NUREG/CR-6745 

INEEL/EXT-01-00183

Dry Cask Storage 

Characterization 

Project-Phase 1: 

CASTOR V/21 Cask 

Opening and Examination 

Idaho National Engineering and Environmental Laboratory 

U.S. Nuclear Regulatory Commission •0l .E.. 

Office of Nuclear Regulatory Research 4 ý 

Washington, DC 20555-0001
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NUREG/CR-6745 

INEEL/EXT-01-00183 

Dry Cask Storage 

Characterization 

Project-Phase 1: 

CASTOR V/21 Cask 

Opening and Examination 

Manuscript Completed: August 2001 

Date Published: September 2001 

Prepared by 

W.C. Bare, L.D. Torgerson 

Idaho National Engineering and Environmental Laboratory 

Idaho Falls, ID 83415-3114 

R.M. Kenneally, NRC Project Manager 

Prepared for 

Division of Engineering Technology 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

NRC Job Code Y6038
NUREG/CR-6745 has been reproduced 

from the best available copy.
ABSTRACT 

This report documents visual examination and testing conducted in 1999 

and early 2000 at the Idaho National Engineering and Environmental Laboratory 

(INEEL) on a Gesellschaft fUr Nuklear Service (GNS) CASTOR V/21 

pressurized water reactor (PWR) spent fuel dry storage cask. The purpose of the 

examination and testing is to develop a technical basis for renewal of licenses 

and Certificates of Compliance for dry storage systems for spent nuclear fuel and 

high-level waste at independent spent fuel storage installation sites. The 

examination and testing was conducted to assess the condition of the cask 

internal and external surfaces, cask contents consisting of 21 Westinghouse PWR 

spent fuel assemblies from Dominion's (formerly named Virginia Power) Surry 

Power Station and cask concrete storage pad. The assemblies have been 

continuously stored in the CASTOR cask since 1985. Cask exterior surface and 

selected fuel assembly temperatures, and cask surface gamma and neutron dose 

rates were measured. Cask externallinternal surfaces, fuel basket components 

including accessible weldments, fuel assembly exteriors, and primary lid seals 

were visually examined. Selected fuel rods were removed from one fuel 

assembly, visually examined, and then shipped to Argonne National Laboratory 

for nondestructive, destructive, and mechanical examination. Cask interior crud 

samples and helium cover gas samples were collected and analyzed. The results 

of the examination and testing indicate the concrete storage pad, CASTOR V/21 

cask, and cask contents exhibited sound structural and seal integrity and that 

long-term storage has not caused detectable degradation of the spent fuel 

cladding or the release of gaseous fission products between 1985 and 1999.

iii
CONTENTS 

ABSTRACT ................................................................................................................................................ 

EXECUTIVE SUM M ARY ....................................................................................................................... xvii 

ACKN OW LEDGM ENTS .......................................................................................................................... xix 

ACRONYM S ............................................................................................................................................. xxi 

I. INTRODUCTION .............................................................................................................................. 1 

1. 1 Regulations ............................................................................................................................... 1 

1.2 Objectives and Scope ....................................................................................................... 1 

1.3 CASTOR-V/21 Cask Description ...................................................................................... 3 

1.3.1 Cask Body ......................................................................................................... 3 

1.3.2 Spent Fuel Basket .................................................................................................. 4 

1.3.3 Prim ary Lid .......................................................................................................... 5 

1.3.4 Secondary Lid ...................................................................................................... 6 

2. INSPECTION S AND TESTS ....................................................................................................... 7 

2.1 Long-Term Spent Fuel Storage Pad .................................................................................... 7 

2.1.1 Description of Pad .................................................................................................. 7 

2.1.2 Pad Inspection ..................................................................................................... 9 

2.2 Cask Exterior .......................................................................................................................... 11 

2.3 Cask Lid Bolts ........................................................................................................................ 11 

2.4 Cask Lid Seals ........................................................................................................................ 14 

2.4.1 Description of the Seals ......................................................................................... 14 

2.5 Rem ote Inspection .................................................................................................................. 14 

2.5.1 Direct Visual Inspection ....................................................................................... 16 

2.5.2 Observations and Results .................................................................................... 16 

2.6 Cask Interior ........................................................................................................................... 19 

2.6.1 M ethod of Inspection ............................................................................................ 20 

2.6.2 Interior Cask Sidewall ......................................................................................... 20 

2.6.3 Interior Cask Bottom and Bottom Sidewall ......................................................... 21 

2.7 Fuel Assem bly Basket ...................................................................................................... 23

v
2.7.1 Basket Condition .................................................................................................. 24 

2.7.2 Basket W elds ....................................................................................................... 25 

2.8 Fuel A ssem blies ..................................................................................................................... 28 

2.8.1 Purpose of the Inspection .................................................................................... 28 

2.8.2 Method of Inspection and Lifting Force Measurements ..................................... 28 

2.8.3 Fuel Rod Length V ariation ................................................................................... 30 

2.8.4 Fuel Rod Bowing .................................... ................ 31 

2.8.5 Cladding Failure .................................................................................................. 31 

2.8.6 Crud Coating ............................................................................................................. 31 

2.8.7 Fuel Rod Rem oval and Exam ination ................................................................... 32 

2.9 Cask Tem peratures ................................................................................................................. 36 

2.9.1 Internal Temperature ........................................................................................... 36 

2.9.2 External Surface Temperature .............................................................................. 37 

3. RAD IATION SURVEY ................................................................................................................... 40 

3.1 Objectives ............................................................................................................................... 40 

3.2 M ethods of M easurement .................................................................................................. 40 

3.3 Observations ........................................................................................................................... 42 

3.3.1 Gamm a Contributors ........................................................................................... 42 

3.3.2 Gamm a Dose Rates on the Sides of the Cask ...................................................... 43 

3.3.3 Neutron Dose Rates on the Sides of the Cask ..................................................... 45 

3.3.4 Gamma Dose Rates on the Top and Bottom Surfaces of the Cask ....................... 49 

3.3.5 Conclusion ................................................................................................................. 55 

4. GAS SURVEY .................................................................................................................................. 58 

4.1 Background ............................................................................................................................. 58 

4.2 1985 Data .............................................................................................................................. 58 

4.3 1986 Post Cask Perform ance Test Gas Sam pling Data ..................................................... 59 

4.4 1994 Through 1999 Data ................................................................................................... 60 

4.5 C-14 and Kr-85 Data .............................................................................................................. 64 

4.6 2000 Data After Refill of CASTOR V/21 ......................................................................... 64 

4.7 Conclusion on Gas Survey ............................................................................................... 65 

5. CONCLUSION ................................................................................................................................ 66 

6. REFERENCES ................................................................................................................................ 68

vi
Appendix A-Dosimeter Radiation Measurements from CASTOR V/21 Cask Performance Test ......... A-1 

Appendix B-Cracked Stitch Welds Identified in Fuel Basket of CASTOR V/21 Cask Basket ...... B-1 

Appendix C-Fuel Assembly Features ............................................................................................... C-1 

A ppendix D - Fuel Tubes ......................................................................................................................... D -1 

Appendix E-Hot Cell Operation and Removal of Fuel Rods from Fuel Assembly T11 .................... E-1 

Appendix F--Cask Interior Wall and Floor ......................................................................................... F-i 

Appendix G-Condition of Existing Cracked Welds Identified in 1985 ............................................ G-1 

Appendix H-Castor V/21 Cask Side Radiation Survey Locations .................................................... H-1 

FIGURES 

1-1. CASTOR-V/21 PWR Spent Fuel Storage Cask ........................................................................... 3 

1-2. CASTOR-V/21 cask cross section ................................................................................................ 4 

1-3. CASTOR V/21 prim ary lid ....................................................................................................... 5 

1-4. CA STOR V/21 cask lid system .................................................................................................... 6 

2-1. Location of fuel storage casks on spent fuel storage pad TAN/TSF-791 ........................................... 8 

2-2. Rebound hammer test locations at CASTOR V/21 storage pad location ....................................... 9 

2-3. Shrink crack indication of CASTOR V/21 concrete pad storage area .............................................. 10 

2-4. Cask bottom cover plate bolt corrosion ...................................................................................... 12 

2-5. Cask bottom cover plate bolt after corrosion removal ................................................................ 12 

2-6. Typical cask primary lid bolt after removal .................................................................................. 13 

2-7. Cask primary lid bolts prior to removal ...................................................................................... 13 

2-8. Primary lid elastomer O-ring crushed dimensional configuration .............................................. 17 

2-9. Elastomer O-ring splice joint configuration ................................................................................ 17 

2-10. Elastom er O -ring splice joint configuration ...................................................................................... 18 

2-11. Primary lid metal O-ring crushed dimensional configuration ..................................................... 18 

2-12. Primary lid metal O-ring .............................................. ..... 19 

2-13. Cask wall at top-typical surface condition ............................................................................... 21 

2-14. Cask floor and wall below fuel tube C8-reflective surfaces free of corrosion .......................... 22

vii
2-15. Cask wall below and between fuel tube A7 and A8-reflective surface free of corrosion

typical of cask wall condition without mineral stains ................................................................... 23 

2-16. CASTOR-V/21 basket crack indication locations and fuel tube arrangement ............................. 26 

2-17. Typical top stitch weld crack at location 14 identified in Figure 2-16 ....................................... 27 

2-18. Typical bottom stitch weld crack at location 14 identified in Figure 2-16 .................................. 27 

2-19. Crack in basket weld below weld crack 4 (120 degree side) identified in Figure 2-13 .............. 28 

2-20. Fuel assembly being lifted out of the basket for visual examination .......................................... 29 

2-21. Fuel assembly T 11 attached to lifting grapple for visual examination ....................................... 29 

2-22. Surry 15 x 15 PWR Fuel Assembly T 11, Serial number LM02JR ............................................... 34 

2-23. Close up of Fuel rod removed from fuel assembly T 1 position 7H identified in 

F ig ure 2 -22 ........................................................................................................................................ 3 5 

2-24. Fuel rod removed from fuel assembly T 1 position 81 identified in Figure 2-22 .................. 35 

2-25. Surface temperatures of the CASTOR V/21 Cask ....................................................................... 39 

3-1. Vertical location of the gamma and neutron dose rate readings on the sidewall of the 

CASTOR V/21 cask. Locations of the readings at the 45 and 90-degree azimuths are 

given as elevation (in millimeters) from the bottom of the cask. The bottom of the cask 

in the fi gure is on the left .................................................................................................................. 41 

3-2. Radial locations of the gamma and neutron dose rate readings on the top and bottom surfaces 

of the cask. Locations of the readings are given as radial displacement (in millimeters) 

from the center on the 45 and 90-degree azimuths .................................................................... 41 

3-3. Comparison of the 2001 and 1985 gamma dose rates on the cask sidewall at the 45 and 

90 -degree azim uths .......................................................................................................................... 44 

3-4. Gamma dose rates in 2001 as a function of elevation for the 45 and 90-degree azimuths on 

the sidewall of the cask. The dose rate profiles for the two azimuths appear to differ slightly 

at the ends of the cask ....................................................................................................................... 4 6 

3-5. Circumferential profiles of gamma dose rates on the cask sidewall, acquired in 2001 as a 

function of azimuth and elevation. Readings were taken at the 4400, 4298, and 2701-mm elevations in 30-degee increments around the entire cask circumference ....................... 47 

3-6. Comparison of the neutron dose rates measured in 1985 and 2001 on the cask sidewall at the 

45 and 90-degree azim uths ............................................................................................................. 48 

3-7. Circumferential profiles of neutron dose rates on the cask sidewall, acquired in 2001 as a 

function of azimuth and elevation. Readings were taken at the 4400, 4298, and 2701-mm 

elevations in 30-degree increments around the entire cask circumference ..................... 50

viii
3-8. Comparison of the 1985 and 2001 neutron dose rates as a function of azimuth and elevation. 

Although the 1985 neutron dosimetry is limited to the 45-120-degree sector, the 2001 

circumferential data compare favorably ...................................................................................... 51 

3-9. Local variations is shielding due to the stacking configuration of the cylindrical polyethylene 

rods. Plot compares the 1985 and the 2001 data for gamma and neutron radiation .................... 52 

3-10. Radial gamma dose rate profile on the top surface of the cask. The plot compares the 2001 

and 1985 readings at the 45 and 90-degree azimuths .................................................................. 53 

3-11. Radial gamma dose rate profile on the bottom surface of the cask. The plot compares the 2001 

and 1985 readings at the 45 and 90-degree azimuths .................................................................. 54 

3-12. Radial neutron dose rate profile on the top surface of the cask. The plot compares the 2001 

and 1985 readings at the 45 and 90-degree azimuths .................................................................. 56 

3-13. Radial neutron dose rate profile on the bottom surface of the cask. The plot compares the 2001 

and 1985 readings at the 45 and 90-degree azimuths ................................................................... 57 

B-I Top weld crack at location 9 identified in B-19. (Triangular air channel between fuel 

tube A 7 & B 8) ................................................................................................................................. B -3 

B-2. Bottom weld crack at location 9 identified in B-19. (Triangular air channel between fuel 

tube A 7 & B 8) ................................................................................................................................. B -3 

B-3. Top weld crack at location 10 identified in B-19. (Triangular air channel between fuel 

tube A 3 & A 6) ................................................................................................................................ B -4 

B-4. Bottom weld crack at location 10 identified in B-19. (Triangular air channel between fuel 

tube A 3 & A 6) ................................................................................................................................ B 4 

B-5. Top weld crack at location 11 identified in B-19. (Triangular air channel between fuel 

tube A 3 & B 6) ................................................................................................................................. B -5 

B-6. Bottom weld crack at location 11 identified in B-19. (Triangular air channel between fuel 

tube A 3 & B 6) ................................................................................................................................. B -5 

B-7. Top weld no crack. Location number 12 identified in B-19. (Triangular air channel between 

fuel tube B 7 & B 8) .......................................................................................................................... B -6 

B-8. Bottom weld crack at location 12 identified in B-19. (Triangular air channel between fuel 

tube B 7 & B 8) ................................................................................................................................. B -6 

B-9. Top weld crack (minor) at location 13 identified in B-19. (Triangular air channel between fuel 

tube B 7 & C S) ................................................................................................................................. B -7 

BIO. Botton weld crack at location 13 identified in B-19. (Triangular air channel between fuel 

tube B 7 & C 8) ................................................................................................................................. B -7 

B-I 1. Top weld crack at location 14 identified in B-19. (Triangular air channel between fuel 

tube C 6 & D33) ................................................................................................................................. B -8

ix
B-12. Botton weld crack at location 14 identified in B-19. (Triangular air channel between fuel 

tube C 6 & D 3) ................................................................................................................................. B -8 

B-13. Top weld crack at location 15 identified in B-1. (Triangular air channel between fuel 

tube D 3 & D 6) ................................................................................................................................ B -9 

B-14. Bottom weld crack at location 15 identified in B-19. (Triangular air channel between fuel 

tube D 3 & D 6) ................................................................................................................................ B -9 

B-15. Top weld crack at location 16 identified in B-19. (Triangular air channel between fuel 

tube A 7 & D 8) .............................................................................................................................. B -10 

B-16. Bottom weld crack at location 16 identified in B-19. (Triangular air channel between fuel 

tube A 7 & D S ) .............................................................................................................................. B -10 

B-17. Crack in basket weld below weld crack 4 (120') identified in B-19 ....................................... B-1i 

B-18. Same as B-17 with complete view of existing weld crack 4 identified in B-19 ...................... B-1 1 

B-19. CASTOR V/21 Basket Crack Indication Locations ..................................................................... B-12 

C-1. Westinghouse gropple lifting fuel assembly V-05 .......................................................................... C-3 

C -2. T op of assem bly V -05 ..................................................................................................................... C -3 

C-3. Oxide mottling noted around spacer #5 on fuel assembly V-05 ................................................. C4 

C-4. Oxide mottling noted around grid spacer #6 on V-05 ................................................................. C-4 

C-5. Fuel assembly V-05-area around upper grid spacer-rods clean and minimal 

b ow in g no ted .................................................................................................................................. C -5 

C-6. Fuel assembly V-05 area around third grid spacer-noted light abrasion marking 

o n sp acer ..... ............................................................ ....................................................................... C -5 

C-7. Fuel assembly V-05 oxide mottling noted around fifth grid spacer ............................................... C-6 

C-8. Fuel assembly V-05 bottom of fuel element V-05-light oxide mottling noted-no loose 

material seen on rods or top of bottom nozzle .................................... C-6 

C-9. Fuel assembly V-05 detail of bottom of element-note oxide rings adjacent to rod 

en d w e ld .......................................................................................................................................... C -7 

C-10. Fuel assembly V-05 top of assembly .......................................... C-7 

C-11. Fuel assembly V-05 oxide mottling noted around grid spacer 5 .............................................. C-8 

C-12. Fuel assembly V-O5-bottom left nozzle-noted mechanical fasteners searing 

guide tubes ........................................... ........................ ............................... C-8 

C-13. Fuel assembly V-05 looking for loose "crud"-non was evident ............................................... C-9

x
C-14. Elem ent T-03 being withdrawn- note fuel rod growth .................................................................. C-9 

C-15. Element T-03 at intermediate grid spacer-rods appear to be relatively clean with a 

light oxide coating ......................................................................................................................... C-10 

C-16. Element T-03 midpoint of assembly noted oxide density increases as we go lower 

on the elem ent ............................................................................................................................... C-10 

C-17. Elem ent T-03 area adjacent to fifth grid spacer ....................................................................... C-1I 

C-18. Element T-03 bottom of fuel element-light oxide-no loose material 

obvious on surfaces ....................................................................................................................... C-11 

C-19. Fuel element T-03-notes serial number-fuel rod growth evident ............................................ C-12 

C-20. Fuel elem ent T-03 bottom ............................................................................................................. C-12 

C-21. Fuel assem bly V -27 being withdrawn .......................................................................................... C-13 

C-22. Fuel assembly V-27 oxide m ottling noted .................................................................................... C-13 

C-23. Fuel assembly V -27 at bottom ...................................................................................................... C-14 

C-24. Fuel assembly V -04 ...................................................................................................................... C-14 

C-25. Fuel assem bly V-14 ...................................................................................................................... C-15 

C-26. Fuel assembly V-i4 oxide m ottling at fifth grid spacer ................................................................ C-15 

C-27. Fuel assembly T-07 note difference in fuel rod growth ............................................................ C-16 

C-28. Fuel assem bly T-07 at sixth grid spacer ........................................................................................ C-16 

C-29. Fuel assembly T-07 bottom .......................................................................................................... C-17 

C-30. Fuel assembly T-12 mid point ...................................................................................................... C-17 

C-31. Fuel assem bly T-12 ....................................................................................................................... C-18 

C-32. Fuel assem bly T-12 ....................................................................................................................... C-18 

C-33. Fuel assembly V-08 tops of fuel rods appear quite even .............................................................. C-19 

C-34. Fuel assembly V-08 at fifth spacer- noted oxide m ottling .......................................................... C-19 

C-35. Fuel assem bly V -08 ...................................................................................................................... C-20 

C-36. Fuel assem bly V-1 I ...................................................................................................................... C-20 

C-37. Fuel assem bly V -i1 at sixth spacer .............................................................................................. C-21 

C-38. Fuel assembly V- 1 ...................................................................................................................... C-21

xi
C-39. Fuel assembly V -01 ...................................................................................................................... C-22 

C-40. Fuel assem bly V -01 at fifth grid spacer ........................................................................................ C-22 

C-41. Fuel assem bly V -01 ...................................................................................................................... C-23 

C-42. Fuel assem bly T-08 near top ......................................................................................................... C-23 

C-43. Fuel assem bly T-08 near sixth grid spacer .................................................................................... C-24 

C-44. Fuel assem bly T-09- note uneven rods ........................................................................................ C-24 

C-45. Fuel assem bly T-09 at sixth grid spacer ........................................................................................ C-25 

C-46. Fuel assem bly T-09 ....................................................................................................................... C-25 

C-47. Fuel assem bly T-16 ....................................................................................................................... C-26 

C4 8. Fuel assembly T-16 at sixth grid spacer ....................................................................................... C-26 

C4 9. Fuel assembly T-16 ...................................................................................................................... C-27 

C-50. Fuel assembly T-16 ....................................................................................................................... C-27 

C-5i. Fuel assembly V-24 ..................................................................................................................... C-28 

C-52. Fuel assem bly V-24 third spacer .............................................................................................. C-28 

C-53. Fuel assembly V -24 near fifth spacer ........................................................................................... C-29 

C-54. Fuel assembly V-24 ...................................................................................................................... C-29 

C-55. Fuel assem bly V -15 third grid spacer .......................................................................................... C-30 

C-56. Fuel assem bly V -15 close ups at sixth grid spacer .................................................................. C-30 

C-57. Fuel assem bly V -15 ...................................................................................................................... C-31 

C-58. Fuel assem bly V -25 ...................................................................................................................... C-31 

C-59. Fuel assem bly V -25 ...................................................................................................................... C-32 

C-60. Fuel assem bly T-13 ................................................................................... ................................ C-32 

C-61. Fuel assem bly T-13 near sixth grid spacer ................................................................................ C-33 

C-62. Fuel assem bly T-1 ..................................................................................................................... C-33 

C-63. Fuel assem bly T-1 1 ................................................................................................................... C-34 

C-64. Fuel assembly T-1 1 attempt to obtain loose material ...................................... C-34

xii
C-65. Fuel assembly V-05 fourth grid spacer oxide mottling noted ....................................................... C-35 

C-66. Fuel assembly V-04 mottling at fifth grid spacer ......................................................................... C-35 

C-67. Fuel assembly V-04--bottom--rod having greatest indication of oxide appears to be 

slightly above the nozzle plate- no conclusion made .................................................................. C-36 

D-1. Fuel Tube A-1 top-appears clean. Typical appearance and cleanliness of all fuel tubes ................ D-3 

D-2. Fuel Tube A-l--Close up of bottom -minor debris noted ............................................................ D-3 

D-3. Fuel tube A-7---closeup of bottom---noted stain-no conclusion-surfaces are clean and 

bright. This view approximates condition viewed at the bottom of all fuel tubes ...................... D-4 

D-4. Fuel tube A-5 bottom-Noted typical absence of quantities of loose material ......................... D4 

D-5. Fuel tube A-6 bottom ...................................................................................................................... D-5 

D-6. Fuel tube A-3 bottom- Surfaces clean and bright .......................................................................... D-5 

D-7. Looking down fuel tube B-5 ........................................................................................................... D-6 

D-8. Fuel tube B-5--bottom ................................................................................................. D-6 

D-9. Fuel Tube B-6 ................................................................................................................................. D-7 

D-10. Fuel tube B-8 bottom ...................................................................................................................... D-7 

D- 11. Inside fuel tube D-5 at the midway of the tube. View as typical of all fuel tubes .......................... D-8 

D-12. M idway inside fuel tube D-6 .......................................................................................................... D-8 

D-13. Fuel tube D-8- bottom .............................................................................................. D-9 

D-14. Fuel tube B4 ................................................................................................................................... D-9 

D-15. Fuel tube A-4 bottom .................................................................................................................... D-10 

D-16. Fuel tube C-6 ................................................................................................................................. D-10 

D-17. Fuel tube C-6--Appeared quite clean ........................................................................................... D-1 1 

D-18. Fuel tube D-3 bottom .................................................................................................................... D-I1 

D-19. Fuel tube D-2 bottom .................................................................................................................... D-12 

D-20. Fuel tube C-5 reflective surfaces and existence of original fabrication/layout characters ........... D-12 

D-21. Fuel tube C-8, bottom ............................................................................................. D-13 

D-22. Fuel tube B-7- near top- showing side plates and basket wall ................................................... D-13

xiii
D-23. Fuel tube E-7, bottom ................................................................................................................... D-14 

D-24. Tube reinforcement plate and associated welds between A-5, A-6, and A-8 .......................... D-14 

D-25. Top of fuel tube A-2 ..................................................................................................................... D-15 

E-1. Fuel assembly and nozzle cutter ..................................................................................................... E-2 

E-2 Fuel assembly T-I 1- after cutting a 3-inch dia. hole in top nozzle to access rods ........................ E-2 

E-3. Fuel assembly T-11- ready to pull fuel rods .................................................................................. E-3 

E-4. Fuel rod being pulled from fuel assembly T-11 .............................................................................. E-3 

E-5. Fuel rod pulled into inspection tray ............................................................................................... E4 

E-6. Oxide mottling on fuel rod .............................................................................................................. E-4 

E-7. Condition of fuel rod-light oxide coating minor-scratches believed from moving rod 

through spacer grids ........................................................................................................................ E-5 

E-8. End of fuel rod ............................................................................................................................ E-5 

E-9. Fuel rod-linear scratches believed from moving rod through spacer grid plates ............................ E-6 

E-10. Fuel rod with linear scratch ........................................................................................................ E-6 

E- 11. Fuel rod with linear scratch ....................................................................................................... E-7 

E- 12. Fuel rod with linear scratch ............................................................................................................ E-7 

E-13. FSV cask interface shipping tube with the 12 fuel rods ........................................................... E-8 

E- 14. Fuel assembly being moved from the hot shop into the hot cell ............................ E-8 

E-15. Camera Resolution Test showing letter size .................................................................................. E-9 

E-16. Typical camera color check ........................................................................................................... E-9 

F- 1. Castor V/21 configuration with primary lid removed ................................... F-3 

F-2. Cask inner wall at top--typical surface condition ..................................................................... F-3 

F-3. Cask floor and wall below fuel tube B7 ..................................................................................... F-4 

F-4. Cask floor and wall below fuel tube C8-reflective surfaces free of corrosion ......................... F-4 

F-5. Cask floor and wall below fuel tube D3- reflective surfaces free of corrosion ........................ F-5 

F-6. Cask floor and wall below fuel tube D6- reflective surfaces free of corrosion ........................ F-5

xiv
F-7. Cask floor and wall below fuel tube A7-reflective surfaces free of corrosion-minor 

granular debris on floor ................................................................................................................... F-6 

F-8. Cask floor below fuel tube A3-reflective surfaces free of corrosion-minor granular 

debris on floor ................................................................................................................................. F -6 

F-9. Cask floor below fuel tube B6-reflective surfaces free of corrosion-minor granular 

debris on floor ................................................................................................................................. F -7 

F-10. Cask floor below fuel tube B8-reflective surfaces free of corrosion-minor granular 

debris on floor ................................................................................................................................. F -7 

F-11. Cask wall below and between fuel tubeA7 and A8-reflective surface free 

of corrosion-typical of cask wall condition without mineral stains-nodular ductile 

cast iron nickel plated ..................................................................................................................... F -8 

F-12. Cask wall below and between fuel tubeD3 and D6-reflective surface free of corrosion

typical of cask wall condition with mineral stains-nodular ductile cast iron nickel plated .......... F-8 

G-1. Existing 1985 weld crack at location No. 1 identified in Figure G-9 ............................................. G-3 

G-2. Existing 1985 weld crack at location No. 2 identified in Figure G-9 ............................................. G-3 

G-3. Existing 1985 weld crack at location No. 3 identified in Figure G-9 ....................................... G-4 

G-4. Existing 1985 weld crack at location No. 4 identified in Figure G-9 ....................................... G-4 

G-5. Existing 1985 weld crack at location No. 5 identified in Figure G-9 ............................................. G-5 

G-6. Existing 1985 weld crack at location No. 6 identified in Figure G-9 ............................................. G-5 

G-7. Existing 1985 weld crack at location No. 7 identified in Figure G-9 ............................................. G-6 

G-8. Existing 1985 weld crack at location No. 8 identified in Figure G-9 ............................................. G-6 

G-9. Castor V/21 basket crack indication location ................................................................................. G-7 

TABLES 

2-1. Characterization of the particulate debris retrieved from the cask floor. Location of the samples 

were taken from the CASTOR V/21 cask for characterization ................................................... 24 

2-2. Fuel assembly examination sequence and lift force measurements ............................................ 30 

2-3. Internal tem perature .......................................................................................................................... 37 

2-4. External temperature data for CASTOR V/21 Cask ................................................................... 37 

3-1. Decay-normalized gamma intensity from radionuclides identified as major contributors to the 

gamma dose rate in 1985, assuming a 16-year decay time .......................................................... 43 

3-2. Maximum and average neutron dose rates for the 1985 and 2001 cask sidewall readings .......... 49

xv
4-1. 1985 casks gas sample composition (volume percent). (Taken from EPRI-4887) ...................... 59 

4-2. 1986 cask gas sample composition (volume percent). (Taken from PNNL-1 1576) .................... 60 

4-3. Cask gas sample composition. Values for 85Kr and 14C are in units of nCi/cm3; values for the 

remaining gas components are in units of volume percent. Except as noted, 85

Kr and 14C were 

determ ined by liquid scintillation ................................................................................................ 61 

4-4. Summary of the nitrogen and oxygen concentrations in the cask cover gas analyses. 

