, load' bearing stab. The slab is designed for normal and postulated accident conditions. The HSM is also designed to maintain its dimensional and structural integrity during postulated environmental and geological events.

Sais storage in the HSM is provided by:

(1) a natural convection heat removal path, (2) the concrete radiation shielding, and (3) the double closure welds of the DSC. The operation of the HSMs and DSCs is totally passive. No active systems are required.

Since the first three units of the ISFSI facility will be used as part of the demonstration program, two of the HSMs and the DSCs will be instrumented for the purpose of collecting data. The instrumentation is limited to placement of a number of thermocouples in these components. The placement of thermo-couples inside the HSMs does not effect its structural and mechanical integrity. The instrumentation of the DSC, however, requires a feed through penetration at its bottom cover plate. This penetrati'on is designed such that the confinement integrity of the DSC is not compromised under both normal operating and accident conditions. Furthermore the instrumentation of these components does not change the total passive nature of the system. Details of the instrument penetration analysis are provided in Chapter 8 of this' report.

h)

A more detailed description of each component's functions is located in the L/

following subsections.

l I

i 3.1-2 Amendment No. 1

ISFSI SAR 3.1.2.2 Handling and Transfer Equipment O

-/

The ISFSI is designed to utilize the existing HBR2 fuel handling equipment and the CE IF-300 shipping cask. The function of the various HBR2 fuel handling equipment which are employed in loading and handling the IF-300 shipping cask.

are described in the plant's existing procedures and Section 9.1.4 of the HBR2 Updated FSAR (Reference 3.2).

The systems and equipment which are unique to the ISFSI for handling and transfer'of IFAs are the DSC, the cask liner and docking collar, the cask skid, the hydraulic ram, and the HSM. The function of each transfer and handling system or piece of equipment along with the waste processing system are briefly described in the following paragraphs. -

t a)

DSC - The DSC will serve as the confinement vessel during transport of the IFAs to and from the HSH as well as during storage of the IFAs in the HSM. The shielded end plugs will provide biological shielding during i

transport of the fuel assemblies and also provide shielding for the front and rear accesses of the HSM.

b)

Cask - The CE IF-300 shipping cask is used to transport the DSC to and from the HSM. The function of the shipping cask is to provide biological shielding along the axial length of the IFAs and a means for removing a sufficient quantity of decay heat so that the mechanical integrity of the DSC or IFAs is not jeopardized. The CE IF-300 is fully documented in Reference 3.5.

As stated in Section 1.3.1.3 a new cask collar and lid are O

used on the IF-300 cask. This addition will provide the minimum cask cavity 1ength requirement and allows for the removal of the cask lid in the m

~

horizontal position.

c)

Cask positioning Skid - The function of the cask positioning skid is to provide a means by which the final alignment of the cask with respect to the HSM can be achieved, and to restrain the shipping cask in a horisontal position during the transfer of the-DSC to and from the HSM. The skid and lthe l

cask will be transported from the fuel building to the HSMs by a trailer.

Both the skid and the DSC are designed to withstand the inertia forces

~

associated with the transportation shock loads.

d)

Hydraulic Ram - The hydraulic ran is used to move the DSC between the l

cask and the HSM. The hydraulic ran is a long hydraulic cylinder with a.

grapple device mounted at the cylinder's head. The hydraulic ran will be positioned at the rear access of the HSM. The ran will then be extended through the rear access and through the entire length of the HSM. Once the grapple is seated within th's grappling plate of the DSC, the arms of the grapple will be extended so that it is securely in place between the DSC and the grapple plate. The ram will then be retracted, pulling the DSC out of the cask cavity and into the HSM.

e)

Horizontal Storage Module-The function of the HSM is to provide protection for the DSC against geological and environmental events as specified in Section 3.2, and serve as the principle biological shield for O

irradiated fuel during storage. The HSM contains shielded air ducts near the bottom of the structure to admit ambient air around the DSC.for cooling purposes. The air, warmed by the canister, is exhausted through shielded 3.1-3 hdment No. 1

ISFSI SAR vents at the top of the HSM. The HSM also provides support for the DSC. The DSC rests on a support rail assembly which is anchored to the walls of the HSM. The rear end of the DSC support rails are equipped with stopping blocks. The purpose of these stopping blocks is to establish the final axial position of the DSC inside the HSM. The fical positioning is achieved when the DSC, being pulled into the HSM by the ram, makes contact with these stopping blocks. After the DSC insertion into the HSM is completed, a seismic retaining assembly will be attached to the grappling plate of the DSC top cover plate. This will prevent any possible axial sliding of the DSC during any postulated accident such as an earthquake. The front access of the RSM is i

covered by a steel plate. The air inlets and outlets are covered with stainless steel wire bird screens to prevent foreign objects from entering the HSM.

f)

Weste Processing - During the normal storage of IFA's in the ISFSI, no i

waste will be generated. However, contaminated water and possit Ly con-taminated gases will be removed from the DSC cavity during the cask drying.

operation. The cask drying operation will take place in the HBR2 decontami-nation facility. The HBR2 utilities and radioactive waste processing system are described in Section 9.1.4. and Chapter 11.0 of the HBR2 Updated FSAR (Reference 3.2).

i 1

l l

i I

1

\\

l 4

3.1-4 Amendment No. 1

L ISFSI SAR TABLE 3.1-1 PHYSICALCpCTERISTICSOFPWR 4

BASED ON NOMINAL DESIGN FUEL ASSEMBLIES Array 15 x 15 Envelope (in) 8.426 Overall Length 2 (in) 162.05 i

Weight (1bs) 1466 i

Fuel Rod Number -

204 Fuel Rod Length (in) 152.0 Active Fuel Length (in) 144.0 Maximum Distance Between 26.19 Crid Straps (in) i C

1 See Reference 3.4.

2 Additional 3/4 inch added to overall length to allow for irradiation' growth.

4 O

% t No. 1

--..-.o..-.-... -.--.....-..

ISFSI SAR 4

TABLE 3.1-2 ACCEPTABLE RADIOLOGICAL CRITERIA FOR STORAGE OF MATERIAL IN THE HBR ISFSI CRITERIA VALUE 10 Mev/sec GAMMA SOURCE PER CANISTER (total) 1.48 x 10 i

Fractional Breakdown

Above 1.3 Nov 0.004 Between 1.3 Mev and 0.8 Mev 0.114 i

Between 0.8 Nov and 0.4 Mev 0.808 Below 0.4 Nov 0.074 9

j NEUTRON SOURCE PER CANISTER (total)** 1.17 x 10 'n/sec Fractional Breakdown

~

Above 5 Mev 5.40 x 10 n/sec = 5.41%

Between 2.5 and 5 Mev 2.43 x 10 n/sec = 24.'32%

Between 1 and 2.5 Nov 4.56 x 10 n/sec =~45.67%

.i Below 1 Mev 2.45 x 10 n/sec = 24.53%

3 I

Fractional breakdown. based on isotopic composition and resulting gasmse spectrum calculated by ORICEN2 analysis.

Spectrum from U-235 fission, total number of neutrons per second from ORIGEN2 analysis.

9 a

l l

3.1-6

'---' se No. 1

'l

~.

ISFSI SAR 3.2 STRUCTURAL AND iSCHANICAL SAFETY CRITERIA j

The H. B. Robinson Independent Spent Fuel Storage Installation is designed to perform its intended function under extreme environmental and geological hasards as specified in 10CFR, Part 72.72(a). The HSMs are installed on a 1

monolithically placed, reinforced concrete sat foundation. The mat foundation is also designed to r.esist forces generated by extreme environmental and geological conditions. Specifics of the foundation design are reported in Section 8 3 The environmental features at the ISFSI site, which are used to define the

]

normal operating design basis for the DSCs and H3Ms, are such that they are th's same as or enveloped by those specified in table 3.2-1 of the NUHOMS Topical Esport (Esference 3 1). Specifically, the highest recorded ambient temperature of 107'F, and the lowest recorded temperature of -$'F, as reported in Section 2 3 2.1 of this SAR, are well within the extreme ambient range of 125*F to -40'F specified in the NUHOMS Topical Report. The maximum diurn'al temperature range for the ISFSI site is 25*F.

This range is also lower than

)

the45'Fdiurnaltemperaturerangespecifiedinthe^abovereferencgdreport.

l The site area design basis solar radiation value of 188 Beu/hr-(f t ) is the same as that reported in the above referenced report.

In general, the structural and mechanical safety criteria of the ISFSI are the same as or enveloped by the criteria specified in Section 3 2 of the NUHOMS j

Topical Report.

)

, O As stated earlier in this report, some design features of the H. B. Robinson ISFSI differ from those of the NUHOMS generic concept.

In particular, the T

NUROMS Topical Report addresses an eight module unit, whereas the H. B. Robinson ISFSI is a three module, monolithically poured in place un(t, with a rear access penetration which allows for the hydraulic ran to be operated from the back of the modules. These unique features of the ISFSI reinforced concrete modules,have no impact on the structural evaluation presented throughout chapter 8 of the NUROMS Topical Report. This is due to the fact that the NUHOMS structural analysis of the HSM utilizes the frame action of the roof slab and the walls of only a single module in the transverse direction. This single module approach conservatively envelopes the structural response of any multi-module units including the three module concept which is the subject of this report. The rear access penetration, however, requires additional shielding evaluation which is presented in chapter 7 of this report.

Some of the design features of the H. B. Robinson's DSCs also differ from those of the NUBOMS Topical Report.

In particular, the bottom region of the DSC has been redesigned to fit into the IF-300 cask.

Furthermore, both the top and bottom regions, the spacer disk and the support rods of the DSC have been redesigned to withstand the inertia forces associated with the cask drop accidents in which the drop heights are significantly greater than the minimum 8 foot drop criteria drop criteria established earlier in this report. This was done for compatability with future shipping options. Another unique feature of the H. B. Robinson DSCs, is the instrument penetration at the O

bottom region of one of the DSC assemblies.

The instrumentation of the DSC is for the purpose of collecting temperature data during the first year of the storage. The penetration assembly is designed to maintain the confinement 3.2-1 Amendment No. 1

j integrity of the DSC during both normal operating and accident conditions.

The structural evaluation of the instrument penetration under various normal i

operating and accident conditions is addressed in chapter 8 of this report.

3.2.1 TORNADO AND WIND LOADINGS 3.2.1.1 Applicable Design Parameters The ISFSI will be constructed within the existing boundaries of the H. B.

Robinson Steam Electric Plant which is located within Region I of the NRC i

i Regulatory Guide 1.76 regionalization, and as such, the intensity of the i

design basis tornado for this region is the same as that assumed in the NUHOMS l

Topical Report (Reference 3.1).

f The design basis tornado (DBT) intensities were obtained from NRC Regulatory Guide 1.76.. Region I intensities were considered since it has the most severe parameters. For this region, the maximum wind speed is 360_ miles per hour, the rotational speed is 290' miles per hour, the maximum.translational speed is 70 miles per hour, the radius or maximum rotational speed is 150 feet, the i

pressure drop across the tornado is 3.0 psi, and the rate of pressure drop is 2.0 psi per second. The maximum transit time based on the specified 5 miles per hour minimum translational speed was not used since the transit time is

]

conservatively assumed to be infinite.

3.2.1.2 Determination of Forces on the Structures The forces due to the design basis tornado and tornado generated missiles are O

enveloped by those reported in the NUHOMS' Topical Report.

The method of analysis for overall and local damage prediction due to a design basis tornado and tornado generated missiles is discussed in Section 3.2.1 of the NUHOMS Topical Report, and is fully applicable to the Robinson site specific analysis.

3. 2 '. l. 3 Ability of Structures to Perform The ISFSI is designed to withstand the design basis tornado wind loads.

Furthermore, all components of the ISFSI with the exception of the air outlet shielding block are designed to. withstand the tornado generated missile forces. The loss of an air outlet shielding block is addressed in Section 8.2.1 of this report.

Since the ISFSI is noti housed in any storage building, there is no possibility of any roof collapse on the facility. However, the possibility of total air inlet and outlet blockage by foreign objects or burial under debris during a j

tornado event is considered. The effect of facility burial under debris is l

presented in Section 8.2.

3.2.2 WATER LEVEI. (FLOOD) DESIGN The maximum flood water level at the Robinson site is 222 ft. elevation (see Section 2.4).

The grade level of the ISFSI foundation is at the 234 ft.

elevation. Therefore, there is no possibility of flooding within the ISFSI.

/

3.2-2 Amendment No. 1

ISFSI SAR 3.2.3 SEISMIC DESIGN i

3.2.3.1 Input Criteria

{

The maximum horizontal ground acceleration specified for HBR2 is 0.20g for safe shutdown earthquake (SSE) (see Section 2.6.2.3 and Reference 3.2).

The maximum vertical ground acceleration is specified at two thirds of the hori-sontal component, or 0.1333 These horizontal and vertical component values I

j are less than the values of 0.25g and 0.17, respectively, specified in the 3

i NUHONS Topical Report (reference 3.1).

Hence, the seismic evaluation con-l tained in Section 8.2.3 of the referenced report is fully applicable to those I

components of the H. B. Robinson ISFSI which have design features similar to those of the NUHOMS generic concept.

It is also applicable to those that can be conservatively enveloped by the NUHOMS generic assumptions, such as the single module approach to HSH evaluation discussed earlier in this section.

i In this manner, the seismic response of the HSM and the DSC support assemblies of the H. B. Robinson ISFSI are enveloped by the responses reported in the NUHOMS Topical Report for these components. The DSC, however, which has some unique features, is analyzed for the site specific seismic event, utililizing the methodology and the analytical approach of the NUHOMS Topical Report. The seismic evaluation of the DSC is contained in Section 8.2 of this SAR.

}

3.2.3.2 Seismic-System Analysis i

j The stresses in the HSH and the DSC support assembly due to the 0.20g horizontal and 0.1333 vertical acceleration are enveloped by the results of i

the generic seismic analysis reported in the NUHOMS Topical Report. The stresses in the DSC due to the horizontal and vertical seismic acceleration i

specified above are evaluated and reported along with the DSC rollover T'

l evaluation in Section 8.2 of this report.

The ISFSI foundation and the HSMs tie down system are also designed to 1

withstand.the. forces generated by the SSE. The details of the foundation design are provided in Section 8.3 of this report.

3.2.4 SNOW AND ICE LOADS

~

l The NUHOMS Topical Report specified a postulated live load of 200 pounds per square foot which conservatively envelopes the maximum snow loads for the l

Robinson site (see Section 2.3 for meteorology of the site area).

)

3.2.5 CONBINED LOAD CRITERIA Load combination criteria established in the NUHOMS Topical Report l

(Reference 3.1) for the HSM, DSC and DSC support assembly are also applicable to the HBR ISFSI. The specific load combination evaluation of the DSC, utilising the NUHOMS criteria, are reported in Section 8.2 of this report.

i The facility's mat foundation is designed to meet the requirements of ACI 349-80 (3.6). The ultimate strength method of analysis was utilized with -

appropriate strength reduction factors. The load combination procedure of a

Section 9.2.1 of the ACI 349.80 was used in combining normal operating loads l

(i.e. dead loads and live loads) with severe and extreme loads (i.e., seismic l

and tornado loads)4 The details of the load combination procedure are i

3.2-3 Amendment No. 1 l

4 w.

T ISFSI SAR described in the NUHOMS Topical Report.

Specific foundation analyses 3

including load combination and anchorage analysis are presented in Section 8.3 i

of this report.

1

\\

l O

e e

i T

S I

I

]

t i

i l

/

)

3.2-4 I

h dment No. 1 i

_ _ _ _. ~. _,.

ISFSI SAR 3.3 SAFETY PROTECTION SYSTEM O

3.3.1 CENERAL The HBR Independent Spent Fuel Storage Installation is designed for safe and secure, long-term containment and storage of IFAs. The equipment which must be designed to assure that the safety objectives are met are shown in Table 3.3-1.

The major festm es which require special design consideration are:

l a)

Double Closure Seal Welds on DSC Upper End b)

Radiation Exposure During DSC Drying and Closure j

c)

Design of DSC Body and Internals for a Cask Drop Event During the Transfer Operation d)

Minimization of Contamination of the DSC Exterior by the Spent Fuel Pool Water e)

Minimization of Radiation Shine During Transfer of the DSC from the Cask to the HSH These items are addressed in the following sub' sections.

3.3.2*

PROTECTION BY MULTIPLE CONFINEMENT BARRIERS AND SYSTEMS 3.3.2.1 Confinement Barriers and Systems T

The ISFSI relies on a system of multiple confinement barriers during all handling and storage operations. Table 3.3-2 from the NUHOMS Topical Report (Reference 3.1) has been included. and summarizes the radioactive confinement barriers and systems employed in the design of the ISFSI.

During transport and storage operations, the IFAs are confined within the DSC. The DSC consists of a cylindrical shell and multiple end plates.

Each end plate will be seal welded to the canister in order to provide redundant seals for the DSC. These redundant seals minimize the likelihood of an uncontrollable release of radioactivity. Detailed discussion of the DSC confinement integrity, including the discussion on helium confinement, is presented in Section 3.3.2.1 of the NUHOMS Topical Report and is fully applicable here.

The criteria for protection against any postulated internal or external natural phenomena are discussed in Section 3.2.and Chapter 8 of this report.

3.3.2.2 Ventilation - Offaas During'the normal storage operations of the ISFSI, there will be no release of radioactivity. Additionally, as discussed in Chapter 8 of the NUHOMS Topical O

Report -(Reference 3.1), there are no credible accidents which could cause a release of radioactivity. Therefore, the HBR ISFSI does not require an offgas system.

/

3.3-1 Amendment No. 1 1

ISFSI SAR During the cask drying operation, water and gas will be removed from the

'l cavity of the DSC. This operation will take place in the HBR2 decontamination i

facility and the water and gas will be routed through Unit 2's existing l

filtration and radioactive vaste processing system.

3.3.3 PROTECTION BY EQUIPMENT AND INSTRUMENTATION SELECTION 3.3.3.1 Equipment The DSC and the CE IF-300 shipping cask are the only equipment that specifically provide protection during normal and off-normal. operations of the ISFSI. The design criteria for the'DSC are provided in Section 3.2 of this l

SAR. The design criteria for the cask are listed in the GE IF-300 Safety i

Analysis Report (Reference 3.5).

