ML20128B098
| ML20128B098 | |
| Person / Time | |
|---|---|
| Issue date: | 05/31/1985 |
| From: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | |
| Shared Package | |
| ML20128B081 | List: |
| References | |
| REF-PROJ-M-34 NUDOCS 8505240406 | |
| Download: ML20128B098 (84) | |
Text
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'. SAFETY EVALUATION REPORT m
RELATED TO THE TOPICAL SAFETY ANALYSIS
' REPORT FOR' CASTOR Ic DRY SPENT FUEL STORAGE CASK-U.S. Nuclear Regulatory Comission Office of Nuclear Material Safety and Safeguards May 1985 g 52 gO6 p 850514 M-34 PDR:
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TABLE OF CONTENTS Page
- 1. 0 General Description. ....................... 1 1.1 Introduction. ........................ 1
- 1. 2 General Description of the Storage Cask . . ......... 2 1.2.1 Cask Design Characteristics. . . . . . . . . . . . . . 2 1.2.2 Operational Features . . . . . . . . . . . . . . . . . 2 1.2.3 Cask Contents. . . . . . . . . . . . . . . . . . . . . 6
- 1. 3 Identification of Agents and Subcontractors . ........ 6 1.4 Generic Cask Arrays . . . . . . . . . . . . . . . . . . . . . 8 _
2.0 Principal Design Criteria . . . . . . . . . . . . . . . . . . . . 9 2.1 Introduction. ........................ 9 2.2 Fuel to be Stored . . . . . . . ... ... . ... ..... 9 2.3 Quality Standards . . . . . . . . . . . . . . . . . . . . . . 9 2.4 Protection Against Environmental Conditions and Natural Phenomena . . . . . . . . . . . . . . . . . . . . 10 2.5 Protection Against Fire and Explosions. . . ......... 11 2.6 Confinement Barriers and System . ..... ......... 11 2.7 Instrumentation and Control Systems . . . . . . . . . . . . . 12 2.8 Criteria for Nuclear Criticality Safety. . . . . . . .. . . . 13 2.9 Criteria for Radiological Protection. . . . . . . . . . . . . 13 2.10 Criteria for Spent Fuel and Radioactive Waste. . . . . . . . 15 2.11 Criteria for Decommissioning . . .............. 15 3.0 Structural Evaluation. . . . . . . ....... ......... 16 3.1 Area of Review. . ...................... 16 3.2 Acceptance Criteria . . . ......... ......... 16 3.3 Review Procedure. . . . .... . ..... ......... 16 3.4 Findings and Conclusions. .... .......... .... 16 3.4.1 Loads. . . . . . . . . . . . . . . . . . . . . . . . . 16 3.4.1.1 Normal Operating Conditions . . . . . . . . . . 17 3.4.1.2 Loads Due to Environmental Conditions and Natural Phenomena . . ........... 17
! 3.4.1.3 Load Due to Postulated Accidents. . ...... 17 3.4.2 3.4.3 Materials. . . . . . . . . . . . . . . . . . . . . . . 17
! Stress Intensity Limits. . . . . . . . . . . . . . . . 18 3.4.4 Structural Analysis. . . ...... . ... ..... 19 l
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o o .a TABLE OF CONTENTS (Continued)
Page 3.4.4.1 Cask Body . . . . . . . . . . . . . . . . ... 19 3.4.4.1.1 Normal Operating Loads . . . . . .... 19 3.4.4.1.2 Environmental Conditions and Natural Phenomena. . . . . ..... 20 3.4.4.1.3 Accident Conditions. . . . . . . . . . . 20 3.4.4.1.4 Fracture Toughness Evaluation. ..... 21 3.4.4.1.5 Cask Thermal Stress Analysis . . . . . . 22 3.4.4.1.6 Tornado - Generated Missiles . ..... 23 3.4.4.2 Primary Lid . . . . . ............ 25 _