Dim ensions of values are in volume percent .............................................................................. 63 

4-5. Cask gas sample composition (volume mole percent) ................................................................ 65

xvi
EXECUTIVE SUMMARY

To augment the technical basis for renewals of licenses and Certificates of 

Compliance for dry storage systems for spent nuclear fuel and high-level waste at 

independent spent fuel storage installation sites, a dry cask storage system, the 

CASTOR-V/21, was examined and tested after it had undergone an extended 

period of service. This report presents the results of visual examination and 

testing conducted in 1999 and early 2000 on a Gesellschaft fuir Nuklear Service 

(GNS) CASTOR V/21 pressurized water reactor (PWR) spent fuel storage cask 

located at the Idaho National Engineering and Environmental Laboratory 

(INEEL). The examination and testing evaluated the condition and integrity of 

the exterior and interior surfaces of the CASTOR V/21 cask and its contents 

consisting of 21 Westinghouse 15 x 15, PWR spent fuel assemblies from 

Dominion's (formerly named Virginia Power) Surry Power Station) and the 

concrete spent fuel storage pad after a 14-year storage duration. Performance of 

activities was conducted under a standard order agreement between the U.S. 

Nuclear Regulatory Commission, Office of Nuclear Regulatory Research 

(NRC-RES) and the U.S. Department of Energy (DOE), Idaho Field Office, the 

INEEL operated by Bechtel BWXT Idaho. The project (Dry Cask Storage 

Characterization) was jointly funded by Nuclear Regulatory Commission-RES, 

Electric Power Research Institute (EPRI), DOE-Office of Civilian Radioactive 

Waste Management (RW), and DOE-Office of Environmental Management 

(EM) to examine the spent fuel and cask in dry storage at the INEEL. 

Examination and testing, including video and photographic scans, were 

performed at the INEEL Test Area North (TAN) Hot Shop facility and consisted 

of pretest preparations, performance testing, and post-test activities. 

Pretest preparations included conducting cask-external examination, 

measurement of cask exterior surface temperature, extraction of helium cover gas 

samples and analysis, and handling and transport of the cask from the concrete 

storage pad to the TAN Hot Shop. Performance examination and testing included 

cask remote primary lid removal, remote indirect visual examination of cask 

internals, primary lid seals, 21 PWR spent fuel assemblies, and measurement of 

selected fuel assembly temperatures. Interior crud samples were taken, and cask 

surface gamma and neutron dose rates were measured. Selected fuel rods were 

removed from one fuel assembly, visually examined, and shipped to Argonne 

National Laboratory West for nondestructive, destructive, and mechanical 

examination. Post examination and test activities included examination and 

removal of one fuel assembly from the TAN Hot Cell and reinstallation in the 

CASTOR V/21, examination of the primary lid and cask sealing surfaces, 

installation of new primary lid seals, primary lid installation, refill of cask with 

fresh helium, leak testing of primary lid seals, and extraction of helium cover gas 

samples and analysis. 

The CASTOR-V/21 PWR spent fuel storage cask consists of a nodular 

cast-iron body. The cast-body material was found to be in very good condition 

and continues to provide effective gamma shielding. The cask is 4.9 m (16 ft) 

high and 2.4 m (8 ft) in diameter and weighs approximately 112 tons when 

loaded with PWR spent fuel. Two concentric rows of polyethylene rods 

incorporated in the cask wall continues to provide the required level of neutron 

shielding. The external surface consists of heat transfer fins oriented

xvii
circumferentially around the cask surface. The fuel basket within the cask is 

configured to hold 21 PWR spent fuel assemblies and is constructed of stainless 

steel and borated stainless steel for criticality control. The Surry spent fuel 

assemblies are of a standard Westinghouse 15 x 15 rod design. The cask is closed 

with a primary lid having both elastomer and metallic O-rings to seal the cask 

cavity from the environment. A secondary lid, used in commercial applications, 

was not used in this testing program (Section 1.3.4). 

Based on 1985 pretest ORIGEN2 predictions, fuel assembly decay heat 

generation rates totaled approximately 28 kW at the start of testing and 27 kW at 

the end of testing in 1985. In 1985, thirteen of the twenty-one fuel assemblies had 

decay heat rates near 1 kW; the remaining eight assemblies had decay heat rates 

of approximately 1.8 kW. The 1999 measured temperature of the exterior cask 

surface ranged from 103 0F (390C) to 141'F (61°C) (ambient temperature 91.2°F 

[32.9°C]). The hottest of selected fuel assemblies measured 309'F (154°C). This 

compares to 109'F (430C) to 221°F (105 0C) on the exterior cask surface, and a 

peak internal cask guide tube temperature of 676°F (358 0C) measured in 1985 

with a nitrogen backfill, with lid in place and cask in the vertical orientation 

(EPRI NP-4887). 

Examination and Testing Findings 

Based on the 1999 examination and testing results, there was no evidence 

of cask, shielding, or fuel rod degradation during long-term (14 years) storage 

that would affect cask performance or fuel integrity. There was no evidence of 

exterior or interior cask deterioration during storage period, nor were there any 

signs of seal or shielding failure. The fuel was intact, and there were no 

indications of any significant cladding creep or rod bow. There was some crud 

adherent to the rods, but no crud appears to have fallen into either the spacers or 

cask bottom. The crud is not a result of the fuel or cladding oxidation during dry 

storage, but rather the results of oxidation of steel components by the coolant in 

reactor operations. However, fifteen of sixteen stitch welds (Appendix B) in the 

triangular air channels at the perimeter of the basket were found to be cracked. 

These welds attached the basket borated stainless partition plates to the basket 

barrel. Based on the cask testing performed in 1985 (EPRI NP-4887 report), it 

was concluded that applicable triangular air channels were not examined back in 

1985. and the defective stitch welds cracked during high temperature testing 

cycles and test configurations that the cask experienced during the 1985 testing 

campaign rather than during the long-term storage period. These welds in their 

current condition do not affect required cask performance parameters.

xviii
ACKNOWLEDGMENTS 

The Dry Cask Storage Characterization Project personnel would like to 

express our appreciation to those that helped make this project successful and to 

those that contributed to this report. 

Our thanks to Roger Kenneally, Charles Interrante, and Louise Lund of the 

Nuclear Regulatory Commission, Tom Brookmire of Dominion, and John 

Kessler and Albert Machiels from the Electric Power Research Institute for their 

technical guidance, reviews, and support during the performance of this project. 

Our thanks to Peter Dirkmaat and David Koelsch of the U.S. Department 

of Energy, Idaho Operations Office, for their contributions and support of the 

project. 

Our thanks to Bruce Hilton, Michael Billone, and Robert Einziger of the 

Argonne National Laboratory for their technical guidance and their assistance in 

shipping the Surry fuel. 

Our thanks to Matthias Ebner at the Idaho National Engineering and 

Environmental Laboratory (INEEL) for providing technical input and preparation 

of this report. 

Our thanks to the personnel at the INEEL Test Area North for their 

dedication and work performance, which allows this report to be possible. These 

personnel include Lyle Albertson, Jimmy Spells, Kevin Streeper, Ilene Harrell, 

John Cummings, John Hansen, Jake Green, and Doug Illum. 

Our thanks to Jim Dobbins for his exceptional work with the design and 

fabrication of the equipment for boring and pulling spent fuel rods and to James 

Rivera for his support with camera systems that allowed the video inspections to 

be completed. 

Our thanks to the Quality Assurance inspectors Lonnie Gilson and Jim 

Dowalo, to Daryl Lopez for his design efforts and to Richard Deaton and 

Raymond Mitchell for fabrication of needed equipment. 

Our thanks to Paul Ruhter, INEEL Dosimetry, who with Rick Sorenson 

selected appropriate radiation monitoring and reporting methods to allow 

garnma/neutron assessments to be made. 

We would also like to thank Tom Bridges for his support with the 

Transport Plan for movement of fuel rods to Argonne National Laboratory, Gary 

Koyama for budget support, Claude Kimball and Baird McNaught for project 

management efforts, Brent Satterthwaite for cask transportation support, Chris 

Morgan for photo and video service support, Jan Montgomery for technical 

editing support, and Louise Judy for text processing support.

xix
Our thanks to Dinesh Gupta of the DOE Office of Environmental 

Management, and David Stahl and Joe Price of the DOE Office of Civilian 

Radioactive Waste Management, for their technical guidance and financial 

contribution to this project.

xx
ACRONYMS 

ANL-E Argonne National Laboratory-East 

ANL-W Argonne National Laboratory-West 

CASTOR Cast Iron Cask for Storage and Transport of Radioactive Material 

DOE U.S. Department of Energy 

EDS Energy Dispersive Spectrometry 

EM Office of Environmental Management 

EPRI Electric Power Research Institute 

INEEL Idaho National Engineering and Environmental Laboratory 

INTEC Idaho Nuclear Technology and Engineering Center 

LLNL Lawrence Livermore National Laboratory 

LTSM Long-Term Storage Monitoring (Program) 

NRC U.S. Nuclear Regulatory Commission 

PNNL Pacific Northwest National Laboratory 

RES Office of Nuclear Regulatory Research 

RW Office of Civilian Radioactive Waste Management 

SEM Scanning Electron Microscope 

SFB Spent Fuel Behavior (Study) 

TAN Test Area North 

TED track etch dosimeter 

TLD thermoluminescent dosimeter

xxi
Dry Cask Storage Characterization Project-Phase 1: 

Cask Opening and Examination 

1. INTRODUCTION 

1.1 Regulations 

Most nuclear power plants in the United States were not originally designed with a storage capacity 

for the spent fuel generated over the operating life by their reactors. Utilities originally planned for spent 

fuel to remain in the spent fuel pool for a few years after discharge and then to be sent to a reprocessing 

facility. Because reprocessing has been eliminated, and no other option for spent fuel disposition currently 

exists, utilities expanded the storage capacity of their spent fuel pools by using high-density storage racks. 

This has been a short-term solution with many utilities having reached, or soon will reach, their spent fuel 

pool storage capacity (Fisher and Howe 1998). Utilities have developed independent spent fuel storage 

installations as a means of expanding their spent fuel storage capacity on an interim basis until the 

geologic repository is available to accept spent fuel for permanent storage. 

The U.S. Nuclear Regulatory Commission (NRC) promulgated 10 CFR Part 72 (Title 10 1999) for 

the independent storage of spent nuclear fuel and high-level radioactive waste outside reactor spent fuel 

pools. Part 72 currently limits the license term for an independent spent fuel storage installation to 

20 years from the date of issuance. In preparation for possible license renewal, the NRC, Office of 

Nuclear Material and Safeguards, Spent Fuel Project Office is developing the technical basis for renewals 

of licenses and Certificates of Compliance for dry storage systems for spent nuclear fuel and high-level 

radioactive waste at independent spent fuel storage installation sites. These renewals would cover periods 

from 20 to 100 years and would require development of a technical basis for ensuring continued safe 

performance under the extended service conditions. An analysis of past performance of selected 

components of these systems is required as part of that technical basis. The components include the spent 

fuel and all structures, systems, and other components with functions important to safety. The safety 

functions, which apply for normal, off-normal, and accident conditions, are as follows: maintain 

subcriticality, maintain confinement, ensure that radiation rates and doses to workers and the public do 

not exceed acceptable levels and remain as low as reasonably achievable, maintain retrievability, and 

ensure heat removal as needed to meet the safety requirements. 

1.2 Objectives and Scope 

In the mid-1980s, the U.S. Department of Energy (DOE) procured three prototype dry storage 

casks for testing at the Idaho National Engineering and Environmental Laboratory (INEEL): MC-10, 

TN-24P, and CASTOR V/2 1. The primary purpose of the test was to benchmark thermal and radiological 

codes and to determine the thermal and radiological characteristics of the three casks. The CASTOR V/21 

cask was loaded with irradiated assemblies from the Surry Nuclear Power Plant and then tested in a series 

of configurations using a variety of fill gases. Because the tests were not intended to be fundamental fuel 

behavior tests, the fuel prior to the tests had undergone only minimal characterization consisting of visual 

examination of the outside of the assemblies and ultrasonic examination to ensure no breached rods 

would be included. During the tests, the temperature at various locations was monitored, and the cover 

gas was periodically analyzed to determine if any leaking rods had developed. None were found. The 

details of these tests have been reported in a number of documents. 

In 1999, a project was jointly funded by NRC-Office of Nuclear Regulatory Research (RES), 

Electric Power Research Institute (EPRI), DOE-Office of Civilian Radioactive Waste Management (RW), 

and DOE-Office of Environmental Management (EM) to examine the Surry spent fuel in dry storage at 

the INEEL. This project is intended to provided confirmatory data to be used by licensees submitting an

I
application (no later than 2004 for the first licensee) for continuing dry storage beyond 20 years and by 

the NRC staff in their technical licensing reviews. The objectives of the Dry Cask Storage 

Characterization Project are to: 

1. Obtain data to confirm the predicted long-term integrity of dry storage cask systems and 

spent nuclear fuel under dry storage conditions 

2. Provide data to augment the technical bases and criteria for evaluating the safety of spent 

fuel storage and transportation systems. and for extending dry cask storage licenses. 

Phase 1 of the overall program involved moving the CASTORV/21 dry storage cask from the 

INEEL Test Area North (TAN) storage area to the INEEL TAN Hot Shop facility; extracting helium 

cover gas samples and analysis; obtaining temperature readings of the cask exteriors; performing a 

radiation survey; performing video and photographic inspection of the cask exterior and interior surfaces; 

examining cask primary lid seals, fuel assembly exteriors and selected fuel rods; and returning the cask to 

storage. The specific tasks executed in the project are as follows: 

Task 1.1--Equipment: Design, fabricate, and purchase of material, equipment, and fixtures 

necessary to (1) move the cask; (2) remove and replace fuel assemblies; (3) remove fuel rods from one 

selected assembly; and (4) videotape and photograph the external and internal surfaces of the cask, fuel 

assemblies, and fuel rods. 

Task 1.2-Procedures and Training: Develop procedures, obtain required reviews and approvals, 

perform needed training, and other pertinent activities associated with (1) movement of the cask, 

(2) removal of the designated fuel assemblies or canisters, (3) removal of fuel rods, (4) returning the fuel 

assemblies or canisters to the cask, and (5) returning the cask to storage. 

Task 1.3--Hot Shop/Cell Rental: Use the TAN Hot Shop and Hot Cell and TAN operations 

oversight. 

Task 1.4-Inspection of Cask and Internals: Activities include (1) temperature readings of the cask 

exterior; (2) radiation survey; (3) video and photographic inspections of the cask exterior and interior, 

seal, and storage pad; (4) video and photographic inspections of sixteen fuel assemblies in general and 

five assemblies in detail and twelve fuel rods selected from one fuel assembly; (5) obtain crud and smear 

samples; and (6) temporary storage of the fuel rods until transported to ANL for Phase 2 evaluations. 

Task 1.5-Transportation: Transport the fuel rods to Argonne National Laboratory-West (ANLL-W) 

and crud and smear samples to the Idaho Nuclear Technology and Engineering Center (N'TEC) for 

analysis. Return the fuel rod pieces and other material (e.g., crud samples) to the TAN facilities for 

permanent storage. 

Phase 2 of the program, to be reported by others in the future, involves nondestructive, destructive, 

and mechanical examinations of dry-stored spent nuclear fuel elements. This will provide quantitative and 

qualitative information concerning the integrity of the fuel. Examples of the type of information that will 

be obtained include: in situ creep; percentage of fission gas release, internal rod pressure; oxide thickness, 

hydride morphology and orientation, residual cladding thickness, cladding microstructure- hydrogen 

content; creep rates, breakaway temperature; tensile strengths; and ductility. These tasks will take place at 

either ANL-W or ANL-East (ANL-E). The Phase 2 task interface between TAN and ANTL-W at the 

LNEEL are activities at ANL-W associated with the transportation of the fuel rods from the TAN facilities 

at the INEEL to ANL-W, the fuel rod segments from ANL-W to ANL-E. and the return of the fuel rod 

pieces and other material from ANTL-W and ANL-E to the TAN facilities for storage.

2
1.3 CASTOR-V/21 Cask Description 

1.3.1 Cask Body 

The cask body is a one piece cylindrical structure composed of ductile cast iron in nodular graphite 

form. This material exhibits good strength and ductility, as well as providing effective gamma shielding. 

The overall external dimensions of the cask body are 4886 mm (16 ft) high and 2385 mm (8 ft) in 

diameter (Figure 1-1). The external surface has 73 heat transfer fins that run circumferentially around the 

cask and is coated with epoxy paint for corrosion protection and ease of decontamination. 

The cask body wall, excluding fins, is 380 mm (15 in.) thick. Incorporated within the wall of the 

body are polyethylene moderator rods to provide neutron shielding. Two concentric rows of these 60-mm 

(2.3-in.) nominal diameter rods are distributed around the cask perimeter (Figure 1-2). Two lifting 

trunnions are bolted on each end of the cask body. 

The diameter of the inner cavity is 1527 mm (5 ft), and the overall inner cavity length is 4152 mm 

(163 in.). Precision-machined surfaces are provided at the open end of the cask cavity for positive gasket 

sealing, and bolt holes are included at these locations to secure the two cask lids. The interior cavity 

surfaces, including sealing surfaces, have a galvanic-applied nickel plating.

Secondary Lid

Figure 1-1. CASTOR-V/21 PWR spent fuel storage cask. (From EPRI NP-4887. Permission to use this 

copyrighted material is granted by EPRI.)

3
Figure 1-2. CASTOR-V/21 cask cross section. (From EPRI NP-4887. Permission to use this copyrighted 

material is granted by EPRI.) 

1.3.2 Spent Fuel Basket 

The spent fuel basket (Figure 1-2) is a cylindrical structure of welded stainless steel plate and 

borated stainless steel plate, having a boron content of approximately 1% for criticality control. The 

basket comprises an array of 21 square fuel tubes/channels that provide structural support and positive 

positioning of the fuel assemblies. The basket overall height is 4110 mm (13.5 ft) including the four 

130-mm-diameter (5-in.) pedestals that support the basket and fuel weight on the bottom of the cask 

cavity. The basket outside diameter of 1524 mm (5 ft) fits tightly in the cask cavity inner diameter of 

1527 mm (5 ft). The depth of each fuel tube is 4050 mm (13.3 ft). A spacing of 74 min (3 in.) is present 

between the top of the basket cavity and the underside of the primary lid, thus accommodating a fuel 

assembly length of 4124 mm (162 in.) and supporting convection heat transfer. The final assembly results 

in a clearance of approximately 60 mm (2.3 in.) between the top of the fuel assemblies and the bottom of 

the primary lid, for a reference fuel assembly of 4064 mm (160 in.). 

The basket layout results in inter-fuel tube spaces that act as flux traps for criticality control and 

channels to support free convection heat transfer. The basket design ensures a subcritical configuration 

under worst-case conditions, and the basket structure physically protects the fuel under normal and 

accident conditions. 

A pipe with an inner diameter of 42 mm (1.6 in.) and a lead-in funnel at the top are welded to the 

side of a fuel tube near the outer circumference of the basket. The pipe location corresponds to a 

penetration in the primary lid and the low side of the slope in the cask cavity bottom. The pipe provides a 

path for a flanged pipe used to fill and drain the cask.

4
1.3.3 Primary Lid

A stainless steel primary lid, 1785 mm (6 ft) in diameter and 290 mm (12 in.) thick, is provided 

(Figure 1-3). Forty-four bolt holes are machined near the lid perimeter to secure the lid to the cask body. 

Two grooves machined around the lid underside, inside the bolt circle, are provided for 0-ring seals 

(Figure 1-4). The inner groove accepts a metal "C" shaped O-ring, which serves as the first barrier 

between stored fuel and the environment. The outer grove accepts an elastomer 0-ring. A 

10-mm-diameter (0.4-in.) penetration through the lid provides access to the annulus between the two seals 

to perform post-assembly leak testing. This penetration is plugged when not in use. 

Three penetrations through the lid are provided for various cask operations. A 35-mam-diameter (1.4-in.) 

straight-through penetration is used for water fill/drain operations. This penetration is located near the 

perimeter of the lid and is normally sealed with two flanges equipped with elastomer O-rings. This 

location corresponds to the pipe attached to the fuel basket. The other two penetrations, spaced next to 

each other and covered by a single flange, are also located near the lid perimeter, but 180 degrees from 

the fill/drain penetration. The through-lid penetration at this location is equipped with a quick-disconnect 

fitting used for vacuum drying and backfilling with gas. The second penetration at this location leads to 

the lower edge of the lid. Although not needed for the CASTOR-V/21, this penetration could be used for 

leak-testing an optional third lid gasket. This penetration is sealed by a gasketed seal plug in addition to 

the top cover flange. 

The primary lid used during 1985 testing was not a standard lid and has 10 additional penetrations 

for fuel assembly guide tube instrumentation (thermocouple [TC] lances). The pattern of the 10 fuel 

assembly instrumentation penetrations was selected to measure radial temperature profiles across the 

basket in the spent fuel assemblies. 

Lid Orientation Mark 

TC Lance 

00 Penetrations Penetration (typical 9 places) 

0 900 

Quick Disconnect 

. D Penetration 

Seal Test TC Lance/Pressure Monitoring/ 

Penetration 1800 Gas Sampling Penetration 

Figure 1-3. CASTOR V/21 primary lid. (From EPRI NP-4887. Permission to use this copyrighted 

material is granted by EPRI.)

5
Pressure Switch 

Penetration

Metal O-Rings

Figure 1-4. CASTOR V/21 cask lid system. (From EPRI NP-4887. Permission to use this copyrighted 

material is granted by EPRI.) 

1.3.4 Secondary Lid 

The Secondary Lid, used in commercial application, was not used during the 1985 CASTOR V/21 

cask performance test or the long term storage monitoring program conducted during 1994 through 1999, 

because of interference with fuel assembly instrument leads and the repeated removal and handling of the 

secondary lid for gas sampling activities. The following functional description of the Secondary Lid is 

provided for completeness. The stainless steel secondary lid is 2007 mm (79 in.) in diameter and 90 mm 

(3.5 in.) thick (Figure 1-4). Forty-eight bolt holes are machined near the lid perimeter to secure the lid to 

the cask body. Two concentric grooves located inside the bolt circle on the underside are provided for a 

metal 0-ring/elastomer O-ring sealing system of the same design as that used on the primary lid. Three 

normally sealed penetrations are provided for various cask operations (Figure 14). A 10-mm-diameter 

(0.4-in.) penetration through the lid provides access to the annulus between the two seals for post

assembly seal testing. A gasketed seal plug is used to close this penetration. 