3.3.3.2 Instrumentation The HBR ISFSI is designed to be totally passive and therefore, no safety i

related instrumentation is required for operation of the facility. However, two of the DSCs and HSMs will be instrumented for experimental purposes only f

for the one year test period (Agreements with DOE and EPRI).*

The instrumentation is limited to placement of a number of thermocouples within these components. Instrumentation of the HSM dogs not effect its structural and mechanicp1 properties. The placement of thermocouples in the DSC, however, requires a feed-through penetration at the DSC bottom region.

This feed-through incorporates the same backup weld-seal philosophy used in the DSC design. The penetration is also designed such that confinement i

integrity of the DSC is not compromised under both normal operating and

{

accident conditions. The instrument penetration analysis is provided in Section 8.4 of this report.

'3.3.4 NUCLEAR CRITICALITY SAFETY The DSC internals are designed to provide nuclear criticality safety during l

wet loading operations. A combination of administrative procedures, materials l

properties, geometry, and neutron poisons are used to assure that suberitical conditions exist at all times. Further details on criteria which ensure suberiticality are presented in Section 3.3.4 of the NUHOMS Topical Report, (Reference 3.1).

3.3.5 RADIOLOGICAL PROTECTION 3.3.5.1 Access Control A locked fence will be placed around the ISFSI for the purpose of designating the area a radiation control area. The key will be controlled by the HBR2 i

Radiation Control unit. Access to the ISFSI will be on an as needed basis.

1 3.3.5.2 Shielding O

An estimate of collective onsite and offsite doses during operations and i

around the ISFSI are presented in Chapter 7.

3.3-2 Amendment No. 1 i

lL

l ISFSI SAR 3.3.5.3 Radiological Alare System j

There are no radiological alarms required.

1 l

3.3.6 FIRE AND EXPLOSION PROTECTION 3.3.6.1 Fire Protection l

The degree of fire protection a structure requires is based on a number of j

factors type and location of combustible materials and their proximity to or location within the ISFSI, type of construction and its fire resistance i

characteristics, fire barriers, and the ability of the plant's fire brigade to i

j reach and effectively extinguish a credible fire.

No combustible materials are stored within the ISFSI or within the ISFSI's boundaries. There is no fixed fire suppression system within the boundaries 1

of the'ISFSI. The facility is, however, located o'utside the confines of any building and is directly accessible to H8R2's fire brigade. The fire brigade 1

l has access to H8R2's eitisting portable fire suppression equipment or the i

site's water fire protection system, as described in Section 9.5.1 of the H8R2 i

FSAR (Reference 3.2).

3.3.6.2 Explosion Protection The DSC and HSM contain no volatile materials and therefore, no credible j

internal explosion is possible. Internal explosions are not considered as part of the design criteria. The design basis for explosions away from the g

s HSM is bounded by the design basis tornado missile described in Section 3.2 of this report and of the NUHOMS Topical Report (Reference 3.1).

3.3.7 MATERIALS HANDLING AND STORAGE 5

3.3.7.1 Irradiated Fuel Handlina and Storate j

The fuel handling systems used in loading the IFAs into the DSC are presented i

in Section 9.1.4 of the HER2 Upated FSAR (Reference 3.2).

Irradiated fuel j

handling outside the spent fuel storage pool will be done with the fuel i

assemblies enclosed in the canister. Criticality safety during handling and storage is discussed in Section 3.3.4.

The criterion for safe configuration l

Is an effective mean plus two-sissa neutron multiplication factor (k.,g) of j

0.95.

Calculations have shows'that the espected k,gg value is well 5eIow this limit.

1 The basic criterion for the cooling of irradiated fuel during storage is a maximum cladding temperature of 380 C (716'F). Higher temperatures may be.

i 0

sustained for brief periods without endangering cladding lategrity. During canister drying and other normal aed abnormal transients, the criterion is a 1

cladding temperature of 570'C for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . For further details, see 1

Section 3.3.7.1 of the NUHOMS Topical ~ Report (Reference 3.1).

l The canister external contamination limits are the same as the external contamination limits for shipping caskst l

3'3-3 A-a==at No. 1 1

,e

--n n~__._---n,-.m.------n.--.

-,,--- - _. _,, _ -,,,, n -.--.~~-

,._nn,-__-.,,---,,.,.-,,_,,_

. _ _ _ =

1 Bets /Gaansa Emitters 10~4 C1/Ca2 Alpha Emitters 10-5 CL/cm2 1

The canister is sealed by double welds prior to storage so that any contamination of the canister interior or its contents will remain confined j

during transfer and storage.

l 3.3.7.2 Radioactive Waste Treatment 1

1 t

The contaminated water removed from the cavity of the loaded DSC will be handled by H882's radioactive vaste treatment system. The site's radioactive I

j waste treatment system is discussed in Chapter 6 of this report and is i

described in Chapter 11 of the HBR2 Updated FSAR (Reference 3.2).

3.3.7.3 Weste Storaae Facilities

)

i No radioactive wastes will be generated during the life of the ISFSI. The contaminated water removed from the cavity of thi loaded DSC will be handled by 88R2's radioactive waste treatment system. The waste storage facility associated with the H8R2 radioactive waste treatment system is described in l

Chapter 11 of the HER2 Updated FSAR (Reference 3.2).

3.3.8 INDUSTRIAL AND CHEMICAL SAFETY 1

i No hasardous or volatile chemicals or chemical reactions are involved in the i

operation of the ISFSI and.therefore, were not considered in any of the facility's design criteria.

I 7

i j

l 1

1 i

I l

,1 i

i

/

1 i

3*3-4 m e y,,

g i

l

3 ISFSI SAR

)

4.2 STORAGE STRUCTURES j

j 4.2.1 STRUCTURAL SPECIFICATIONS e

i I

4.2.1.1 Desian Basis i

j The ISFSI and all its components are designed in accordance with the requirements of 10CFR part 72.

The components are designed to maintain their j

dimensional and structural integrity and safely perfore their intended i

functions under normal and off-normal operating conditions and during postulated geological or environssotal events. The loading conditions l

associated with normal and accident events are specified in Section 3 2 and i

Chapter 8 of this report.

f 4.2.1.2 Construction. Fabrication, and Inspection i

a)'

Eorisontal Storane Module - The RSM is constructed of reinforced concrete. The RSM is designed in accordance with the requirements of the ACI 349-80. The applicable American Society of Testing and Material (ASTM) standards referred in Section 3 8 of.the ACI code was used as part of the 1

fabrication and construction requirements of the RSM. Construction of the RSM 1

is in accordance with ACT 301-84; Specification for Structural Concrete for i[

Buildings.

The concrete materials used for construction of the RSM and the foundation i

consist of: Type II cement conforming to ASTM C150, fine and coarse aggregates conform (as to ASTM C33, concrete air-entraining adeixture 1

conforming to ASTM C-260, and reinforcing bars conforming to ASTM A615

?

Grade 60. The concrete compressive strength specified is 4000 pet at 28 days. Minimum specified density of concrete is 145 pounds per cubic l

l

foot. The RSM walls are tied to the foundation by reinforcing dowels to

}

prevent possible overturning or sliding during any accident condition specified in Section 8 2 of this report.

b)

Dry Shielded Canister - The DSC is designed and fabricated in i

accordance with the requirements of the American Society of Mechanical-Engineers ( ASM) Boller and pressure Yessel Code,Section III, Division 1, Subsection NB, 1983 Edition. Material selection, weldlag, and inspection of l

the DSC was per requirement of this code and the applicable ASTM, ANSI, and i

other codes and standards invoked by the ASM code.

1 Details of the material and construction of the DSC and its laternals are lf contained in Section 4.2 3.1 of the NUBOMS Topical Report (Esference 4.1).

422 INSTALLATION LAYOUT 1

4.2.2 1 Buildina plans There is no building associated with the EBR2 ISFSI other than the horisontal storage module wh'ich is discussed in Section 1 3 1.2.

!O t

i 1

j 4.2-1 Amendesat No. 1

ISFSI SAR 4.2.2.2 Confinement Features The confinement features for each unit of the ISFSI are provided in Section 3.3.2.1.

4.2.3 INDIVIDUAL UNIT DESCRIPTION Each unit of the ISFSI is composed of two components, a dry shielded canister and a horisontal storage module. The DSC is a weld-sealed stainless steel container that provides confinement.of contaminants associated with the irradiated fuel assemblies, encloses the fuel basket in an inert atmosphere and provides biological shielding at the ends of the canister. The HSM is a reinforced concrete structure that' serves as the primary biological shield for the irradiated fuel assembli'es along with providing protection for the DSC against environmental and geological hasards. A detailed discussion and description of the function, design basis, and safety assurance considerations of each component are provided in Section 4.2.3.1 of the NUHOMS Topical Report (Reference 4.1).

The engineering drawings of the DSC and the HSM, showing plan views, sections, ar.d elevations of these components are provided in ligures 4.2-1 and 4.2-2.

Two of the DSCs and the HSM will be lastrumented for the purpose of collecting data during the first year of storage. The instrumentation is 11mited to j

placement of thermocouples in the DSC and the HSM. The HSM thermocouples are cast in place or attached to the concrete surfaces, and as such, have no i

impact on the integrity of the structure. The DSC thermocouples will be 5

O connected to an esternal cable by means of a specially designed feed-through. This feed-through penetration incorporates the same backup weld-seal philosophy used in the DSC containment design. After the penetration plug assembly has been welded to the bottom of the DSC cover plate, a sleeve will be welded over the plus, forming a redundant seal. Thermocouple sheaths till likewise be brased to the plus assembly at inner and outer penetrations.

Leakage through the individual sheaths will be prevented by meta 11 sing the I

aluminum oside insulating sheath at the inner and outer penetrations and l

brasing the T/C wire to the sheath.

1 1

i l

1 l

I i

!O i

I i

4.2-2 Amendment No. 1 l

4.3 AUIILIARY SYSTEMS The design of the HBR ISFSI IJ based on the NUHOMS system for irradiated

)

l

{

fuel. Each unit of this dry storage system is totally passive and self j

contained, requiring no auxillary systems other than a transfer cask for l

transport operation.

J l

4.3.1 VENTILATION AND OFFGAS SYSTEM 4.3.1.1 Ventilation System The decay heat rejection system for each module is based on natural 4

circulation convection cooling. The IFAs are confined within a double weld sealed DSC. The decay heat from the IFAs is transferred by radiation, l

conduction and convection to the surface of the canister. Air inlets near the l

bottom and air outlets at the top of each HSM allow air to circulate around i

the DSC. Decay heat is removed by convection and radiation from the surface l

of the DSC. The driving force for circulation of the air is thermal buoyancy. An analysis of the HSH ventilation system is described in Section 8.1.3 of the NUHOMS Topical Report (Reference 4.1).

Since the IFAs are 1

confined within the double weld sealed DSC, no filtration system for contamination is required.

1 4.3.1.2 offmas System l

)

The offgas system used during the,DSC drying and backfitting operations is the same as that described in Chapter 6 of this report and Chapter 11 of the H8R2 Updated Final Safety Analysis Report (Reference 4.2).

7j 4.3.2 ELECTRICAL SYSTEM The ISTSI is totally passive and requires no electrical system. However, two

{

of the first three DSCs will contain thermocouples for research purposes j

only. The instrumentation will utilise HBR2's existing power supplies

The

~

l l

power supply systems are described in Chapter 8 of the H8R2 Updated FSAR j

(Reference 4.2).

Any instruments which may be us.ed are for experimental and i

data collection purposes only and therefore require no emergency power source I

or means of ensuring an uninterruptible power source.

1 The BBR2 electrical system associated with the fuel handling area is utilised during DSC drying and backfill operations.

I 4.3.3 AIR SUPPLY SYSTEM i

j 4.3.3.1 Compressed Air i

1 The ISFSI requires no compressed air supply system. The H8R2 compressed air l

supply system will be utilised during the DSC drying operation.

l j

4.3.3.2 Breaching Air The ISFSI is located in an outside environment and requires no breathing air supply.

Provisions for breathing air supply during an emergency situation exist at HBR2.

l l

4.3-1 Amendment No. 1 IIO U

ISFSI SAR i.

4.6 CATHODIC PROTECTION The ISFSI is dry and above ground so that cathodic protection in the form of impressed current is not required. The normal operating environment for all metallic components is well above ambient air temperatures so that there is no opportunity for condensation on those surfaces.

4 The austenitic canister body requires no corrosion protection for any forseeable event. The DSC support assembly components are Type A36 carbon steel. The top surface of the DSC rail is to be machined and plated with a high-phosphorus electroless nickel finish having a minimum thickness of

.001 inch. The remaining surfaces of the rail and other components of the support assembly shall be painted with Carbo-Zine 11 (Raference 4 3).

Consequently, the A36 structural steel are protected against corrosion.

1 4

t f

e I

i i

i I

4.6-1 Amendment No. 1-4

4.7 FUEL HANDLING OPERATION SYSTEM 4 O The HBR ISFSI is based on the NUHOMS system for storage of irradiated nuclear fuel. The basis and engineering design for the various fuel handling systems j

4 used during the operation of the ISFSI are described in the following 1

sections.

l I

4.7.1 STRUCTURAL SPECIFICATIONS j

The bases and engineering design of HBR2's fuel handling systems are described in Section 9.1.4 of the HBR2 Updated FSAR (Reference 4.2).

Other handling and transport operation equipment (cask positioning skid, hydraulic ram, and trailer) are designed to meet the criteria established in Chapter 3 of the j

NUBOMS Topical Report (Reference 4.2).

This equipment is designed to safely perform its intended functions under both normal and off-normal ope' rating conditions. However, this equipment does not impact the safety features of l

the ISFSI facility and as such is not safety related.

4.7.2 INSTALLATION LAYOUT i

4.7.2.1 Buildins Plans

!j j

~

Fuel handling operations will occur within the existing HBR2 Fuel Handling Building (see Figure 4.5-1).

4.7.2.2 Confinement Features

}

The irradiated fuel assemblies will only be handled while in the spent fuel i

pool or in the confines.of the DSC which is placed inside the cavity of the shipping cask or the HSM. The confinement features of the H8R2 spent fuel pool are provided in Section 9.1 of the HBR2 Updated FSAR (Reference 4.2) and 3

I the confinement features of the NUHOMS system are discussed in Section 3.3.2

}

of the NUHOMS Topical Report.

i l

4.7.3 INDIVIDUAL UNIT DESCRIPTION 1

l l

4.7.3.1 Shipping Cask Preparation During the preparation, the DSC will be placed into the shipping cask cavity. This operation will take place in the HER2 decontamination facility j

(see Figure 4.5-1).

After loading, the cask and the DSC will be filled with domineralised water and then lifted into the spent fuel pool.

The following components will be used for this operationt I

a)

Spent Fuel Cask Handlina crane - The spent fuel cask handling crane is used to place the DSC into the shipping cask cavity. The design basis and safety assurance features of the crane are discussed in Section 9.1.4 of the HBR2 Updated FSAR (Reference 4.2).

i i

b)

Spent Fuel Cask Lifting Yoke - The spent fuel cask lifting yoke is used in transporting and handling the shipping cask while in the plant and loading the shipping cask onto the transport skid. The design basis and safety l

l 4.7-1 h d=aat No. 1

i features of the lifting yoke are provided in Section 9.1.4 of the HBR2 Updated l

FSAR (Reference 4.2).

4.7.3.2 Spent Fuel Loading i

Loading of the IFAs into the DSC takes place in the existing spent fuel j

pool. The components which are used are described below.

s a)

Spent Fuel Cask Handling Crane - The spent fuel cask handling crane is used to transport the shipping cask to and from the spent fuel pool. The

. design basis and safety assurance features of the crane are discussed in Section 9.1.4 of the HBR2 Updated FSAR (Reference 4.2).

b)

Spent Fuel Pit Bridge - The spent fuel pit bridge is used to move the i

fuel assemblies within the spent fuel pool. The design basis and safety features are descrited in Section 9.1.4 of the HBR2 Updated FSAR (Reference 4.2).

l 4.7.3.3 DSC Drying, Backfilling and Sealing i

l Once the IFAs have been placed into the DSC, the top lead shield plug is-placed on the DSC. Using the spent fuel cask handling' crane, the loaded DSC, j

sitting in the spent fuel cask cavity, will be removed from the spent fuel pool and moved to the decontamination facility where the cask will be drained. The DSC will be drained, vacuum dried, and backfilled with helium.

The top lead shield plug and cover plates will be seal welded to the DSC 4

body. After these operations, the cask will be placed onto the transport skid and taken to the ISFSI site where the DSC will be loaded into the HSM.

i Throughout all of the above operations, the fuel will be confined within the DSC and the DSC will be seated within the shipping cask. The design basis and safety assurance features of the DSC are discussed in Sections 3.2 and 3.3 of i

the NUHOMS Topical Report (Reference 4.1).

More details on the drying and sealing operations are provided in Chapters 4 and 5 of Reference 4.1.

1 l

i

)

f 4.7-2 a-u-em.1 L

REFERENCES:

CHAPTER 4 4.1 NUTECH Engineers, Inc., " Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel," NUH-001, Revision 1, November 1985.

i 4.2 Carolina Power and Light Company, "H. B. Robinson Steam Electric Plant Unit No. 2 Updated Final Safety Analysis Report," Docket No. 50-261, License No. DPR-23.

4.3 Carboline Company, St. Louis, MO.

/

7 l

l

~

Amendment No. 1 m

ISFSI SAR 5.0 OPERATION SYSTEMS 5.1 OPERATION DESCRIPTION The following sections describe the operating procedures which are unique to the operations of the H. B. Robinson (HBR) Independent Spent Fuel Storage Installation (ISFSI) such as loading, unloading, and surveillance. Standard fuel and cask handling operations which are currently being employed at H. B. Robinson Steam Electric Plant Unit No. 2 (HBR2) will be incorporated with these procedures.

5.1.1 NARRATIVE DESCRIPTION The,following steps describe the operating procedures for the ISFSI.

5.1.1.1.