3.4.4.2.1 Normal Operating Loads . . . . ..... 25 3.4.4.2.2 Environmental Loads and Natural Phenomena. . ........ 25 3.4.4.2.3 Accidents. . . . . . . . . . . . . . . . 25 3.4.4.3 Seconda ry Lid . . . . . . . . . . . . . . . . . 25 3.4.4.3.1 Normal Operating Loads . . . . . . . . . 26 3.4.4.3.2 Environmental Loads and Natural Phenomena. . . . . . .... 26 3.4.4.3.3 Accidents. . . . . . . . . . . . . ... 26 3.4.4.4 Fuel Basket . ................. 26 3.4.4.4.1 Normal Operating Loads . . . . . . . . . 26 3.4.4.4.2 Environmental Loads and Natural Phenomena. . . . . . . . . . 27 3.4.4.4.3 Accident Loading . . . . . . . . .... 27 3.4.4.5 Trunnions. . . . . . . . . . . . . . . . . . . . 27 3.4.4.5.1 Normal Operating Loads . . . . . . . . . 27 3.4.4.6 Primary Lid Bolts . . . . . . . . . . . . . . . 28 3.4.4.6.1 Normal Operating Loads . . . . . . . . . 28 3.4.4.6.2 Environmental Loads and Natural Phenomena. . . . . . .... 28 3.4.4.6.3 Accident Conditions. . . . . . . . . . . 28 3.4.4.7 Secondary Lid Bol ts . . . . . . . . . . . . . . 29 3.4.4.7.1 Normal Operating Loads . . ....... 29 3.4.4.7.2 Environmental Loads and Natural Phenomena. . . ....... 29 3.4.4.7.3 Accident Conditions. . ......... 29 i
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TABLE OF CONTENTS (Continued)
Page 3.4.4.8 Trunnion Bolts. . . . . . . . . . . . . . . . . 29 3.4.4.8.1 Normal Operating Loads . . . . . . . . . 29 3.4.4.8.2 Environmental Loads and Natural Phenomena. . . . . . . . . . 30 3.4,4.8.3 Accident Conditions. . . . . . . . . . . 30 3.4.4.9 Fuel Rods . . . . . . . . . . . . . . . . . . . 30 3.4.4.9.1 Area of Review . . . ... . . . . . . . 30 3.4.4.9.2 Acceptance Criterion . . . . . . . . . . 30 3.4.4.9.3 Review Procedure . . . . . . . . . . . . 31 3.4.4.9.4 Findings and Conclusions . . . . . . . . 31 4.0 Thermal-Evaluation . . . . .............. . . . . . 34 4.1 Normal Conditions . . . . . . . . . . . . . . . . . . . . . . 34 4.1.1 Area of Review . . . . . ..... ... . . . . . . . 34 4.1.2 Acceptance Criteria. . . . . . . . . . . . . . . . . . 34 4.1.3 Review Procedure . .................. 34 4.1.4 Findings and Conclusions . . . . . . . . . . . . . . . 35 4.2 Accident Conditions . . . . . . . . . . . . . . . . . . . . . 35 4.2.1 Explosion. . . . . . . . . . . . . . . . . . . . . . . 35 4.2.2 Fire . . . . . . . . . . . . . . . . . . . . . . . . . 36 4.2.2.1 Area of Review. . . . .. . . . . . . . . . . . 36 4.2.2.2 Acceptance Criteria . .. . . . . . . . . . . . 36 4.2.2.3 Review Procedure. . . . . . . . . . . . . . . . 36 4.2.2.4 Findings and Conclusions. . . . . . . . . . . . 36 i 5.0 Shielding Evaluation . . . . . .................. 38 5.1 Area of Review. . . .
.................... 38 5.2 Acceptance Criteria . . . .................. 38
- 5. 3 Shielding Review Procedure. . . . . . . . . . . . . . . . . . 39
- 5. 3.1 Source Specification . . . . . . . . . . . . . . . . . 39 5.3.1.1 Gamma Source. . . . . .. . . . . . . . . . . . 39 5.3.1.2 Neutron Source. . . . . . . . . . . . . . . . . 40 5.3.2 Model Specification. . . . . . . . . . . . . . . . . . 40 5.3.2.1 Description of the Radial and Axial Shielding . 40 5.3.2.2 Shield Regional Densities . . . . . . . . . . . 41 t
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TABLE OF CONTENTS (Continued)
Page 5.3.3 Shielding Evaluation . . ... . . . . . .. . . . .. 41 5.4 Findings and Conclusions. . . . . . . . . . . . . . . . . . . 42 6.0 Criticality Evaluation . . . . . . ..... . . . ... .. .. . 43 6.1 Area of Review. . . . . . . .. .... . . .... . . . . . 43 6.2 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . 43 6.3 Review Procedure. . . . . . . . . . . . . . . . . . . . . . . 43 6.4 Findings and Conclusions. . . . . . . . . . . . . . . . . . . 44 7.0 Confinement. ................ .. . ... .. . .. 45 m 7.1 Area of Review. .. . . . . ... ......... . .. . . 45
- 7. 2 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . 45 7.3 Review Procedure. . . . . . . . . . . . . . . . . . . . . . .