A second penetration is equipped with a quick-disconnect fitting, which is used for vacuum drying 

and gas backfilling of the primary/secondary inter-lid space. A 130-mm-diameter (5-in.) cover plate and 

gasket secured by six 12-mm (0.5-in.) bolts is in place when this penetration is not used. The third 

penetration provides a pressure sensing port between the inner-lid space and a pressure switch mounted in 

the secondary lid. The pressure switch is the primary component of the cask seal monitoring system. Due 

to the secondary lid not being used dose rates discussed in Section 3 were obtained on the primary lid 

exterior surface. Addition of the secondary lid will greatly reduce measured gamma dose rate values.

6
2. INSPECTIONS AND TESTS 

The following inspections and tests are described in this section: 

"* Long-term surveillance of the concrete pad upon which the casks rest. Objective: look for 

any degradation of the concrete pad. 

" Cask exterior visual inspection. Objective: look for any visual indications of degradation. 

"* Cask lid bolts. Objective: look for degradation such as rust. 

"* Cask lid seals. Objective: look for signs of corrosion, wear, or scoring of the seals that may 

lead to loss of the pressure boundary. 

"* Cask interior and basket inspections. Objectives: look for signs of corrosion, presence of 

crud spalled from fuel assemblies; gouging; and integrity of basket welds. 

"* Fuel assembly visual inspections. Objectives: look for signs of additional corrosion or other 

cladding degradation; determine if assemblies have bowed or otherwise corroded that may 

cause difficulty removing the assemblies from the cask. 

2.1 Long-Term Spent Fuel Storage Pad 

2.1.1 Description of Pad 

Facilities were constructed in 1985 directly to the west of TAN-607, adjacent to the rail track that 

exits the TAN Hot Shop, for the long-term surveillance of several dry storage casks. The facilities 

consisted of a concrete pad for the dry storage casks and a data acquisition building. The facilities 

originally include a weather station that was part of the 1985 thermal testing program and was removed 

from service between 1990 and 1991. 

The concrete pad was designed to hold several spent fuel storage casks (Figure 2-1). Analysis for 

pad loading show that the concrete slab can support the weight of ten 102,454 kilograms (225,400-pound) 

casks arranged in two rows and spaced 1.52 meters (5 ft) apart and located 1.52 meters (5 ft) from the 

edge of the slab. The size of the pad is approximately 28.7 m (94 ft 4 in.) long by 12.1 m (39 ft 8 in.) 

wide. The pad consists of 0.6 1-m (2-ft) thickness of concrete on top of a minimum of 30 cm (12 in.) of 

compacted subbase of pit run gravel. The concrete was reinforced with two mats of #6 steel reinforcement 

bar spaced 18 cm (7 in.) on center (each way); the mats were each embedded 10 cm (4 in.) below and 

above the top and bottom surfaces of the pad, respectively. The concrete was covered and kept wet during 

the first 7 days of the curing period to ensure maximum strength and durability. The design strength of the 

concrete was 28 MPa (4000 psi); the 28-day post-cure compression strength averaged 30 MPa (4400 psi). 

Although it was designed to hold several storage casks, the pad held four dry storage casks, 

including the CASTOR V/21. The CASTOR V/21 cask was located approximately 13.9 m (45 ft 6 in.) 

from the west edge and 4.3 m (14 ft) from the north edge of the pad (Figure 2-1).

7
V/21 

96"- O.D. 

LOADED WEIGHT: 225, 419 LB 

FULLY LOADED WITH 21 

WESTINGHOUSE PWR 15X15 

INTACT FUEL ASSEMBLES 

TN-24P 

9o00 O.D. 

LOADED WEIGHT: 177, 728 LB 

PARTIALLY LOADED WITH 7 

CONSOLIDATION FUEL CANISTERS 

EACH CONTAINING THE RODS 

FROM 2 WESTINGHOUSE PWR 

FUEL ASSEMBLES 15X15 

24 SPACES AVAILABLE IN CASK

MC-10 

88-0 O.D. 

LOADED WEIGHT: 214, 102 LB 

PARTIALLY LOADED WITH 18 

WESTINGHOUSE PWR 15X15 

INTACT FUEL ASSEMBLES 

CASK WILL HOLD 24 

VSC-17 

105"'0 O.D. 

LOADED WEIGHT: 210, 926 LB 

FULLY LOADED WITH 17 

CONSOUDATION FUEL CANISTERS 

EACH CONTAINING THE RODS 

FROM 2 WESTINGHOUSE PWR 

FUEL ASSEMBLES 15X15

Figure 2- 1. Location of fuel storage casks on spent fuel storage pad TAN/TSF-791.

8
2.1.2 Pad Inspection

The evaluation of the integrity of the 15-year-old storage pad consisted of the testing of the 

structural soundness of the surface of the concrete and a visual assessment of the physical condition of the 

concrete surface, particularly at and immediately around the cask location. 

The structural soundness of the concrete was determined by ASTM test standard C805-94 (Test 

Method for Rebound Number of Hardened Concrete, also known as the Swiss hammer test). The Swiss 

hammer test was performed on the concrete surface in 9 places in a 37.2 m (122 ft ) area centered on the 

placement of the CASTOR V/21 cask, of which 5 places were selected in the area under the cask. The 

nine test results (Figure 2-2), which ranged from 28 MPa (4050 psi) to 41 MPa (5900 psi), averaged at 

33 MPa (4800 psi) and demonstrated that the structural integrity of the concrete pad still meets or exceeds 

the 28 MPa (4050 psi) design strength of the concrete.

(20'-o")

b 

b c'4

CASTOR V/21 STORAGE -' 

LOCATION ON SPENT FUEL 

STORAGE PAD - TAN/TSF-791

* *=TEST LOCATIONS - 1999

Figure 2-2. Rebound hammer test locations at CASTOR V/21 storage pad location.

9
The pad was also visually inspected for evidence of degradation and structural failure. The surface 

of the whole pad did not exhibit any evidence of structural failure of the concrete, such as open cracks or 

cracks with displacements in elevation of the surface. The surface of the concrete did not exhibit any 

evidence of spallation of the surface, exposed aggregate, or aggregate pop-out from the surface. The 

surface was solid and exhibited only minor wear and environmental weathering, well within the extent of 

weathering expected for the cold and windy climate of Idaho. The broom-finished unpainted surface 

exhibited only a network of faint, fine surface shrinkage cracks (Figure 2-3), less than 0.8 mm (1/32 in.) 

wide and of superficial depth, and a few rust stains under the cask from lightly rusted bolts on the cask. 

Similar cracks were prevalent across the entire surface of the pad, and were not associated with the cask 

locations. Tests with a straight taught line across the 6.1 m x 6.1 m (20 x 20 ft) grid indicated that there 

was no sag or vertical displacement in the concrete associated with the crack network; measurements with 

a straight edge and the taught line indicated only localized variations in the elevation of the concrete that 

were less than 3.2 mm (1/8 in.). The localized variations in elevation were not associated with the location 

of the cask and were most likely an artifact of the screeding and finishing when the concrete was 

originally poured.

CASTOR V/21 STORAGE 

LOCATION ON SPENT FUEL 

STORAGE PAD - TAN/TSF-791

AK

Figure 2-3. Shrink crack indication of CASTOR V/21 concrete pad storage area.

10
2.2 Cask Exterior 

The cask exterior (Figure 1-1) consists of the cask cooling fins painted with epoxy paint, upper and 

lower exposed portions of the cask body painted with epoxy paint, upper and lower lifting trunnions, 

bottom cover plate, and an auxiliary top weather cover. These items have been exposed to all seasonal 

outside weather conditions for the past 14 years. The Primary Lid (Figure 1-4) was protected by the 

auxiliary top weather cover from weather conditions. 

The entire cask exterior surface of items described above were video examined and supplemented 

by direct visual examination for any evidence of degradation. The resolution and color rendition of the 

cam recorder was checked with a resolution chart (with alphabet characters ranging in height from 

0.063 in. to 0.10 in.) and a Kodak color resolution chart. The observed conditions are: 

"* The cask cooling fins contained small superficial corroded areas where the epoxy paint had 

been chipped and peeled due to possible combination of improper mixing or application of 

paint, or interface with cask handling equipment. At two separate fin locations on the outside 

fin edge were gouged as a result of cask movement activities. The operational performance or 

structural integrity of the fins are not impacted by the gouges. 

"* The upper and lower exposed portions of the cask body and upper and lower lifting trunnions 

and bolts contained minor superficial rust areas resulting from exposure to outside weather 

conditions and contact with dissimilar metals and cask handling equipment. 

"* The cask bottom cover plate that interfaces with the concrete spent fuel storage pad during 

storage contained minor discoloration from exposure to rain and snow water, surface 

scratches from cask movement activities, superficial rust areas around the circumference of 

the cover plate, and cask body interface. There was evidence of heavy corrosion deposits on 

one of the plate recessed attachment bolts at the cask 180-degree position as shown in 

Figure 2-4. Removal of the corrosion deposits from the affected area supplemented with 

visual examination showed bolt head pitting and surface irregularities as presented in 

Figure 2-5. Based on the condition and function of the bolt, the bolt and surrounding area was 

coated with moisture resistant sealant and retained for service. The cause of the heavy 

corrosion deposits appeared to be from long-term exposure to moisture resulting from wet 

weather conditions and condensation possibly being generated on the top side of the cover 

plate and resulting excess water passing through the bolt thread path. 

The primary lid was removed using an overhead crane. The operation was sensitive because of the 

tight clearance between the lid and the cask. The torque values on the bolts were not recorded during their 

removal. 

2.3 Cask Lid Bolts 

The cask is sealed by the inner primary lid. The outer secondary lid was not installed or used 

during the long-term storage monitoring program for reasons described in Section 2.2. The primary lid is 

secured with 44 bolts. The 44 bolts of the primary lid were individually inspected visually for their 

physical condition, specifically for evidence of cracks, pitting corrosion, general corrosion, thread 

damage, and any discoloration. The exposed portion of the bolt heads of the 44 bolts were examined prior 

to removal and the entire bolt surface after removal. Included in the examination was the exposed portion 

of the primary lid penetration flange bolts. 

All bolts were in satisfactory condition. The exposed portion of all bolt heads were found to be in 

very good condition. None of the primary lid bolts as shown in Figures 2-6 and 2-7 had any indications of

11
Figure 2-4. Cask bottom cover plate bolt corrosion.

Figure 2-5. Cask bottom cover plate bolt after corrosion removal.

12
Figure 2-6. Typical cask primary lid bolt after removal. 

Fu . k p a 

Figure 2-7. Cask primary lid bolts prior to removal.

13
pitting or general corrosion, cracks, thread damage, discoloration, or any defects or indications of 

potential failure. All bolts had a deposit of graphitic thread lubrication over the threaded area of the shaft. 

Nine of the bolts also had a residue of a light gray/white material on the threads that resembled tape 

residue, and a residue of a red compound, probably a thread compound. 

The lid was removed using an overhead crane. The operation was sensitive because of the tight 

clearance between the lid and the cask. The procedure for opening the V/21 cask was checked and torque 

values on the bolts were not recorded for opening the cask. 

2.4 Cask Lid Seals 

The CASTOR V/21 primary lid is sealed by a metal and an elastomer O-ring housed in separate 0

ring grooves. The O-rings of the primary lid were inspected immediately after opening of the cask and 

after replacement of the O-rings. The objectives of the inspection were to evaluate the condition of the 

seals for potential degradation due to: 

* Oxidation of the elastomer and metal seals; thermal degradation of the elastomer seal 

* Embrittlement or hardening, including cracking, crazing, and evidence of loss of elasticity or 

ductility 

* Physical damage to the seals, such as scratches across the seal surfaces. dents, and seal 

deformation. 

2.4.1 Description of the Seals 

The metal O-ring seals are, 1600 mm (5.25 ft) outside diameter x 1580.2 mm (5.19 ft) inside 

diameterx 9.9 mm (0.39 in.) cross section diameter, Helicoflex type HN 200, manufactured from the 

following materials: 

"* Outside lining: aluminum 

"* Inside lining: 304 L or 316 L stainless steel 

"* Spring: Nimonic or Inconel. 

The elastomer O-ring seals are, 1580 mm (5.19 ft) outside diameter x 1560.2 mm (5. 12 ft) inside 

diameter x 9.9 mm (0.39 in.) cross section diameter, EPDM type 50049-2.2. supplied by the Helicoflex 

Company. Columbia, SC 

2.5 Remote Inspection 

The O-rings of the primary lid were inspected by remote video camera immediately after the 

opening of the cask and by direct visual examination after replacement of the O-rings prior to resealing 

the cask. The initial remote camera inspection occurred on September 8, 1999, when the primary lid was 

first opened. The O-rings on the primary lid were reinspected by remote camera on September 20, 1999. 

Obviously, the remote camera inspection of the seals (before removal from lid) permitted the inspection 

only of the compression surface contacting the cask. rather than the whole surface of the seal. The remote 

inspection used three video cameras mounted on work stand (work platform) at 120 degree intervals 

around the top perimeter of the cask.

14
The resolution and color rendition of the cameras were checked daily with a resolution chart (with 

alphabet characters ranging in height from 0.063 in. to 0.10 in.) and a Kodak color resolution chart. The 

magnification and resolution of the remote cameras were sufficient to discern fine defects; in the initial 

inspection immediately upon opening the cask, we were able to identify clearly a long fine hair 

(presumably human hair) that was looped across the two O-rings of the primary lid. The O-rings were 

also inspected by direct visual examination on March 28, 2000, after they were removed from the primary 

lid for a complete examination of the whole surface of the seals. 

The O-rings in the primary lid were in excellent condition. The remote visual inspection 

immediately upon opening the cask and removal of the primary lid indicated that the elastomer and the 

metal O-rings were free of breaks, cracks, crazing, delamination, pull-outs, oxidation, or other evidence of 

degradation of the O-rings. 

The compression area of each seal was in excellent condition with no visible damage to the seating 

(compression) area of the O-ring seal. The only defect that was visible by remote camera was an 

imperfect splice joint in the elastomer O-ring. The compression sealing area of the metal O-ring was quite 

reflective, glinting in the natural illumination in the hot shop and indicating that no corrosion or excessive 

oxidation had occurred. The compression area of the metal O-ring was textured due to the impression of 

the machining marks from the mating metal seal surface of the cask body. The metal compression surface 

did not show any evidence of breaks, scratches, dents, distortion, or corrosion. 

The O-rings, particularly the elastomer, did exhibit random, crisply delineated patches of light 

surface discoloration, which appeared gray against the black color of the elastomer. These patches were 

often associated with similar areas of light discoloration on the metal flange of the lid, especially in the 

bolt circle area of the flange. In several areas on the bolt circle of the lid flange, the discoloration patches 

formed particularly thick surface films and had partially peeled. From the peeled sections, it was obvious 

that the discolored areas were polymeric films and most likely are deposits of excess antigalling and 

antiseize compounds used on the lid bolts. Slight amounts of the antiseize compound were also evident on 

the land of the flange between the two O-ring grooves. The compound may be responsible for the random 

particles of "dirt" on the sides and a few dark discolorations or "smudges" on the metal O-ring 

compression surface. 

The discolorations on the compression surface of the metal O-ring were usually associated with 

similar gray discoloration on the elastomer seal and with deposits/films of material on the metal flange of 

the bolt circle. It seems possible that excess fluid antiseize compound may have run off the lid bolts onto 

the solid sealing surface of the cask body and wicked onto the elastomer and the metal O-rings before the 

bolts and the lid were torqued down. 

The O-rings were reinspected by remote video approximately two weeks after the initial inspection. 

In the interim, the primary lid was removed and replaced daily (without bolting) on the cask as part of the 

fuel and cask inspection process. On the second remote inspection, the condition of the O-rings was 

unchanged, except that the metal O-ring exhibited a few pin-point indentations. The indentations were 

typically less than I mm (0.04 in.) in dimension in the compression area of the seal. Quite likely these 

pin-point indentations were the consequence of small amounts of grit on the cask body seal surface, 

deposited there by the seal protector or by air currents. A swab collected a small but visible amount of 

grime from the seal surface on the cask body. 

The precision-machined seal seat of the cask body was in excellent condition. The surface of the 

seat was clean, brightly reflective, and free of corrosion, scratches, cracks, dents, or other forms of 

degradation that could affect the quality of the seal. However, the surface was lightly coated with a thin 

film whose optical density varied and seemed to be associated with patches of excess thread lubricant in 

the bolt circle area. Except for occasional small dense spots, the film was generally faint and difficult to

15
define. However, two continuous, concentric bright circular lines were discernable on the seal seat, 

equivalent to the imprint of the O-ring seals, where the film was absent. 

2.5.1 Direct Visual Inspection 

Finally, the O-ring seals were examined by direct visual inspection after they were replaced with 

new O-rings prior to resealing the cask. The whole surface of the seals, including the edges, was 

examined. 

The elastomer was still firmly resilient in consistency, flexible, and limber, with no evidence of 

embrittlement, stiffness, or depolymerization. Bending, pulling, twisting, and coiling the elastomer into a 

30 cm (12 in.) diameter coil did not cause fracture or stress failure. There was no physical evidence of 

delamination or pullout of material, spallation, or chalking. The black elastomer surface exhibited a satin 

matte sheen and was free of breaks, cuts, scratches, cracks, or craze defects. There was no evidence of 

oxidation or physical degradation of the elastomer. The crushed dimensional configuration taken along a 

segment of the elastomer O-ring after removal (14 years of service) is presented in Figure 2-8. 

The elastomer O-ring was fabricated with a 45-degree scarf splice joint. As was noted in the remote 

camera inspections, the splice joint was slightly misaligned and partially open; the glue did not 

completely fill the gaps in the joint as shown in the Figures 2-9 and 2-10. However, the joint still had 

good strength and could not be pulled apart manually. 

The elastomer exhibited numerous sharply defined gray patches on all surfaces. These patches 

were not associated with surface relief or differences in flexibility, resiliency, or firmness of the polymer 

and had a graphitic sheen, suggesting that they are probably caused by excess antiseize lubricant that was 

used on the lid bolts. These graphitic patches were much more numerous on the back side of the 

elastomer. which contacted the seat of the O-ring groove in the lid, suggesting that the antiseize lubricant 

was used as an aid to hold the seal in place during lid assembly. 

The metal O-ring was also in good condition and was still ductile, as indicated by a few slight 

kinks, bends. and fresh surface scratches imparted by handling during removal from the lid. The metal 

showed no evidence of gross embrittlement such as cracks either in the body of the seal or at the handling 

defects. In general, the metal surface exhibited a metallic luster and showed no evidence of corrosion or 

extensive oxidation. The flattened contact surfaces bore the imprint of the machine marks of the O-ring 

groove and the mating seal surface of the cask body. The crushed dimensional configuration taken along a 

segment of the metal O-ring after removal (14 years of service) is presented in Figure 2'-11. 

The top surface of the metal O-ring, which was in contact with the seat of the O-ring groove, did 

not have any defects except for a few fresh (unoxidized) shallow scratches that have been attributed to 

handling damage during removal from the lid. The bottom surface as shown in Figure 2-12, which was in 

contact with the seal surface of the cask body, exhibited approximately 20 small pin-point indentations 

into the surface. The indentations were typically less than 1 mm (0.04 in.)in lateral dimension and did not 

appear to puncture the thickness of the metal 0-ring. Because these were not spotted in the initial remote 

camera inspection, these features are being attributed to compression damage from environmental grit on 

the seal surface of the cask body. The grit was probably transfer from the bottom of the seal protector to 

the seal area during the fuel examination operations. 

2.5.2 Observations and Results 

On the metal seal there was observed: 

0 Several pinpoint indentations, some on sealing surface; typically <0.5 mm dia, probably 

<0.5 mm deep

16
DMENSON A 

L0r- MEAS. 

L0" 

6" 0.-.• 

S M0.34" 

ir r0 0.337 0.340" 

12" 0.338" 

14' 0.342? 

16" 0.34T 

AVERAGE 0.339" 

IT: 0.002"

Figure 2-8. Primary lid elastomer O-ring crushed dimensional configuration.

Figure 2-9. Elastomer O-ring splice joint configuration.

17

DIMENSION 8 

Mm~ 

0.427" 

0.422" 

0.423" 

0.421" 

0.421" 

0.421" 

0.420" 

0.420" 

0.422" 

0.002"

B 

A 

- g -0
Figure 2-10. Elastomer O-ring splice joint configuration.

DIMN•4N A DIMENSON B D 0ENSION C 

LOC. MEAS. MEA& MEAS. 

2' 0.355' 0.445' 0.085' 

e .35f" 0.445" .080' 

S. " .34" 0.446' 0.065' 

18 D.353" 0.450 0.081' 

!w 0.355m 0.452' 0.086' 

A C 12" 0.353 0.454' 0.095' 

14' D.353' a.456' 0.088T 

W1" 0.354' 0.450' 0.074' 

18a 0.354' 0.454' 0.073' 

AVA E_ 0.354 0.450' 0.081' 

C: o0.001' 00o4" 0.009' 

Figure 2- 11. Primary lid metal O-ring crushed dimensional configuration.

18
Figures 2-12. Primary lid metal O-ring. 

"* Several light scratches, a few across the sealing surface; most have metal glint in scratch 

root, indicating fresh scratches 

"* No evidence of significant general corrosion, pitting corrosion, stains 

"* Sealing surface defined by the impression of the machining marks from the seal surface on 

cask 

"* No evidence of significant oxidation or spallation of these fine line features 

"* No deformation of the individual line impressions despite numerous cycles of lid 

removal/replacement during fuel inspection. 

2.6 Cask Interior 

The CASTOR V/21 cask is a one-piece cylinder manufactured from ductile cast iron in nodular 

graphitic form. The external dimensions of the cask body is 4886 mm (16 ft) high and 2385 mm (8 ft) 

external diameter. The internal diameter is 1527 mm (5 ft), and the cavity depth is 4152 mm (13 ft 7 in.). 

The internal surfaces of the cask, including the sealing surfaces, were galvanically coated with nickel 

plating. 

The fuel basket is a cylindrical barrel that is partitioned into an array of 21 square fuel tubes in a 

quadrant layout. The basket is fabricated from welded stainless steel and borated stainless steel plate for 

criticality control; the borated steel contains approximately 1% boron. Each fuel tube is separated from 

the adjacent tube by channels that act as flux traps for criticality control and as channels for convective 

heat transfer. The basket has an outer diameter of 1524 mm (5 ft) and a gross depth of 4110 mm (13.5 ft), 

including the four 130 mm (5 in.) diameter pedestals at the bottom of the basket. The basket barrel fits 

snugly within the cask with a clearance of only 3 mm (0.12 in.) total and 74 mm (3 in.) between the 

bottom of the primary lid and the top of the basket. 

The objectives of examination of the cask interior were to inspect the exposed, accessible internal 

surfaces of the cask structure for evidence of the cask and basket degradation caused by long-term

19
storage. For the cask cavity, the visual inspection focused on evidence of corrosion and crack formation in 

the sidewalls and the bottom of the cask, particularly in the bottom comer. It also focused on the failure of 

the nickel coating by blistering, delamination, corrosion, or discoloration. For the fuel basket, the 

inspection focused on evidence of new cracks in welds or in walls of fuel tubes,' propogation of existing 

cracks in welds, corrosion and discoloration of fuel tube walls, and accumulations of oxide particles on 

bottom support brackets and at the bottom of the cask in each fuel tube. 

2.6.1 Method of Inspection 

Most of the cask inner wall and bottom was not accessible to visual inspection due to the size and 

tightly fitting characteristics of the basket. At the top of the cask, only approximately 8 cm (3 in.) of 

sidewall were exposed above the top of the basket and the rebate below the seal area of the cask body 

(i.e., the sidewall area between the top of the basket and the bottom of the lid), the 5 cm (2-in.) step of the 

rebate, and approximately 25 cm (10 in.) of sidewall between the primary and secondary seal seats. The 

floor of the cask was accessible only through the 21 fuel tubes. The bottom comer and 2-5 cm 

(1-2 in.) height of cask sidewall was partially accessible through a few of the larger flux traps 

(only = 9 cm [3.5 in.] at the widest) at the periphery of the basket. 

The inspection of the inner wall at the top of the cask was performed by remotely using three video 

cameras mounted on work stand (work platform) at 1200 intervals around the top perimeter of the cask. 

The floor and the bottom corner of the cask were examined with a radiation-tolerant miniature (pencil) 

camera and light mounted at the end of a 4.5 m (15 ft) pole. The pencil camera head was a cylindrical unit 

approximately 1.3 cm (0.5 in.) in diameter and 6.4 cm (2.5 in.) long. As with the video cameras, the 

pencil camera resolution was checked daily with the resolution and color charts. 

Because of the tight clearances for access to the bottom comer and sidewall of the cask, the 

examination was attempted initially with a borescope, but failed because of the narrow field of view, short 

working distance, short depth of field, and poor dynamic response of the borescope camera. 

2.6.2 Interior Cask Sidewall 

The upper exposed area of the inner sidewall of the cask was in very good condition as shown in 

Figure 2-13. The galvanically applied nickel coating was still intact and did not show any evidence of 

blistering, peeling, cracking. delamination, or corrosion. 