Preparation of the Transfer Ctsk and Canister a)

Prior to the start of the operation, the fuel assemblies to be placed in dry storage will be visually examined (e.g., by television monitors) to insure that no visible defects exist and that the assembly structure is intact. The assemblies will also,be checked (by analysis or by examination of appropriate records) to verify that they meet the physical, thermal and radiological criteria described in Chapter 3.

Measures will be taken to ensure that no known failed fuel will be placed in dry storage. This process will be independently verified to ensure that fuel assemblies meeting the fuel specifications (see Chapter 10)'are selected for storage.

b)

Prior to the loading of fuel, the CE IF-300 shipping cask cavity will be cleaned or decontaminated as necessary. The fuel basket used during the transportation operation will be' removed.

c)

Place the cask in the vertical position in the decontamination facility.

d)

Using the spent fuel cask handling crane, lower the docking collar (Figure 5.1-1) onto the cask. Once the docking collar (cask extension) is properly oriented onto the transportation cask, bolt the docking collar into place and tighten.

e)

Using the crane, lower the dry shielded canister (DSC) into the cask cavity.

f)

Fill the DSC and the cask-canister annulus with clean, domineralized l

I water.

g)

Seal the top of the gap between the DSC exterior and the cask interior.

l h)

Place the lid on the cask and lift the cask into the spent fuel pool.

l 5.1.1.2 Fuel Loading O

a)

Remove the cask lid and place the irradiated fuel assemblies (IFAs) in the DSC (which is inside the shipping cask) using the existing HBR2 fuel

/

handling equipment and procedures.

5.1-1 Amendment No. 1 i

ISFSI SAR b)

When all seven of the IFAs have been loaded into the DSC and the cask lid has been secured, the cask will be moved to the decontamination facility.

5.1.1.3 Cask Dryina Process a)

Place the cask in a vertical position in the decontamination stand.

b)

Remove the cask lid.

c)

Lower the water level (about 2 in.) by removing approximately 15 gallons from the DSC. Lower the water level (about 2 in.) in the cask-canister gap by removing 0.25 gallons of water from the cask.

d) '

Seal weld the upper steel cladding plates of the top lead plug to the canister body.

e)

Connect a compressed air supply to the vent tube and' another hose from the siphon tube to the decontamination facility's radioactive waste system.

Activate the air supply forcing the remaining water out of the DSC cavity.

(See Figure 4.7-1 of the NUHONS Topical Report (Reference 5.1) for a schematic

~

of the piping system).

f)

Once the water stops flowing from the DSC, remove the piping from the l

siphon line.

3)

Seal veld the prefabricated plug over the siphon tube connection.

l h)

. Connect the vent tube piping system to the intake of:the vacuum pump.

l A hose should be connected from the discharge outlet of the vacuum system to the site radioactive vaste system.. Connect the vacuum system to the helium source.

1)

Start the vacuum system and draw a vacuum of 3mm He within the DSC cavity.

j)

Once a vacuum of 3mm Hg has developed in the DSC cavity, disengage the vacuum pump and backfill the DSC with 1.5 sta (22 peig) of helium.. Verify with pressure gauge that pressure is holding and helium leak test the entire length of the primary and plus closure weld.

k)

Seal weld the prefabricated plus over the vent tube connection.

5.1.1.4 DSC Sealina Operations a)

Place the top cover plate onto.the DSC and seal weld the cover plate to the body of the DSC.

b)

Place the cask lid onto the cask and bolt the lid into place.

5.1.1.5 Transport of the Cask to the Horisontal Storate Module (HSM)

O a)

Move the cask to the transport, skid and lower the cask onto the skid.

I 5.1-2 Amendment No. 1

ISFSI SAR b)

Transport the cask to the horizontal storage module.

5.1.1.6 Loadina of the Canister into the HSM a)

Inspect all air inlets and outlets on the HSH to ensure that they are clear of debris.

Inspect all screens on the air inlets and outlets for damage. Replace screens if necessary. Leave the front access of the HSM open.

b)

Position the cask so that the docking collar is within i foot of the HSM.

c)

Remove the cask lid.

d)

Using the optical alignment system and the targets on the cask and HSM, adjust the position of the cask until the :ssk is properly positioned with respect to the HSM.

e)

Using the optical alignment system, align the ram with the cask centerline.

f)

Estend the hydraulic ran and activate the grapple to grab the canister.

l g)

Nove the cask against the HSM, so that the docking collar is positioned in the HSM recess.

h)

Retract the piston of the hydraulic ram. If the ran fails to retract O

when the load on the hydraulic system exceeds 6,200 lbs, stop the RAM. Check the orientation of the cask with respect to the HSM and reorient the cask if necessary. Continue this step until the DSC contacts the stopping blocks mounted on the top of the DSC support rails at the rear end. These blocks will assure correct asial positioning of the DSC within the HSM.

1)

Withdraw the hydraulic ram.

j)

Pull the skid away from the HSM.

k)

Install the steel plate over the front access of the HSM.

1)

Insert through the rear access opening the seismic retainer assembly until it couples with the DSC grapple assembly. Bolt down the retainer assembly to the rear access cover place. Bolt down the cover plate to the rear access embedded plate.

5.1.1.7 Monitorina operations a) on a daily basis, site personnel will walk around the perimeter of the HSMs to visually inspect the air inlets and outlets to insure that they remain unblocked and the integrity of the screens remains intact. Any debris present will be removed.

If damage is evident, appropriate remedisi action will be.

taken. The level of action will be responsive to the observation and will be O

sufficient to ensure safe operation of the facility. For exampist if superficial damage to a screen has occurred, an effort will be made to

' determine the cause and prevent its reoccurrencel if the screen has lieen i

5.1-3 Amendment No. 1 l

t

i punctured, it will be replaced, the air passage it covers will be examined O

using a boroscope and any obstructing material removed.

If the boroscopic examination indicates a possibility of further blockage, that too will be checked.

i b)

On a yearly basis, the site personnel will visually inspect the j

internals of one of the loaded HSMs. This inspection will include the inlet chamber, the outlet pathways, inside concrete surfaces, the embedded bolts, the floor, and the DSC support rails. The purpose of this inspection is to verify lack of blockage, organic matter growth, concrete degradation, and steel oxidation.

Response to any finding potentially affecting the safe operation of the i

facilii.y will be appropriate to the finding. If a generic problem is indicated, the remaining modules will be inspected and appropriate remedial action taken. It is anticipated that the many conservatisms and safeguards l

inherent in this design will ensure safe spent fuel storage over the lifetime of the facility.

5 1.148 Unioadina the DSC from the HSM a)

Eamove front access cover and rear cover plate and seismic

retainer of the HSM.

b)

Using the optical alignment system, align the cask and ran with respect.

to the HSM.

c)

Align the ran with the cask centerline. Extend the ran through the T

t rear access of the HSM until'it contacts the DSC.

d)

Activate the grapple of the ram.

e)

Activate the ran and push the DSC lato the cask.

f)

Retract the ran piston out of the cask and HSM.

I g)

Slowly Pull the cask forward 1 foot away from the HSM.

l h)

Place the cask lid,onto the cask and bolt the lid into place.

i)

Close the front and rear accesses to the HSM.

The fuel is in a safe configuration in the DSC within the IF-300 shipping cask. During the one year demonstration phase, provisions will be made for the DSC to be returned to the spent fuel pool, if necessary. Possible conditions upon which the DSC would be returned include exceeding the design limits shown in Sections 10.2 and 10.3.

Additionally, if shipping were required, the DSC would be returned to the decontamination area or to the fuel pool for removal of the cask collar and placement of the BWR head on the IF-300.

5 1-4 Amendment No. 1

_,.. - - - ~. -

ISFSI SAR 5.1.2 FLOW SHEET A flow sheet for the handling operations is presented in Figure 5.1-3.

j 5.1.3 IDENTIFICATION OF SUBJECTS FOR SAFETY ANALYSIS 5.1.3.1 Criticality Prevention Criticality is prevented by geometrical separation of the guide sleeves and by boron poison contained in the boral guide sleeves of the canister basket. All DSC baskets will include seven boral guide sleeves.

5.1.3.2 Chemical Safety i

There are no chemicals used during the operation of the ISFSI that require i

special precautions.

5.1.3.3 operation Shutdown Modes i

The ISFSI is a totally passive system and therefore this section is not applicable.

5.1.3.4 Instrumentation The ISFSI is a totally passive system requiring no instrumentation. However, some of the units will be instrumented for experimental purposes only. The description of the instrumentation is provided in Section 4 of this report.

5.1.3.5 Maintenance Techniques The ISFSI is a tot' ally passive system and therefore will not' require maintenance. However, to insure that the airflow is not interrupted, the module will be periodically inspected to insure that no debris is in the airflow inlet or outlet. This inspection will be performed daily (see Chapter 10).

O

/

I 5.1-5 Amendment No. 1 1

ISFSI SAR 2

i 4

5.2 FUEL HANDLING SYSTEMS 5.2.1 SPENT FUEL HANDLING AND TRANSFER The ISFSI is a modular storage system which provides for the dry storage of irradiated fuel in a horizontal position with natural draft cooling of the dry storage canister. The ISFSI is located within the boundaries of Unit 2 of the H. 8. Robinson Steam Electric Plant and utilises HBR2's existing system for handling the irradiated fuel and irradiated fuel cask. The DSC is designed to i

be used for transporting the spent fuel to a federal reposicory and can be l

removed from the HSH as described in Section 5.1.1.8.

5.2.1.1 Functional Descripclon J

Figure 5.1-3 presents the flow diagrams for the transfer, loading and retrieval operations. The transfer system is composed of the HBR2 fuel handling system, the GE IF-300 irradiated fuel cask, a transport skid, an j

optical alignment system, the hydraulic ram, and the HSM T-section guides.

j Table 5.2-1 lists these major systems and their important subsystems.

i a)

H8R2 Fuel Handlina System - The ISFSI is designed to utilise the existing.HBR2 fuel hand 1:,ng system. The majer components of this system that

'will be employed during the cask loading operation are the spent fuel pic bridge, the spent fuel cask handling crane, and the spent fuel cask lifting l

yoke. A description of these components is provided in Section 9.1.4 of the l

HBR2 Updated FSAR (Reference 5.2).

i i

b)

HBR2 Decontamination Facility - The HBR2 cask decontamination facility consists of a heated water tank, spray header, recirculation pump, permanent j

scaffolding, associated wiring, instrumentation, controls, piping, and power supply cubicles.

c)

Irradiated Fuel Cask - The General Electric IF-300 shipping Cask is j

used to transfer the loaded DSC to and from the HSM. The cask provides j

shielding along the axial length of the fuel during the transfer, loading, and

]

retrieval operations. A description of the cask's cooling and shielding l

capabilities is provided in the IF-300 Shipping Cask Safety Analysis j

Report (Reference 5.3).

The surfaces of the cask which come in contact with the DSC will be treated with a lubricant tha't is compatible with'che spent fuel pool chemistry. The cask docking collar is a circular ring of steel which -is bolted to the top of j

thg cask. The top six inches of the cask docking collar are seated inside of the HSM walls. The cask, HSM, and docking collar serve as the shield for radiation during the transfer operation.

a The cask auxiliary components described above aid in the horisontial transfer of the DSC and were previously described in sections of this document and the WUHOMS Topical Report (Reference 5.1).

l d)

Cask Positioning Skid - The purpose of the skid is to transport the l

cask in a horizontal position to the HSM and,to maintain the cask in'the properly aligned position during the loading and retrieval operations.

l 3.2-1 heendment No. 1

..-.--.-..--.l..

e)

Optical Alignment System - Once the loaded skid has been positioned at the HSM front access, the cask will be aligned with the HSM. The alignment j

system consists of a precision transit and optical targets on the cask ande HSM. Once the cask is aligned with the HSM, the jack system and cask clamping system will insure that the alignment is maintained throughout the transfer or retrieval operation.

The cask position control system physically moves the cask into precise l

alignment with the HSM. It consists of a group of hydraulic jacks to adjust vertical position and a set of hydraulic cylinders to control horizontal position.

f)

Ras and Grappling Apparatus - The ran is a telescopic hydraulic cylinder which extends from the back of the HSM through the length of the HSM. The grappling apparatus is mounted on the front of the piston. Figures 5.2-1 and 5.2-2 show drawings of the hydraulic ran and the grappling l-i apparatus, respectively. The hydraulics for the grappling apparatus are then activated and the arms move out between the cover plate and grappling plate.

j Once the arms are positioned, the ram is retracted, pulling the DSC out of the cask and into the HSM. For retrieval of the DSC, the process is reversed.

g)

HSM T-Section Guide - During the transfer operation, the DSC will slide l

out of the cask and onto the T-section guides which are within the cavity of the HSM. The T-section guides serve as both the sliding surfaces during the i

transfer operation as well as supports during storage of the DSC. The surface.

of the T.-section guide which comes in contact with the surface of the DSC will be covered with solid film tubricant.

5.2.1.2 Safety Features Racept for the transfer of the DSC from the cask to the HSM, the loaded DSC will always be seated inside the cask cavity until it is inside the HSM. The safety features of the HER2 fuel handling systems are described in Section 9.1.4 of the H8R2 Updated FSAR (Reference 5.2).

The safety features of the shipping cask are described in the CE IF-300 SAR (Reference 5.3).

To ensure that the minimum amount of force is applied to the DSC during the

{

transfer operation, the surfaces of the cask and T-section guides which are in i

contact with the DSC will be treated with a solid film lubricant. A low j

coefficient of friction will minimise the amount of force applied to the DSC, j'

thus minimizing the possibility of damage to the DSC.

i The maximum force which may be exerted by the ran is 22,000 lbs. All i

components of the DSC, ram, grappling assembly and DSC supports are designed j

to withstand.this force. Materials and lubricants are specified so that an operating force of 6200 pounds should easily accelerate the DSC from rest and i

L move it into the HSM. A pressure limitation device in'the hydraulic pump will limit the ran force to less than 6200 pounds. The operator can increase the l

ran force to the 22,000 pound maximum design pressure only by dropping the ram and resetting the limit. Operating procedures will prohibit resetting the 1

limit without approval of the supervising engineer. It should be noted that it is not espected that a 22,000 pound force will ever be required. However, i

the system was designed to take such a force if it is ever needed.

/

+

5.2-2 Amendment No. 1 i

1

i

'l ISFSI SAR o

5.2.2 SPDIT FUEL STORACE l

A description of the operations involved in the transfer and retrieval of the DSC to and from the HSH are presented in Section 5.1.

During storage, the i

ISFSI area will be patrolled and the HSM will be visually inspected once per day. He removal of the DSC from storage was described in Section 5.1.1.8 of this docuswat and the NUHOMS Topical Report (Reference 5.1).

5.2.2.1 Safety Features The features, systems and special techniques which provide for the safe loading and retrieval operations are described in Section'5.2.1.2.

1

~

O q

1

\\

l

- I O.

/

5.2-3 Amendment No. 1 4

4

,.,__m.

..,._-,_-,.,,...._v

,._.,,,m

. - - -, -ry-m.,

ISFSI SAR TABLE 5.2-1 TRANSFER SYSTEM COMPONENT LIST HBR2 Fuel Handling System H8R2 Cask Handling System HBR2 Decontamination Facility Irradiated Fuel Shipping Cask l

Cask Docking Collar i

Cask Lid Cask Positioning Skid Tilting Cradia Skid Body (with roller's)

Transport Trailer Optical Alignment System Precision Trenait Optical Targets Cask Position Control System Hydraulic Jacks Hydraulic Cylinders Control Unit Cask Clamping System Hydraulic Ram Crappling Device O

A 6

f

'\\

O

/

5.2-4 Amendment No. 1 l

l

ISFSI SAR 5.3 OTHER OPERATINC SYSTEMS s

The ISFSI is a totally passive storage system which requires no additional operating systems other than those systems associated with the loading and retrieval of the DSC.

5.3.1 OPERATING SYSTEM No operating systems are required other than those used in transferring the DSC to and from the HSM.

5.3.2 COMPONENTS / EQUIPMENT SPARES The only component postulated to be damaged during the life of the in-stallation is the air outlet shielding block. As described.in Section 8.2.1, a tornado induced missile could damage or knock off the shielding blocks.

Consequently, two additional shielding blocks will be precast during construc-tion and maintained as spares at the site. The screens on the air inlets and outlets will be inspected periodically for damage or blockage by debris.

If the screens appear to be damaged they will be replaced. Additional or alternate responses to any event affecting_the integrity of the screens will be appropriate to the level of damage or disturbance observed. For example, if a tree branch is seen to penetrate a screen, the screen and branch would be removed,/ he air passage borescoped, any blockage removed, and the screen t

replaced.

O 5.3-1 Amendment No. 1

ISFSI SAR 54 OPERATION SUPPORT SYSTEM The ISFSI is a self contained system and requires no instrumentation and control systems to monitor any of the safety-related variables. For research purposes, however, some of the DSCs and the HSMs to be installed at the H. B. Robinson facility have been designed to accept instrumentation.

Instrumentation was included as part of an agreement between CP&L, EPRI and the DOE to augment the U.S. data base on LWR fuel rods in dry storage.

The instrumentation of these components is limited to placement of the thermocouples. The DSC thermocouples will be connected to an external cable by means of a specially designed feed-through. This feed-through incorporates the same backup weld-seal philosophy used in the DSC containment design.

l Details of the feed-through are shown in Figure 5 4-1.

After the penetration plug assembly has been welded to the bottom of the DSC cover plate, a sleeve will be welded over the plug, forming a redundant seal. Thermocouple sheaths

~

will likewise be brased to the plug assembly at inner and outer penetra-tions. Leakage through the individual sheaths will be prevented by metalizing the aluminua oxide insulating crystals and brazing the metalized sections at the inner and outer penetrations.

HSM instrumentation wi11 consist of thermocouples cast in place in the concrete and others attached to the surface and at various locations on the heat shield.

5 4-1 Amendment No. 1 l

l

REFERENCES:

CHAPTEa 5 5.1 NUTECH Engineers, Inc., " Topical Report for the NUTECH Horizontal Modular Storage System For Irradiated Nuclear Fuel," NUH-001, Revision 1, November 1985.