7.4 Findings and Conclusions. . ...... .. . ... ..
45
. . . 46 7.5 Confinement Requirements for the Hypothetical Accident Conditions . . . . . . . . . . . . . . . . . . . . . 46 7.5.1 Area of Review . . . . ..... . .. .. . ... . . 46 7.5.2 Acceptance Criteria. . . . . . . . . . . . . . . . . . 47 7.5.3 Review Procedure . . . . .... . . .. ... . .. 47 7.5.3.1 Maximum Gaseous Activity Within the Cask. . . . 47 7.5.3.2 Maximum Dose From Gaseous Activity Release. . . 47 7.5.3.3 Maximum Dose From Gaseous and Particulate Activity Release ....... .. .... .. 48 7.5.4 Findings and Conclusions . . . . . . . . . . . . . . . 49 8.0 Operating Procedures . . . . . . . . . . . . . . . . . . . . . . . 51 9.0 Acceptance Tests and Maintenance Program . . . . .. .... . . . 52
, 9.1' Acceptance Tests. . . . . . . . . . . . . . . . . . . . . . . 52 L 9.2 Maintenance Program . . . . . .. .. .... ... ... .. 52 10.0 Radiation Protection. . . . . . . ....... .. .. . .. . . 53 l
10.1 Area of Review . . . .. . ... ... . ... .. ... . . 53 l 10.2 Acceptance Criteria. . . . . . . . . . . . . . . . . . . . . 53 10.3 Review Procedure . . . . . . . . . . . . . . . . . . . . . . 54 10.3.1 Ensuring that Occupational Radiation Exposures are As Low As Is Reasonably Achievably (ALARA). . .. 54 l
10.3.2 Radiation Protection Design Features. . .. . .. . . 54
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10.3.3 Estimated Onsite Dose Assessment . . . .. . . .. . 55 P
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TABLE OF CONTENTS (Continued)
Page 10.4 Findings and Conclusions . . . . . . . . . . . . . . . . . 56 11.0 Accident Analysis . . .. . . . . . . . . . . . . . . . . . . . . 57 11.1 Area of Review . .. . . . . . . . . . . . . . . . . . . . . 57 11.2 Acceptance Criteria. . . . . . . . . . . . . . . . . . . . . 58 11.3 Review Procedure . . . . . . . . . . . . . . . . . . . . . . 58 11.3.1 Off-Normal Operations . . . . . . . . . . . . . . . . 58 11.3.1.1 Event. . . . . . . . . . . . . . . . . . . . . 58 11.3.1.2 Radiological Impact frrm ,
Off-Normal Operations. . . . . . . . . . . . . 58 11.3.2 Accidents . . . . . . . . . . . . . . . . . . . . . . 59 11.3.2.1 Accidents Analyzed . . . . . . . . . . . . . . 59 11.4 Findings and Conclusions . . . . . . . . . . . . . . . . . . 60 12.0 Decommissioning . . . . . . . . . . . . . . . . . . . . . . . . . 61 12.1 Decommissioning Plan . . . . . . . . . . . . . . . . . . . . 61 12.2 Decommissioning Considerations . . . . . . . . . . . . . . . 61 12.2.1 Cask Body Activation. . . . . . . . . . . . . . . . . 61 12.2.2 Unloading the Cask. . . . . . . . . . . . . . . . . . 62 12.2.3 Findings and Conclusions. . . . . . . . . . . . . . . 62 13.0 Operating Controls and Limits. . . . . . . . . . . . . . . . . . 63 13.1 Area of Review . . . . . . . . . . . . . . . . . . . . . . . 63 13.2 Acceptance Criteria. . . . . . . . . . . . . . . . . . . . . 63 1~.3 Review Procedure . . . . . . . . . . . . . . . . . . . . . . 63 13.4 Findings and Conclusions . . . . . . . . . . . . . . . . . . 63 14.0 Quality Assurance . . . . . . . . . . . . . . . . . . . . . . . . 64 15.0 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . 65 I
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.- 2 TABLE OF CONTENTS (Continued)
LIST OF FIGURES Page 1.2-1 CASTOR Ic Cask Completed . ......... .......... 3 1.2-2 CASTOR Ic Cask . . . . . . . . . . . . . . . . . . . . . .... 4 m
vii
Safety Evaluation Report of:
Topical Safety Analysis Report (TSAR) for CASTOR Ic Cask Independent Spent Fuel Storage Installation (Dry Storage), Revision 3, September 1984.