The nickel-plated sidewall was free of significant defects. However, a few isolated minor, 

superficial features or imperfections were visible in the visual inspection: these appeared to be light scuff 

and faint scrape marks that were most likely created during the initial installation of the basket in the 

cask. Adjacent to fuel tube D3 (at the 2700 mark with respect to the 0' orientation mark as shown in 

Figure 2-16), the sidewall had an imperfection that initial inspection identified as a blister or dimple. 

However, close examination of the illumination shadows indicated that the feature was a shallow 

depression (dimple) about 2 cm (0.8 in.) in diameter and probably only approximately a millimeter deep, 

with the nickel coating still intact. The visible surface of the sidewall also had several isolated, randomly 

oriented superficial lines that could be surface deposits (from abrasion by a softer material) or superficial 

scrapes. These features are quite faint, with no discernible vertical dimensions, burrs, ridged edges, or 

plow marks that usually are associated with scratches that penetrate coatings or gall a surface. 

a- Cracks were formed in some of the basket welds during the initial thermal testing conducted in 1985 (prior to long-term 

storage). The intent here is to determine if new cracks may have formed during long-term storage.

20
Figure 2-13. Cask wall at top--typical surface condition.

Considering the tight fit of the basket within the cask body, the nickel coating on the upper 

sidewall shows little evidence of damage due to insertion of the basket. The only discernible feature that 

might constitute significant coating damage was a black mark on the sidewall at the level of and 

coincident with the corner of fuel tube D3. However, the black surface mark appeared to be superficial, 

and did not have any burrs or dimensional relief indicative of substantial abrasion damage, corrosion 

product formation, or cracks. While the feature is coincident in location with the corner of fuel tube D3, it 

cannot be the result of abrasion by the corner of the fuel tube (by vibration from cask handling), for the 

fuel tube is separated from the wall by the thickness of the steel barrel plate comprising the outer rim of 

the basket. Instead, this feature may be the result of abrasion during insertion of the tightly fitting basket 

into the cask or from vertical thermal expansion of the tightly fitting basket barrel wall during the 1985 

thermal tests. The upper sidewall has several similar, less distinct blemishes that could be construed as 

light scuffing or abrasion of the nickel coating from contact during the insertion of the basket into the 

cask body. These features consist of black 'scuffs' and spots on the nickel surface, as if the nickel plating 

was lightly abraded from the high points of the rough as-machined surface of the cask body. These 

features have no discernible relief, implying negligible superficial damage at worst, and have no evidence 

of more than possibly superficial surface corrosion as might have occurred prior to sealing the cask in 

1985. Furthermore, there was no evidence of delamination and peeling of the nickel layer around these 

features or of subsurface corrosion or blistering in the areas surrounding the features, which could be the 

expected effect from a corrosive, oxidizing environment. 

2.6.3 Interior Cask Bottom and Bottom Sidewall 

The cask bottom and bottom sidewall could be inspected only to a limited extent by access through 

the 21 fuel tubes and the eight channels at the perimeter of the basket. Access via the remaining flux traps 

was prevented by the tight dimensions of the traps and the structural gussets and spacers in the cavities of 

the traps. In addition, the inspection of the whole area of the cask floor was hampered by the small 

clearance (3.8 cm [1.5 in.]) between the bottom of the basket and the cask floor.

21
The floor of the cask was of roughly grained as-cast texture, overcoated with the nickel plating. 

The floor of the cask turned smoothly up into the sidewall, so that the first centimeter or two of sidewall 

was also generally of rough as-cast texture. The sidewall above the bottom comer radius was smoother 

than the floor as if it had been machined to remove the as-cast texture prior to nickel plating. 

The nickel plating on the floor and bottom sidewall was generally clean and quite reflective despite 

the as-cast texture as shown in Figures 2-14, 2-15, and Appendix F. There was no evidence of any 

corrosion, cracks, or flaws in the nickel plating, such as blistering or delamination in the floor, comer or 

sidewall of the cask. In general, the bottom sidewall was quite clean and reflective, particularly those 

areas that were machined prior to nickel plating. There were, however, isolated areas that appeared to be 

covered with light-colored spots of material that were not reflective and had no relief. These patches 

appeared to be mineral spots (Appendix F, Figure F-12), as if deposited from residual water in the cask 

(as from evaporation of residual plating solution or rinse water). These flat, light-colored spots did not 

appear to contain much material as they had no relief (depth); neighboring areas were free of these 

deposits. They did not appear to be caused by corrosion or oxidation, nor was the integrity or adherence 

of the nickel plating affected by them. 

Small grains of debris were thinly scattered over most of the cask floor as shown in Figure 2-14 

and Appendix F. The debris ranged from sandlike particles of submillimeter to several millimeter size to 

long slivers of material several millimeters in length. These generally appeared to have been deposited 

after the nickel plating of the surface, because the larger particles were dark in color and not reflective. 

Similar material had accumulated on the horizontal bars at the bottom of each fuel tube on which the fuel 

assemblies rested. Much of the sand-like debris probably consists of welding slag or grinding swarf from 

the basket. However, some of the debris appeared to consist of slivers of metal. The slivers may be slivers 

of stainless steel gouged from the fuel tube walls by insertion and extraction of the fuel assemblies, 

because the fuel tube walls exhibited much evidence of scraping by the fuel assemblies.

Cask bod 

wall

-7Z- -

Cask body 

floor

Minor 

granular 

debris on 

"floor (worst 

case)

Figure 2-14. Cask floor and wall below fuel tube C8-reflective surfaces free of corrosion.

29
-. -÷ •Cask body - .. > wall-nodular 

9 <- ~ ~ ~ ductile iron 

nickel plated. 

Figure 2-15. Cask wall below and between fuel tube A7 and A8-reflective surface free of corrosion

typical of cask wall condition without mineral stains. 

Samples of the particulate debris on the cask floor were retrieved with tape swabs. The locations of 

the samples are summarized in Table 2.1. 

The visual inspection, scanning electronic microscope (SEM), and energy dispersive spectrometry 

(EDS) results suggest that the visible grains are primarily steel slivers or steel oxide particles from the 

steel fuel basket. The slivers appeared to be the result of scraping or abrasion of the steel fuel tube walls 

by the fuel assemblies during past insertions and extractions. The walls of the fuel tubes have long, 

straight, deep axial (vertical) scratches that are consistent with abrasion by the fuel assembly components 

during insertion or removal. Moreover, the roots of those scratches glint in the camera lighting, indicating 

that the scratches are relatively unoxidized and not a characteristic of the mill finish of the plate steel. The 

larger oxide particles appeared to be residual pieces of welding slag that were dislodged from the basket; 

the smaller particles of oxide and steel are probably grinding residue dislodged from the basket. 

The radiation/chemical analysis of the crud samples were analyzed at the INTEC in November 

1999. Samples include those taken from fuel assemblies and the bottom of the cask interior. The test for 

"iron-phase" was completed and was not conclusive due to the insufficient amount of loose material for 

sample analysis. 

All of the samples show low levels of radioactivity, primarily due to the presence of Co-60, a 

component of crud. The other common radioisotopes associated with crud were below detection limits for 

the test. In addition, Cs-137, one of the primary gamma-emitting fission products, was also below 

detection limits. The gamma analyses suggest that the debris on the floor of the cask may contain some 

crud particles. Because the crud constituents other than Co-60 were below detection limits, the mass of 

crud contamination is small and probably consistent with the limited amount of crud that might have been 

scraped off the fuel assemblies during insertion into and extraction from the fuel tubes. 

2.7 Fuel Assembly Basket 

The accessible portions of the fuel assembly basket inside the cask were inspected visually, using 

the three remote video cameras positioned around the top rim of the cask and the pencil camera used to 

inspect the internal cask floor and exposed portion of the cask wall. The fuel basket was examined for

23
Table 2-1. Characterization of the particulate debris retrieved from the cask floor. Location of the samples 

were taken from the CASTOR V/21 cask for characterization. 

Sample Number Method of Samplinga Location of Sample Material (see Figure 2-16) 

"V/21-1 Tape pad on rod (1) Cask floor at fuel tube Al 

V/21-3 Tape pad on rod Cask floor at fuel tube C5 

"V/21-10 Tape pad on rod Cask floor at fuel tube D5 

V/21 -12 Tape pad on rod Cask floor at fuel tube D6 

V/21-14 Tape pad on rod Cask floor at fuel tube A4 

"V/21-15 Tape pad on rod Control, exposed to Hot Shop ambient environment and 

manipulated like samples 

V/21-16 Tape pad on rod Control, exposed to Hot Shop ambient environment and 

manipulated like samples 

"V/21-17 Tape pad on rod Cask floor between fuel tube C8 and barrel wall 

V/21-18 Tape pad on rod Cask floor between fuel tube D8 and barrel wall 

a- For retrieval of small eranular debris from the bottom of the cask, white nuclear grade duct tape was fastened to a flat foam 

pad mounted to the end of a 4.6 m (15 ft) steel rod

evidence of further corrosion of the plate surface, the welds and associated heat-affected zone, the 

junction between stainless steel and borated stainless, and contact points between the stainless steel 

structure and the zircaloy fuel assembly structure, such as on the steel brackets at the bottom of each fuel 

tube that support the weight of the fuel assemblies. In addition, the welds in the basket structure were 

inspected for failure, both for propagation of the cracks in the known broken welds (prior to long-term 

storage) and for initiation of new cracks in other welds that may have occurred during long-term storage. 

2.7.1 Basket Condition 

Only the surfaces of the basket directly accessible to the video and pencil cameras were inspected. 

The basket was inspected while in place within the cask. The extremely tight diametral clearance between 

the basket and cask wall (3 mm) prevented the unloading and extraction of the basket from the cask. The 

top surfaces of the basket were inspected by the three video cameras mounted around the top of the cask. 

With the fuel assemblies removed, the interior surfaces of the 21 fuel tubes (Appendix D) and eight 

ungussetted air channels were inspected with the pencil camera system. The interior surfaces of the flux 

traps and the triangular air spaces at the perimeter of the basket could not be inspected with the pencil 

camera. These air spaces were obstructed by welded spacers and gussets or were too narrow to permit 

insertion of the pencil camera. 

The fuel basket was in good condition, comparable with the surface condition noted in the 1985. 

(EPRI NP-4887). In fact, some of the images of the tops of the basket in the 1985 video tapes looked 

worse (more oxide scale) that in the 1999 inspections, an effect attributed to the difference in lighting 

conditions. The basket structure showed no evidence of corrosion beyond the mill surface finish and the 

heat tarnish in the heat-affected zones of the welds. The fabricator of the basket had left the mill surface 

finish on the steel plate components of the basket. No attempt had been made to remove the native oxide, 

stencils, construction layout marks, or environmental stains on the as-supplied steel stock. 

Therefore. most of the surfaces of the basket structure had a light-gray nonreflective surface as well 

as superficial oxide tarnish in the region of many of the welds. However, some of the interior surfaces of

24
the fuel tubes bore the marks of spot (rotary) surface grinding that 'skinned' the flat surfaces and original 

fabrication layout markings. These ground surfaces were still brightly reflective under the camera 

illumination and markings were still legible, indicating that neither significant air oxidation nor corrosion 

had occurred since the fabrication of the basket as shown in Appendix D (Figure D-20 and D-25). There 

was no evidence of corrosion due to incompatibility between the stainless steel and the borated steel, nor 

was there evidence of corrosion or degradation at the contact between the stainless steel 304 bottom 

nozzle of the fuel assemblies and the bottom support plates in the fuel tubes. No cracks or similar 

degradation was seen in the steel plate components, except for some of the welds as noted below. 

2.7.2 Basket Welds 

The 1985 inspection of the basket after the completion of the heat transfer performance tests 

identified eight broken welds in the top of the basket (EPRI NP-4887). The affected welds are identified 

as welds 1 through 8 in Figure 2-16. The welds cracked as a consequence of the stresses created by the 

differential thermal expansion of the tightly fitting basket within the cask during the tests (EPRI NP

4887). An objective of the 1999 inspections was to reexamine the affected welds for any changes in 

configuration and to examine other accessible welds in the basket structure for cracks or corrosion. 

Unfortunately, the stitch welds of the structure are located in the flux trap and spacer channels, not inside 

the 21 fuel tubes. Therefore, the only accessible welds were the welds visible at the top of the basket and 

a few others. The present condition of these welds are as shown in Appendix G. 

The eight known existing broken welds appeared to be the same as in the 1985 inspection. The four 

welds in the comers of the central fuel tube Al (cracks 4, 5, 6 and 7 in Figure 2-16) all involved welds of 

stainless steel to borated stainless steel, whereas the four welds joining the fuel tubes in clusters 

A8/A5/A6, B6/B5/B8, C8/C5/C6, and D6/D5/D8 (Cracks 1, 2, 3, and 8 in Figure 2-16) involved only 

stainless steel. The 1999 inspection confirmed that five welds were broken clear through, and three had 

substantial cracks that propagated partially through the top stitch weld. The narrowness of the flux traps 

and the supports within blocked the views of the stitch welds below the top welds from the top-side video 

cameras, and prevented the insertion of the pencil camera assembly. Therefore, the condition of those 

welds could not be determined. The top stitch welds throughout the top of the basket were inspected as 

well as the welds of the top-most struts within the flux traps. Except for the eight known cracked welds, 

the remaining welds appear to be in good condition and free of cracks. 

The stitch welds in the triangular air channels at the inside perimeter of the basket barrel 

(Figure 2-16, crack locations 9 through 16) were examined with the pencil camera system. The gusset

and strut-free channels were just large enough to permit insertion of the pencil camera for viewing the 

stitch welds attaching the fuel basket borated stainless steel partition plates to the stainless steel basket 

barrel. There were two stitch welds approximately three inches long located in each air channel, one at the 

top and one at the bottom. 

The inspections found that all eight of the bottom stitch welds were cracked and seven of the eight 

top welds. The weld free of cracks was the top weld at location 12 as shown in Figure 2-16. The top weld 

at location 13 was only partially cracked at the bottom (1/4 in. long) including several fine lateral cracks. 

A typical view of the remaining cracked stitch welds with full length longitudinal center line cracks are as 

shown in Figure 2-17 and 2-18. A view of all stitch welds examined in the triangular air channels are as 

shown in Appendix B. 

Additional cracks in basket welds (as shown in Figure 2-19.) were associated with stainless/borated 

steel junctions welds in fuel tubes.

25
0, I BORATED STAINLESS STEEL 

(TYPICAL FOR CRACK LOCATIONS 9-16)

BASKET BARREL (304 33) JOIT PE 2 T •/ V25 D8/ /V27 A'8 

JOINT TYPE JGAP 2 

270' T9 0131 T1E 02 /%/VO 05Al ý1T12 A2 T07 A3 X - o 

FUEL ASSEMBLY 10J/•9• FUEL TUBE 10 

O LOCATIONS OF CRACK INDICATIONS 

1 ~IDENiFiED IN 1885 

180" ( LOCATIONS OF CRACK INDICAllONS 

IDENTIFIED IN 1999 

1 FLLET WELDS EXAMINED IN 1999 

IN ADDITION TO EXISTING CRACK 

LOCATIONS IDENTIFIED IN 1980 

Figure 2-16. CASTQR-V/21 basket crack indication locations and fuel tube arrangement.

26
Cask Body Wall -\

Basket 

Barrel Top

Figure 2-17. Typical top stitch weld crack at location 14 identified in Figure 2-16.

Basket Barrel at Bottom

Fuel Basket 

Borated 

Stainless 

Steel 

Partition 

Plate

Cask Body 

Wall

Figure 2-18. Typical bottom stitch weld crack at location 14 identified in Figure 2-16.

27

Fuel Basket 

Borated 

Stainless 

Steel 

Partition 

Plate
Figure 2-19. Crack in basket weld below weld crack 4 (120 degree side) identified in Figure 2-13. 

The stitch welds in the triangular air channels appear to be nonstructural welds and were only 

intended to add additional stability for holding the fuel basket in place during 1985 fuel loading and 

vertical and horizontal thermal testing. The 1985 cask heat transfer performance testing (EPRI NP-4887) 

thermal analysis, concluded that the tightly fitted fuel basket expanded and came in contact with the 

basket barrel and the fuel barrel came in contact with the inner wall of the cask. The results of the thermal 

analysis indicated that thermal stresses in the weld joints at the eight crack locations identified in 1985 

were well above yield stresses measured during tensile testing of applicable weld samples. Based on these 

results and speculations. it appears that all new cracks identified in 1999 occurred during the 1985 

thermal testing. Further. it is concluded the cracks are not relevant to normal long-term storage and 

presents no adverse safety implications on the cask or components to perform their safety related 

function. 

2.8 Fuel Assemblies 

2.8.1 Purpose of the Inspection 

The inspection searched for visual evidence of change in the structure and integrity of the fuel 

assemblies after 14 years of dry storage. The primary characteristics examined were: changes in corrosion 

and crud deposits on nozzles, grid spacers, rod cladding; loose or lightly adherent corrosion product or 

crud: evidence of spallation or flaking; physical degradation or damage to nozzles, grid spacers, fuel rods; 

cracks, bowing of rods. or distortion. 

2.8.2 Method of Inspection and Lifting Force Measurements 

The visual inspection requires the removal of the fuel assembly from the basket (Figures 2-20 

and 2.21). 

Table 2- 2 provides the fuel assembly weight and the force required to start lifting each assembly 

out of the V/21 cask. The lifting force measurements indicate little sticking of the assemblies during 

removal, suggesting that no significant bowing of the assemblies or development of corrosion products 

causing adherence to the cask floor occurred. 

Once the assembly was lifted out of cask basket, the inspectors identified the assembly serial 

number and its orientation with respect to basket. They checked the relative uniformity of fuel rod lengths 

by clearance between tops of rods and top nozzle. then scanned the four sides using the three remotely 

operated cameras. Each scan permitted a clear-through view between the fuel rod array. After the visual 

inspection, the fuel assembly was returned to its original fuel tube in the cask, maintaining the original 

orientation.

28
Figure 2-20. Fuel assembly being lifted out of the basket for visual examination. 

Figure 2-2 1. Fuel assembly T 11 attached to lifting grapple for visual examination.

29
Table 2-2. Fuel assembly examination sequence and lift force measurements. 

Fuel Inspection Selected for Closer Grapple + Fuel wt Force to Start Lifting 

Assembly ID Sequence Examination (lb) (kg) (lb) (kg)

Vy5 

T03 

V27 

V04 

V14 

T07 

T12 

V08 

VII 

Vol 

T08 

V09 

V12 

V13 

T09 

T16 

V24 

V 15 

V25 

T13 

TI1

1 YES

2 

3 

4 

5 

6

7

8 

9

10 

11 

12 

13 

14 

15 

16 

17 

18 

19 

20 

21

YES

YES 

YES

YES

14 642.9 

141 645.9 

14 642.7 

14 642.7 

14 643.2 

14 642.7 

14 643.2 

14 641.8 

14( 639.5 

14 643.2 

14 640.9 

14 641.8 

14( 637.7 

14 644.1 

14 640.9 

14 642.3 

14 640.9 

14( 640.0 

14 642.3 

14 641.8 

14 640.9

1436 

1472 

1457 

1439 

1446 

1434 

1442 

1431 

1421 

1433 

1431 

1440 

1420 

1442 

1428 

1441 

1430 

1433 

1426 

1427 

1420

652.7 

669.1 

662.3 

654.1 

657.3 

651.8 

655.5 

650.5 

645.9 

651.4 

650.5 

654.5 

645.5 

655.5 

649.1 

655.0 

650.0 

651.4 

648.2 

648.6 

645.5

Once the fuel assembly was removed, the entire length of the four external surfaces were inspected 

by three remotely operated video cameras with zoom capabilities. Specification requirements for camera 

resolution was 2.6 mm (0.105 inches). maximum lower case character height, actual was demonstrated as 

better than 1.5 mm (0.060 inches). Appendix C shows an example of the condition of the assemblies. 

2.8.3 Fuel Rod Length Variation 

The uniformity of the gaps between the tops of the fuel rods and the top nozzle was evaluated for 

all fuel assemblies. This information was required to evaluate the potential for interference by the fuel 

rods in drilling the top nozzle for fuel rod extraction. In all of the fuel assemblies, the gap varied 

considerably, by as much as 0.5 inch (1.2 cm). 

The variations in rod length did not show any recognizable pattern, except that the rods within the 

array, near the center of the assembly, often appeared to be slightly longer on the average than the rods at 

the periphery of the array. In the outer peripheral rows, the rod length varied sharply from one rod to the

30
next, with no discernible pattern. The variation in the gap was previously ascribed to uneven fuel rod 

growth during irradiation (EPRI 4887). 

2.8.4 Fuel Rod Bowing 

The fuel rods in the fuel assemblies did not exhibit any detectable deformation except for slight 

localized bowing in the fuel rod sections between spacer grids. The extent of bowing was relatively small, 

amounting to less than a 50% deviation from the nominal 0.141 inch (3.58 mm) space between rods, 

based on a nominal rod pitch and diameter of 0.563 inches (14.30 mm) and 0.422 inches (10.72 mm), 

respectively. The views through the spaces between the fuel rods rows in the arrays indicated that the fuel 

rods within the arrays were also generally straight and without significant bowing or deformation. 

2.8.5 Cladding Failure 

The cask fill gas has been monitored for the last 14 years for gaseous fission product release, as 

evidence of cladding failure. No fission products have been found in the cask backfill gas samples (as 

presented in Section 4), indicating good integrity of the rod cladding. In the 1999 inspection, the surfaces 

of the rods were examined directly for cracks, cladding defects, general corrosion, and pitting corrosion as 

indications of established or incipient failure of the cladding. The inspections found no evidence of 

cracks, cladding damage or defects, general corrosion, pitting corrosion, or other indicators of established 

or incipient failure of the cladding. 

2.8.6 Crud Coating 

The oxide coatings on the visible rods were remarkably similar from rod to rod and assembly to 

assembly. The top nozzle, the exposed spacer grids surfaces, and the upper portions of the rods (above the 

third spacer grid) had a thin, uniform coating of light-colored adherent oxide (light gray to light tan/brown 

in the ambient lighting of the Hot Shop). Generally, below the third spacer grid (numbered from the top of 

the assembly), the oxide coating became increasingly mottled in appearance, with increasing areal 

fractions of darkgray oxide patches mixed with the light brown oxide. The gray mottling usually peaked 

on the rod surfaces near the fifth or sixth spacer grid (numbered from the top). The mottling was most 

pronounced on the rod surfaces just above the spacer grids, and often was the dominant oxide. 

The 1985 examinations ascribed the uniform light gray/brown oxide coating on the upper portion 

of the fuel assembly to crud deposits (EPRI 4887). Although characterization attempts on the crud were 

not successful, it is likely that the crud is composed of Fe30 4, and probably the NiFe3_,O4 spinel that is 

typical of PWR systems. The axial gradation of oxide uniformity (judged by oxide color), with greater 

color uniformity at the top of the assembly, suggests a slightly thicker accumulation of light tan/brown 

crud on the upper surfaces, with thinning of the crud in the lower sections. It is possible that the gray 

mottling that was usually prominent above the fifth and sixth spacer grids may represent an area of very 

thin crud deposition, that the dark gray mottled phase consists of exposed substoichiometic ZrO, (which, 

unlike the white stoichiometric ZrO2 , is dark gray or black in color). However, this identification is 

speculative, since the color of PWR crud is dependent on reactor conditions, and can vary from light 

colors to gray/black (Sandoval et al., 1988). It is highly unlikely that the mottling is due to spallation of a 

lighter-colored surface coating, with exposure of an underlying dark-colored oxide. Mottled oxide 

structures have been reported previously on PWR fuel (Sandoval et al., 1988). Furthermore, the mottled 

oxide structure was never associated with any visual evidence of spallation, flaking, or surface relief in 

these inspections, nor was there anyevidence of accumulated spalled material on the upper surface of the 

bottom nozzle, the upper surfaces of the fuel tube brackets on which the assemblies rested, or on the 

bottom floor of the cask. 

Most of the fuel rods in all of the assemblies exhibited two faint white rings of material just above 

and below the bottom weld that sealed the bottom end plug to the cladding. The rings were typically thin

31
and narrow, on the order of 0.05-0.1 inches wide. The most pronounced of these features is shown in 

Figure C-59 of Appendix C, which shows some rods with typical white rings, as well as one rod with one 

of the most pronounced oxide rings seen in the inspections (fuel assembly V-25, LM040H) Similar 

features were not seen on the upper ends of the rods, nor were these features observed on the individual 

rods that were extracted from fuel assembly T 11. This suggests that these features, if real, are very thin 

and their detection is sensitive to the lighting conditions. Since the two white rings correspond to the heat

affected zone of the bottom weld, they may consist of stoichiometric ZrO2 , which can form as a white, 

discontinuous and non-adherent oxide. The formation of the non-adherent stoichiometric oxide is usually 

associated with low levels (parts per million) of contaminants, such as nitrogen or carbon, in the zircaloy 

(Berry, 197 1). Because these features were so faint, they were probably present in the 1985 tests but not 

detected in the visual inspections, and were not considered significant in the 1999 inspections. 

The only difference between the 1985 and the 1999 inspections was one of color perception; the 

1985 examination reported the oxide as a 'typical orange/reddish crud' (EPRI 4887) whereas in 1999 the 

oxide appeared light tan/brown and light gray, depending on the viewing angle (i.e.. the lighting 

conditions). The uniform light-colored oxide coating on the upper surfaces of the fuel assemblies and the 

mottled coatings on the lower portions seen in 1999 were similar in appearance to the video-taped images 

acquired in the 1985 inspections. As in the 1985 inspections, the oxide coatings on the fuel assemblies 

were thin and very tightly adherent. There was no evidence of growth of oxide scale on the top or bottom 

nozzle, the grid spacers, and the cladding. The wear marks on the surfaces of the grid spacers (from 

abrasion during extraction of the fuel assembly) indicated that the oxide coating was still very thin and 

tightly adherent. The edges of the rods were crisply-defined, with no evidence of loosely adherent 

material. In addition, the upper surfaces of the bottom nozzle did not have any accumulations of spalled 

or loose material. There was no visual evidence that the dry storage period had caused a change in the 

oxide coatings, either with respect to thickness, oxide structure (surface patterns), or the adherent nature 

of the oxide. The tiny amount of crud that was collected was still tightly adherent after 14 years, and 

consisted of a few grains a few microns in size. Too little material was collected to obtain quantitative 

values for 60Co. 