5.2 Carolina Power and Light Company, "H. B. Robinson Steam Electric Plant Unit No. 2 Updated Final Safety Analysis Report," Docket No. 50-261, License No. DPR-23.

5.3 Ceneral Electric Co., "IF-300 Shipping Cask Consolidated Safety Analysis Report," NEDO-10048-2, Nuclear Fuel and Special Products Division, March 1983.

T 1

9 5.R-1 Amendment No. 1-

.o...

I ISFSI SAR C) existing experience.

e)

Recess in HSH front for cask docking to reduce scattered radiation during transfer.

f)

Double seal welds on each and of DSC to provide redundant radioactive material containment.

g)

Placing domineralized water in the cask and DSC and sealing the DSC-cask gap to reduce contamination of the DSC exterior during loading.

h)

Placing external shielding blocks over the HSH air outlets to reduce direct and streaming doses.

i)

Passive system design that requires minimum maintenance.

j)

Insertion of internal shielding blocks around air inlets to reduce direct and streaming doses.

k),

Use of existing. shipping procedures and experience to control contamination during handling and transfer of fuel.

7.1.3 OPERATIONAL CONSIDERATIONS Operational considerations at HBR2 that promote th'e ALARA philosophy include the determination o.f the origins of radiation exposures, the proper training of personnel, the preparation of radiation protection procedures, the development of conditions for implementing these procedures, and the formation of a review system to assess the effectiveness of the ALARA philosophy.

Operational radiation protection objectives deal with access to radiation areas, exposure to personnel, and decontamination. Working at or near highly radioactive components requires planning, special methods, and criteria directed toward keeping occupational radiation exposure ALARA. Job training and debriefing following selected high exposure jobs contribute toward reduced exposures. Decontamination also helps to reduce exposure. Procedures and techniques are bas'ed upon operational criteria and experience that have worked to keep radiation exposure ALARA.

Procedures for the ISFSI will be integrated into current HBR2 procedu'res and will incorporate the same ALARA philosophy.

The ISFSI is considered a radiation control area. Therefore, a locked fence is located around the ISFSI. The HBR2 Radiation Control unit controls the key to this area.

O 7.1-2 Amendment No. 1

7.3 RADIATION PROTECTION DESIGN FEATURES O

7.3.1 INSTALLATION DESIGN FEATURES The ISFSI is a passive outdoor storage system.

Each HSM is capable of providing sufficient ventilation to ensure adequate cooling of the DSC and its contents. The convective cooling system is completely passive and requires no filtration system.

7.3.2 SHIELDINC 7.3.2.1 Radiation Shielding Design Features Radiation shielding is an integral part of both the DSC and HSM designs. The features described in this section assure that doses to personnel and the public are ALARA.

The DSC body is a section of 0.5 inch thick, 36 inch inside diameter stainless steel pipe. Two lead-filled end plugs and three steel plates provide shielding at the ends of the DSC.

During handling operations, shielding in the radial direction is provided by the IF-300 shipping. cask.

Two penetrations in the top lead plus allow water draining," vacuum drying and helium bdekfilling of the DSC. The penetrations are located away from fuel assemblies and contain sharp, non-coplanar bends to reduce radiation streaming. Table 7.3-1 lists relevant dimensions of the shielding materials present at the ends of the canister.

The ESM provides shielding in both the radial and axial directions during the 7

storage phase. Forty-two inch thick, portland-cement concrete walls and roofs i

provide the shielding. The module's front access is covered by a two-inch I

thick steel plate.

Four penetrations in the module allow convective air cooling of the DSC and module internals. Two identical intake vents at the bottom of the front HSM wall draw air into a shielded box inside the module. The exit vents are l

placed at both ends of the module roof. Openings to the HSM interior are placed above the end shield regions and not directly over the. active fuel region. Sharp duct bends and precast concrete shielding caps over the exhaust exits assure that radiation streaming is reduced to a minimum. Figure 4.2-2 shows details of the module penetrations.

Further details of the radiation shiel. ding design features are presented in Section 7.3.2 of Keference 7.2.

4 7.3.2.2 shielding Analysis The shielding analysis methodology for the NUHOMS generic design (Section 7.3 of Reference 7.2) is applicable to the HBR ISFSI as described in this section. However, due to the increased burnup and decreased enrichment of the HBR fuel, the neutron and gamma source, terms are slightly different from those O

used for the generic design.

Source terms are 11.4% lower for gamma rays and 17.2% higher for neutrons. This causes the dose rates calculated in Reference 7.2 to require scaling for use in this SAR. Also, because of the S

7.3-1 Amendment No. 1

l ISFSI SAR revised structure criteria and the necessity of fitting into an existing shipping cask, some steel and lead plate thickness on the DSC have changed.

i i

The HSH also has a rear ram access. These design changes required complete analyses of some of the HER ISFSI shielding.

Figure 7.3-1 shows the locations at which dose rates are presented.

Table 7.3-2 shows the resulting surface doses for HBR fuel. HBR ISFSI-unique shielding calculations have been performed for points shown in Figure 7.3-1.

Dose rates at other points were scaled from the values in the Topical Report. The following paragraphs provide a brief description of the analysis at each point that was reanalyzed.

Due to the different design of the bottom shield plug on the DSC the shielding analysis of the HSM front air outlet was redone (points 2.1 and 3.1).

This analysis used the DOT code to calculate the neutron and samma dose rates at point 3.1.

Hand calculation was then used to calculate the attenuation by the shield cap of the radiation beam coming out of the air outlet slot. The results are shown in Table 7.3-2.

Points 4, 5, and 7 provide estimates of the dose rate at the front of the module with the door open or closed. The dose rate at point 5 was obtained with the DOT analysis of the front half of the module (same analysis used for point 3).

Dose rates of point 7 (4.5 feet away from the open door) and point 4 (surface of closed door) were obtained by hand calculations.

Because the dose at point 6 is primarily due to radiation leaving the canister O

longitudinal surface (and the design for the topical and the HBR' canisters are identical for the canister body), the dose rate here was obtained by scaling 7._

the topical dose races by the factors of the source strengths. The dose rates at points 8 and 9.1 were obtained by using the dose race on the top cover plate surface of the DSC from the DOT analysis and a hand calculation to determine the attenuation due to the 3 1/2 feet distance to the outside HSM surface and the attenuation of the above plate (point 8).

The dose rate at points 8 and 9.2 were obtained from a DOT analysis of the rear of the HSM with the, canister half way in the HSM (a scoping study was done by hand albedo i

methods to determine that this was the case for the worse dose at the uncovered rear ram exit.

The dose rates for all points on the DSC top and transfer cask were' calculated using the QALHOD (gamma only), DOT and the MORSE (Monte Carlo) programs. The three different codes were used to assure that the radiation streaming through the cask-canister gap was adequately modeled and estimated. The dose ra fromallthreemethods'yieldedsimilarresult's(reductionfactorsof10~ges).

1 The QAD and DOT analysis methodologies were described in the Topical Report.

The NORSE calculations are described below.

MORSE, a three-dimensional Monte Carlo shiel' ding code (Reference 7.5), was used to assess the severity of the neutron and gamma ray streaming which occurs when the loaded DSC is inside the transport cask. The shielding calculations were performed in 22 neutron groups and 18 gamma ray groups with 3 expansion of the angular distributions using a coupled cross section O

aP set.

Thus', secondary gamma rays are included when primary neutrons are taken l

as the source.

t l

3~

Amendment No. 1 l

1

ISFSI SAR The HORSE model was constructed in three dimensions using the MARS geometry j

package. The PICTURE code was used to verify the model. One octant of the l

cask / canister system was modeled with reflective boundaries at symmetrie planes. Fuel assemblies were modeled discretely as two-sone regions containing an outer layer of (homogenized) 0.125 inch thick Boral and a homogenized interior composed of irradiated fuel and cladding. The upper two spacer disks were modeled discretely due to their significant effect of local dose rates in the area of interest. Additional spacer disks were omitted to reduce computational time and ensure conservatism. The lead region in the top shield plus was modeled as a disk with reduced diameter to account for the steel siphon line region. The cask / canister system was modeled with the nominal 0.25 inch annular gap. A boundary crossing routine was employed to determine the average dose rate in the annular gay containing air only. The choice of boundary crossing, rather than point detector collision estimating, was made in order to allow octant modeling and to improve the statistical deviation.

The source terms were obtained in a fashion similar to the WUHOMS Topical Report and will not be discussed here. Russian Roulette, path length stretching, and source energy biasing were all used to minimize statistical deyistion in the area of, interest. Russian Roulette weighting parameters were established based on the number of mean free path's from significant sources to the cask / canister gap. Path length stretching was in the forward direction.

Adjoint ISDENPM-S (Reference 7.5) calculations were used to determine source energy biasing parameters.

Dose rates were calculated from the fluxes by using the Snyder-Neufeld factors for neutrons and the Henderson factors for gamma rays (Reference 7.8).

i The application of MORSE *to the cask / canister stressing problem represents a i

refinement in the albedo technique used in the NUHOMS Topical Report. Sixty-four thousand, eight hundred neutron histories and 1,920,000 primary gamma ray histories were executed to obtain streaming dose rates. The neutron dose rate 1

(which includes a negligible contribution due to secondary gamma rays) was calculated to a one-sigma deviation of 6.6%.

The primary gamma ray dose rate was calculated to a one-sigma deviation of 11.9%

The dose rates reported for the DSC in th's cask include numerous combinations of the presence of water in the DSC and DSC-cask gap (the gap-hereafter).

These combinations are present due to the operational procedures. The various cases are described below.

Point 1.1, lead plug on DSC with water in the DSC and gap is the condition during the welding of the lead plus assembly to the DSC. After welding the DSC will be drained, dried and backfilled with He.

The calculations for point 1.2 reflect the estimates for the dose. rate af ter the DSC is drain ~ed.

However, it should be noted that personnel will not be required to be directly over the canister during this time.

All operations will be done from the side of the cask and only the forearms and hands of the personnel will be over the cask for a short time during the connecting and disconnecting (with " quick connect" Swigelok fittings) of 4

water, air and He lines. The dose rate in the gap during this initial welding 7.3-3 Amendment No. 1

ISFSI SAR and drying is shown for point 3.1.

Point 2 gives the dose rates on the top cover plate during the welding of that plate to the DSC body.. The dose rate in the gap is given for point 3.2.

After the top cover plate is welded on the DSC, the cask lid will be lowered 1

into position, bolted on, and then the water de'ained from the gaps. Point 5 gives the estimated dose rate for the top of cask lid.

Point 4 gives the dose rate at the cask side where the operating personnel will be working.

Point 6 gives the dose rate at the cask collar side during the insertion of the DSC. While the DSC is being sesi welded, the lead plug is next to the inner surface of the cask collar and, hence, the dose at the outside of the collar is small. The large dose race shown for point 6 is only present during the loading of the DSC into the HSM. This dose rate is present over a 6-inch wide ring of the collar which is not inserted inside the HSM cask docking recess. Although no personnel will be within 20 feet of this area during loading, it would be prudent, from an ALARA standpoint, to usg portable lead /

polyethelyne shielding to reduce the dose rate in the vicinity of the HSM/ cask interface.

7.3.3 VENTILATION The HSM has a ventilation system to provide for natural draft cooling of the DSC. Howev*r, no off gas treatment system is required due to the low exterior

()

contaminati'on level of the DSC.

The ISFSI is designed for zero release of radioactive material during normal storage of the DSC in the HSM. Th'erefore, additional design work and i

equipment would not result in a reduction of radioactive materials released.

Furthermore, no credible site accident will result in a radioactive leak because of the integrity.of the double seal welds at each end of the DSC, the 1

passive nature of the system, the operational limits and controls used during handling and the integrity of the stainless steel body of the canister.

7.3.4 RADIATION MONITORING INSTRUMENTATION The operation of the ISFSI will be monitored under the HBR2 radioactivity monitoring program. No additional radiation monitoring instrumentation is required.

l i

O 7.3-4 a==ad==at No. 1 l

ISFSI SAR TABLE 7.3-1 DSC END SHIELDING MATERIAL THICKNESSES DSC Top End Shields Lead Shield Plug 0.50 in. Steel + 3.5 in. Lead + 1.75 in. Steel l

Cover Plate 1.25 in. Steel 1

DSC Bottom End Shields Pressure Place 2.00 in. Steel Lead Shield Plug 0.25 in. Steel + 4.75 in. Lead O

~

l 4

1 " Top" and "Botton" refer to the top and bottom ends of the irradiated fuel assemblies.

/

7.3-5 Amendment No. 1

ISFSI SAR TABLE 7.3-2 SHIELDING ANALYSIS RESULTS Nominal Surface Dose Rates (ares /hr)

ISFSI Location Neutron Gamma Total DSC in HSM 1.

HSM Wall or Roof 0.03 2.5 2.5 2.

HSM Air Outlet Shielding Cap 0.03 10 10 2.1 Front HSM Shield Cap 0.06 103 103 2.2 Rear HSM Shield Cap 0.06 10 10 3.

HSM Air Outlet (no ' shielding cap) 3.1 Front HSM Air outlet 3.0 4450 4450 (no shielding cap) 3.2 Rear HSM Air Outlet 1.6 440 440 (no shielding cap) 4.

Center of Door 35 81 126 3 78 428 5.

Cecter of Door opening 50 6.

Center of Air Inlets 27 29 56 7.

4.5 Ft. from HSM Door 7.1 17 24 8.

Ram Opening with Access Pl' te 2.9'

.37 3.3 a

(fully inserted DSC) 9.

Ram Opening Without Access P.'. ate 9.1 Fully Inserted DSC 3.8 0.92 4.7 9.2 Half Inserted DSC 1.2 605 606 DSC in Cask 1.

Centerline Top of DSC Plug 1.1 No Water in DSC, Water in Gap 296 92 390 1.2 Water in DSC and Gap 0.83 36 37 2.

Centerline Top of DSC Cover Plate i

(no water in DSC, water in gap) 252 62 314 3.

Cask / Canister Annular Gap 3.1 Water in Gap and DSC 0.34 390 390 3.2 Water in Gap (no water in DSC) 190 365 555

(~)

4.

Transfer Cask 3.3 2.8 6.1

) U Side Surface (no water in DSC) 7.3-6 Amendment No. 1

~

ISFSI SAR TABLE 7.3-2 (Continued) 5.

Transfer Cask Top Cover Surface 194 24 218 i

6.

Cask Collar During DSC 37 620 657 Transfer from Cask to HSM

{

Note:

1.

These values for worst case situation where no water is presette in the i

gap. For ALARA purposes, water should be present in the ga; which will reduce streaming exposures by approximately 1/20.

i j

l O

T O

i Amendment No. 1

s ISFSI SAR 7.4 ESTIMATED ONSITE COLLECTIVE DOSE ASSESSMENT Estimated doses for tL'e fuel loading, drying, sealing and transfer are provided in Section 9 1.2 of the HBR2 Updated FSAR (Reference 7.1) and Section 7.4.1 of the NUHOMS Topic 01 Report (Reference 7.2).

Onsite radiation dose rates due to the storage ot the fuel'in the first three HSMs are shown in Figure 7.4-1.

If all eight modules are built and filled, the dose rate will be 8/3 times the values listed in Figure 7.4-1.

The resulting dose for a person located at the ISFSI fence for eight hours a day for 250 days per year would be approximately 10 mram. For a person located at the offices, the yearly dose would be less than 1 area.

(Actual dose inside the office is less due to shielding from the building.)

i

\\

l l

7.4.1 OPERATIONAL DOSE ASSESSMENT l

This section establishes the expected cumulative dose delivered to site pertonnel during the fuel handling and transfer activities associated with one ISFSI module. Chapter 5 describes in detail the ISFSI operational procedures, a number of which involve radiation exposure to personnel.

A summary of the operational procedures which result in radiati,on exposure to personnel is given in Table 7.4-1.

.The cumulative dose is calculated by a l

four step process. First, the number of personnel required to perform the i

task is estimated'. Then the time required to perform the task is estimated based on the operational guidelines presented in Chapter 1, engineering judgement, and previous esperience with similar or identical operations. An ambient dose rate is obtained for the operation. It is based on an estimated distance from the individual's trunk to the most significant radiation source. The dose rate is conservatively calculated by modeling radiation T

sources as line, cylinder, or plane sources, as is appropriate to the

~

particular geometry of the operation. With the number of personnel, the time required, and the local dose rate, individual and collective exposures may be calculated.

7.4.2 STORACE TERM DOSE ASSESSMENT Figure 7.4-2 is a graph of the dose rate versus distance from the face of a filled storage array of three ISFSI modules. Direct neutron and gamma flux, as well as the air-scattered radiation from the module surfaces, are con-sidered. Air-scattered dose rates are determined with the computer" code SKYSHINE-II (Reference 7.3).

Initial loading of all modules with the NUHOMS Topical Report design basis five year post irradiated fuel is assumed. If five additional modules are added, the dose rate will be 8/3 times that shown in Figure 7.4-2.

Estimates of cumulative doses to site personnel from three filled ISFSI modules are given in Table 7.4-2.

The dose' rates are based on data shown in Figure 7.4-2.

Occupancy information for number of personnel, location, and time is estimated based on the five year plan for facilities' layout at the HBR2 site. Because of thu very rapid decrease of dose rate with distance from l

the storage facility, a maximum distance of 600 feet was used in these O

analyses.

No credit is taken for shielding of personnel by buildings, or for radioactive decay.

l i

7.4-1 Amendment No. 1

ISFSI SAR TA8LE 7.4-1 StpetARY OF ESTIMATED ONSITE DOSES DURING FUEL HANDLING OPERATIONS I

Average Distence Total

)

From Source Dose Oose Per Personnel Number of Time Surface Rate Person Dose Operation Personnel (Hours)

(Feet)

(aree/hr)

(arem)

(arem) l Location: Fuel Pool

)

i 5 4 40 80 Loed fuel into canister 2

8.0 Bolt Ild assembly onto cask 2

0.5 1.5 22.3 11.2 22.3.