- 1. 0 General Description 1.1 Introduction This Safety Evaluation Report (SER) documents the. staff's review and evaluation of the Topical Safety Analysis Report (TSAR) for CASTOR Ic Cask ~
Independent Spent Fuel Storage Installation (Dry Storage) Revision 3, September 1984. (Ref. 1). The TSAR was originally generated in the Reg. Guide 3.48 (Ref. 2) format as applicable. This SER utilizes the format of Reg. Guide 3.
(CE-306-4) (Ref. 3) with some differences in the section numbering.
The staff's review of this TSAR addresses the handling, transfer and storage of spent fuel for an at reactor site independent spent fuel storage installation (ISFSI). Such storage in an ISFSI would be licensed under 10 CFR Part 72, " Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation (ISFSI)." In this TSAR a single dry storage cask design, the model CASTOR Ic is presented. The term CASTOR is an acronym for Cast Iron Cask Storage and Transport of Radioactive Materials.
The staff's assessment is based on the proposed design's meeting the applicable requirements of 10 CFR Part 72, found under Subpart E, " Siting Evaluation Factors," and Subpart F, " General Design Criteria," and of 10 CFR Part 20 for radiation protection for o site receipt and storage of spent fuel in an ISFSI. Decommissioning, to the extent that it is treated in this TSAR, presumes unloading of a CASTOR Ic cask at the site reactor (s) and subsequent decontamination of the cask prior to its disposition or disposal. Use or certi-fication of CASTOR Ic cask under 10 CFR Part 71, for offsite transport of radio-active materials including spent fuel, is not a subject of this safety evaluation.
1
This review also does not address requirements for physical protection under Subpart H, " Physical Protection," of 10 CFR Part 72 or under 10 CFR Part 73, " Physical Protection of Plants and Materials."
1.2 General Description of the Storage Cask 1.2.1 Cask Design Characteristics The CASTOR Ic Cask (see Figure 1.2-1) was developed by the Gesellschaft fur Nuklear-Service mbH (GNS), and is designed for the storage and shipment of irradiated spent fuel assemblies. The cask fulfills the International Atomic Energy Agency international specifications for Type B(U) packaging correspond-ing to Nuclear Safety Fissile Class I. However, as noted above, this review addresses spent fuel handling, transfer, and storage on a NRC-licensed nuclear reactor site and not any use or certification of this cask design for offsite transport of radioactive materials.
The CASTOR Ic is a thick-walled cast ductile iron cask that is approx-imately 5.56 m (218.9 in) high, 1.72 m (67.7 in) in diameter, and weighs approx-imately 80 tonne (88.2 ton). The cask has a rectangular cask cavity which holds a fuel basket and is designed to accommodate 16 BWR fuel assemblies.
The cask is sealed with two lids bolted to the cask body with 36 bolts and one protection plate which are installed one on top of the other. Both lids are sealed with multiple seals consisting of metal seals and elastomer 0-rings.
Four trunnions are bolted on, two at the head end and two at the bottom end of the body. (See Figure 1.2-2).
1.2.2 Operational Features The CASTOR Ic cask consists of a thick-walled ductile fron casting. The overall length is 5560 mm (218.9 in) (including protection plate), and the side wall thickness without fins is 440 mm (17.3 in). The cross-section of the rec-tangular cask body is 1720 mm (67.7 in). The cask cavity has a square form, with a width of 666 mm (26.2 in) and a length of 4560 mm (179.5 in).
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O e For neutron shielding, two concentric rows of axial holes in the wall of the cask body are filled with polyethylene rods. The bottom and the secondary cover each have a slab of the same material inserted for the same purpose.
The body of the cask is cast in one piece. The material used is ductile cast iron (referred to as nodular cast iron in the TSAR). This material exhibits higher ductility than grey cast iron and has high resistance to corrosion.
In the area of the fuel assemblies, the body has cooling fins on the outside; four trunnions, two each at the top and bottom ends, are attached with bolts.
~
The cask is sealed, to maintain a helium atmosphere, with a multiple-cover system consisting of a primary lid, a secondary lid, and a protection plate.
The primary lid is constrected of stainless steel. The overall thickness is 340 mm (13.4 in). It is fastened to the body with 36 bolts. The primary lid has three penetrations, used for flushing and venting of the cask cavity as well as the performance of the leak test. The flushing and venting connections are sealed with separate lids. One leak test connection is sealed with a bolt closure. The secondary lid is also made of stainless steel. The overall thick-ness is 130 mm (5.1 in) including neutron-moderating material. It is bolted to the body. The protective cover plate is made of carbon steel and is bolted to the cask body. It provides general mechanical protection against actions from outside, as well as against dust and humidity. A combination of multiple elast-omer and metal seals for each lid provide leak tightness. However, no credit is claimed in the TSAR (see Section 3.3.2.2) or given in this evaluation for elastomer seals for the 20 year storage period.