2.8.7 Fuel Rod Removal and Examination 

At the conclusion of the examination of the fuel assemblies, fuel assembly T l lwas selected for 

extraction of 12 individual fuel rods for visual examination and shipment to the Argonne National 

Laboratory for nondestructive, destructive, and mechanical examinations. Fuel assembly T 11, located in 

fuel tube A4 (see Figure 2-16), was selected on the basis of maximum burnup and decay heating. The 

entire surface of each of the 12 individual fuel rods was visually inspected for the condition of the 

cladding including cracks, pitting, corrosion, bowing, imperfection, discoloration, stains and general 

surface deterioration. 

The tops of the fuel rods were accessed by cutting a 3 inch diameter hole through the top fuel 

assembly nozzle, centered on the fuel rod array, using a portable hole saw as shown in Figure E-1 of 

Appendix E. 

The fuel rods selected and removed are as shown in Figure 2-22. To withdraw the individual fuel 

rods, a rod puller grapple and load cell were attached to the top end of the rod. The rods were pulled from 

the assembly with the in-cell remote manipulators, taking care that the load limits were not exceeded. The 

applied pull force required to extract each rod from the assembly was approximately 35 to 45 pounds 

mass (15.9 to 20.5 kg) to dislodge each rod from the grid spacers: thereafter the applied force quickly 

dropped to 20 to 25 pounds (9.1 to 11.4 kg) for the duration the extraction. The fuel rods were pulled 

individually into a horizontal aluminum alloy channel tray for visual examination. After the visual 

examination, each fuel rod was transferred to the shipping container (Appendix E. Figure E-13)

32
The twelve fuel rods were free of physical damage, except for long, continuous longitudinal 

scratches in the cladding from initial installation into the fuel assembly (Figure 2-23, 2-24 and 

Appendix E). The longitudial scratches were old scratches, for the roots of the scratches were thinly 

coated with oxide (crud). Several of the rods had many (6-12) parallel longitudinal scratches running most 

of the rod length; none of the scratches had perforated the cladding and are not caused by creep failure. 

There was no evidence on any of the rods of extensive corrosion, pitting, or creep failure. 

While the cladding of all fuel rods exhibited numerous longitudinal, continuous scratches due to 

the initial installation into the fuel assembly, the extractions created only a few new scratches in the 

cladding surfaces. The new scratches did not break through the cladding. The relative ease of extraction, 

which caused less distress to the cladding than the original assembly operation, indicated that the cladding 

had not experienced any significant radial creep in any of the fuel rods, nor any significant reaction with 

the inconel spacers. 

Each rod was covered with a thin but uniform coating of oxide. The oxide coating seemed to 

comprise two layers, a light tan and a red-brown layer with sharp delineations between them, especially in 

the vicinity of the spacer grids (Figure 2-23 & Appendix E). These rods, drawn from within the bundle, 

appeared to have a more uniform oxide coating than the peripheral rods in the array. In general, the oxide 

layers appeared to be very thin and tenacious. On the upper portions of the rods, the incident lighting 

reflected diffusely from the thin oxide, imparting a pearly or vitreous satin sheen to the surfaces. 

The dark gray mottled oxide structure that was so characteristic in the lower portions of the 

peripheral rods in the array was limited to a few inches of rod length, generally only on areas covered by 

the fifth and sixth spacer grid. As with the peripheral rods, these rods also exhibited the very thin white 

bands of oxide above and below the weld on the bottom end caps of the rods (Appendix C, Figure C-13). 

The oxide layers were tightly adherent to the surface of the rods. The adhesion was tested with a 

steel wire brush and a steel spatula scraper. Neither tool was able to remove detectable amounts of oxide 

or to change the surface appearance of the rod. Only with application of considerable force and with 

considerable flexing of the spatula blade was the oxide layer broken and a small amount of oxide 

removed. 

All of the fuel rods exhibited small spots (<0.060 inches [ 1.6 mm] diameter), often occurring in 

sets of two, with similar sets of spots spaced 90 degree azimuthally around the rod. These features 

corresponded to the positions of the spacer grids and are assumed to be the contact points between the 

spacer grid springs and the cladding. These features were not pinholes or pitting corrosion, for several of 

these black spots gleamed with reflected light when oriented appropriately. The contact points were 

usually associated with the long longitudinal old scratches in the cladding surface and frequently were the 

termination of an old scratch. 

The assemblies were in a generally good condition, which had not changed since the 1985 

inspection. The general visual survey revealed a dark gray oxide layer under ambient cell lights, and light 

tan by video. The inspection found no increase in the oxide layer thickness. There was no formation of a 

loose oxide scale or particles between the fuel rods of the grid spacers or on the bottom nozzles. A general 

view of fuel assemblies is shown in Appendix C.

33
FUEL ASSEMBLY 

ID. NUMBER 

REFERENCE POINT A B 

(upper left 

hand comer 

on top nozzle

06

Fuel assembly 

serial number 

side

11 

12 

1'

LIP, A, ME X, PX i1f, INIX I 15 I* ir lip M I P, 50IN PI4 9 X X 9 W: 

5 1 A P5 W; W. 'i6 AL.AL.AW WV AN ýjl 

.4 . -1 $1 

ME 156i. 5 5 

L f-f, -, , r r - ,r .,L L U Ab, All At. A 

r*,, rf r 

Wriwv. Mr, "-I 

,4 f, rL L L L L L L. L

1-

C D E F G H I J K L M N O ; r r K' N rw r-'r r' rP()KJK)Y WANAM )%AA) ira W54155 L 5.K5.A1A54X4.A15A-A. )

IpC ccipJt'C tQcc Control rod guide tubes (20) 

- Fuel rod (2G4) 

(.4212-) 

Instrument 

thimble (1) 

Grid spacer 

Fuel rods (12 ea.) removed for 

examination and testing in 1999

- I 7 1 1 T I I 1 1 TNT N1 TN ¶1.J .- '" fA )¶ 

8.425"

Cross Section

Figure 2-22. Surry 15 x 15 PWR Fuel Assembly Ti 1. Serial number LM02JR.

34

I

I

D 1

T •

face)
.- -



Area housed by 

fuel assembly 

spacer grid

/

Area outside 

, spaced grid

Figure 2-23. Close up of Fuel rod removed from fuel assembly T 11 position 7H identified in Figure 2-22. 

-2 -~.

Figure 2-24. Fuel rod removed from fuel assembly T 1 position 81 identified in Figure 2-22.
2.9 Cask Temperatures

Storage systems must be designed to allow ready retrieval of spent fuel from the monitored 

retrievable storage installation (10 CFR Part 72). The spent fuel cladding must be protected against 

degradation by thermally activated processes by keeping the storage temperature down. Spent fuel storage 

or handling systems must be designed with a passive heat-removal capacity without active cooling 

systems. The conditions in the second dry cask storage period will be less severe than in the first storage 

period because the decay heat decreases with time. The decreasing decay heat requires less heat removal 

capacity during the extended licensing period. 

2.9.1 Internal Temperature 

The thermocouple lance system used in 1985 was no longer available for the tests in 1999. Thus, 

the temperatures inside the cask had to be measured with the lid off. Therefore, the internal temperature 

measurements obtained during the 1999 testing would only provide a general indication of temperatures 

inside the cask. 

Internal temperatures were recorded on September 29, 1999, between approximately 4 pm and 

5 pm. The cask lid was removed every workday morning at approximately 7:30-8 am, and by procedure, 

it was replaced nightly at the end of the day's activities (generally between 7 and 9:30 pm). When these 

temperature measurements started. the lid had been off the cask for approximately 8 hours. 

The bolts were removed from the primary lid on September 7, 1999. and the primary lid was first 

removed on September 8, 1999. Between September 8 and 29, the lid had been open generally 5 days per 

week and about 10 hours per day, with at least 8 hours of convective cooling on that day. The gradual 

convective cooling was achieved during the working days prior to that measurement. Therefore, the 

contents would have cooled considerably. 

The temperature was measured with a Type J thermocouple inserted into the control rod guide 

tube. It was expected that the upper portions of the fuel assembly would exhibit the highest temperatures 

due to convection. To approximate the best position, the temperature in V05 (fuel tube Al, Figure 2-16) 

were quickly measured at three positions: 

o After 12 minutes, 0.6 m (2 ft) below the top nozzle, 152.10 C (305.8°F) 

o After 5 minutes. 1.5 m (5 ft) below the top nozzle, <140'C (284°F) 

o After 10 minutes, 0.3 m (1 ft) below the top nozzle, 146.5°C (295.7°F). 

Within 10 minutes or so, the temperatures equilibrated to within 0. 10C/min (0.2°F/min) rise; the 

temperature readings were recorded for at least the last 5 minutes of equilibration. The final readings at 

10 minutes (12 minute for VQ5). measured approximately 0.6 m (2 ft) beneath the top nozzle, were as 

follows in Table 2-3. 

Readings at 1-min intervals indicated that the equilibration was close to completion at 10 min; the 

final true equilibration temperature might be at most 2- 30C (5 degrees Fahrenheit) higher.

36
Table 2-3. Internal temperature. 

Assembly ID Fuel Tube ID Temperature 

V05 Al (center of basket) 152. 1°C (305.80F) 

T 11 A4 (between center and outer tubes) 154.6°C (310.30F) 

T03 A7 (outer tube, at 0° mark on cask) 122.80 C (253.00$) 

The hottest zone in the hottest of the three measured assemblies was 154°C (309'F) at the end of 

10 minutes equilibration, when the rate of rise was still 0.10C/min (0.2 °F/min). A plot of the data 

indicated that the temperature would eventually equilibrate between 155 and 160'C (311 and 320'F). 

It must be emphasized that these results pertain only to the conditions at the time of measurement 

and represent an estimate of the maximum temperature of the assembly. It was considered as satisfactory 

that the air temperatures were well below 200'C (3920F) to allow the use of aluminum for the laydown 

fixture used to transport the fuel assembly from the hot shop to the hot cell. More rigorous measurements 

of the actual rod surface temperatures to get a better assessment of the impact on using aluminum for the 

laydown fixture were not executed. It was assumed that fuel assembly Ti1 with its high burn up was 

thermally the hottest or at least representative of the hottest assemblies It is nevertheless true that with the 

cask lid in place, the final equilibrium temperature will be higher than the value recorded and reported. 

2.9.2 External Surface Temperature 

Measurements of the external surface temperature of the cask were taken by portable pyrometry. 

The temperatures were recorded on August 2, 1999, at approximately 4:30 pm. The results are 

summarized in Table 2-4 along with the 1985 measurements. 

As expected for the reduced decay heat, the surface temperatures of the cask were substantially 

cooler (by approximately 10'C [50'F]) in 1999 than in 1985. In addition, the 1999 readings qualitatively 

track the 1985 readings quite well. The 1999 data show less scatter in the various azimuthal readings at a 

given elevation and describe a flatter temperature profile than the 1985 data. These data are summarized 

in Figure 2-25.

Table 2-4. External temperature data for CASTOR V/21 Cask. 

Temperature 1999 

No. Elevationa (mm) @ Azimuth"b F 0 C 

1 175 @0° 103.9 39.9 

2 460@3530 105.2 40.7 

3 1148@00 107.7 42.1 

4 1130@00 110.9 43.8 

5 2348@00 118.9 48.3 

6 2330@00 113.6 45.3 

7 3148@0O 119.3 48.5 

8 3130@00 115.4 46.3 

9 3898@0° 117.9 48.8 

10 4380@0° 116.4 46.9 

11 4621 @0° 115.1 46.2 

12 1148@450 110.6 43.7 

13 2348@450 117.5 47.5 

14 3898@450 117.7 47.6

Temperature 1985 

OF °C Location

148.2 

154.9 

167-9 

161.0 

178.8 

166.6 

179.0 

166.4 

172.5 

169.7 

169.8 

161.7 

172.7 

174.5

64.6 

68.3 

75.5 

71.7 

81.6 

74.8 

81.7 

74.7 

78.1 

76.5 

76.6 

72.1 

78.2 

79.2

Cask sided" 

Cask sided'e 

Cask sided'e 

Cask side d,, 

Cask sided,, 

Cask sided,, 

Cask sidede 

Cask sided,, 

Cask sided'e 

Cask side d,, 

Cask side d,, 

Cask side d'e 

Cask side d*, 

Cask side d•e

37
"Table 2-4. (continued). 

Temperature 1999 Temperature 1985 

No. Elevationa (mm) @ Azimuthb OF 0C OF 0C Location

15 1148@ 1800 

16 2348 @ 1800 

17 2330@1800 

18 3880@1800 

19 2348@2250 

20 3898@2250 

21 1148@270c 

22 2348C@270° 

23 3898@2700 

RadiusC (mm) @ azimuth5 

24 487@0` 

25 1100@0° 

26 487@90° 

27 1100@90° 

28 Center 

29 1100@ 180 

Radiusý (mm) @ azimuth5 

30 487@0c 

31 1100@00 

32 487 C90° 

33 1100@90 

34 Center 

35 1100@ 1800 

36 Ambient Temperature

111.4 

118.5 

114.0 

120.3 

117.8 

118.5 

112.3 

119.2 

117.4 

131.3 

113.0 

133.9 

115.6 

141.0 

113.9

44.1 

48.1 

45.6 

49.1 

47.7 

48.1 

44.6 

48.4 

47.4 

55.2 

45.0 

56.6 

46.4 

60.6 

45.5

168.4 

184.2 

175.1 

178.8 

173.3 

199.9 

160.1 

178.5 

174.2 

207.5 

165.3 

217.2 

166.1 

221.5 

109.5

- 76.8 

- 79.7 

91 32.9 82.7TF

75.8 

84.6 

79.5 

81.6 

78.5 

93.3 

71.2 

81.4 

79.0 

97.5 

74.1 

102.9 

74.5 

105.3 

43.1

Cask sided'e 

Cask side de 

Cask sidede 

Cask sided'e 

Cask sided,, 

Cask sided, 

Cask sided,, 

Cask sided.e 

Cask sided-e 

Cask topsde 

Cask topd6 e 

Cask topd,, 

Cask topd.C 

Cask topde 

Cask topdt,

24.9 Cask bottom d" 

26.5 Cask bottomd,, 

- Cask bottomde 

-- Cask bottomdC 

- Cask bottomd,, 

-- Cask bottomd*, 

28.2

a. From exterior bottom of cask. 

b. From 0' orientation mark. Figure 2-16. 

c. From cask centerline. 

d. Cask orientation vertical with helium backfill. 

e. Exterior cask surface

38
250

200 

* * * 9 o1985,00 "LL . 1985, 45

0150 A1985, 1800 

+ 1985, 225' 

C- )K 1985, 270' 

01999,00 

U o • [31999, 450 

:100 A 1999, 1800 

u, • 1999, 2250 

K 1999, 2700 

50 

0 I r I I 

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 

Bevatlon of Weading, mm

Figure 2-25. Surface temperatures of the CASTOR V/21 Cask.
3. RADIATION SURVEY 

3.1 Objectives 

Radiation shielding and confinement features that are sufficient to meet all necessary requirements 

of 10 CFR Part 72 must be maintained as long as the spent fuel is to be stored. Because the radioactive 

and thermal source terms are decreasing, a dry storage cask that met the requirements for the first license 

period will also meet those for the second license period provided that material alterations have not led to 

unexpected behavior or decreased capabilities of factors important to safety. 

A radiation survey was conducted in 2001 to evaluate the performance of the cask shielding for 

neutron and gamma radiation at the sidewall, top and bottom surfaces of the cask. The dose rate readings 

were compared to the 1985 measurements to evaluate the performance of the cask shielding. 

3.2 Methods of Measurement 

In the 1985 EPRI study (EPRI NP-4887), gamma doses were measured at 69 locations on the 

surface of the cask with lithium fluoride (LiF) thermoluminescent dosimeters (TLDs). Neutron doses 

were measured with track etch dosimeters (TEDs), which consisted of allyl diglycol polycarbonate 

containing Cr-39 and were fabricated specifically for this test. Average dose rate values were generated 

by normalizing to exposure time. 

Portable hand-held survey instruments were also used in the 1985 surveys to measure both gamma 

and neutron dose rates at some of the same locations as the TLD and TED readings. However, the 

coverage with portable instruments was not as complete as with TLD and TED measurements. The 

gamma dose rate measurements were made with two ion chamber instruments, an Eberline RO-3A 

(INEL) and an Eberline RO-3B (PNL). The 1985 neutron dose rate measurements were made using an 

Eberline PRN-4 and a SNOOPY instrument. Both instruments had BF 3 detectors with polyethylene 

moderators: the Eberline PRN-4 had a 9-inch diameter moderator sphere (INEL) and the SNOOPY had 

an 8.5-inch moderator cylinder (PNL) (EPRI NP-4887). The results of these measurements were 

presented in detail in EPRI NP-4887 and summarized in Appendix A. 

Similar measurements were obtained in May 2001 at the same locations as the 1985 readings, as 

detailed in Appendix H. The 1985 and 2001 readings were obtained on the side wall at intervals on two 

vertical lines spaced at the 45 and 90-degree azimuths with respect to the 0-degree cask reference, as well 

as on the 45 and 90-degree radius of the lid and bottom of the cask. The locations of the readings on the 

cask are shown in Figures 3-1 and 3-2. An additional 39 gamma dose rate measurements were made in 

2001 to determine the uniformity of the shielding around the circumference of the cask at several 

elevations. These additional readings were obtained at 30-degree increments circumferentially at the 

4400 mm, 4298 mrm, and 2701 mm elevations, as measured from the bottom of the cask. The locations of 

the 2001 measurements are detailed in Appendix H. and the results are given in Appendix A. 

The 2001 readings were made with environmental TLDs and an Eberline RO-20 for the gamma 

dose rates, and an Eberline E-600/NRD digital portable meter for the neutron dose rates. The 

environmental TLDs consist of LiF and a magnesium copper phosphate phospor. The Eberline RO-20 is 

an air-filled ion chamber instrument, comparable to the RO-3A used in 1985. The RO-20 was calibrated 

against a Cs-137 source to an accuracy of ±5% (at 2 standard deviations); the accuracy in field operation 

is considered to be no better than ±10% and is more likely +20%. The digital Eberline E-600 meter was 

fitted with an Eberline Neutron rem Detector (NRD). which consists of a BF 3 detector tube and a 9-inch 

polyethylene sphere. The accuracy was calibrated against a Cf-252 source to within +10% (at 2 standard 

deviations): the accuracy in field operation is considered to be no better than +30%. The neutron response 

of Eberline NRD decreases nonlinearly with decreasing neutron energy, and is approximately

40
CO

LO (7 0%

900 900 900. 900 900 900 

450 450 450 450 450 450 

600 

750 

1050 

1200

(Y)

0 

IO 

co Cv)

00000 

00 00

900 900 900 

450 450 450

Figure 3-1. Vertical location of the gamma and neutron dose rate readings on the sidewall of the 

CASTOR V/21 cask. Locations of the readings at the 45 and 90-degree azimuths are given as elevation 

(in millimeters) from the bottom of the cask. The bottom of the cask in the figure is on the left. 

00 00 

450 450 

2700- - 9 2700. 1" 

k'~\\34 1-51 -0484 

V3 I 

1800 1800 

Cask Top Cask Bottom 

Figure 3-2. Radial locations of the gamma and neutron dose rate readings on the top and bottom surfaces 

of the cask. Locations of the readings are given as radial displacement (in millimeters) from the center on 

the 45 and 90-degree azimuths.

41

E 

0 

0-U,

50mm 

Spacing

900 

450

01-GAOO183-01
3 rem/lncm- at 150 keV (the approximate energy of the unmoderated neutrons of the spent fuel 

[EPRI NP-4887]) and 0.7 rem/n/cm2 for thermal neutrons. The sensitivity of the neutron instrumentation 

used for the 1985 measurements was not included in the EPRI NP-4887 report and is unknown at this 

time. 

The 2001 neutron dose measurements were not attempted with TEDs, because the 1985 TED 

configuration could not be duplicated at the INEEL. An attempt was made to measure the neutron dose 

with Cr39 TEDs supplied by ICN Worldwide Dosimetry Service, Inc. However, these TEDs significantly 

underestimated the neutron dose equivalent because the ICN TED was not sensitive to neutron energies 

less than 250 keV. the range of neutron energies measured for the cask in 1985. (EPRI N-P-4887) 

For the sidewall contact readings, the TLD dosimeters were held by an external frame at the 

appropriate elevations, with the TLDs in contact with the cask cooling fins (i.e., positioned between the 

frame and the cask wail). The TLD dosimeters for the cask bottom were placed in precut cavities in two 

thicknesses of plywood placed under the cask. The dosimeters for the top of the cask were suspended 

from a frame above the primary lid; as in the 1985 measurements, the secondary lid was not in place 

when the measurements were made in 2001. Unfortunately, high winds dislodged some of the suspended 

TLDs during the course of the exposure, dropping the dosimeters onto the primary lid and scrambling 

their sequence. Although they were reattached to the frame 12 to 15 hours later (after discovery of the 

situation), they yielded inconsistent results. Therefore, the analyses of dose rates on the cask top relied 

only on the RO-20 readings. Except for the readings on the cask top, the TLD and RO-20 gamma 

readings compared favorably. 

3.3 Observations 

In general, the gamma dose rates in 2001 were substantially lower than the 1985 measurements, as 

was expected from decay considerations. Peak sidewall gamma dose readings were less than 12 mR/hr, 

substantially less than the peak gamma rate of 140 mR/hr that was observed in 1985. The peak readings 

on the cask sidewall occurred at locations corresponding to the position of the top and bottom flow 

nozzles of the fuel assemblies within the cask. The gamma dose rates for the active fuel region were less 

than 2 mR/hr. The gamma dose rates for the top and bottom surfaces of the cask peaked in the center, at 

approximately 10 and 7 mR/hr. respectively. 

Within the uncertainties of the measurements and differences in the responses of the 1985 and 2001 

instruments, the 1985 and 2001 neutron dose measurements resulted in reasonably similar readings, with 

sidewall readings peaking at approximately 15 mren/hr. The sidewall neutron dose rate peaks occurred at 

the same sidewall elevation in the 2001 and 1985 measurements and corresponded to the top of the 

polyethylene rods in the cask wall. The neutron dose rates at the top and bottom surfaces peaked at the 

center at approximately 45 and 55 mreml/hr, respectively. Within the uncertainty in the measurement, the 

2001 neutron dose rate readings were comparable to the 1985 readings. 

The gamma and neutron dose rate profiles showed no evidence of failure or degradation of the 

shielding. 

3.3.1 Gamma Contributors 

In 1985, the gamma energy spectrum of the radiation from the cask sidewall, top, and bottom 

surfaces was analyzed by germanium spectrometry. Spectral analysis identified the source radionuclides 

for the gamma radiation at the cask surfaces as 0

oCo. l•Ce/Pr. I-'7Cs, 14Cs, and I54]Eu. The relative 

intensity distribution of the various gammas from these radionuclides was not given. However, from the 

gamma spectra. ' 0Co and >"Ce/Pr were identified as the primary components to the radiation at the top 

and bottom of the cask, and "Ce/FPr as the primary component on the sidewall. The dominance of 60Co at 

the top and bottom was due to the activation of the stainless steel flow nozzles on the ends of the fuel

42
assemblies. The remaining radionuclides, including the '"4Ce/Pr, are fission products present in the spent 

fuel.(EPRI NP-4887) 

Because of the decay time and the relatively short half-lives of some of these radionuclides, the 

spectrum of contributors to the 2001 dose rate readings was expected to shift significantly from the 1985 

spectrum, and the 2001 dose rate readings were expected to be substantially lower than the 1985 readings. 

Table 3-1 summarizes the decay-normalized gamma intensity of these radionuclides in 2001, assuming a 

16-year decay time. The decay adjustments indicate that the primary contributors in 1985 to the gamma 

dose rate at the external surfaces of the cask were substantially reduced by 2001. The gamma contribution 

from 6°Co in 2001 is expected to be only approximately 10% of the 1985 levels, and the '44Ce/'44pr has 

effectively decayed away and no longer contributes to the dose rate at the external surfaces of the cask. 

Table 3-1. Decay-normalized gamma intensity from radionuclides identified as major contributors to the 

gamma dose rate in 1985, assuming a 16-year decay time. 

Decay-Adjusted Normalized 

Radionuclide Half-lifea Intensity in 2001 

60Co 5.271 y 0.12 

144Ce/44 Pr 284.6 d/17.28 m 6.7 x 10-7 

1

37Cs 30.17 y 0.69 

134CS 2.065 y 4.7 x i0-3 

1

54Eu 8.59 y 0.28 

a- F. W. Walker, J. R. Parrington, and F. Feiner, Nuclides and Isotopes, Fourteenth Edition, General Electric Company, San Jose, 

CA, 1989. 