Location: Cask Handling Aree 6.l(2) 48.8 146.4 Decontaminate outer surface of cask 3

8.0 Place' scaffolding around cask 4

0.8 4.0 9.5 7.6 30.4 Unholt Ild, remove tid and spacer 2

0.75 1.5 22.3 16.73 33.5 Remove approscimetely 15 gal. of 2-0.5 1.5 22.3 11.2 22.3 water from DSC and lower water

- O level la canister cask gap T

Wold leed plug to canister 2

3.0 4.0 3.5 10.5 21.0 Hydrotest canister 2

2.0 1.5 22.3 44.6 89.2 IU Remove water,from canister 2

2.3 4.0 29.8 68.5 137.0 cavity Seel wold prefabricated plug to 2

0.5 1.5 210 105 210 siphon tube connection IU vacuum dry canister and 2

4.0 4.0 29.5 118.0 236 backfill with Hellum Hollum leek test veld 2

1.0 1.5 210 210 420 Seel weld prefabricated plug to 2

0.5 1.5 210 105 210 vent tube Perfore ICE (PT) 1 1.0 1.5 210 210 210 install end cap 2

0.5 1.5 210 105 210 0

Wold end cap to canister 2

2.3 4.0 29.5 67.9 135.8 Install cask Ild and bolt into place 2

0.5 1.5 118.3 59.2 118.3 7.4-2 Annendment No. 1

__..-...-_. - _ _. -. -. ~.. _ _, - - - _.., _ - _ - _ _

ISFSI SAR TABLE 7.4-1 (Continued)

Remove scaffolding from around cask 4

0.8 4.0 9.5 7.6 30.4 2

1 2

Transport cask to skid and trailer 2

0.5 Location: Traller Attach skid tle down to cask 2

0.5 1.0 o.l(2) 3.1 6.1 l

Transport cask to HSM 5

0.5 5

6.1 (2) 3.1 15.5 l

Remove cask lid 2

0.5 1.5 118.3 59.2 118.3 Install cesk Jacking system and 4

1.5 3.0 6.l(2) 9.3 37.2 align cask with HSM Transfer canister from cask to HSM 4

0.5 3.0 6.l (2) 3.1 12.4 Install steel plate over front 2

0.5 3.0 2.5(2 )

1.3 2.6 access of HSM Tack weld front access door 2

0.,5 1.5 70.2 35.1 70.2 Install seismic retainer assembly 2

0.5 1.0 4.7 (2 )

2.4 4.8 install cover plate to rear access 2

0.5 1.5 2.5(2) 1.65 3.3 TOTAL 42.95 1366 2635 l

(1) Monitoring operation - personne could leave radiation field.

(2) Conservatively assumed to be su* face dose rate.

1 j

7.4 3 Admendiment No. 1 e

,n--.

.--,,...-._.-nc-,

,-..n-.

i ISFSI SAR l

TABLE 7.4-2

~

ESTINATED ANNUAL ONSITE 00SES OURING STORAGE PHASE Average Distance Total From Dose Dose Por Personnel Number of Time Facility Rate Person Dose Area Personnel (Hours /yr)

(Feet)

(ares /hr) (ares /yr) (ares /yr )

EARC Sullding 30 2000 560 1.8 x 10'3 3.7 110 I

Operations support Building 225 2000 360 4.3 x 10'3 8.9 2000 I

Yard Work Area 13 2000 120 4.9 x 10-2 100 1300 I

3 Yard work Area 12 2000 350 4.4 x 10'3 9.2

- 110 2

I TSCAOF Sullding 25 2000 580 1.6 x 10*3 3.3 83 1

Pr.Isery Access Point 10 8760 290 7.5 x 10'3 66 2

660 Bulk Storage Warehouse 25 2000 360 4.3 x 10*3 8.9 220 I

Dolly Visual inspection 1

60 4,

26 1560 1560 HSM Interior inspection 1

0.5 1.5 4.0 2.0 2

6045 Ione 8-hour shift per day, 5 days per week.

2Con'tInuous occupancy (44 people at one 8-hour shif t por day, 5 days per week, 50 weeks per

...r,.

7.4-4 Amendment No. 1

ISFSI SAR 7.6 ESTIMATED OFFSITE COLLECTIVE DOSE ASSESSMENT j

g Because the ISFSI provides containment yielding no radioactive gaseous or liquid effluents, assessment of offsite collective dose is limited to one of direct and reflected radiation to the nearest residence.

7.6.1 EFFLUENT AND ENVIRONMENTAL MONITORING PROGRAM 4

The ISFSI is located within the protected area of HBR2. The HBR2 environmental program is described in Technical Specification 3.17 (Reference 7.4).

In addition, environmental TLDs are maintained at air sampling sites adjacent to the p.lant boundary. These are located at

.170 degrees, 830 feet and 150 degrees, 1500 feet from the ISFSI. The nearest residence is located at 160 degrees, 1350 feet from the facility. These TLDs are changed quarterly.

7.6.2 ANALYSIS OF MULTIPLE' CONTRIBUTION The contribution of maximum dose to a member of the general public from the ISFSI is calculated to be 1 ares per year. Such a dose is insignificant when compared with an average nat. ural background dose of 100 meen per year, and with the 25 meen per year allowed by 40CFR90 for combined operation.

7.6'.3 ESTIMATED DOSE EQUIVALENTS The maximum exposed member of the public would receive 0.3.arem per year, and O

all others within a 50-mile radius would receive significantly smaller amounts. Meteorological conditicas will not alter the direct radiation doses and are not considered in this analysis. Figure 7.6-1 shows the total yearly 7

dose as a function of distance from the center of the first three HSMs. If five additional modules are added,.the dose will be.8/3 times the dose shown a

in Figure 7.6-1.

I 4

o t

O e

a 7.6-1 hdment No.1

-, =. - _ -.... - - - _... _... _ -. - - -... -... - - _. -. _ _ -..

REFERENCES:

CHAPTER 7 7.1 Carolina Power and Light Company, "H. B. Robinson Steam Electric Plant Unit No. 2 Updated Final Safety Analysis Report," Docket No.

50-261, License No. DPR-23.

7.2 NUTECH Engineers, Inc.,." Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel," NUH-001, Revision 1, November 1985.

7.3 C. M. Lamplay, "The SKYSHINE-II Procedure: Calculation of the Effects of Structure Design on Neutron, Primary Camma-Ray and Secondary Gamma-Ray Dose Rates in Air," NUREC/CR-0781, REA-T7901, USNRC, 1979.

7.4 Carolina Power & Light Company, " Technical Specifications and Bases for H. B. Robinson Unit No. 2," Appendix A to Facility Operating License DPR-23, Docket No. 50-261, Darlington County, SC.

7.5 Oak Ridge National Laboratory, " SCALE: A Modular Code Sy. stem for Performing Standardized Computer Analyses for Licensing, Evaluation,"

ORNL/NUREC/CSD-3.

7.6 Oak Ridge National Laboratory, "ANISN-ORNL Multigroup One Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering," RSIC CCC-254, 1973.

7.7 J. H. Price, N.C.M. Blattner, " Utilization Instructions for QADMOD-C," RRA-N7914, RSIC CCC-396, 1979.

7.8 Oak' Ridge National Laboratory, " Cask 40 Croup Coupled Neutron and Camma-Ray Cross-Section Data," RSIC DLC-23, 1978.

i 4

O

{

f O

/

7.R-1 haaht No. 1

ISFSI SAR 8.o ANAI,YSIS OF DESICN EVENTS l

The purpose of this section is to evaluate the safety of the H. B. Robinson

)

l (HBR) Independent Spent Fuel Storage Installation (ISTSI). The safety evaluation is accomplished by analyzing the response of the various components of the ISFSI to normal'and off-normal conditions and a range of credible and hypothetical accident conditions.

j In accordance with NRC Regulatory Guide 3.48, design events identified by ANSI /ANS 57.7-1981, are used in the safety evaluation of the-ISFSI. In ANSI /ANS 57.7-1981, four categories of design events are defined. Design events of the first and second type are addressed in Section 8.1, and design events of the third and fourth type are addressed in Section 8.2 of this report.

Many of the design events in the above four categories have been addressed in the NUTECH Horizontal Modular Storage (NUHOMS) System Topical Report (Reference 8.1) using enveloping criteria. Whenever the site specific load is l

enveloped by that of the NUHOMS Topical Report, it will be noted and will j

reference the appropriate section of the Topical Report. Additional site specific analysis which,has not been covered in the NUHOMS Topical Report will j

be discussed in' detail in the following sections.

1 As discussed in Section 3.2 of this Safety Analysis Report (SAR), some design features of the HBR ISFSI are unique and differ'from those of the NUHOMS generic concept. In particular,,the HSM is a three module unit with a rear O

without any rear access.

access penetration, whereas the generic concept is an eight module unit However, as discussed earlier the methodology of the 7

structural evaluation of the HSM under ther above categories of design events as utilized by the referenced report is such that it will conservatively i

envelop any modular stacking arrangement including the three unit concept of the HBR site. Hence, the stress evaluation and the analytical results presented in Chapter 8 of the referenced report for the NUHOMS modules are fully applicable to the site specific HSM.

Some design features of the HBR DSC are also different than those of the NUHOMS generic concept. Specifically the DSC is designed to withstand inertia forces associated with cask drop accidents in which the drop height is significantly higher than the soft drop criteria established earlier in this report. Because of these design features, additional structural evaluation of-the DSC is required. The method of analysis, however, for many of the design event cases is the sasa as the methodology utilized in the NUBOMS Topical Report. For these cases the appropriate sections of the referenced report containing the applicable methodology will be referenced. In other cases where a c.^w methodology is utilized, such as the drop accident case, the analytical approach will be presented. In either, case the results DSC stress evaluation will be tabulated and reported throughout this chapter.

The design of the DSC support assembly for the BBR ISFSI is identical to the NUHOMS generic concept and as such the stress evaluation presented in the referenced report is fully applicable to this component.

l 8.o-1

,,,,,,,,,,,, 1

ISFSI SAR Since a foundation design was not included in the NUHOMS Topical Report,

[-

Section 8.3 is included in this Safety Analysis Report (SAR) to describe the foundation design and analysis using the four categories of described above.

As descritad earlier in this report two of the DSCs will be instrumented for the purpoao of coll' acting data. Section 8.4 of this report addresses the safety features of the instrument penetration.

1 I

I e

i 8.0-2 Amendment No. 1

ISFST SAR 8.1 NORMAL AND OFF-NORMAL OPERATIONS Design events of the first type consist of a set of events that occur regularly in the course of-normal operation of the ISFSI. These events are addressed in Section 8.1.1 of this report. Design events of the second type consist of events that might occur with moderate frequency (on the order of once during any calendar year of operation). These off-normal events are addressed in Section 8.1.2 of this report.

l 8.1.1 NORMAL OPERATION ANALYSIS l

The loads associated with the normal operating condition of the ISFSI are as followst dead weight loads, design basis internal pressure loads, design basis operating temperature loads, operation handling loads, and design basis live loads. The structural components effected by these loads are the dry shielded canister (DSC), DSC internals, horizontal storage module (HSM), DSC support assembly and the foundation. The following paragraphs discuss these loads and compare them to the generic assumptions reported in Section 8.1.1 of the NUHOMS Topical Report (Reference 8.1).

a)

Dead Weight Loads - Dead weight analysis contained in Chapter 8 of the NUHOMS Topical Report for the HSM and the DSC support, assembly envelops the Robinson ISFSI analysis. Hence, the analysi's of dead weight in the Topical Report is applicable to the HBR ISFSI' analysis for these components.

i The DSC component weights are tabulated in Table 8.1-1.

The dead weight analysis of the DSC shell is based on the same analytical approach specified in Section 8.1.1.2, Page 8.1-17 of the referenced report. Furthermore, since the total. weight of the DSC is approximately the same as that of the NUHOMS i

generic DSC, the resulting DSC shell stresses are the same. For the dead j

weight analysis of the spacer disk the results of the finite element analysis I

reported in Section 8.'1.1.3, page 8'1-32 of the referenced report can be directly ratioed for the effect of the weight redistribution and the change in the spacer' disk thickness. The site specific spacer disks are 2 inch thick compared to the 1.25 inch of the NUBOMS spacers. Also, the maximum total weight distributed on one spacer is 2034 pounds compared to the 1834 pounds for the NUHOMS. Based on these differences the maximum resulting stress reported in Table 8.1-7 of the referenced report can be ratioed by the relation W

t S

= (SNU)(W) (E )

sp NU sp Wheret S,, = kai, the site specific spacer disk membrane stress SNU = 1.58 ksi, the NUHOMS spacer disk membrane stress W,p = 2,034 lb, weight per site specif. spacer disk WNU = 1,823 lb, weight per NUHOMS spacer disk 8.1-1 A==ad===t No. 1 l

l.

ISFSI SAR NU = 1.25 in, NUHOMS spacer disk thickness t

t,p = 2.0 in, site specific spacer disk thickness Therefore S,, = 1.10 kai The results of the above analyses are tabulated in Table 8.1-2 of this report. The stresses caused by own weight of other components of the DSC and.

j, Its internals are insignificant and do not warrant extra analysis.

b)

Design Basis Internal Pressure Loads - The HER DSCs are operated with l-0.0 pois pressure. However, the DSC is designed for 25.0 psis operating.

pres'sure at off-normal conditions. This pressure is the same as that specified in the referenced report. Since the HER DSC has the same shell thickness as the NUHOMS, the resulting primary membrane stress will remain.

unaffected. However, for the secondary stresses at the discontinuities the i

analysis reported in Section 8.1.1.2, Pages 8.1-21 through 8.1-24 of the I

3 referenced report is reworked to incorporate the change in the effective thickness of the cover plates. The NUHOMS analysis is based on an effective thickness of 1.5 inch, whereas the minimum available cover plate thickness of l

the site specific DSC is 1.75 inch at the bottom region. With all other j

conditions and assumptions being identical, the analysis yields a maximum secondary membrane plus bending stress of 7.64 kai.

For the bending stress on the cover place itself,'the re'sult of the analysis contained in Pages 8.1-24 and 8.1-25 of the referenced report is multiplied by the square of the ratio of the thicknesses. In this manner, the maximum bending stress on the 1.75 inch thick cover plate is 3.27 kai.

The results of the above pressure analysis of the DSC and comparison against i

code allowables are contained in Table 8.1-2 of this report. The maximus DSC internal pressure under accident' conditions is 39.7 peig, which is the same as that specified in the'NUHOMS Topical Report.

c)

Design Basis Operating Temperature Loads - The extreme range of ambient temperature at the Robinson site, as specified in the Updated FSAR (Reference 8.2) is -5'F to 105?F. For the NUHOMS Topical Report design, a range of -40*F to 125'F was assumed. Consequently the thersel analyses of the HSH and the DSC support assembly reported in Sections 6.1.4 through 8.1.5 of the NUHOMS Topical Report conservatively envelopes thos0 of the HBR ISFSI.

The DSC thermal analysis contained in Section 8.1.1.2 of tee referenced report conservatively envelopes the site specific DSC thermal analysis. This is due to the fact that the maximum.shell bending stress reported in that report is based on the generic assumption that no gaps exist between the spacer disk and the inside cavity of the DSC (Section 8.1.1.2, Pages 8.1-26 through 8.1-28).

The HBR DSC, however allows for a nominal radial gap of 0.13 inch. This amount of gap is larger than the differential thermal expansion of the disk.

O Other thermal stress evaluations of the NUHOMS canister, such as the shell stress evaluation due to temperature variation in circumferential direction, 1

and due to dissimilar material, indicated stresses far below the 20.9 ksi

/ obtained for the case discussed above. Hence, for the sake of conservatism 8.1-2 Amendment No. 1 a --. - -.--.--.

. - ~

ISFSI SAR c

I O

and in order to envelope the actual state of thermal stress in the HBR DSC, the thermal stress obtained from the differential expansion of the spacer disk will be reported herein and is tabulated in Table 8.1.2 of this report.

For the thermal expansion evaluation of the DSC internals, the evaluation reported in Section 8.1.1.3, Pages 8.1-33 and 8.1-34 of the referenced report is also fully applicable. This is due to the fact that the gap existing between the top of the fuel region and the bottom of the HBR DSC lead plug is the same as that reported in the referenced report.

i d)

Operation Handling Loads - The handling loads on the DSC, DSC support I

assembly, and the HSM are based on the maximum capacity of the hydraulic ram of 22000 pounds. This capacity is the same as that specified in the NUHOMS Topical Report. Therefore, the handling load analysis of the Topical Report, Section 8.1.1, covers the site specific design. Since the ram mounting plate assembly at the rear access of the HSM is site specific, the loading from the ran on this assembly was investigated. The ram loads are transferred to the wall through the embedded pipe and plate which have velded stud anchors. The 22000 pound loading was found to have a negligible effect on the HSM rear wall. The not effect of the tornado generated missile impact considered in j

the topical report is to load tihe side well with over 1000 kips. The much '

narrower end wall, during operational loading, is easily enveloped by the j

previous analysis. The results of the operational handling load analysis of 22000 pounds are tabulated in Table 8.1-2 of this report.

e)

Design Basis Live Loads - The maximum snow load (or other live loads)

.(

for the Robinson site as derived from the Updated FSAR (Reference 8.2) is i

bounded by the NUHOMS Topical Report which assumes a live load of 200 psf.

.? '

8.1.2 0FF-NORMAL OPERATION ANALYSIS This section describes the design basis off-normal events associated with the operation of the H8R ISFSI. The events which are considered here are expected to occur on a moderate frequency.

8.1.2.1 Transport Off-normal events associated with the transport operation of the DSC may occur due to malfunctioning of the auxiliary components (i.e., crane, transporter ram, etc.), or by misalignment of the DSC with respect to the HSM.

Malfunctioning of the auxiliary components does not relate to the safe functioning of the DSC and can be rectified without any impact to the

+

. operation of the system. As described in Section 1.3.1.7 of this report the only time the cask crane is operating without the redundant yoke is during the i

cask lowering on the skid assembly. A postulated malfunction or more specifically a yoks failure during this operation is considered as part of the cask drop accident which is reported in Section 8.2.4 of this report. The DSC and the ram grappling assembly are designed to the maximum ram capacity loading of 22,000 pounds. Hence, any off-normal event such as misalignment or ram malfunctioning will not cause any damage to any component of the ISFSI.

O The misalignment of the DSC may also cause jasusing or binding of the canister casing. The analysis of the DSC under assumed januming and binding conditions is covered in Section 8.1.2 of the NUHOMS Topical Report (Reference 8.1), and is applicable to Robinson ISFSI operation.

8.1-3 y,,

1

ISFSI SAR All auxiliary components used during the transport operation (i.e., the cask positioning skid, the cask tie-down system, the cradle support, the saddle and the transporter) are designed to withstand the inertia forces associated with

~

transport shock loadings. The DSC and the cask are designed for the postu-laced drop accident. The inertia forces of a drop accident is significantly greater than the transport shock forces and, hence, inertia forces associated with transportation shock for these components are enveloped by the 8 ft drop accident.

8.1.2.2 Air Flow Blockage 1

Another off-normal event that may occur is the possibility of air inlet j

blockage. Because the air inlets are close to the ground, there is a chance i

that they could become blocked with blowing paper, dirt, snow or other debris. Due to the height of the air outlets, their separatior. and since hot air is blowing out of the exits, it is less likely that both the exits would become blocked. Furthermore, blockage of one exit alone would not be as j

severe as blockage of both inlets. Therefore, this off-normal event is defined as complete blockage of the HSH inlets. Blockage of all inlets and outlets'is considered highly unlikely and is presented in Section 8.2 of this SAR.and in Section 8.2 of the NUMONS Topical Report.

l i

The blockage of the air inlets has been addressed i.n the NUHOMS Topical Report in Section 8.1.2.

The results of this analysis indicate that the rise in temperature and pressure in various components of the storage system is well within the acceptance limits. The blockage of the air inlets would be discovered during the normal surveillance of the modules. As the analysis shows, excessive temperatures are not reached and, hence, if the blockage were to occur just after one inspection and not be discovered until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months later, l

no threat'to the public health and safety would result. Once detected, the air inlets will be cleared of the blockage.

8.1.3 RADIOLOGICAL IMPACT FROM OFF-NORMAL OPERATIONS 4

Based on the off-normal operations described in Section 8.1.2, there is no additional radiological impact from the ISFSI beyond what is described in Chapter 7.

f 8.1-4 Amendment No. 1

TABLE 8.1-1 DRY SHIELDED CANISTER AND HORIZONTAL STORAGE MODULE COMPONENT WEIGHTS COMPONENT DESCRIPTION CALCULATED WEIGHT (Pounds) 1.

Dry Shielded Canister:

Casing 2758 Top Grapple Assembly 49 Top Cover Plate 360 Top Lead Casing 703 Top Lead Plug 1384 Top Ring Plate 31 Bottom Cover Plate 601 Bottom Lead Casing 159 Bottom Lead Plug 1549 Total 7594 2.

Canister Internals:

Spacer Disks 1731 4 x 2 1/2" # Support Rods 897 7 x Boral Tubes 843 Total 3471 3.

15 x 15 PWR Spent Fuel Assembly 10154 Total Three Loaded Canisters Weight, 63657 4.

3 Canister Support Assemblies 4725 5.

3-Bay Reinforced Concrete Module 800770 6.

3 x 2" Steel Door 6290 7.

6 x Shielding Blocks 10556 Total (3 Bay ESM Weight Loaded) 885998 9

Amendment No. 1

m.---vmie->M%-en"

-t g >r ee-r-yWNway-awW-Pg r T-

l 4

TABLE 8.1-2 MAXIMUM DRh STORACE CANISTER SHELL STRiSSES FOR NORMAL OPERATINC LOADS STRESS (ksi)

WAD ASME TYPE DESIGN DESIGN OPERATION CODE DSC DEAD BASIS BASIS HANDLING ALLOWABLES COMPONENTS STRESS WEICHT PRESSURE TEMPERATURE (3)

(ksi)

TYPE 1

l 0.21 0.91 7.4 0.38 18.7 Local I

Cani: ster Primary N/A 1.38 7'. 4 N/A 28.05 Shell Membra'ne Primary 11.55 7.64 20.90 10.99 56.10

,7

(_.e Bendtng 7 j j]

N/A N/A N/A N/A 18.70 Cover Plate Primary Membrane +

N/A 3.27 0.45 13.36 28.05 Bending

{#

1.10 N/A N/A N/A 18.70 Notes:

1.

Values shown are maximwns irrespective of location.

2.

Allowable stresses are conservatively taken at 400*F.

3.

Values are based on ram capacity load of 22,000 lb.

8.1-6 Amendment No. 1

ISFSI SAR 8.2 ACCIDENT ANALYSIS This section addresses design events of the third and fourth types specified by ANSI /ANS 57.7-1981 and any other credible accident that could affect safe operation.of the H. B. Robinson ISFSI. The postulated accidents are:

o Loss of air outlet shielding blocks o Tornado and tornado generated missiles o Earthquake g

I o Eight foot drop o Lightning o Blockage of air inlets and outlets o Accident pressurization of the DSC o Fire o Leakage of the DSC o Load Combination In the following paragraphs, the accident analyses for various components of the ISFSI are described. When the accident loads or conditions are the same se (or enveloped by)'those addressed in the NUHOMS Topical Report (Reference 8.1), reference will be made to the appropriate section of that

~

report.

8.2.1 LOSS OF AIR OUTLET SHIELDING This pos,culated accident assumes the loss of both air outlet shielding blocks O

ISFSI are assumed to be in normal condition.

from the top of the horizontal storage module. All other components of the l

The air outlet shielding blocks are designed to remain in place and remain completely functional for all postulated accidents except tornado generated missiles. There are no structural or thermal consequences to the.ISFSI as a result of the loss of the shielding blocks; however, there ara radiological consequences which have been addressed and analysed in the NUHOMS Topical Report, Section 8.2.1.

The resulting increase in air scattered (sky shine) doses or direct radiation as reported in the Topical Report are within 10CFR100 dose limits.

Recovery To recover from a lost or damaged shielding block caused by a tornado projectile, one of the spare blocks is transferred to the HSM. After the sh$sid block is transferred to the HSM, a truck-mounted crane is used to lift the block into position. The block is then bolted in place. The entire renounting operation should take less than 30 minutes, during which a cachanic will be on the HSH roof for approximately 15 minutes. During this time he will receive less than 50 ares. The dose to the crane operator and the mechanic on the ground while putting the shield block in place will be approximately 20 mres each (assuming an average distance of 15 ft from the center of the module roof).

8.2.2 TORNAD0/ TORNADO CENERATED MISSILE i

The most severe tornado wind l'oadings as specified by NUREC 0800 (Reference 8.3) and NRC Regula ory cuide 1.76 (1974) are selected as a design basis for this accident condi' tion. The applicable design parameters of the

}

8.2-1 headmen: No. 1 I

,fN v

ISFSI SAR design basis tornado (D8T) are the samu ao those specified in Section 3.2.1 and 3.2.2 of the NUHOMS Topical Report. The accident analysis of the ISFSI under the DBT is covered by the analysis presented in Section 8.2.2 of the referenced report. Given the fact that the HSH method 'of' structural analysis l

as utilized by the referenced report conservatively endolopes any stacking arrangement of the modules, including the three modular concept at the H. B.

Robinson ISFSI, the maximum moment and shear for the design basis wind pressure and missiles are also enveloped by the values given in Table 8.2-3 of this referenced report. Furthermore, the walls of the horizontal storage modules are anchored into the concrete foundation and as such, there is no possibility of overturning or sliding of the modules due to the impact of a massive high kinetic energy missile. The uplift forces generated by the impact of tho' massive missile and tornado wind loads are included in the foundation design presented in Section 8.3 of this report. The design and analysis of the anchorage system is also presented in Section 8.3.

I The result of this accident analysis indicates that all components of the ISFSI are capable of withstanding the tornado wind loads and tornado generated missiles with the exception of the air outlet shielding blocks. The loss of the shielding blocks is addressed in Section 8.2.1 of,this report.

8.2.3 EARTHQUAKE 8.2.3.1 Accident Analysis As specified in Section 3.2 of this report, the maximus ground horizontal' acceleration is 0.20s and the maximum ground vertica'l acceleration is O

0.1333 The NUHOMS Topical Report assumes a value of 0.25g for maximum horizontal acceleration and 0.173 for maximum vertical acceleration. In the Topical Report, for the seismic stress analysis of various components, a multiplier of 2 is'used.co account for multimode excitations. Since the values of the vertical and horizontal acceleration of the referenced report are higher than the H. 8. Robinson site accelerations, the seismic analysis for the HSM and the DSC support assembly presented in Section 8.2.3 of this referenced report is fully applicable and the results of these analyses envelop the site specific design. To establish the actual seismic response of the HBR DSC additional analysis is performed. However, the methodology is the same as that reported in Section 8.2.3.2, Pages 8.2-15 through 8.2-19 of the referenced report. Since the site specific design is different than that of the Topical Report, the longitudinal horizontal seismic loadLag on the HSM was reviewed. First of all, the DSC loads are transferred to the seismic retainer during a seismic event. The rttainer is connected to the ran mounting assembly plate through eftfovn solts in the two inch cover plate.

Consequently the loadi4 1 3 u a tsfer ed to the embedded pipe and plate which are anchored into the TF; m.: esli. The maximum loading of 8800 pounds generated by this event was found to have a negligible effect on the HSH rear wal1.

In the referenced report, Secti.on 8.2.3.2, the DSC shell ovaling mode was Cound to yield the lowest natural frequency. Since the HBR DSC shell parameters (i.e., the thickness and nominal diameter) have not been changed Os the lowest natural frequency remains the same at 37.2 Hz.

The stresses induced on the canister casing and the basket due to the 0.20s horizontal and 0.1333 vertical seismic accelerations are calculated on the basis of 8.2-2 Amendment No. 1 e

,.,-,.--,,,m,rw-m.,-,--e---

-w,-ve.---wev,

--~---Ww-=--r--,,-w-e~~w

l 1

l equivalent static method. The static stresses obtained are increased by a factor of 2.0 to account for multimode excitation. To obtain the DSC stresses due to the vertical component of the seismic load, the bending stresses l

calculated for the dead weight analysis can be factored directly by 0.266.

The maximum stress obtained in this manner is 3.07 ksi. For the horizontal seismic analysis both the longitudinal and the transverse directions are considered. For the horizontal acceleration in the transverse direction, the method of analysis presented in Page 8.2-16 of the referenced report was employed and a bending stress intensity of 7.53 kai was obtained. The stresses in the DSC shall and outer top plate due to the restraining action of the seismic restraint assembly under the longitudinal seismic loading was also investigated and found to have negligible effect. The shall stresses obtained for the vertical and horizontal cases were summed absolutely and.a combined stress of 9.32 kai was obtained.

Additionally, using the same methodology as that presented in Section 8.2.3.2, Page 8.2-17 of the referenced report, a margin of safety against a DSC roll over during a seismic event was established. A value of 2.5 was obtained for this margin of safety against the DSC roll over.

In summary, the ISFSI seismic analysis using site specific accelerations is i

enveloped by that reported in the NUHONS Topical Report. Furthermore, the i

HSMs are anchored to the foundation and as such, no overturning or sliding of the modules is possible. However, the overturning effects on the foundation are included in the foundation design which is presented in S,ection 8.3 of this report. The anchorage design is also presented in Section 8.3 of this O

report.

7 8.2.3.2 Accident Dose Calculation The mejor components of the HBE ISFSI are designed and analyzed to withstand the forces generated by the safe shutdown earthquake, hence there are no dose consequences.

8.2.4 DROP ACCIDENT l

8.2.4.1 Postulated Cause of Events As described in Section 1.3.1.7 of this report, the only time during the transfer operation that the IF-300 cask is operating without its redundant l

yoke is during the cask lowering into the cradle of the skid assembly. As shown in Figure 8.2-1 the maximum height that the cask is raised during this operation is 8.0 feet. Hence, the maximum height of a postulated drop accident *is ILaited to this value. Furthermore, since the cask is always lifted from the trunnions located at the upper regions of the cask, the postulate failure of the single yoke can only cause a cask bottom end or a corner drop. Consequently, if the yoke fails during the tilting operation the cask will either land on the bottom and fins or on the side steel rings located near the upper and lower regions of the cask outer shell.

Based on the above discussion, an 8-foot drop criteria in either horizontal or O

vertical bottom end orientation will bound any possible drop orientation during the transfer operation, including a corner drop orientation. The skid assembly and the e,ask/ skid / trailer tie down systems are designed to withstand

~

8.2-3 Amendment No. 1

.n

- - - - - - ~. - - - -

ISFSI SAR the inertia forces associated with the transportation shock loads, and as such there is no possibility of a cask drop during the transport operation from the decon area-to the HSM site. Even if such unlikely event occurs or the i'

cask / skid / trailer tip over as a unit, the height of this drop condition is enveloped by the 8 foot drop height criteria.

8.2.4.2 Drop Accident Analysis As stated earlier in Section 1.3.1.3 of this report, the IF-300 cask requires i

an additional extension collar and a new cask lid, in order to meet the cask cavity minimum length requirement and meet the criteria for cask lid removal in horizontal orientation. In this modified configuration the cask's impact limiters which are the radial fins attached to the cask's original head are removed. Hence, the energy absorbing properties of the cask is significantly reduced at upper regions. However, as discussed earlier the cask is always handled in upright position and no postulated failure mechanism can produce a top and drop. Additionally, the 8 feet rise of the cask is not sufficient for the cask to rotate 180 degrees in mid' sir to land on its head or upper i

i

. corner. The remaining part of the cask impact limiters, i.e.,

the bottom radial fins and the ring and both ends are not altered and will provide the energy absorption mechanism needed for the vertical bottom end and the I

horisontal' drop.

The IF300 cask energy absorbing properties are contained.in the cask Safety Analysis Report (Reference 8.4).

This SAR contains extensive data concerning l

s 30-foot drop a,ccident.

v The latest de'celeration time history development work of the IF300 cask is contained in Appendix V-1 of the above referenced document. These particular impact time histories'contain peak deceleration values, at early time of impact. These peak acceleration values are associa~ted with the dynamic yield stress characteristic of the stainless steel fins (strain rate dependency).

These time histories which envelop the previous histories reported in the referenced document, include 3 horizontal and 2 vertical drop orientations.

These selected time histories were modified to reflect the 8 ft drop criteria i

described earlier. Since the overall geometry and the weight of the loaded cask are not significantly changed, these deceleration time histories were linearly scaled to ref' lect the 8 ft drop criteria. Figure 8.2-2 shows the modified deceleration time histories used in the DSC drop analyses.

Horizontal Drop

)

Principle structures effected by the horizonta'l drop are the spacer' disk and the boral tubes. The bdral tubes serve only as a guide for the fuel assemblies and are not considered load bearing members, except own weight.

In the NUHONS Topical Report, Section 8.2.9, Page 8.2-35, the stresses in the boral tube under the inertia forces of a 343 drop criteria were evaluated by a finite analysis technique. Since the boral tube design is not changed, the result of this analysis can be directly ratioed for the higher deceleration value of 54.4. In this manner a maximum stre'es of 4.24 kai is obtained.

3 O

The DSC basket is designed such that the locations of the spacer disk coincide with the fuel assembly grid strap. Therefore the weight of the fuel assemblies is directly transmitted to the disk. For the analysis of this 8.2-4 haandment No.1

i i

l ISFSI SAR j

l l

member, the finite elements analysis reported in Section 8.2.4, pages 8.2-31 through 8.2-34 of the referenced report can be utilised directly. This is due s

to the fact that the overall configuration of this member has not been changed from that of the NUHOMS generic concept, with the exception of thicker disks, and the analysis is linear elastic. Consequently, the results of the referenced analysis can be factored to include the effect of the mass, thickness, length and deceleration value changes. Additionally, a factor was added to include the additional weight of the support rods in relation to the mass used in the STARDYNE Model.

M t

3 1

M sp) (

) ( nse) sp nu (M,,) (t,p) (g "a

S

=S sp nu nu u

wheret S,,= Maximum Stress, kai

~

5 " = 38.52 kai (from NUHOMS)

M"[,=2034lbMass,CP&Lwithoutrods M

= 1818 lb Mass, NUHOMS without rods g,, = 54.4, site deceleration value i

3 go,= 34, NUHOMS drop value

)

3 I,= 26.19 in, actual cell length 1,= 26.00 in, NUHOMS length M

= 1962.1 lbs Mass, NUHOMS with lids'

= 2133.1 lbs Mass, STARDYNE Model therefore:

l 5,= 39.93 kai 1

It must be noted that this stress intensity is mainly due to the shear force j

developed near the imposed artificial support boundary, and as such is not representative of the actual stress of the disk. A more critical stress i

location of the disk is at the spacer beams adjacent to che fuel assemblies.

The maximum membrane stress intensities is 23.5 kai which is obtained by the same ratioing technique discussed above. The results of the horizontal drop analysis are contained in Table 8.2-1 of this report.

Vertical Botton End Drop The components of the DSC that are critically effected during a vertical i

bottom and drop are the DSC shell; the top and botton DSC regions, the support rods and related DSC welds. The vertical drop analyses utilize both hand calculations and finite element technique. For the DSC shell and the end regions ANSYS program was employed for its axisymmetric and linear or nonliner features. Other components of the DSC are analysed by. hand calculation techniques.

4 As stated earlier in this report, the HBR DSC configuration is different from that o'f the NUHOMS generic concept. The DSC has been redesigned to fit into

]

the IF300 cask, and also is designed to withstand a drop accident in which the height of the drop is significantly greater than the 8-foot criteria. This is done for compatibility with future shipping options.

8.2-5 Amendment No. 1 l

l _

i ISFSI SAR i

l l

For the DSC bottom region analysis a model consisting of 131 elements and 183 O

nodes was developed.

The model is shom in Figure 8.2-3.

Both lead and steel are modeled as 2-D, 4-node isoparametric axisymmetric finite elements (STIF42). The interface of the lead and steel is modeled with coincident nodes which are coupled in vertical direction only. In this manner, only normel forces are transmitted between the two surfaces, and the shear and friction forces are conservatively released.