3.3.2 Gamma Dose Rates on the Sides of the Cask 

The gamma dose rates were acquired in 2001 at the same locations as the 1985 readings. The 

gamma results for 2001 and 1985 are summarized in Appendix A. As shown graphically in Figure 3-3, 

the 2001 gamma dose rate profiles for the 45 and 90-degree azimuths on the sidewall are qualitatively 

similar to the corresponding 1985 profiles, and show distinct gamma dose rate peaks near the bottom 

(480 mm elevation) and top areas (=4400 mm elevation) of the side wall. The peaks correspond to the 

locations of the bottom and top flow nozzles of the PWR assemblies within the cask and are presumed to 

be due to 60

Co in the steel flow nozzles EPRI NIP-4887, GNS Drawing 49001-E-01, GNS CASTOR SAR 

Figures 7.3-5 and 7.3-6). 

The peak gamma dose rate (TLD) values decreased more than 90% from the 1985 values in 

agreement with the above decay-based projection. At the 480-mm elevation at the 45 and 90-degree 

azimuths, the peak dose rate dropped from 140 and 125 mR/hr in 1985 to 4.4 and 8.1 mR/hr, respectively, 

in 2001. At the 4400-mm elevation, the peak dose rates decreased from 118 and 134 (for the 45 and 

90-degree azimuths) in 1985 to 8 and 11.1 mR/hr, respectively, in 2001. Similarly, the readings in the 

active fuel zone, corresponding to elevations of 1048 through 3850 mm, decreased approximately 95%, 

from 36.0 mR/hr (1 cy = 4.05) and 30.5 mR/hr (1 (T = 19.4) in 1985 to 1.32 (1 c7 = 0.19) and 2.25 mR/hr 

(1 a = 1. 17) for the for the 45 and 90-degree azimuths, respectively, in 2001.

43
0 

A 

00

0

A A A

0 0 0

A 

0

A

A 

0 

A A

140 

120 

100 

80

* a *

1000 2000 3000

Elevation from Bottom of Cask, mm

Figure 3-3. Comparison of the 200)1 and 1985 gamma dose rates on the cask sidewall at the 45 and 90-degree azimuths.

160 1

A 

0

A 2001,450 azimuth] 

02001, 90' azimuth 

A 1985, 450 azimuth 

0 1985, 900 azimuth

60

E 

0 

E 

E 

4ý. CU

40

20 2

A 

0

a I

0

0 4000 5000 6000

SLO

ia gh i6 i
The 1985 gamma data in Figure 3-3 show a slight difference in the gamma profiles for the 45 and 

90-dgree azimuths, primarily in the shoulder below the peak at 4400 mm in the 90-degree profile. The 

shoulder in the 90-degree profile was reproduced in the 2001 gamma dose rate measurements, as shown 

in Figure 3-4. The shoulder at 3850-4400-mm elevation of the 90-degree profile may be a consequence of 

the decreased shielding due to the flat milled seat at that location for the upper, bolt-on trunnion. As can 

be seen from the error bars in Figure 3-4, the dose rate peaks at each end show some differences in 

intensity between the two azimuths that may be statistically different, with slightly higher readings at the 

90-degree azimuth, and is probably due to the proximity of three fuel assemblies to the cask wall at this 

azimuth. There is no significant difference in the two azimuthal sets of readings in the active fuel zone of 

the cask (1048 through 3850 mm). 

In 2001, 39 additional gamma dose readings were acquired at three sidewall elevations around the 

full circumference of the cask to determine the circumferential and axial uniformity of the cask shielding. 

These readings were taken at 30-degree intervals at the 4400, 4298, and 2701-mm elevations. The full 

circumference readings are summarized in Appendix A and in Figure 3-5, along with 45-120-degree 

sector readings at other elevations. The gamma data show a distinct but short-range axial non-uniformity 

in dose rates comparable to the profiles shown in Figures 3-3 and 3-4. The highest gamma dose rate 

readings occurred at 4400-mm elevation, with readings that varied circumferentially from approximately 

10 to 25 mR/hr. These high readings are due to the 6Co in the upper flow nozzle of the fuel assemblies. 

The readings at 4298-mm and 2701-mm elevation are substantially lower, generally less than 2 mR/hr. 

However, the readings at the 4400 and 4298-mm elevations exhibited some azimuthal variations. The set 

of readings at 4400 mm exhibited peaks at 0, 90, 180, and 270-degree azimuth, while the readings at 

4298 mm exhibited two peaks at 90 and 270-degree azimuth. These readings correspond to locations with 

three fuel assemblies in close proximity to the cask wall. 

3.3.3 Neutron Dose Rates on the Sides of the Cask 

The neutron dose rate profiles for the cask sidewall in 1985 and 2001 are shown in Figure 3-6. The 

neutron dose rate data for both 1985 and 2001 show considerable scatter due to the measurement 

uncertainty of ±30%. Because of the extremely long decay half-life of the fissile radionuclides in the fuel, 

the neutron dose rates for 1985 and 2001 should be comparable within the uncertainty of the 

measurements. The maximum neutron dose rates in 1985 and 2001 are comparable at approximately 

16 mrem/hr. However, as indicated by Figure 3-6 and the summary in Table 3-2, much of the 2001 

neutron dose rate data for the active fuel zone are approximately half the 1985 values. It is unlikely that 

the lower values in 2001 pertain to changes in shielding; the difference most likely is due to the use of 

two different methods for neutron dosimetry. The lower readings in 2001 were obtained with the Eberline 

E-600/NRD and suggest a lower response sensitivity to thermalized neutrons than the TED used in 1985. 

This possibility is suggested by the shape of the 2001 profile in Figure 3-5 in which the E-600/NRD 

readings are lower than the TED in the polyethylene-shielded fuel zone, yet comparable to the TED 

readings at elevations above 4400 mm, above the polyethylene rods within the cask wall. The 

manufacturer's response curve shows a decreasing sensitivity with decreasing neutron energy, from 

approximately 3 rem/n/cmr2 at 150 keV to 0.7 rem/n/cm2 at I eV (Eberline). Depending on the extent of 

thermalization of the neutrons by the polyethylene shield rods, the E-600/NRD response for the neutrons 

from polyethylene-shielded zone of the side wall could easily be only 50% of the response for the less 

moderated neutrons above the polyethylene shield. 

The 2001 data in Figure 3-6 do not show any increase in the neutron dose rate readings for sidewall 

elevations below 4400 mm, which corresponds to the top of the polyethylene shielding rods. This 

indicates that major degradation or slumping of the polyethylene shielding rods has not occurred.

45
14 

12 

10

E 

(I) 

0 

0

8 

6

ON

4 

2 

0

0

error bars: +/- 20% I

{

i

1000

i

2000 3000

I

Al

4000 5000

Agamma, 450 

0 gamma, 900]

6000

Elevation from Bottom of Cask, mm 

Figure 3-4, Gamma dose rates in 2001 as a function of elevation for the 45 and 90-degree azimuths on the sidewall of the cask. The dose rate profiles 

for the two azimuths appear to differ slightly at the ends of the cask.
N

I&

M

M

0

N

E

0 M E

M

A

A

50 100

A

a A& A

150 200

A

250 300

*4400 mm 

A4298 mm 

A4100 mm 

[33370 mm 

*2701 mm 

02443 mm

350

Azimuth from 00 Cask Reference, degrees 

Figure 3-5. Circumferential profiles of gamma dose rates on the cask sidewall, acquired in 2001 as a function of azimuth and elevation. Readings were 

taken at the 4400, 4298, and 2701-mm elevations in 30-degree increments around the entire cask circumference,

30 

25 

20

15

E 

"5' 

cr 

(/) 

0

10 

5 

0 

0

I
25

A 

20 

E 15 0 

"•A A2001, 450 azimuth 

E 0 40A 02001,900 azimuth 

"a Q A A1985, 450 azimuth 

A A 0 1985, 90 azimuth rrA A o g ... 

S1OA 

o 0 

OoA A • 

A A 0 

0 0 

A 

5 0' 

00 

0 1000 2000 3000 4000 5000 6000 

Elevation from Bottom of Cask, mm

Figure 3-6, Comparison of the neutron dose rates measured in 1985 and 2001 on the cask sidewall at the 45 and 90-degree azimuths.
Table 3-2. Maximum and average neutron dose rates for the 1985 and 2001 cask sidewall readings. 

1985 Neutron Dose Rates 2001 Neutron Dose Rates 

450 Azimuth 900 Azimuth 450 Azimuth 90' Azimuth 

Maximum Reading, mrem/hr 15 16 14 16 

Average of Full Profile (1 a), mrem/hr 7.83 7.96 6.56 7.06 

(5.17) (4.95) (3.11) (4.05) 

Average over Fuel Zone (1 a), mrem/hr 10.18 12.42 5.5 6.0 

(1.77) (2.85) (1.38) (1.90) 

In 2001, 39 additional neutron dose readings were acquired at three sidewall elevations around the 

full circumference of the cask to determine the circumferential and axial uniformity of the cask shielding. 

The readings were obtained at the same locations as the circumferential gamma readings discussed above. 

The results for the neutron dosimetry are given in Appendix A and in Figure 3-7. The average 

circumferential readings at 4400, 4298, and 2701 mm are 11.3 mrem/hr (ay = 2.8 mrem/hr), 9.9 mrem/hr 

(a = 1.8 mremrhr), and 6.5 mrem/br (a = 1.6 mrem/hr) respectively. The readings at 4400 and 4298-mm 

elevation appear to be statistically identical and are in good agreement with the neutron dose rate data 

generated in 1985 as shown in Figure 3-8. The lower average at 2701 mm is most likely a function of the 

response curve of the NRD unit. 

At each elevation, the circumferential readings show substantial scatter that is consistent with the 

uncertainty in the measurements (±30%). While the circumferential readings at 4298 and 2701-mm 

elevation appear to have similar profiles, the similarity is most likely a statistical artifact. There is no 

obvious trend in the circumferential readings that would indicate problems with the uniformity or 

integrity of the shielding. The range of neutron dose rates in these tests is consistent with the other 

neutron dose rate values obtained 2001 and 1985. 

In 1985, neutron dose rate readings were made at 1-degree intervals from 72 to 77-degree azimuth 

(with respect to the 0-degree cask reference) at the 2149-mm elevation of the sidewall of the cask. The 

purpose of the original readings was to test for local nonuniformity of shielding due to the circumferential 

spacing of the cylindrical polyethylene shielding rods in the cask wall. At the time, no significant 

nonuniformity in neutron dose rate was observed. The test was repeated in 2001; the results are 

summarized in Figure 3-9 along with the 1985 results. Again, no significant nonuniformity was found in 

2001. The measured neutron dose rates varied from 5 to 7 mrem/hr. However, as shown in Figure 3-9, the 

azimuthal profile of the 2001 neutron dose readings appears to duplicate the minor variations in the 1985 

gamma and neutron dose rate profiles. The data do indicate that the stacking configuration of the 

polyethylene rods at this location has not changed between 1985 and 2001, and the effectiveness of the 

shielding has not diminished. 

3.3.4 Gamma Dose Rates on the Top and Bottom Surfaces of the Cask 

The gamma dose rate data for the top and bottom surfaces of the cask are summarized in 

Appendix A. Figures 3-10 and 3-11 graphically compare the gamma dose rate readings for the 45 and 

90-degree azimuths in both the 1985 and the 2001 tests. The 2001 gamma dose rate values in the fuel 

region are approximately 15-20% of the corresponding 1985 values, consistent with 60Co decay 

considerations. (Both the 1985 and 2001 measurements were taken with the secondary lid removed from 

the cask.) As shown by Figures 3-10 and 3-11, the dose rate profiles are fairly flat, with no spikes. The 

data for the 45 and 90-degree azimuths for the top and bottom are in very good agreement, within the 

uncertainty of the measurement.

49
UN / "N-

/ \ 

N / 

""/"

/ 

/ 

N N

/

/

/

3!

/

/

/

50 100 150 200 250 300

-O-4400 mm 

-- M-- 4298 mm 

[ --. - 2701 mm 

350

Azimuth from 00 Cask Reference, degrees 

Figure 3-7. Circumferential profiles of' ncutron dose rates on the cask sidewall, acquired in 2001 as a function of azimuth and elevation. Readings were 

taken at thc 4400, 4298, and 2701 -mm elevations in 30-degree increments around the entire cask circumference.

16

14 

12

10 

8 

6

L-. E 

E 

0 

0 0

4 

2 

0

0
18

16 

14 

12

E 10 0 A A2001, 4400 mm 

E 02001, 2701 mm 

A1985, 4400 mm 

x A O 01985,2701 mm 

4) 

Cl, 0 0 0 

6 0 0 0 0 0 

0 0 0 

4 0 

2 

0

0 50 100 150 200 250 300 350 

Azimuth from 00 Cask Reference, degrees 

Figure 3-8, Comparison of the 1985 and 2001 neutron dose rates as a function of azimuth and elevation. Although the 1985 neutron dosimetry is 

limited to the 45-120-degree sector, the 2001 circumferential data compare favorably.
[]

0

11

0

13

M

v

0

Figure 3-9. Local variations is shielding due to the stacking configuration of 

data for gamma and neutron radiation.

0 1985, gamma 

[ 1985, neutron 

02001, gamma 

U 2001, neutron

78

the cylindrical polyethylene rods. Plot compares the 1985 and the 2001

0

0

30

25

E 

cc 

0

0

0

0

0

U

5 

0 

71

U

0

0

72 73 74

03 

0

75 77

Azimuthal Location from 00 Reference, degrees

76

,*%t%

Q•q i U
0

03

0

50 

40 

E 

+ 30 

0 

20

0 

0 

0 

0 

] 0 

wE

0 200 400 600 800 

Location from Center of Lid, mm

1000

02001,450 

02001, 90

01985, 450 

01985, 90-

1200

Figure 3-10. Radial gamma dose rate profile on the top surface of the cask. The plot compares the 2001 and 1985 readings at the 45 and 90-degree 

azimuths.

60

03 03 

0 

S 

[]

10 

0

(-A Uj
30 

25

0

0 0

01

0

20 

E 

S15 

0 

0 

10

5

0

0

0 0

a a M U

03

U 0

[]

0

S 

U]

0 1 1 - I_ 

0 200 400 600 800 

Location from Center of Lid, mm 

Figure 3--l 1. Radial gamma dose rate profile on the bottom surface of the cask. The plot compares the 2001 

azimuths.

U

1000

02001,45' 

02001, 900 

01985, 450 

011985, 900

1200

and 1985 readings at the 45 and 90-degree

4 I
Figures 3-12 and 3-13 compare the 2001 and 1985 neutron dose rate data for the top and bottom 

surfaces of the cask. As shown in Figure 3-12, the 2001 and 1985 neutron dose rates for the cask top are 

identical both in profile shape and in magnitude. As in the 1985 measurements, the neutron dose rate 

peaks near the center of the cask lid at approximately 45 mrem/hr. 

Figure 3-13 compares the neutron dose rate data for the bottom of the cask from the 2001 and 1985 

tests. Unlike the top surface, the 2001 neutron dose rates are consistently higher than the corresponding 

1985 data. Within the radius of the fueled area, the 2001 neutron dose rates are approximately twice as 

high as the 1985 values; the difference is even greater beyond the fueled area. The 2001 neutron dose 

rates at the 90-degree azimuth ranged from 53 mrern/hr near the center to 24 mrem/hr at the outer edge; 

whereas the 1985 readings ranged from 31 to 1.6 mrem/hr, respectively. In addition, unlike the top 

surface readings, the 45-degree azimuth readings are consistently lower than the readings at the 90-degree 

azimuth. The 45-degree azimuth readings in 2001 ranged from approximately 55 mrem/hr in the center to 

12 mrem/hr at the outer edge. By comparison, the 1985 neutron dose rates at the 45-degree azimuth 

ranged from 41 to 3 mrern/hr. 

The disparity in the 1985 and 2001 readings may be in part an artifact of the difference in the cask 

orientation in 1985 and 2001 and the proximity of the detector to structures that thermalize and scatter 

neutrons. In 1985, the dose rate measurements were made with the cask in a horizontal orientation resting 

on a rail car in the TAN Warm Shop. Floors and structures that could thermalize and scatter or reflect 

neutrons back to the detector were many feet from the cask bottom and from the detector. In 2001, the 

measurements were made with the cask upright on the storage pad. The cask was hoisted approximately 

43 to 48 cm (17 to 19 inches) above the concrete pad for access to the bottom of the cask. This also put 

the detector relatively close to the concrete surface. The proximity of the concrete surface allows it to 

thermalize neutrons and/or backscatter the neutrons to the detector. For the same neutron source, the 

neutron dose rates measured in the configuration on the pad could be 40 to 50% higher due to the 

thermalization and/or backscatter than the dose rates measured on the bottom of a horizontal cask in the 

Warm Shop. The scattering process may have also broadened the neutron field, increasing the apparent 

dose rate at the periphery of the bottom surface. 

3.3.5 Conclusion 

The 1985 radiation measurements demonstrated that the shielding performance of the CASTOR 

V/21 cask exceeded the design goal of less than 200 mrenmhr activity at the external surface of the cask. 

The 2001 radiation measurements have demonstrated that the shielding performance of the cask has not 

degraded since 1985. 

The 2001 dose rate data supports the conclusion that gamma dose rates have decreased since 1985, 

as would be expected due to radioactive decay. Because of differences in cask configuration and location, 

some of the 2001 neutron dose measurements appear higher than those taken in 1985. This is caused by 

backscatter from the concrete pad. There is no obvious evidence of degradation in the performance of the 

gamma or neutron shielding between 1985 and 2001 that calls into question the long-term ability of the 

cask to maintain adequate shielding.

55
50

M

M

A

M

0

0 O 0

45 

40 

35 

E 30 

E 

6 25 

cc, 

c 20 

0 

15 

10 

5 

0

M

0I

0

[] U]

* 0 

A 

U]

02001,450 

* 2001, 900 

01985,450 

01985,900 

A 1985, Snoopy

0 200 400 600 800 1000 1200 

Location from Center of Lid, mm 

Figure 3- 12. Radial neutron dose rate profile on the top surface of the cask. The plot compares the 2001 and 1985 readings at the 45 and 90-degree 

azimuths.

M

A 

0

N

U

0

01 E0
60

54 

40

E 02001,45o (D S0m2001, 90o 

o30 0l O1985,45

"031985, 900 

njA 1985, Snoopy 

0 0 

20-0 

10 

0 O0 

0 200 400 600 800 1000 1200 

Location from Center, mm 

Figure 3-13. Radial neutron dose rate profile on the bottom surface of the cask. The plot compares the 2001 and 1985 readings at the 45 and 90-degree 

azimuths.
4. GAS SURVEY 

4.1 Background 

A primary concern of the examination of the CASTOR V/21 cask in late 1999 and early 2000 was 

the integrity of the cask lid seals and the containment of fission products. Evidence of failure of the cask 

lid seals may be the loss of the helium backfill and the ingress of ambient air. This condition can cause 

the corrosion of the spent fuel cladding during long-term storage. As part of the present cask examination, 

samples of the cover fill gases were analyzed prior to opening the cask and compared with earlier gas 

analysis results. 

The fill gas in the cask was sampled and analyzed several times in 1985 during performance testing 

of the cask to determine the integrity of the cask lid seals and the spent fuel cladding. Subsequently, the 

fill gas was sampled and analyzed in 1986 and in 1994 through 1999 prior to opening the cask for the 

present examination. Gas sample analyses included mass spectroscopy and radiochemical analysis. The 

results of those analyses are summarized in the following sections. The analyses found primarily helium, 

indicating satisfactory performance of the cask seals. Further, the absence of fission products in the assay, 

indicated complete integrity of the fuel cladding for the entire 14-year storage period. Because of the lack 

of evidence of a fuel breach, no volatile gases or fines were assumed to have been released, so analysis 

was not done for their presence. 

4.2 1985 Data 

In 1985, the cask cover gas was sampled several times during cask performance testing to evaluate 

the integrity of the spent fuel rods and the cask lid seals. Each sample was collected in a separate 500-cc 

stainless steel cylinder sealed only with quick disconnect-fittings (without bellows-sealed valves). The 

cylinders were checked for leaks prior to sampling. (PNNL-11576) However, the 1985 cover gas analyses 

indicated contamination of the samples with air. (EPRI NP-4887) The contamination was probably caused 

by the quick disconnect-fittings on the sample cylinders and may have occurred during sampling, 

shipment. or handling prior to analysis at Lawrence Livermore National Laboratory (LLNL). Argon was 

probably introduced at LLNL when it was necessary for valves to be fitted to the cylinder 

quick-disconnects in an argon atmosphere to perform the sample analyses. In many cases, the 

contamination was made more severe by the collection of only small amounts of cask cover gas, 

presumably due to short equilibration times between the cask and the sample bottle during the gas 

collection procedure. The end effect of small, diluted samples on the cover gas analyses was to increase 

detection limits, increase measurement uncertainties, and introduce questions of sample validity. 

(EPRI NP-4887. PNNL-1 1576) 

The results of the 1985 LLNL gas analyses, originally presented in EPRI NP-4887, are summarized 

in Table 4-1. The common fixed gases with masses less than 100 were analyzed by mass spectrometry. 

Only N2, 0,• He. Ar, and CO2 are detected at concentrations above 0.01% in any of the samples. Analyses 

of the other species reported are of marginal reliability. In the EPRI NP-4887 report, water was reported 

as a contaminant at a level greater than 0.01%. Because the water probably was due to residual 

physisorbed moisture on sample vessel walls and in the mass spectrometer inlet system, it was omitted in 

Table 4-1. The accuracy of the mass spectra measurements is noted in Table 4-1. The magnitude and the 

variations in the N- and 0, gas analysis data clearly indicate that significant amounts of air were 

introduced in each gas sample.

58
Table 4-1. 1985 casks gas sample composition (volume percent). (From EPRI NP-4887. Permission to 

use this copyrighted material is granted by EPRI.) 

LLNL 

Sample No. He N2 02 Ar CO2 CO N20 H2 

1 89.05 6.02 1.35 3.50 0.05 <0.I <0.01 <0.01 

2 69.30 21.90 5.97 2.80 0.03 <0.01 <0.01 <0.01 

3 62.51 25.57 6.69 5.15 0.03 <0.1 <0.01 <0.01 

4 59.59 17.59 4.11 18.66 0.05 <0.01 <0.01 <0.01 

5 <0.01 95.34 1.18 3.46 0.02 <0.01 <0.01 <0.01 

6 <0.01 90.58 6.81 2.58 0.02 <0.01 <0.01 <0.01 

7 0.03 68.79 3.13 27.95 0.06 <0.01 0.019 <0.01 

8 0.04 70.08 8.43 21.29 0.15 <0.01 <0.01 <0.01 

9 70.59 14.23 5.95 9.09 0.06 <0.1 <0.01 0.018 

10 32.38 50.55 13.31 3.70 0.05 <0.01 <0.01 <0.0I 

11 65.59 1.56 0.16 32.58 <0.01 <0.01 <0.01 <0.01 

12 <0.01 76.94 20.26 2.75 0.05 <0.01 <0.01 <0.01 

13 0.05 74.92 11.44 13.43 0.12 <0.01 0.015 0.016 

14 <0.01 73.95 18.77 7.22 0.06 <0.01 <0.0I <0.01 

15 0.048 85.63 5.56 8.53 0.06 <0.01 <0.01 0.16 

16 0.068 91.06 0.16 8.46 0.04 <0.01 <0.01 0.21 

a. Species present in mass spectra at 0.01% or more. Accuracy of these measurements is + 0.2% of the reported values. 

The 1985 gas samples were also analyzed for 85Kr and 1

4C content. The 1985 report identified that 

9

5 Kr concentrations in the samples ranged from <0.02 to 0.4 pCi/cc, which was marginally detectable in 

the 1985 analyses, while the 1

4C concentrations ranged from 0.6 to 7.9 pCi/cc. The amounts of 85Kr and 

14C are relatively low and are what would be expected to result from crud, not a leaking fuel rod. 

(EPRI NP-4887) 

4.3 1986 Post Cask Performance Test Gas Sampling Data 

Post Cask Performance Test Gas Sampling was conducted in 1986. Table 4-2 presents results of 

the analyses of CASTOR V/21 cask gas samples performed by mass spectrometry and gamma 

spectrometry (for "

5Kr) in 1986. (PNNL-1 1576) 

With the exception of one sample (GN-8A), the analysis results in Table 4-2 (taken from Table 5-3 

in reference PNNL-1 1576) indicate that the cask fill gas consisted primarily of helium with only small 

amounts of residual air constituents. Sample GN-8A, however, obviously was grossly contaminated with 

air during sampling or subsequent handling. The scatter in the nitrogen data (excluding GN-8A) is 

approximately an order of magnitude. The N2 data scatter indicates that the air contamination is most 

likely an artifact of the sampling process and is not representative of the cask cover gas composition. Four 

of the samples in Table 4-2 were also analyzed for H2 and 9

5Kr; the 85Kr concentration was less than 

0.06 pCi/mL. The relatively low amounts of 85Kr detected indicate that no leaking fuel rods were present.

59
Table 4-2. 1986 cask gas sample composition (volume percent). (Taken from PNNL-1 1576) 

Radionuclide 

Sample Sample Concentration, pCi/mL 

No. Date He N2 02 Ar CO, H, 5Kr 

GN-3A 3-05-86 99.0 0.82 0.2 <0.01 <0.01 

GN-3B 3-05-86 99.7 0.27 0.06 <0.01 <0.01 -

GN-4A 3-17-86 99.86 0.11 0.03 <0.01 <0.01 -

GN-4B 3-17-86 99.81 0.15 0.04 <0.01 <0.01 -

GN-5A 3-24-86 99.84 0.13 0.03 <0.0I <0.01 -

GN-5B 3-24-86 99.83 0.14 0.03 <0.01 <0.01 -

GN-8A 6-27-86 65.3 29.1 4.2 0.36 1.1 -

GN-9A 8-15-86 99.8 0.05 0.002 0.001 0.001 0.15 <0.027 

GN-9B 8-15-86 99.78 0.05 0.006 0.001 0.002 0.16 <0.028 

GN-10A 9-26-86 99.8 0.03 <0.01 <0.01 <0.01 0.17 <0.024 

GN-11A 12-08-86 99.8 0.06 0.01 <0.01 <0.01 0.10 <0.057 

4.4 1994 Through 1999 Data 

The fill gas in the CASTOR V/21 was periodically sampled and analyzed from 1994 through 1999 

under the INEEL Spent Fuel Behavior (SFB) Study project beginning in FY 1994. This project was 

combined with the Long-Term Storage Monitoring (LTSM) Program beginning in FY 1996. The 

objective of the SFB Study was to provide a database on the behavior of spent nuclear fuel in long-term 

dry storage through gas sampling to determine the cover gas composition and to infer cladding integrity. 

The CASTOR V/21 was one of two casks selected for enhanced monitoring in the SFB Study project. 

This cask was transferred into the SFB Study project in 1994. The CASTOR V/21 Cask was selected 

because it is fully loaded with intact spent nuclear fuel assemblies that have been undisturbed since 

July 1985. 

Table 4-3 summarizes the results of the mass spectrometric analyses of CASTOR V/21 cask gas 

samples from 1994 through 1999. Approximately four 300 std-cc of gas samples were taken from the 

cask for mass spectrometric analysis in each sampling cycle (only two samples were taken after the 

backfill of fresh helium in 1994). In addition, the samples were also analyzed by gamma spectrometry for 

S5Kr and by liquid scintillation counting for 85

Kr and 34

C. 

As shown in Table 4-3, the initial analysis of the cover gas in the cask in 1994 indicated that the gas 

sample consisted of approximately 2% air. with the balance consisting of helium. The introduction of air 

into the cask was caused by an inadvertent breach during sampling on September 20. 1994. On 

October 11. 1994, the CASTOR V/21 was refilled with fresh helium. (PNNL-1 1576) The gas analysis 

after the fresh helium backfill indicated that the cask cover gas contained 0.5% residual air contamination. 

Subsequent gas analyses through 1999 suggested slightly rising levels of N- The average N, and O0 

concentrations in the gas samples. as summarized in Table 4-4, indicated a three-fold increase in the N

concentration from an average of 0.4 to 1.4 vol% while the 02 concentration did not change. The N2 

concentration appeared to increase steadily after the backfill of the cask with fresh helium in 1994 to 1996 

and appeared to remain constant thereafter. At first glance, these data suggest a slow' air infiltration into 

the cask. However, the steady N2 levels from 1996 onwards are not consistent with a continuous leak

60
Table 4-3. Cask gas sample composition. Values for 85Kr and 14C are in units of nCi/cm3; values for the 

remaining gas components are in units of volume percent. Except as noted, 85Kr and 14C were determined 

by liquid scintillation.

Cask V/21 Cask V121 Cask V121 Cask V/21

Gas Sample 8A Sample VEPCO#24 Sample 13D Sample 14A 

Type INEEL INEEL ANL-W ANL-W 

H2 0.04 0.04 <0.01 <0.01 

He 97.70 97.75 97.84 97.87 

N2 1.83 1.77 1.72 1.69 

02 0.43 0.41 0.38 0.36 

Ar 0.02 0.02 <0.01 <0.01 

C02 <0.01 <0.01 <0.01 <0.01 

85 r a a a a

b b b

1 ______ -r

Gas 

Tvye

Sample 14A 

INEEL

Sample VEPCO#24 

INEEL

b

ANLWc

H2 <0.01 0.01 -- -

He 99.4 99.6 -

N2 0.53 0.35 -

02 0.05 <0.01 -

Ar <0.01 <0.01 _ 

C02 <0.01 <0.01 - _ 

851(1. a a

Gas 

Type

b

Sample 1 

INEEL

b

Sample 2 

INEEL

Sample 3 

ANL-W

Sample 4 

ANL-W

H2 <0.01 <0.01 <0.01 <0.01 

He 99.11 99.60 99.70 99.60 

N2 0.77 0.39 0.33 0.37 

02 0.11 0.01 <0.01 0.01 

Ar <0.01 <01.01 <0.01 <0.01 

C02 <0.01 <0.01 <0.01 <0.01 85

1K,,r a a a a

b b b b

Gas Sample I Sample 2 Sample 3 Sample 4 

FY 1995 Type INEEL IN-EEL ANL-W ANL-W 

2nd Qtr H2 <0.01 <0.01 <0.01 <0.01 

He 99.43 99.28 99.37 99.42 

N2 0.56 0.68 0.60 0.56 

02 <0.01 0.03 0.02 0.01 

Ar <0.01 <0.01 <0.01 <0.01 

C02 <0.01 0.01 <0.01 <0.01 

85Kr 0.0067d -a 

14C b 0.0057d b b

61

FY 1994 

4t Qtr 

(prior to 

purging 

with 

fresh He)

FY 1994 

4 'b Qtr 

(after 

purging 

with 

fresh He)

FY 1995 

ISt Qtr

ANL-W'c

ItC
Table 4-3. (continued).

FY 1995 3'd Qtr

Cask V/21 Cask V/21 Cask V/21 Cask V/21

Gas Sample 1 Sample 2 Sample 3 Sample 4 

Type INEEL INEEL ANL-W ANL-W 

H2 <0.01 <0.01 0.02 0.02 

He 98.52 99.40 99.40 99.41 

N2 1.28 0.59 0.55 0.55 

02 0.20 <0.01 0.01 0.01 

Ar 0.01 <0.01 <0.01 <0.01 

C02 0.01 <0.01 0.01 0.01 

85Kr <6 x 104 a a -a

:4C b 0.0108' b b

Gas Sample 1 Sample 2 Sample 3 Sample 4 

FY 1996 Type INEEL INEEL ANL-W ANL-W 

H2 ND ND 0.02 0.01 

He 98.51 98.86 98.42 98.93 

N2 1.33 1.07 1.42 1.02 

02 0.11 0.04 0.10 0.02 

Ar 0.01 0.01 0.01 0.01 

C02 0.03 0.03 0.02 0.02 85Tz<_ 104 a a a 

KKr <6 x0 0.4e 14

C -b 0.048 e b 

Gas Sample I Sample 2 Sample 3 Sample 4 

FY 1997 Type INEEL INEEL ANL-W ANL-W 

H2 0.01 <.01 <0.01 0.01 

He 98.77 98.83 98.87 98.85 

N2 1.14 1.10 1.07 1.08 

02 0.02 0.01 <0.01 <0.01 

Ar 0.01 0.01 0.01 0.01 

C02 0-05 0.05 0.05 0. 05 

S5Kr 5.9 E1-- a a3 

1b 0.061' --- __ b 

Gas Sample 1 Sample 2 Sample 3 Sample 4 

FY 1998 Type INEEL INEEL ANL-W ANL-W 

H2 ND ND 0.01 0.01 

He 98.48 98.80 98.99 98.98 

N2 1.36 1.11 0.99 1.99 

02 0.02 0.01 <0.01 <0.01 

Ar 0.01 0.01 0.01 0.01 

C02 0.07 0.07 0.01 0.01 

95Kr <6 x 104 -a a 

b 0.0749 b b

62
Table 4-3. (continued).

Table 4-4. Summary of the nitrogen and oxygen concentrations in the cask cover gas analyses. 

Dimensions of values are in volume percent. 

N2 Content, Vol% 02 Content. Vol% 

Average Std. Dev. Average Std. Dev. 

1994, after He purge 0.44 0.13 0.025 0.035 

1995, 1st quarter 0.47 0.20 0.032 0.052 

1995, 2nd quarter 0.60 0.06 0.015 0.013 

1995. 3 rd quarter 0.74 0.36 0.055 0.097 

1996 1.21 0.20 0.067 0.044 

1997 1.10 0.03 0.008 0.010 

1998 1.36 0.45 0.008 0.010 

1999 1.14 0.16 0.020 0.040 

into the cask. The variations in N2 content from sample to sample in some of the sample sets (see 

Table 4-3) suggest instead the contamination of the samples during or after the sampling process, possibly 

by the incomplete removal of residual air from the sample vessel prior to sampling. 

The nitrogen to oxygen ratios of the analytical results were consistently higher than the ratio in 

ambient air, varying from 14 to 180, indicating depletion of the oxygen relative to the nitrogen. The low 

oxygen results would seem more consistent with air infiltration into the cask than with sample 

contamination, for the oxygen could be depleted by rapid oxidation of the metal components and possibly 

the crud (from Fe30 4 to Fe20 3) within the cask. However, the high nitrogen to oxygen ratios do not 

preclude contamination of the gas samples. The apparent reduction of oxygen in the gas samples could

63

Cask V121 Cask V/21 Cask V/21 Cask V/21 

FY 1999 Gas Sample 1 Sample 2 Sample 3 Sample 4 

(Prior to Type INEEL INEEL ANL-W ANL-W 

opening H2 ND ND <0.01 <0.01 

V/21) He 98.45 98.82 98.88 98.86 

N2 1.38 1.08 1.03 1.05 

02 0.08 <0.01 <0.01 <0.01 

Ar 0.01 0.01 0.02 0.02 

C02 0.08 0.08 0.07 0.08 

85K1 <6 x 10-4 a a a 

It b 0 .0 0 9 5 h b b 

a. Analyzed by gamma spectroscopy (INEEL) or mass spectroscopy (ANL-W), not by liquid scintillation. Results were below detection limits 

for gamma spectroscopy (0.3 nCi/crn3) and mass spectroscopy (0.01 vol%) 

b. Not analyzed. 

c. There were no gas samples submitted to ANL-W after purging with fresh He. 

d. Detection limits not available. 

e. Detection limit: 4 x I10' nCi/cm3

. 

f. Detection limit: 6 x 10-4 nCi/cm3. 

g. Detection limit: 5 x 10-4 nCi/cm3

. 

h. Detection limit: 2 x 104 nCi/cm3

.
instead be an artifact of the mass spectrometric analysis, caused by the oxidation of the rhenium or 

tungsten filaments and other hot surfaces in the mass spectrometer source. 

4.5 C-14 and Kr-85 Data 

Table 4-3 presents results of the radiochemical analysis of CASTOR V/21 cask gas samples for the 

years 1994 through 1999. For all of the samples, krypton was analyzed by mass spectrometry and gamma 

spectroscopy. However, the detection limits for these methods were 0.01 volume % and 300 pCi/cm5. 

respectively. From 1995 through 1999, radiochemical analyses were also performed by liquid scintillation 

counting, which is approximately 10' times more sensitive than gamma counting. 

The 85Kr concentration data in Table 4-3 show some inconsistency, with results at or below the 

detection limits for five of the test periods (3rd quarter 1995, 1996 through 1999) and ten times greater 

than detection limits for the 2nd quarter of 1995. Of particular interest is the decrease in value from the 

second to the third quarter of FY 1995. Because krypton is a noble, nonreactive gas, it cannot react with 

internal components in the cask. Therefore, once krypton is released by a breached fuel rod, its 

concentration in the backfill gas should remain constant unless it is preferentially physisorbed on the 

surfaces of the internal components. It is possible that the variations in the krypton results are a result of 

contamination of the sample containers or inadvertent desorption of krypton from other surfaces during 

sample processing and analysis. A review of the 1995-1999 data and the method of analysis has 

substantiated the data in Table 4-3. However, the same review of the krypton analysis method found 

evidence of low-level cross-contamination of sampling vessels by physisorption of noble gases in other 

work. At present, the anomalous results in 1995 have not been satisfactorily explained. 

The 14C data in Table 4-3 show an order of magnitude increase between 1995 and 1998 to a level 

that is two orders of magnitude greater than the detection limit, followed by an order of magnitude 

decrease in the 1999 analysis. A reevaluation of the original data has indicated no significant change in 

the detection limits or statistics that could account for the variation in the data. There is a possibility that 

the low values may be an artifact of sample preparation prior to analysis. However, the variation in the 

results has not been satisfactorily explained yet. 

It is likely that the 1

-C is a real constituent of the cask cover gas. However, the presence of '

4C in 

the cover gas is not a conclusive indicator of breached cladding in the fuel. It is possible that the '

4C was 

generated by neutron capture by an (n.alpha) process from 170 in the crud or other oxides external to the 

fuel rod cladding, or by an (n,p) process from '

4

N contamination in the cask cover gas. The absence of a 

corresponding increase in the 85 Kr concentration indicates that the increase in the '

4C concentration in the 

gas samples is not an indicator of cladding failure. 

4.6 2000 Data After Refill of CASTOR V/21 

Table 4-5 presents results of the mass spectrometric analysis of CASTOR V/21 cask gas samples 

taken on March 23, 2000, immediately after refill of Cask V/21 with fresh helium of 99.98% purity. In 

addition to the analysis of helium and air components as shown in the table, the samples were also 

analyzed for Ne, Kr, and Xe; measurable quantities were not detected. Mass spectrometric data shown as 

ND (not detected) were interpreted as less than 0.005 mole percent. Detection limits for the 35Kr are based 

on counting statistics of the gamma spectrometry and are 0.3 nCi/cm-. 

The samples of the helium backfill gas in the cask, which were sampled on the day of completion 

of the backfill, showed approximately 2% air contamination, comparable to the composition data reported 

in 1994 and 1996 through 1999. Residual air on the order of 0.7% can be expected for the backfill process 

consisted of one evacuation to 5 mbar followed by helium backfill to 700 mbar. The observed 2% residual

64
Table 4-5. Cask gas sample composition (volume mole percent).

March 23

Sample 1 Sample 2 

Gas Type INEEL INEEL 

H2 ND ND 

He 98.31 98.19 

N2 1.31 1.40 

02 0.35 0.38 

Ar 0.02 0.02 

CO 2 0.02 0.01 

85

Kr ND ND

_ Xe ND ND 

ND---Not detected above detection limit of 0.3 nCi/cmr. 

air may be caused by additional residual air in the sample bottles, which are routinely evacuated to 

5-10 mbar prior to taking the sample. 

4.7 Conclusion on Gas Survey 

The analyses of the cask backfill gases have indicated that there was no air ingress in the 

1994-1999 time period and substantiated that the cask lid O-rings were in adequate condition to maintain 

a good seal. The gas analyses did indicate small amounts of air components, totaling approximately 0.5% 

after the 1994 helium backfill and as much as 2 vol% air prior to the 1994 backfill with fresh helium and 

after the 2000 helium backfills. The residual air in the gas samples is most likely an artifact of the backfill 

process and the sampling process, which used only one evacuation (to approximately 5 mbar) prior to 

helium backfill. The gas analyses subsequent to the 1994 backfill did not show any significant increase in 

the total concentration of the air components. 

The presence of the residual air appears to have had no detrimental effect on the fuel or cask 

components, based on visual inspection of the cask internals and the krypton gas analyses. The analytical 

results for krypton from 1996 through 1999 have been below the detection limits. For the six analyses in 

the 1995-1999 period, five analyses have resulted in krypton data at or below the detection limits. In the 

remaining test (2nd quarter of 1995), the krypton concentrations were approximately an order of 

magnitude greater than the detection limits. Because of the radical variation in the time series of readings, 

the high krypton results are probably the result of slight contamination of the sample containers or 

contamination during preparation of the samples for analysis. The krypton results are not consistent with 

failure of the fuel rod cladding during dry storage. 

The 14

CO 2 data also show somewhat erratic behavior. While the 1

4C results rose by an order of 

magnitude from 1995 to 1998. the 1999 analysis results were comparable to the 1995 results. The 

available analysis data do not offer a satisfactory explanation for this behavior. Taken at face value, the 

erratic results, coupled with the krypton results, are not consistent with failure of fuel rods during dry 

storage.

65

FY 2000 1 1 Cask V/21 I Cask V/21
5. CONCLUSION 

A series of examinations in 1999 and early 2000 to investigate the integrity of the CASTOR V/21 

cask were undertaken. The examinations reported in this interim report include: 

* Radiation survey of the cask surface and at I and 2 m (3 and 6 ft) with comparison to the 

1985 radiation survey 

Gas analysis of the internal atmosphere to check for the presence of air ingress into the 

container past the lid seal and for the presence of fission gases that would suggest cladding 

failure 

* Integrity of the concrete pad upon which the CASTOR V/21 has rested since 1985 

o Integrity of the inner and outer O-rings on the lid for signs of corrosion or wear 

* Assembly lifting force measurements to see if there is any resistance to pulling the 

assemblies back out of the cask basket channels due to corrosion or excessive rod bowing 

* Visual inspection of the outside of each assembly for indications of additional corrosion, 

crud spallation or other damage 

* Visual inspection of the cask basket welds and internals for indications of additional 

corrosion or degradation. 

There is no evidence of degradation of the CASTOR V/21 cask systems important to safety from 

the time of initial loading of the cask in 1985 up to the time of testing in 1999. Supporting evidence for 

this lack of significant degradation are summarized as follows. 

The 1999 and 2001radiation survey suggests that doses are generally lower than in 1985-as would 

be expected. Doses are now well below the 200 mrem/hr contact limit. Due to potential errors or 

uncertainties in some of the 1985 and 1999 and 2001 measurements, trends in doses between 1985 and 

1999 are not always easy to determine, however. Nevertheless, the dose trend is down suggesting that 

radiation shielding materials in the cask-both gamma and neutron-are maintaining their function. 

Gas analyses show neither signs of air ingress into the container nor signs of cladding failure 

leading to fission product release. 

Visual examination of the cask lid 0-rings suggest they were, indeed, in adequate condition to 

maintain a seal. 

There was no evidence of major crud spallation from the fuel rod surfaces. 

The concrete pad did not show any sign of failure. Strength measurements did not show any 

evidence of strength loss. Only small cracks typical of a normal, small amount of shrinkage were noticed. 

Concrete conditions immediately below the cask were similar to that of other areas. 

All materials inside the cask-including the assemblies themselves-appeared the same as they did 

in 1985. 

The stitch welds in the triangular air channels appear to be nonstructural welds and were only 

intended to add additional stability for holding the fuel basket in place during 1985 fuel loading and

66
vertical and horizontal thermal testing. The 1985 cask heat transfer performance testing (EPRI NP-4887) 

thermal analysis, concluded that the tightly fitted fuel basket expanded and came in contact with the 

basket barrel and the fuel barrel came in contact with the inner wall of the cask. The results of the thermal 

analysis indicated that thermal stresses in the weld joints at the eight crack locations identified in 1985 

were well above yield stresses measured during tensile testing of applicable weld samples. Based on these 

results and speculations, it appears that all new cracks identified in 1999 occurred during the 1985 

thermal testing. Further, it is concluded the cracks are not relevant to normal long-term storage and 

presents no adverse safety implications on the cask or components to perform their safety related 

function. 

The nondestructive and destructive examination of the cladding from assembly T- 11 will be 

reported by others in a later interim or final report.

67
6. REFERENCES 

10 CFR Part 72.122, U.S. Code of Federal Regulations, Licensing Requirements for the Independent 

Storage of Spent Nuclear Fuel and High-Level Radioactive Waste. 

Berry, W. E., Corrosion in Nuclear Applications, The Electrochemical Society, Inc., John Wiley and 

Sons, Inc., New York, 1971. 

EPRI NP-4887, The CASTOR-V/21 PWR Spent Fuel Storage Cask: Testing and Analyses, EPRI NP-4887, 

PNL-5917, Interim Report, prepared by Virginia Power Company, published by the Electric Power 

Research Institute, Palo Alto CA, November 1986. 

Fisher., L. E., and A. Howe, 1998, Qualification of Independent Spent Fuel Storage Installation, 

Proceedings of the 7th International Symposium on Current Issues Related to Nuclear Power Plant 

Structures, Equipment and Piping, Raleigh, North Carolina, page IV-5. 

General Nuclear System, Inc., CASTOR Vb Dry Spent Fuel Storage Cask, GNS Drawing 49001-E-01, 

sheet 1 of 2. Drawing date illegible. Similar drawings are in the 1984 to 1986 time frame. 

GNSI, Topical Safety Analysis Report for the CASTOR V/21 Cask Independent Spent Fuel Storage 

Installation (Dry Storage), Revision 1, General Nuclear Systems, Inc., Avon, Connecticut, January 

1985, figures 7.3-5 and -6. 

McKinnon, M. A., and A. L. Doherty, Spent Nuclear Fuel Integrity During Dry Storage- Performance 

Tests and Demonstrations, PNNL-1 1576, Pacific Northwest National Laboratory, Richland, WA, 

June 1997. 

Sandoval, R. P., R. E. Einziger, H. Jordan, A. P. Malinauskas, and W. J. Mings. Estimate of CRUD 

Contribution to Shipping Cask Containment Requirements, SAND88-1358. Sandia National 

Laboratories, Albuquerque, NIM, January 1991.

68
Appendix A

Dosimeter Radiation Measurements from 

CASTOR V/21 Cask Performance Test

A-I
Appendix A 

Dosimeter Radiation Measurements from 

CASTOR V/21 Cask Performance Test

- ..VIN/1- UASqjPERFORMA1

N'CIETEST , iLOCATIt6N> Doeacit&~ - se mte At contact V 

Angle, { a} Elevation TLD Dose Rate TED Dose Rate TLD RO-20 E-600 

Ib}

D3e~rnees•

or radius 

{c} 

mm1T

Gamma 

mR/hr

Neutron

mRlhr

Gamma Gamma Neutron 

mR/hr mR/hr mRlhr

SIDE 45 175

-480

23 2.1 2.6 4 2 

*140~ 21_):2

9.7 1.3

12 1.6 

10.6 1.3 

10.3 31.4 

7. 1* 

5.3 3 

185 8 

9.7 6.2 

7,3 . 3.8 

5.6 2.1 

3.8. 1.1 

3.8 0.8 

2.1 0.5

2.6 

1.5 

13 

11 

1.5 

3.5 

8.7 

15 

12 

13

0.5 

0.3

1.3 

1.2 

3.8 

2.2 

1.5 

1.7

1.3 7 1.4.•?-!7•

1.4 

-1.5 

1.5 

1.2 

2.5 

4 

8 

7 

4 

2.5 

1.5 

: I[

0.6 

0.5 

1.4 

1.4 

6 

7 

2 

1.7 

1.5 

1.5 

1.5

6 

4 

14 

13 

8 

6 

8 

6 

5 

4 

6 

2 

6 

3 

6 

8 

5

1048 

* 1597 

2149 

3298 

3850 

4350 

4450 

4550 

4600 

4650 

47 00

36 

3•8, 

38 

37 

39 

28 

22 

118 

68 

26 

9.4 

4,.3" 

2.5 

1.6 

1.1 

.0.7 

31 

23 

26 

"125 

25 

21 . 

21 

24 

22

60 

75 

90

4750 

4800 

2701 

2701 

175 

480 

1048 

1597 

2149 

2701 

3298

A-3

De ees min mR/hr

( :
LOCATION Dose rates-at contact - Dose~rate atecontact.' 