Both physical and syuusetrical boundary conditions are imposed at appropriate locations. The material properties are conservatively taken at 400'F to envelop peak temperatures of the DSC shell. The entire weight of the basket and the fuel assemblies are included as added mass elements along the top surface of the 2 inch cover plate. The weight of the top region of the DSC and that portion of the DSC shell that is not included in the model, is also included as added mess at the appropriate location on the shell. The response of the DSC bottom region under the drop impact time. history was conservatively approximated by an equivalent stati.c analysis. The impact time history has a very short duration and essentially behaves like a very short triangular impulse. Frequency analyses performed on various DSC components indicated that the longest natural period was much greater than the duration of the impulse. Thus, the dynamic impact loads cannot produce a response that exceeds the static response, and as such the dynamic amplification factor is less than unity.

Therefore, the static analysis performed is more conservative than a-dynamic analysis. An acceleration value of 76.5g was imposed statically on the model. Both membrane and extreme fiber stress intensities at critical i

locations including the weld elements reported in Table 8.2-1 of this report.

For the top region of the DSC, another ANSYS finite element model as shown in l

Figure 8.2-4 was developed. This model consists of 298 isoparametric STIF42 I

elements and 409 nodes. Similar assumptions and modeling technique as discusi..' for the bottom region model were employed. Static acceleration of 76.5s was applied to the model to obtain the membrane and bending stresses for various components and welds. The results of this analysis is also included in Table 8.2-1 of this report along with the ASME code allowables.

l The 21/2 inch diameter steel support' rods were analyzed under this postulated vertical drop. These rods extend the entire inside cavity of the DSC. The i

main function of these rods is to provide resistance to axial loads for the spacer disks.

Each of the seven spacer disks is welded to these rods by means of fillet welds. One inch clearance is provided between the support tods and the top lead plus of the DSC. This clearance is provided so that thermal expansion of the components and deflection of components during accident loading condition's, such as a drop accident, will not cause interference.

The 2 1/2 inch diameter support rods are designed so that they will resist the weight of the spacer disks under the postulated drop. The most critical segment of the support rod is between %e two bottom spacer disks. For this analysis, the weight imposed on a single rod at this critical location was the weight of six spacer disks divided by 4 plus the self weight of 1 rod.'

The axial stress at this 26 inch segment of the rod was found from the following relationships

/

8.2-6 Amendment No. 1

l

,(W) (a) l 3ex A

l where:

S,, = Axial stress I

W

= 651.7 lb, total weight imposed on the rod

=76.53,geskverticaldeceleration a

A

= 4.91 in, cross sectional are of the rod therefore 4

S,, = 10.15 kai The support rod material has been changed from SA304 stainless steel to SA-479 i

Type IM19 material which has a yield stress of 40.8 kai at 400*F.

The allowable compressive stress for this material is established by the rules of the ASME Appendix XVII, and Appendix F, which include the effect of slenderness ratio.

The results of the support rod analysis along with the compressive stress allowable are tabulated in Table 8,.2-1 of this report.

]

The results of the horizontal and vertical drop analysis as shown in Table l

8.2-1 indicate that the stresses in all components of the DSC and its internals are within the ASNE acceptance limits and are capable to withstand inertia forces associated with the 8 foot drop accident condition. A corner drop accident was also cbasidered. However the deceleration values as established by the IF300 cask SAR are significantly lower than the values of either the horizontal or the vertical deceleration components. Therefore the g

stresses for corner drop analysis are bounded by.the analyses presented above.

8.2.5 LIGHTNING 8.2.5.1 Postulated Cause of Events Since the ISFSI is outdoors, there is a likelihood that lightning could strike the ISFSI. Section 2.3.1 of the H8R2 Updated FSAR (Reference 8.2) provides information on the frequency of cloud-to ground lightning strokes (the only type of lightning stroke which poses a hasard to the ISFSI) at the site.

In order to protect the ISFSI fran any damage which could be caused by a lightning discharge, a lightning protection system is installed on the ISFSI. The lightning protection system is designed in accordance with NFPA l

No. 78-1979 Lightning Protection Code (Reference 8.5).

This system will prevent any damage to the HSH and its internals. Therefore, lightning striking the HSM and causing an off-normal condition is not a credible accident.

8.2.5.2 Analysis of Effects and Consequences Lightning protection systems have proven to be an effective means of

~

protecting a structure and its contents from the effects of a lightni'ng discharge. The lightning protection system'does not prevent the occurrence of a lightning discharge; however, the system does intercept the lightning discharge before it can strike the HSH and provides a continuous path for the discharge to the earth. In the event of lightning striking the HSM, the air 8.2-7 Amendment No. 1

ISFSI SAR I

f O

terminal located on top of each air outlet opening would latercept the lightning discharge. The current will follow the low impedance path of the air terminal, conductor's, and ground terminals to the earth. Since the system diverts the current, the HSM and its contents will not be damaged by the heat or mechanical forces generated by the current passing through the HSM.

In addition, since the ISFSI requires no electrical system for its continuous operation, the resulting current discharge will have no effect on the l

operation of the ISFSI.

l 8.2.6 BLOCEAGE OF AIR INLETS AND OUTLETS This accident is the complete and total block' age of the air. inlets and outlets of the horizontal storage module. Since the ISFSI is located outdoors, it can be postulated that the module is totally covered by debris from such an unlikely event as a tornado. The ISFSI's design features, such as a perimeter fence and separation of air inlets and outlets, minimise the probability of such an accident occurring under normal conditions. Nevertheless, such an accident is postulated and analyzed.

There are no structural consequences under this event. The thermal consequence of this accident results from heating cf the DSC and HSH due to the blockage of air flow. Section 8.2.7 of the.NUHOMS Topical Report addresses this accident condition. The results of the analysis indicate that there is no structural or dose consequence if the air inlets and outlets are cleared within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months time limit for clearing the air inlets and outlets is specified in the HBR2 ISFSI operation and limits criteria (See Chapter 10).

8.2.7 ACCIDENT PRESSURIZATION OF DSC Internal p'ressurisation of the DSC results from fuel cladding failure and the subsequent release of fuel rod fill gas and free fission gas. To establish the maximum accident pressurization, it is assumed that all fuel rods in the DSC are ruptured.and that the fission gas release fraction is 25%, and the 4

original fuel rod fill pressure is 500 psig.

(HBR fuel actually has a fill pressure of 300 psig.) The resulting internal pressures at HBR's maximum 1

ambient temperature of 105'F and at the minimum ambient temperature'of -5'F are below the accident pressures reported in Section 8.2.9 of the NUHOMS Topical Report (for temperature extremes of 125'F and -40*F).

The limiting accident for canister pressurisation is the blockage of air flow to the DSC.

Under these conditions, the gas temperatures in the DSC will rise to 413'c (775'F) producing a DSC internal gauge pressure of 2.76 bar (39.7 pois). The canister shell stresses due to acci' dent pressurisation are enveloped by those l

reported in the Topical Report.

The DSC has a safety margin of gre.acer than 3 under this accident condition and as such, there are no dose consequences.

8.2.8 FIRE l

No flammable or combustible substances are stored within the ISFSI or within i

the ISFSI's radiation control area. Additionally, the ISFSI is constructed of non-flammable heat-resistant materials (concrete and steel). The only credible accident which could expose.the ISFSI to a flammable substance would 8.2-8 Amendment No. 1 l

ISFSI SAR be the accidental spillage of a flammable liquid, either through human error j

i or equipment malfunction, at the perimeter of the ISFSI. However, the sandy soil and concrete between the ISFSI's perimeter fence and the HSM, are highly porous. Most of the flammable liquid would be absorbed by these surfaces, greatly reducing the intensity or duration of the fire.

2 j

j The only other time in which a component of the ISFSI would be exposed to a j

potential fire hasard would be during the DSC drying and transport operations. Throughout these operations, the DSC is located within the cavity of the CE IF-300 shipping cask.

Based on the above discussion, exposure of the ISFSI to a long or intense fire is not considered a credible accident.

8.2.9 DRY STORACE CANISTER LEAKAGE The DSC is designed for no leakage under any normal or credible accident conditions. The accident analyses in previous sections show that none of the events could breach the canister body. However, to show the ultimate safety of the ISFSI, a total and instantaneous leak was postulated. The postulated 1

accident assumes that one DSC ruptured and all fuel rod claddings, failed f

simultaneously such that 25% of all fission gases in the irradiated fuel assemblies (mainly Kr-85) are instantaneously released to the atmosphere. The dose consequences from the leaking DSC are evaluated in the NUHOMS Topical l

Report, Section 8.2.8, and the resulting accident dose is found,to be well below the 10CFR Part 72.68 acceptable limit of 5.0 rem.

i l

8.2.10 LOAD COMBINATION

}

Normal operating and postulated accident loads associated with various components of the ISFSI are either the same as or are enveloped by those reported in the NUHOMS Topical Report, except for the DSC and the i

foundation. Hence, the combined effect of various. accident and normal operating loads for the DSC support assembly and the HSM are enveloped by the load combination results presented in Section 8.2.10 of the Topical Report.

The methodology used in combining normal operating and accident loads and their associated over load factors for various components of the ISFSI, with the exception of the foundation, is presented in the aforementioned report.

Load combination procedures for the foundation addressed in Section 8.3 of j

this report. The DSC analysis load combination utilises the same methodology as in the Topical Report, but due to design differences the results are l

changed slightly. The results of the DSC load combination for the worst case, i.e.,. drop accident, are contained in Table 8.2-2 of this report.

j Furthermore, the DSC fatigue analysis due to normal operating pressure loads, accident pressure loads, seismic loads, seasonal temperature loads, and daily temperature cycling as presented in Section 8.2.10 of the Topical Report, l

envelops the H8R site specific analysis. This is because the extreme ambient temperatureselectedforgenericdesignoftheDSC(-40'Tto125'F) envelops j

the HER ambient temperature range (-5 F to 105'F) and the HBR has a lower seismic acceleration.

'O 8.2-9 Amendment No. 1

I Table 8.2-1 MAXIMUM DSC STRESSES FOR 8-F00T BOTTOM END DROP ACCIDENT STRESS (ksi)

SHESS COMPONENTS TYPE CALCULATED ALLOWABLE Primary *!embrane, 10.11 44.88 Canister

- Shell Primary Membrane 16.56 64.40

+ Bendtng Bottom Primary Membrane 6.09 44.88 Cover Plate Primary Membrane 13.40 64.40

+ Bending Top Primary Membrane 2.77 44.88 Co'ver Plates Primary Membrane 5.47 64.40

+ Bendtng 4

Support Ring Primary Membrane 1.71 44.88 O

for Top Lead Plug Primary Membrane 5.43 64.40

+ Bending Lead Compressive 1.86 6.80 1

Lead Casing

Primary Membrane 17.29 44.88 Spacer Disk Primary Membrane 43.10 44.88 Primary Membrane Boral Tubes

+ Bending 4.24 64.40

n. d Compression 10.16 21.13 g p g gg 1

Fillet Primary

,11.68 22.44 J-Weld Primary 6.18 29.20 Notes t

1.

Allowable stresses shown correspond to service Level D limits, unless noted otherwise.

2.

Material properties taken at 400*F design temperature.

3.

Compressi,ve stress allowable of the support rods is based on Appendices XVII and F rules and for Level A limits.

8.2-10 Amendment No. 1

_-...,,,,.,-,,_,.-,.,---,-.-,,,,---r--_,_,--_--,,-,.,,,,-,-.-.._---,..,,,.-en.mn_,,

.n._w.+

,_--ww w._-c.r-.- -, - -..,

l I

l Table 8.2-2 DSC ENVELOPINC LOAD COMBINATION DSC STRESS COMPONENTS TYPE LJMBINED ALLOWABLE Primary Membrane 11.02 44.88 Cuiste Shell Primary Membrane 48.35 64.40

+ Bending Bottom Primary Membrane 6.09 44.88 Cover Plates Primary Membrane 15.91 64.40

+ Bendtng Top Primary Membrane 2.77 44.88 Cover Plates Primary Membrane 8.74 64.40

+ Bendtng Support Ring Primary Membrane 1.71 44.88 for Top Lead Plug Primary Membrane 5.43 64.40

+ Bending Lead Compressive 1.86 6.80 Lead Casing Primary Membrane 17.29 44.88 Spacer Disk, Primary Membrane 43.10 44.88 3

Primary Membrane Boral Tubes

+ Bending

'4.24 64.40 212 n.

Compression 10.16 21.13 g

g Rod 1/4 i Fillet Primary 11.68 22.44 J-Weld Primary 6.18 29.20 Notest 1.

When applicable, stresses due to the drop accidents are combined with that of pressure and dead weight.

2.

Stresses for each DSC components are conservatively combined irrespective of location.

3.

Allowable stresses shown ' correspond to service Level D limits, unless i

noted otherwise.

l 4.

Material properties taken at 400*F design temperature.

5.

Thermal stresses need not be included under service Level D limits.

6.

Compressive stress allowable of the support rods is based on Appendices XVII and F rules for Level A limits.

8'2-11 Amendment No. 1

l l

8.3 FOUNDATION DESIGN O

4 To provida a means of transmitting the reaction loads of the ISFSI modules to the ground, a rectangular, flat plate type, mat foundation was selected. The mat foundation is ideally suited for the ISFSI since it spreads out the loadings and consequently reduces the soil bearing pressure and at the same time minimizes the differential settlements.

To accommodate the ISFSI modules, the front cask unloading area and the hydraulic ran area behind the modules, an overall foundation size of 28'-9" by 60'-0" was selected. The HSM foundation slab is 3 feet thick. A construction joint connects this slab to the cask unloading slab which is 2 feet thick starting from a point 5 feet from the module front. The ram mounting slab at the rear of the modules is 8 inches thick and connects to the 3 foot founda-tion by an expansion joint. The foundation concrete is 4000 psi normal weight l

concrete poured on.a 4 inch mud slab. The HSM foundation and the cask unload-ing slab are interlaced with continuous two-way reinforcing top and bottom.

Number 9 bars are used for tensile reinforcement and as dowels to anchor the HSM walls to the foundation. The ram mounting slab has a number 5 bar continuous.two-way reinforcing at the bottom only. Welded wire fabric is placed at the top of the 8 inch slab.

For analysis purposes, a STARDYNE rectangular plate finite element model as shown in Figure 8.3-1 was developed. The model consists of 255 nodes and 224 plate elements. At each node, a ground support spring was added to simulate the soil elastic properties. The elastic soil spring is obtained by modifying O

the esperimental modulus of subgrade reaction by an appropriate size factor of the foundation. The modified modulus of subgrade reaction is then multiplied 3~

by the tributary area associated with each node. The resulting values of the

~

spring stiffnesses were used as input to the finite element model asi a ground i

stiffness matriz. The method for finding the stiffness K is shown below (Reference 8.7):

y (8

1 )

K=K A

Where l

K, espergnengalmodulusofsubgradereaction

=

= 100 /in (for granular, soil) i foundation width = 28.75 feet i

8

=

nodal tributary area (varies)

A

=

Five separate load cases were considered in the foundation design 1)

Center module loading 2)

Outside module loading 3)

Dead Weight + Live Load 4)

Dead Weight + Tornado Wind / Impact (lengthwise) 5)

Dead Weight + Tornado Wind / Impact (widthwise)

O Since cask unloading and ram mounting slabs are cast in place after HSM construction, the differential settlement due to HSM dead weight will not be esperienced by the cask unloading slab. Consequently, for load cases 1 and 2 8.3-1 Amendment No. 1

ISFSI SAR G

loading of 175 k, which includes the saddle, canister, the dead weight was not included.

rollers, trunnion, and cradle is For load cases 1 and 2 the total trailer t

locations in the unloading areas. applied as concentrated loads at nodal operation will occur at a time, the two load caseSince only one loading or unloading dently.

taining the DSCs. Load case 3 consists of the dead weight of the ths were eval weight of three DSCs.This total 3 bay module weight of 800 8 k i ree modules con-contact with the foundation to get an equivalentThe total loading is divided s added to the a live load of'200 psf is postulated for the HSM area in pressure load.

also divided by the contact area to get a pressur Additionally, roof.

This total load is foundation.

plate elements of the STARDYNE model.These loads are applied as press uplift load combinations caused by tornado loadiLoad cases 4 and 5 are the maxi e appropriate directions.

module walls and roof plus the reaction load of 458Using a con applied on the s

automobile traveling'at 184.8 ft./s applied to th

.2 k caused by a 3967 lb.

same direction combined with the module dead wei he top of the module in the were calculated.

A simple frame model was used to calculatLive loads are exclu g t, loads.

Figure 8.3-2.

uce uplift using the contact area of one wall.The maximum uplift force of 38.80 k is conve u 3.1 psi which is used in the analysisThis yleids a maximum uplift pressure of seismic loadings shows that tornado loads are Comparison of' seismic loads will not be included in the foundmuch more severe tornado l uplift forces are calculated they are applied O.

ation analysis.

Consequently, appropriate plate elements corresponding to th Once the as negative pressures on the surface.

Tornado wind and impact loads are evaluat d fe HSM/ foundation connect Since an uplift force is created by the t or both directions.

e will have~a negative bearing or uplift alonornado~1oads the foundation itself itself does not resist uplift.

g the edge of the module.

for the effects of the uplift-alon experienced in low stressed areas.g the foundation edge.Results from The' soil ewed Minimal uplift was not in that vicinity..tre not significantly affected by the uplift since thTherefore, re

{

e high stressed areas are The maximum calculated bearing stress is

{

bearing pressures in the range of 3000 to 4000 soils pres For sandy oundation, allowable soil recommended by the Southern Standard Buildi pounds per square foot are exploration performed at the Robinson site by L ng Code per the geotechnical Company.

Since the maximum soil contact pressureaw Engineering Testing foundation analysis is 2210 psf the bearin produced by the HSM normal dead weight and live loads the bearig strength is sufficient.

For The reinforcement design was based on the ng pressure is only 1605 psf.

the finite element analysis.

Using the ultimate design method, the r i felemen wi'th a conservative load factor of 1 7 ament was designed to wit e

e n orce-pplied to envelope all load combina-oad com tion factors specified in Section 9 2 iO results for the HSM Foundation and the cask unloading slab i of ACI 349-80.

Table 8.3-2.

A tabulation of the s presented in 8.3-2 l

haandment No.1

TABLE 8.3-1 FOUNDATION BEARING STRESS C,0AD LOAD BEARING CASE DESCRIPTION STRESS (KSF) 1 Center Module 0.247 Loading 2

Outside Module 0.463 Loading 3

Dead Weight +

1.605 Live Load 4

Tornado Wind +

2.210 Impace j

(Widthwise)

.,y 5

Tornado Wind +

0.671 Impact (Lengthwise) i Note:

1.

Dead weight of module not included in load cases 1 and 2.

l I

8.3-3

(

Amendment No. 1

TABLE 8.3-2 FOUNDATION SLAB ~

MAXIMUM BENDING MOMENTS MAXIMUM ALLOWABLE LOAD LOAD SLAB MOMENT MOMENT CASE DESCRIPTION THICKNESS (K-in./in.)

(K-in./in.)

Center 2'-0" 13.8 87.0 1

Module Loading 3'-0" 28.9 186.0 Outside 2'-0" 13.1 87.0 2

Module 24.5 186.0 Loading 3'-0" l

2'-0" N/A 87.0

,,,,,,1,,,

+ L ve ad 3'-0" 46.4 186.0 i

Tornado Wind 2'-0" N/A 87.0 4

+ Impact (Widthwise) 3'-0" 80.1 186.0 4

Tornado Wind 2'-0" 57.6 87.0' 5

+ Impact (Lengthwise) 3'-0" 155.8 186.0 Notest 1

1.

Dead weight.of module not included in load cases 1 and 2.

2.

All moments conservatively factored by 1.7 to envelop all ACI 349 load combination factors.

I 1

i 8.3-4 Amendment No. 1

h moment capacity of the sections are calculated per methods identical to the NUHOMS Topical Section 8.1.1.5, Equation 8-1-32.

The 3'-0" slab with number 9 bars at 9 inches yields an ultimate strength of 186 k.in per inch I

section. The 2'-0" slab with number 9 bars at 12 inches yields an ultimate strength of 87 k.in per inch section. Therefore all bending moments esperienced by the foundation are below ultimate capacity.

As calculated before, the maximum uplift pressure exerted by the module wall is 3.1 psi for load case 5.

For a l'-0" section of module the resulting uplift is 1.56 k/fc. section. m dowel area required can be calculated by:

A = (UPLIFT) (1.7)

(9) (f )

y b res UPLIFT = 1.56 K f = 0.9 = Factor for Tension ~

{

f = 60 kai y

I h refo e, Area A = 0.05 in. conservatively 2 number 6 bars with an area of 2

O.88 in will be'used every 12 inches to prevent uplift. Keyways will be used between the module foundation interface to prevent sliding. Assuming the i

i maximum horisontal tornado loads are shared by the walls perpendicular to load i

yields a nazimum shear force of 24 h/ft including the load factor of 1.7.

The j

nominal shear strength of the keyway and dowels can be found from l

V, = (V, + V,) G h eet 2

F!,

b,= concrete shear strength (k)

.V,

=

l f',

4000 psi

=

9 inches b,

=

l 12 inches d

=

l V,

(A,) (f ) = steel reinforcement shear strength (k)

=

.88 in27 A,

=

i 60 kai f

=

y i

8

=.85 = shese factor l

Consequently, V, = 56.49 k which exceeds the nazimum factored shear force of 24 k.

h s, the module will neither slide nor overturn. Table 8.3-3 presents-l foundation anchor loads and capacities for load cases 4 and 5.

N 8 inch ram mounting slab was designed by hand calculations suggested by Teng (Reference 8.7) and sowies (Reference 8.8).

sy applying the maximum factored spider les loadings from the hydraulic ran a simple span is appr~omi-mated by treating the soll as a uniform load and the spider les as reaction points. A maximum factored moment of 32.3 k-in/ft is calculated. Using the ultimate strength method with number 5 bars at 12 inches, the ultimate strength of the 8 inch slab is 64.2 k-in/fc. Welded wire fabric was placed at l

the top of the slab as shrinkage and temperature reinforcement. Additionall'y j

all punching shear from the ran supports were found to be negligible.

l l

8.'3-5 Amendment No.*1 i

ISFSI SAR Furthermore, the cask unloading slab was analyzed for bearing and punching shear due to the hydraulic cylinder. A maximum bearing stress of 2.34 ksi was calculated which is less than the allowable of 4.76 kai calculated from ACI

'349-80 Section 10.16. The maximum punching shear was also found to be under code allowables.

I i

i 8.3-6 Amendment No. 1

i TABLE 8.3-3 O

FOUNDATION ANCHOR LOADS LOAD LOADING LOAD CAPACITY DESCRIPTION CASE TYPE (K/FT)

(K/FT)

Tornado Wind Shear 13.00 56.5 4

+ Impact (W1.dthwise)

Uplift 0

28.0 Tornado Wind Shear 24.00 56.5 5

+ Impact (Langthwise)

Uplift 1.56 28.0 Notest O

1.

All Shear and Uplift loads factored by 1.7 to envelop aLL ACI 349 load combination factors.

2.

Stiear capacity based on' concrete keyway plus embedded dowels.

3.

Uplift capacity calculated from embedded dowel area.

1 O

8.3-7 Amendment No. 1

ISFSI SAR f

REFERENCES:

CHAPTER 8 8.1 NUTECH Engineers, Inc., " Topical Report for the NUTECH Horizontal Modular Storage System for Irradiated Nuclear Fuel," NUH-001, Revision 1, November 1985.

8.2 Carolina Power and Light Company, "H. B. Robinson Steam Electric Plant Unit No. 2 Updated Final Safety Analysis Report," Docket No. 50-261, License No. DPR-23.

8.3 U.S. Nuclear Regulatory Commission, " Missiles Generated by Natural l

Phenomena," Standard Review Plan NUREC-0800, 3.5.1.4, Revision 2, July i

1981.

8.4 Ceneral Electric Company', "IF-300 Shipping Cask Consolidated Safety Analysis Report," NEDO-10048-2, Nuclear Fuel and Saecial Products Division.

8.5 National Fire Protection Association, National Fire Codes, No. 78, 1979 j

i Edition.

I

.8.6 Cybernet Services, STARDYME User Idformation Manual, Control Data Corporation, Minneapolis, Minnesota, Revisi9a C, April 1980.

l 8.7 W. C. Tens, " Foundation Design," Prentice-Hall, Inc., Englewood Cliffs, N.J., 1962.

g i

8.8 J. E. Bowles, " Foundation Analysis and Design," McGraw-Hill, New York, N.Y., 1977.

~

i 8.9 R. J. Roark and W. C. Young " Formulas for Stress and Strain," Fifth Edition, McGraw-Hill, New York, N.Y.,

1975.

l-1 i

l l

t l

l 8.R-1 l

Amendment 180. 1

10.0 OPERATING CONTROLS AND LIMITS The H. B. Robinson (HBR) Independent Spent Fuel Storage Installation (ISFSI) is a totally passive system which requires minimum operating controls.

In the following sections, the operating controls and limits that are pertinent to the ISFSI are specified. The-conditions and other items to be controlled are based on the safety assessments for normal and postulated accident conditions j

provided in Chapter 8 of this report.

1 The following operating controls and limits are specified:

)

10.1 Fuel Specifications 10.2 Limits for the Surface Dose Rate of the HSH While the DSC is in Storage 10.3 Limits for the Maxierum Air Temperature Rise After Storage 10.4 Surveillance'of the HSH Air Inlets 10.5 Surveillance of the HSH Inside cavity 4

I

![

?

}

i l

10.0-1

% % g y,,

g

.ISFSI SAR 10.1 FUEL SPECIFICATIONS

1.1 Title

Fuel Specifications

1.2 Specifications

Type 15 x 15 PWR Fuel Burnup

< 35,000 mwd /MT Initial (Beginning of Life)

Enrichment 1 3.5% U-235 Post Irradiation Time

> 5 years i

Weight Per Distance Between Adjacent Spacers Per Assembly

$ 106.56 kg Distance Between Spacers 1 0.65 m Any fuel not specifically filling the above requirements may still be stored in the ISFSI if all the following requirements are met:

$1kw/assp"bly Decay Power 1.67 x 10

" * " *

I Neutron Source 15 Camma Source 5.73 x 10 photons /sec/ canister With spectrum bounded by that shown in Table 10 1-1 End of Life 0.8% U-235 Fissile Content 0.5% Pu-239 0.1% Pu-241

1.3 Applicability

This specification is applicable to all-fuel to be stored in the ISFSI.

1.4 Objective

This specification was derived to insure that the peak fuel rod temperatures, surface doses and nuclear subcriticality are below the design values.

1.5 Action

If this specification is not met, additional analysis and/or data must be presented before the fuel can be placed in the DSC.

1.6 Surveillance The fuel selected for storage must have Requirements:

the parameter values specified in l'.2 above verified prior to fuel loading. No other surveillance is required.

1.7 Basis

The fuel parameters specified in this operating control and limit were selected to bound the types of PWR fuel which are currently in use or planned to be in use at HBR2.

O i

I 10.1-1 Amendment No'. 1

l ISFSI SAR 10.2

' LIMITS FOR THE SURFACE DOSE RATE 0F THE HSM WHILE THE DSC IS IN STORACE

2.1 Title

Surface Dose Rates on the HSM While the DSC is in Storage

2.2 Specification

Surface dose rates at the following locations I

1) Outside of HSM door on j

centerline of DSC 200 ares /hr l

1

2) Center of air inlets 200 meen/hr j'

3) Center of air outlets 200 mees /hr Average Dose rates for the following surfaces i

1) toof 50 ares /hr

2) Front /Back 50 ares /hr

3) Side 50 meem/hr Dose rates one meter from the center of the following surfaces of a unit of modules

1) Front /Back 20 meen/hr

2) Side 20 ares /hr j

2.3 Applicability

This, specification'is applicable to the ISFSI.

2.4 Objective

The objective of this specification is to maintain as-low-as-reasonably-achievable dose rates on the i

modules.

2.5 Action

If the dose rates are exceeded, temporary shielding l

must be placed so as to reduce the dose rates to the

?

specified levels. When temporary shielding is used, the outlet. air temperature must be measured after the j'

shielding is installed to verify that the air flow has not been restricted.

l 2.6 Surveillances The HSM shall be monitored to verify that this specification has been met immediately after the DSC is placed in storage and the HSM front and rear accesses are closed.

i

2.7 Basis

The dose rates stated in this specification were selected to maintain as-tow-as-reasonably-achievable exposures to personnel performing air duct clearing on the HSM. These dose rates are within Industry accepted standards for contact handling, j

operation and maintenance of radioactive material.

Maintenance personnel will be required to remove any potential air blockage. At 200 meem/hr the dose for a i

one hour job of unblocking the air inlets (or outlets) l-would be less than 200 meen (whole body) and hence l

would be only 4% of the total yearly burden.

Furthermore, analysis provided in Chapter 7 of the HBR ISFSI SAR shows that the espected dose rates around the HSH surface will be well below the specifications listed above.

l 10.2-1 Amendment No. 1

~.

ISFSI SAR 10.3 LIMITS FOR THE MAXIMUM AIR TEMPERATURE RISE AFTER STORACE O

1

3.1 Title

Maximum Air Temperature Rise from HSH Inlet to outlet

3.2 Specification

Maximum air temperature rise 100'F (55.6'C)

3.3 Applicability

This specification is applicable to the ISFSI.

3.4 Objective

To limit the maximum air temperature around the DSC.

L

3.5 Action

If the temperature rise is greater than 100*F (55.6'C), the air inlets and exits should be checked for blockage.

If the blockage is cleared and the temperature still exceeds this limit, the DSC must be removed from the module or addition information and analysis shall be provided that'will prove the exist-ing condition does not represent an unsafe condition.

3.6 Surveillance The temperature rise shall be checked at the time the DSC is stored in the HSM, again 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months later, and again after 7 days.

0

3.7 Basis

The 100 F (55.6*C) temperature rise was selected to Limit the hottest rod in the DSC to below 716'F O.

(380'C). If this temperature rise is maintained, then the hottest rod will be below the 716*F (380'C) limit even on the hottest day conditions of 125'T (51.7'C). The expected temperature rise is less than 100'F (i.e. 82*F (45.5'C); see Section 8.1.3 of HBR ISFSI SAR) and hence, the current design provides adequate margin for this specification.

If the.

temperature rise is within the specifications, then the HSM and DSC are performing as designed and no i

further temperature measurements are required during i

normal surveillance.

l j

4

\\

O f

10.3-1 Amendment No. 1 s

l ISFSI SAR 10.5 SURVEILLANCE OF THE HSM INSIDE CAVITY 1.

Title:

Surveillance of the HSM Inside Cavity 2.

Specifications:

Visual inspection of the inlets air pathways, the outlets air pathway, and the concrete surface, and embedded bolts.

i 3.

Applicability:

10% of the HSMs (but not less than one per installation) shall be inspected. Inspection frequency shall be once a year.

4.

Objective:

To assure lack of blockage of the inlets 'and outlets pathway by any organic or foreign material, and to 4

j assure lack of concrete deterioration or steel l

oxidation.

)

5.

Action:

If the inlet and outlet air' pathways are plugged by organic matters, they should be cleared.

If the inside surfaces of ths HSM concrete show sign of deterioration or spelling the DSC should be removed j

and affected surfaces repaired.

f l

6.

Surveillance:

This is a survelliance specification.

7.

Basis:

This surveillance wi11' provide added assurance that O

4 the concrete normal operating temerature does not cause any deterioration or degradation of the material, and also assures lack of growth of any I

~

organic matters which may casue blockage of the air I

pathways.

i i

i l

l l

Bl Amaad=aat No. 1 10.5-1

...-.._L.,.---.-,_---.--._-----.,--.------.---.--..

I j

1 i

11.1 CORPORATE QUALITY ASSURANCE It is the policy of CP&L to design, construct, and operate nuclear power plants without jeopardy to its employees or to the public health and safety.

The QA programs shall be developed, implemented, and updated as necessary to assure that the Company's nuclear facilities will be managed such that all systems used to produce, convey, or use nuclear generated steam and all systems used to treat, store, or convey waste produced by the generation of nuclear steam will be designed, constructed, and operated in a safe manner.

i Deviations from this program shall be permitted only upon written authority from the corporate management position originally approving the program or implementing procedures.

The design, construction, and operation of nuclear facilities shall be accomplished in accordance with the U. S. Nuclear Regulatory Coassission (NRC) l Regulations specified in Title 10 of the U.S. Code of Federal Regulations.

All cossaiteents to the NRC Regulatory Guides and to engineering and 1

construction codes shall be carried out.

I The operation of the Company's'ISFSI shall be in accordance with the terms and conditions of the ISFSI material license l'ssued by the NRC (10 CFR Part 72).

4 Changes in operating procedures, experiments at the ISFSI, modifications to the ISFSI hardware or systems, shall be made in accordance with the terms and conditions of the ISFSI material license.

l The Dry Storage Canister and Transfer Cask are considered safety-related and are subject to a QA program in conformance with the requirements of 10 CFR 50, Appendix 8 of the Corporate QA Program described in Reference 11.1.

g The Corporate QA Program also provides a specific program for Radioactive Waste Management Systems. This program parallels the appropriate requirements of the 18 criteria of 10 CFR 50, Appendix B and has been developed to match the level of control necessary for equipment, meterials, and processes which are important to maintain the lategrity of.the Radioactive Wasta Management System. This program will be applied to the Horizontal Storage Module and Foundation.

This program requires that design, procurement, installation, and testing be accomplished in a planned and controlled manner in accordance with approved procedures. These procedures include provisions, as necessary, to ensure 1

i thatt i

a)

Design and procurement documents include applicable d.esign requirements and are reviewed for adequacy. Deviations are controlled.

l b)

Purchased material, equipment, and services conform to the requirements j

of drawings, specifications, and purchase order documents.

l c)

Material is inspected'to the extent necessary to assure conformance

{

with technical and QA requirements of the' purchase order documents.

1 d)

Material and equipment are handled and stored to prevent damage and deterioration.

11.1-1 Amendment No. 1

e)

Items which have passed the required inspections and tests established

]

by the specification are identified.

f)

Conditions adverse to quality are properly identified and corrected.

g)

Instructions, procedures, and drawings include appropriate qualitative and quantitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

h)

Tools, sages, instruments, and other measuring and testing devices used in activities affecting quality are properly calibrated and controlled.

1 i)

Sufficient records including results of reviews and inspections are maintained.

j)

The issuance of documents such as instructions, procedures, and i

drawings, including changes thereto which prescribe all activities affecting quality are controlled.

Furthermore, following construction of the Horizontal Storage Modules (HSMs) and the initial loading of the Dry Shielded Canisters (DSCs), the H. B. kobinson ISFSI will be deemed " operational." Keeping in mind that the NUH0NS system being utilized for the H. 5. Robinson ISFSI is a totally passive j

installation, there are no pieces of equipment which require operation nor a requirement for data collection and reporting.

However, as stated in Section 5.1.1.7 of this document, a daily visual inspection of air inlets and outlets'will be made to insure that they remain 7;

1 unblocked and the integrity of the ' screens remains intact. On a yearly basis, site personnel will visually inspect the internals of one of the loaded

'HSMs.

In addition, it is stated in Section 7.3.4 of this document that "The operation of the ISFSI will be monitored under the H. B. Robinson Unit 2 Radioactivity Monitoring Program. No additional radiation monitoring instrumentation is required." Operation of the.H. 8. Robinson ISFSI will be as established by the requirements of the Technical Specifications contained in the Safety Analysis Report and will be carried out in accordance with the QA program for radioactive vaste management systema, i

Survalliance will be accomplished by the existing Robinson Plant QA Program i

and will be subject to audit by the Corporate QA Department at least every two years. Surveillance, audit, and other management oversight activities shall be performed in accordance with the Corporate QA Program for safety-related activities.

l 1

1 i

1 l

\\

11.1-2 Amendment No. 1 i