1985 __20011 

Angle, { a } Elevation TLD Dose Rate TED Dose Rate TLD RO-20 E-600 

{b} 

or radius Gamma Neutron Gamma Gamma Neutron 

{c} 

Degrees mm mR/hr mR/hr mR/hr mR/hr mRlhr 

SIDE 3 850 70 9.8 ~ 4.~6 4S .5" 8 : 

4350 79 11.5 7.8 8 13 

.< 4400' 134_ ~ 1 3-. rJ yI __: 8_< 

4_450 81 10 8.5 10 16 

:4500, 33 < 4..79~_ 6.4-. 12 

4550 12 6.9 3.2 4.5 8 

90 4650 3.1 8.8 1.1 9 

47 00 1.) 318 0.7 1.2, _ 3 

4750 1.3 0.9 0.5 0.8 4 

4800- J_2 0.7 0-4 0.7 _ 3 

105 2701 24 14 1.4 1.3 10 

120 21701 31 9.3 .1126 

72,73 2149 23 15 1.1 1.2 6 

73,74 2 149 24 17 1.1 1.2• 7 

74,75 _2149 23141.117 

75,76 2149 22 11 1.1 1 5 

76,77 2149 25 13 1.1 1 7 

77,78 :21,49 247 1'2 1.3 5 

Top Trunnion, -9 8.1 2.6 2 10 

90 

Bottom Trunnion, 42 27 0.9 L.5 11 

90 

TOP 

45 203 32.5 44 0.2 12 36 

406 . 43 35 3.9 8 34 

585 26 32 3.5 8 28 

928 6.4 5.1 9.8 1 14 

1100 1 1.3 7.2 0.5 7 

90 170 30 44 0.2 6 42 

340 36 36 0.2 6 38 

487 29.5 32 0.3 5 47 

634 29 27 1.8 3.5 28

706 22 19 2.6 26

A-4
-'os t -a contact-u-., LOCATIO - atst ---- t, 

Angle, { a } Elevation TLD Dose Rate TED Dose Rate TLD RO-20 E-600 

{b} 

or radius Gamma Neutron Gamma Gamma Neutron 

5C2 

Degrees mm mR/hr mR/hr mR/hr mR/hr mR/hr 

TOP 780 56 16 2.3 2 22

928 2.5 3.1 If! 3.3 1.5 11

1042 1.6 0.7 4.2 0.5 7 

0 40 44 10.1 90 45 

Hole at 180 - 32 13 1 1 1

JBOL¶TOWM

90 0 22 

170 i1 

340 24 

* .487 22

634 

928

20 

26

1100 7 

203 .17.5 

406 25 

585 25 

928 18 

1100 3.4

41 6.8 4 55 

25 6.3 3.8 54 

23 5.5 . 3.9, __5 

21 4.4 4 45 

5 0.5 3.5' _3 

1.6 0.8 2.5 24 

33 6.1 4 "47' 

22 6 4 43 

J19 51 3.82 2 

___1_1.9 2.5 24--2 

3.4 2 12

2001 Supplemental Cask Side Survey Locations {g}

SIDE 0 

30 

60 

120 

150 

180 

.210 

240 

270 

300 

330 

0

4400 

4400 

4400 

4400 

4400 

4400 

4•400 

4400 

4400 

4400 

4400 

4298

45

16 

12 

12.9 

15 

24.7 

12.3 

13.9 

17.3 

12.6 

1.4

9 

8 

9 -97 

13 

10 

8 

12 

10 

1.8

'10 

14 

9 

12 

14 

13 

15 

7 

8

A-5
LOCTIN os rtes at conet Dose rate at contact: 

Angle, {a} Elevation TLD Dose Rate TED Dose Rate TLD RO-20 E-600 

{b} 

or radius Gamma Neutron Gamma Gamma Neutron 

{c} 

Degees mm mR/hr mRlhr mR/hr mR/hr mR/hr

30 4298 -- -

45 4298 -

0.8 16 9 

3.6 2 10

____ -60 _4298-- > ;- - .;25< 1 

90 4298 18.8 3 13 

120- - : 4,298. -. 6_ 

150 4298 1.2 2.2 10 

18 ~ .4298-- - -- *.13..?~ .9 

210 4298 1.4 2 1 10 

240 4298-,-.' - . -2- - 1 _ 27.: - -11 7 

270 4298 5.7 2.5 12 

300 4298 -1 -.. 3 2.8- _ .12 

330 4298 0.8 1.5 7 

90 4100 0.7 1.5 9 

:! ::-:(••: ._ . 43' . ...: ,.337.0.:. '": : -• >f -) - :_ :.- "_ 1,2 - •- :.42 1 Z -:: --_ 5 :.: 

90 3370 1.5 1.4 8 

0 2701 - L- ,4 .8- 6 

30 2701 1.1 1.4 

1 50 2701 -- 1.3 1.4 6 

180 2701 1.6 1.8 _ 6 

210 - 2701 -- 13 LS.8 . 5 

240 2701 1.5 1.6 6 

270 2701 -- 1.7 2 7 

300 2701 1.2 1.2 5 

330 2701 - -i . 1.3 7 

45 2443 1.4 1.5 6 

90 2443 -1.4 2 6

A-6

(a} from 0 orientation mark 

{b) from exterior bottom of cask 

c } from cask centerline 

If problem with processing dosimeter 

{g} survey locations were not obtained in 1985
Appendix B

Cracked Stitch Welds Identified in Fuel Basket 

of Castor V/21 Cask Basket

B-1
Figure B-1. Top weld crack at location 9 identified in Figure B-19. 

(Triangular air channel between fuel tube A7 & B8)

Figure B-2. Bottom weld crack at location 9 identified in Figure B-19. 

(Triangular air channel between fuel tube A7 & B8)

B-3
Figure B-3. Top weld crack at location 10 identified in Figure B-19. 

(Triangular air channel between fuel tube A3 & A6)

r 

- - 77-. 

-K 

* r Žr;4S'-

Figure B4. Bottom weld crack at location 10 identified in Figure B-19. 

(Triangular air channel between fuel tube A3 & A6) 

B-4
Figure B-5. Top weld crack at location 11 identified in Figure B-19. 

(Triangular air channel between fuel tube A3 & B6)

Figure B-6. Bottom weld crack at location 11 identified in Figure B-19. 

(Triangular air channel between fuel tube A3 & B6)

B-5
•U •!i• • !•:: •- 

7 :- •-•-'-_ ,-_aI 

S• :•7 , .. .-1 

-t--•- ._• ••.

Fi ueB 74 o ed n rc .L ctonn m e 2i etfe nFg r -9 

(Tinua ai chne¶ewe ue1ueB 8

z -i - -y 

Figure B-8. Bottom weld crack at location 12 identified in Figure B-19. 

(Triangular air channel between fuel tube B7 & B8) 

B-6
Figure B-9. Top weld crack (minor) at location 13 identified in Figure B-19. 

(Triangular air channel between fuel tube B7 & C8)

Figure B 10. Botton weld crack at location 13 identified in Figure B-19. 

(Triangular air channel between fuel tube B7 & C8)

B-7
Figure B-1il. Top weld crack at location 14 identified in Figure B-19. 

(Triangular air channel between fuel tube C6 & D3)

Figure B-12. Botton weld crack at location 14 identified in Figure B-19. 

(Triangular air channel between fuel tube C6 & D3)

B-8
Figure B-13. Top weld crack at location 15 identified in Figure B-1. 

(Triangular air channel between fuel tube D3 & D6)

Figure B-14. Bottom weld crack at location 15 identified in Figure B-19. 

(Triangular air channel between fuel tube D3 & D6)

B-9
Figure B-15. Top weld crack at location 16 identified in Figure B-19. 

(Triangular air channel between fuel tube A7 & D8)

Figure B-16. Bottom weld crack at location 16 identified in Figure B-19. 

(Triangular air channel between fuel tube A7 & D8)

B-10
Figure B-17. Crack in basket weld below weld crack 4 (120') identified in Figure B-19.

Figure B-18. Same as Figure B-17 with complete view of existing weld crack 4 identified in Figure B-19.

B-11

"qqq
BORATED STAINLESS STEEL 

(TYPICAL FOR CRACK LOCATIONS 9-16)

BARREL (304 SS)

-CASK 

BODY

90"

S'- FUEL TUBE ID 

O LOCATIONS OF CRACK INDICATIONS 

IDENTIFIED IN 1985 

c0 LOCATIONS OF CRACK INDICATIONS 

IDENTIFIED IN 1999

9 FILLET WELDS EXAMINED IN 1999 

IN ADDITION TO EXISTING CRACK 

LOCATIONS IDENTIFIED IN 1985 

Figure B-19. CASTOR V/21 Basket Crack Indication Locations.

B-12

0, I

1 180"
Appendix C 

Fuel Assembly Features

C-1
Figure C-1. Westinghouse gropple lifting fuel assembly V-05.

Figure C-2. Top of assembly V-05.

C-3
Figure C-3. Oxide mottling noted around spacer #5 on fuel assembly V-05.

Figure C4. Oxide mottling noted around grid spacer #6 on V-05.

C-4
Figure C-5. Fuel assembly V-05-area around upper grid spacer-rods clean and minimal bowing noted.

Figure C-6. Fuel assembly V-05 area around third grid spacer-noted light abrasion marking on spacer.

C-5
Figure C-7. Fuel assembly V-05 oxide mottling noted around fifth grid spacer.

Figure C-8. Fuel assembly V-05 bottom of fuel element V-05-light oxide mottling noted-no loose 

material seen on rods or top of bottom nozzle.

C-6
Figure C-9. Fuel assembly V-05 detail of bottom of element-note oxide rings adjacent to rod end weld.

Figure C-10. Fuel assembly V-05 top of assembly.

C-7
I

Figure C- 11. Fuel assembly V-05 oxide mottling noted around grid spacer 5.

Figure C-12. Fuel assembly V-05-bottom left nozzle-noted mechanical fasteners searing guide tubes.

C-8
Oxide 

bands

Figure C-13. Fuel assembly V-05 looking for loose "crud"-non was evident.

Figure C-14. Element T-03 being withdrawn-note fuel rod growth.

C-9
Figure C-15. Element T-03 at intermediate grid spacer-rods appear to be relatively 

clean with a light oxide coating.

Figure C-16. Element T-03 midpoint of assembly noted oxide density 

increases as we go lower on the element.

C-10
Figure C-17. Element T-03 area adjacent to fifth grid spacer.

-lo i 

ZFigure C- 18. Element T-03 bottom of fuel element-light oxide-no loose material obvious on surfaces.

C-1I
Figure C-19. Fuel element T-O3-notes serial number-fuel rod growth evident.

Figure C-20. Fuel element T-03 bottom.

C- 12
Figure C-21. Fuel assembly V-27 being withdrawn.

rr 

Figure C-22. Fuel assembly V-27 oxide mottling noted. 

C-13
-t 

Figure C-23. Fuel assembly V-27 at bot 

Figure C-24. Fuel assembly V-04.

torn.

C-14

'C
Figure C-25. Fuel assembly V-14. 

Figure C-26. Fuel assembly V-14 oxide mottling at fifth grid spacer. 

C-15
Figure C-27. Fuel assembly T-07 note difference in fuel rod growth.

Figure C-28. Fuel assembly T-07 at sixth grid spacer.

C-16
Figure C-29. Fuel assembly T-07 bottom.

Figure C-30. Fuel assembly T-12 mid point.

C-17
Figure C-31. Fuel assembly T-12

Figure C-32. Fuel assembly T-12.

C-18
Figure C-33. Fuel assembly V-08 tops of fuel rods appear quite even.

Figure C-34. Fuel assembly V-08 at fifth spacer-noted oxide mottling.

C-19
Figure C-35. Fuel assembly V-08.

71 

A...

I 

'C

Figure C-36. Fuel assembly V-11. 

C-20

K
Figure C-37. Fuel assembly V-i1 at sixth spacer.

Figure C-38. Fuel assembly V-1 i.

C-21
Figure C-39. Fuel assembly V-0l.

Figure C-40. Fuel assembly V-01 at fifth grid spacer.

C-22
Figure C41. Fuel assembly V-01.

Figure C-42. Fuel assembly T-08 near top.

C-23
Figure C-43. Fuel assembly T-08 near sixth grid spacer.

Figure C-44. Fuel assembly T-09--note uneven rods.

C-24
Figure C-45. Fuel assembly T-09 at sixth grid spacer.

Figure C-46. Fuel assembly T-09.

C-25
Figure C-47. Fuel assembly T-16.

Figure C-48. Fuel assembly T-16 at sixth grid spacer.

C-26
Figure 049. Fuel assembly T-16.

Figure C-50. Fuel assembly T-16.

C-27

5- Zvi -4
I 

U 

-J 

-,1 

P1

Figure C-51. Fuel assembly V-24.

Figure C-52. Fuel assembly V-24 third spacer. 

C-28
Figure C-53. Fuel assembly V-24 near fifth spacer.

Figure C-54. Fuel assembly V-24.

C-29
Figure C-55. Fuel assembly V-15 third grid spacer.

Figure C-56. Fuel assembly V-15 close ups at sixth grid spacer.

C-30
Figure C-57. Fuel assembly V-15.

Figure C-58. Fuel assembly V-25.

C-31
Figure C-59. Fuel assembly V-25.

Figure C-60. Fuel assembly T-13.

C-32
Figure C-61. Fuel assembly T-13 near sixth grid spacer.

Cf 

Figure C-62. Fuel assembly T- 11.

C-33
Figure C-63. Fuel assembly T-11.

&

Figure C-64. Fuel assembly T-I 1 attempt to obtain loose material.

C-34
Figure C-65. Fuel assembly V-05 fourth grid spacer oxide mottling noted.

Figure C-66. Fuel assembly V-04 mottling at fifth grid spacer.

C-35
Figure 67. Fuel assembly V-04-bottom-rod having greatest indication of oxide 

appears to be slightly above the nozzle plate-no conclusion made.

C-36
Appendix D 

Fuel Tubes

D-1
Figure D-1. Fuel Tube A-I top-appears clean. Typical appearance 

and cleanliness of all fuel tubes.

Fuel tube 

wall 

Cask body 

floor

Figure D-2. Fuel Tube A-I - Close up of bottom - minor debris noted.

D-3

Fuel 

assembly 

support 

plate (typ)
Fuel tube 

wall 

Cask body 

floor

Fuel 

Assembly 

- support 

plate (typ) 

Original 

- fabrication 

layout 

markings

Figure D-3. Fuel tube A-7 - closeup of bottom - noted stain - no conclusion - surfaces are clean and 

bright. This view approximates condition viewed at the bottom of all fuel tubes.

Figure D-4. Fuel tube A-5 bottom - Noted typical absence of quantities of loose material.

D-4
I

Figure D-5. Fuel tube A-6 bottom.

Figure D-6. Fuel tube A-3 bottom - Surfaces clean and bright.

D-5
Figure D-7. Looking down fuel tube B-5.

Figure D-8. Fuel tube B-5 - bottom.

D-6
Figure D-9. Fuel Tube B-6.

Figure D-10. Fuel tube B-8 bottom. 

D-7
Figure D-1 1. Inside fuel tube D-5 at the midway of the tube. View as typical of all fuel tubes.

Figure D-12. Midway inside fuel tube D-6. 

D-8
Figure D-13. Fuel tube D-8 - bottom.

Figure D-14. Fuel tube B-4.

D-9
Figure D-15. Fuel tube A-4 bottom.

Figure D-16. Fuel tube C-6.

D-10
- 2 

, 

-- t 

- -

4 

--'--,

Figure D-17. Fuel tube C-6 - Appeared quite clean.

-4-

Figure D-18. Fuel tube D-3 bottom. 

D-11

1⁄2
Figure D-19. Fuel tube D-2 bottom.

Fuel tube 

wall (typ)

Fuel 

,,~-.assembly "support 

plate

Cask body 

floor

Figure D-20. Fuel tube C-5 reflective surfaces and existence 

of original fabrication/layout characters.

D-12
Figure D-21. Fuel tube C-8, bottom.

Cask body 

Fuel basket barreFel bsewall • •i 

Figure D-22. Fuel tube B-7 - near top 

D-13

- showing side plates and basket wall.
Figure D-23. Fuel tube E-7, bottom.

Fuel

Fueltube -A 

A6 with ;-" fuel" fuel assy ..... V27 

V14 

Figure D-24. Tube reinforcement plate and associated welds between A-5. A-6, and A-8.

tube 

ith 

assy

D-14
Reflective =-OM x- Camera 

surface holding 

grindingtool 

marks 

Fuel tube A2 

wall 

(typ)

Figure D-25. Top of fuel tube A-2.

D-15
Appendix E 

Hot Cell Operation and Removal of Fuel Rods from 

Fuel Assembly Ti 1

E-1
Figure E-1. Fuel assembly and nozzle cutter.

"Fuel 

assembly 

, laydown 

: fixture 

Fuel assembly : 

center -

instrument . 

tube location 

and center of I 

3 inch dia. - -• 

hole cut. I.. •

Figure E-2. Fuel assembly T-1 I-after cutting a 3-inch dia. hole in top nozzle to access rods.

E-2
Figure E-3. Fuel assembly T-I 1-ready to pull fuel rods.

Fuel rod 

puller

Y

Figure E-4. Fuel rod being pulled from fuel assembly T-11 .

E-3
Figure E-5. Fuel rod pulled into inspection tray.

Figure E-6. Oxide mottling on fuel rod.

E-4
i• ! ;? i!•• • i •!~iii!iiiiii•iiiiii~i~i !~~i~i i~l' ̧ • • li i!i: ! !!i • •: ............

Area 

outside 

spacer grid

Area housed 

by fuel 

assemble 

spacer grid

2

Figure E-7. Condition of fuel rod-light oxide coating minor-scratches believed 

from moving rod through spacer grids.

3⁄4

N

Figure E-S. End of fuel rod.

E-5
Figure E-9. Fuel rod-linear scratches believed from moving rod through spacer grid plates.

Figure E-10. Fuel rod with linear scratch.

E-6
Figure E- 1. Fuel rod with linear scratch. 

Figure E-12. Fuel rod with linear scratch. 

E-7
Plugged 

spare 

tubes (typ) 

Fuel rods 

Empty 

"- spare 

tubes (typ)

-- Figur 13Zcaknerac hipping tubewi •os e 12. Z fuel 

Figure E-13. FSV cask interface shipping tube with the 12 fuel rods.

Figure E-14. Fuel assembly being moved from the hot shop into the hot cell.

E-8
Figure E-15. Camera Resolution Test showing letter size.

Figure E-16. Typical camera color check.

E-9
Appendix F 

Cask Interior Wall and Floor

F-1
Cask basket 

barrell 

Cask body 

wall 

Fuel tube and 

fuel assembly i!S:ai, :-•(typ) 

Figure F-1. Castor V/21 configuration with primary lid removed. 

Fuel tube 

wall

Figure F-2. Cask inner wail at top-typical surface condition.

F-3
Cask body 

wall 

Cask body 

floor 

Cask body n• <. 

wall 

Cask body . 

floor

Figure F-3. Cask floor and wall bel

Fuel tube 

Fuel tube fuel 

•-' "assembly 

, .support plate 

edge 

Minor granular 

debris on floor 

low fuel tube B7.

Minor 

granular 

debris on 

floor (worst 

case)

Figure F-4. Cask floor and wall below fuel tube C8-reflective surfaces free of corrosion.

F-4
Fuel tube

Fu 

asu

Cask body ph 

de 

Cask body 

floor 

Figure F-5. Cask floor and wall below fuel tube D3 reflective surfaces free of corrosion. 

Cask body 

wall 

01 

Cask body 

floor 

Figure F-6. Cask floor and wall below fuel tube D6-reflective surfaces free of corrosion.

el tube 

sembly 

pport 

ate 

inor 

anular 

bris on 

or 

inor 

anular 

ebris 

n floor

F-5
Figure F-7. Cask floor and wall below fuel tube A7-reflective surfaces free 

of corrosion-minor granular debris on floor.

Figure F-8. Cask floor below fuel tube A3-reflective surfaces free 

of corrosion-minor granular debris on floor.

F-6
Fuel tube 

assembly 

support plate 

Cask body 

floor

Figure F-9. Cask floor below fuel tube B6-reflective surfaces free 

of corrosion-minor granular debris on floor.

Figure F-10. Cask floor below fuel tube B8-reflective surfaces free 

of corrosion-minor granular debris on floor.

F-7

Fuel tube 

wall 

Cask body 

wall
:. -<!..... . 

S• •-. . • - - ., .",2•~ 

•2 -- r" : ,-4" : .•-- =- -•. _• -7-- - • u •--• : 

•7•• ~--- ="" •-'~--•-w •- • • -

Figure F- 11. Cask wall below and between fuel tubeA7 and A8 reflective surface free 

of corrosion-typical of cask wall condition without mineral stains- nodular ductile cast iron nickel plated. 

:.. :- 7 

" • y. - " "-7• 

Z,

Figure F-12. Cask wall below and between fuel tubeD3 and D6-reflective surface free of corrosion

typical of cask wall condition with mineral stains-nodular ductile cast iron nickel plated. 

F-8
Appendix G 

Condition of Existing Cracked Welds Identified in 1985

G-1
Figure G-1. Existing 1985 weld crack at location No. I identified in Figure G-9.

Figure G-2. Existing 1985 weld crack at location No. 2 identified in Figure G-9. 

G-3
Figure G-3. Existing 1985 weld crack at location No. 3 identified in Figure G-9.

Figure G4. Existing 1985 weld crack at location No. 4 identified in Figure G-9.

U-4
Figure G-5. Existing 1985 weld crack at location No. 5 identified in Figure G-9.

Figure G-6. Existing 1985 weld crack at location No. 6 identified in Figure G-9.

G-5
Figure G-7. Existing 1985 weld crack at location No. 7 identified in Figure G-9.

Figure G-8. Existing 1985 weld crack at location No. 8 identified in Figure G-9.

G-6
STAINLESS STEEL 

FOR CRACK LOCATIONS 9-16) 

1 

/BASKET BARREL (304 SS)

U() LOCATIONS OF CRACK INDICATIONS 

IDENTIFIED IN 1985 

180U 0 LOCATIONS OF CRACK INDICATIONS 

IDENTIFIED IN 1999 

9 FILLET WELDS EXAMINED IN 1999 

IN ADDITION TO EXISTING CRACK 

LOCATIONS IDENTIFIED IN 1985 

Figure G-9. Castor V/21 basket crack indication location.

G-7
Appendix H 

Castor V/21 Cask Side Radiation Survey Locations

H-1
Appendix H 

Castor V/21 Cask Side Radiation Survey Locations 

Elevationa MM 00 300 450 600 750 900 1050 1200 1500 1800 2100 2400 2700 300- 3300 

4886 (192.36") 

[3.39"] 

4800 (188.97") _ [] 

[1.97"] 

4750 (187.00") I_ E[] 

[1.97"] 

4700 (185.03") ) LI 

[1.96"] 

4650 (183.07") )] [] 

[1.97"] 

4600 (181. 10") [] 5 

[1.97"] 

4550 (179.13") ____ 

[1.97"] 

4500 (177.16") 5 [] 

[1.97"] 

4450 (175.19") E 11 

[1.97"] 

4400 (173.22") 0 0 j 0 __ 0 0 0 0 0 0 0 0 

[1.97"] 

4350 (171.25") 0 _ _ 

[2.04"] 

4298 (169.21") 0 0 0 0 0 0 0 0 0 0 0 0 0 

[7.79"] 

4100 (161.41") 0 0 

[9.84"] 

3850 (151.57") E __ 

[18.9"] 

3370 (132.67") 0 0 

[2.83"1 

3298 (129.84") 1 0 

[23.5"] 

2701 (106.33") 0 0 ] : S S E F 0 0 0 0 0 0 0 

[10.1-] 

2443 (96.18") 0 0 

[11.5"1 

2149 (84.6") 5 11 

[21.7"] 

1597 (62.8") E: 

[21.6"] 

1048 (41.25") 01 

[22.4"] 

480 (18.8") _ [] 

[12.0"] 

175 (6.8") :___

-Flevationl trom exltm~or bottom ot cask 

1985 Radiation survey locations taken on exterior side of cask 

O Radiation survey locations to be taken in addition to 1985 locations on exterior side of cask 

(--) Inches 

[--] Inches between elevation points 

H-3

WCB 12/00
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-89) (Assigned by NRC, Add Vol., Supp., Rev., 

NRCM 1102, and Addendum Numbers, if any.) 

3201, 3202 BIBLIOGRAPHIC DATA SHEET 

(See instructions on the reverse) NUREG/CR-6745 

2. TITLE AND SUBTITLE INEELJEXT-01-00183 

Dry Cask Storage Characterization Project - Phase 1: 

CASTOR V/21 Cask Opening and Examination 3 MONTH DATE REPORT PUBLISHED YEAR 

September 2001 

4. FIN OR GRANT NUMBER 

Y6038 

5. AUTHOR(S) 6. TYPE OF REPORT 

Bare, W.C., Torgerson, L.D. Technical 

7. PERIOD COVERED (inclusive Dates) 

07/99 thru 06/01 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor, 

provide name and mailing address.) 

Idaho National Engineering and Environmental Laboratory 

Idaho Falls, ID 83415-3114

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type 'Same as above'; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, 

and mailing address.)

Division of Engineering Technology 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-001

INTEC Programs Division 

U.S. Department of Energy 

Idaho Operations Office 

Idaho Falls, ID 83401-1563

Spent Fuel Storage and Disposal Program 

Electric Power Research Institute 

3412 Hillview Avenue 

Palo Alto, CA 94304-1395

10. SUPPLEMENTARY NOTES

D !iA V,,,etli; MPCPrriimct Mranranr

11. ABSTRACT (200 words or less)

This report documents visual examination and testing conducted in 1999 and early 2000 at the Idaho National Engineering and 

Environmental Laboratory (INEEL) on a Gesellschaft fur Nuklear Service (GNS) CASTOR V/21 pressurized water reactor 

(PWR) spent fuel dry storage cask. The purpose of the examination and testing is to develop a technical basis for renewal of 

Licenses and Certificates of Compliance for dry storage systems for spent nuclear fuel and high-level waste at independent 

spent fuel storage installation sites. The examination and testing was conducted to assess the condition of the cask internal 

and external surfaces, cask contents consisting of 21 Westinghouse PWR spent fuel assemblies from Dominions (formerly 

named Virginia Power ) Surry Nuclear Power Plant and Cask Concrete Storage Pad. The assemblies have been continuously 

stored in the CASTOR cask since 1985. Cask exterior surface and selected fuel assembly temperature, and cask surface 

gamma and neutron dose rates were measured. Cask external/internal surfaces, fuel basket including accessible weldments, 

fuel assembly exteriors and primary lid seals were visually examined. Selected fuel rods were removed from one fuel assembly, 

visually examined, and then shipped to Argonne National Laboratory for nondestructive, destructive and mechanical 

examination. Cask interior crud samples and helium cover gas samples were collected and analyzed. The results of the 

examination and testing indicate the concrete storage pad, CASTOR V/21 cask and cask contents exhibited sound structural 

and seal integrity and that long-term storage has not caused detectable degradation of the spent fuel cladding or the release of 

gaseous fission products between 1985 and 1999.

12. KEY WORDSIDESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT 

CASTOR V/21 cask, visual examination, technical basis, renewal of Licenses and unlimited 

Certificates of Compliance, dry storage systems, spent nuclear fuel, independent spent fuel storage 14. SECURITY CLASSIFICATION 

installation, Surry Nuclear Power Plant, cask external surface, cask internal surface, fuel basket, (This Page) 

weldments, primary lid seals, concrete storage pad, long-term storage, degradation of spent fuel unclassified 

cladding, Idaho National Engineering and Environmental Laboratory. (This Report) 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (2-89) This form was electronically produced by Elite Federal Forms, Inc.

0 .Rll . V -- pll Z., KIPq P Pe-t Mana 32 er
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NUCLEAR REGULATORY COMMISSION 

WASHINGTON, DC 20555-0001 

OFFICIAL BUSINESS 

PENALTY FOR PRIVATE USE, $300