The fuel basket accepts the spent fuel assemblies and ensures that criticality will not occur. In addition, it ensures exact positioning of the individual fuel assemblies. It is of welded construction and is made of stainless steel and borated stainless steel sections. At the top end of the cask there is a flushing connection for rinsing, cleaning, and drying of the interior during loading and unloading procedures at the nuclear power plant.
The flushing channel runs inside the wall of the body; it has one end at the 5
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top and the other er.d at tile bottom of the inside of the cask. Gas intake and exhaust are v.ia the valve in the p'rimary lid. The lid system is fitted with a leak-testing de' ice a pressure gauge, which is also a cask component classified as important to safety in Section 3.4 of the TSAR.
The inside~ of the cask, including the sealing surface, has a nickel coating for corrosion protection. On the outside, the cask is protected by an epoxy resin coating in the fin area and nickel coating elsewhere. The internal heat-transfer medium is an inert gas (helium), which also serves to inhibit corrosion. Gamma and neutron radiation is shielded by the. cast iron wall of the cask, which includes rods of neutran moderating material (polyethylene).
1.2.3 Cask Contents The type of spent fuel to be storad in the CASTOR Ic cask is light water reactor (LWR) fuel of the boiling water reactor (BWR) type. BWR fuel is made of short cylinders (pellets) of high-fired ceramic uranium dioxide (UO2)-
Depending upon the specific GWR-reactor design, these pel'Tets are in the order of 1.25 cm (0.49 in) to 1.45 cm (0.57 in) in diabeter and about 1.50 cm (0.59 in) long. Typically a 366 cm (144.1 in) long stack or about 250 of these pellets are loaded and hermetically sealed into a zirconium alley tube.
Fuel rods are assembled into bundles in a square array, each spaced and supported by grid structures.
The assembly has a t6ttom fitting and a top fitting with a handle.
Typically, a BWR assembly consiith of a 7 x 7 (49 total) or 8 x 8 (64 total) array of individual rods. The overall dimensions are approximately 14 cm (5.51 in) square by 432 cm (170 1 in) long. Each assembly contains about 200 kilograms (440 lbm) of uranium in the form of UO2 -
1.3 Identification of Agents and Sub' contractors General Nuclear Systems, Inc. (GNSI) is a United States c6mpany' incorporated in the State of Delaware. GNSI is:an affiliation of th'e-United States company - Chem-Nuclear Systems, Inc. (CNSI) and the West Germpny company, 6 -
1 Gesellschaft fur Nuklear-Service mbH (GNS). An executive committee represented by members of both GNS and CNSI act to coordinate the staff
, and resources of their respective companies to support the activities of
, GNSI.
r'*
The shareholders of Gesellschaft fur Nuklear-Service mbH, (GNS) are:
STEAG Kernenergie GmbH ,
45%
-VKR, VEBA KRAFTWERKE RUHR 27.5%
DWK, Deutsche Gesellschaft fur g
Wiederaufarbeitung von Kernbrenn-stoffen mbH 27.5%
STEAG Kernenergie is working in the nuclear fuel cycle and performs design and project work in the planning and construction, of nuclear facilities.
VKR, the owner and operator of several coal-fired power stations, is a fully owned subsidiary. of the VEBA concern which owns several nuclear- power stations.
- g. DWK was f unded by the twelve leading German (nuclear based) utilities for the construction of the German Integrated Nuclear Fuel Cycle Center,. and operates-the,small-scale reprocessing facility,,WAK, in Karlsruhe.
1 The following consultants, testing labs, technical inspection and super-vision associations, governmental institutes,.etc., are involved besides GNS.
Federal Institute f$r Materials Testing, Berlin.
Performance of,the cask testing according to the packaging require-ments and as consultant for the German licensing authority, PTB.
Federal Physical and Technical Institute, Braunschweig.
Licensing authority for dry cask storage and transportction of spent nuclear fuel.
Materials ~ Testing Institute operated by the state North Rhine -
- Westphalia,Dbrtmund. Performance of corrosion tests.
Reactor Safety Association, Munich. Performance of the criticality and shielding analysis.
7
4
. Technical Inspection Association, Hannover. Consultant to the German licensing authority, PTB.
. Reactor Safety Commission, Bonn. Governmental consultant for the
~
Minilter of the Interior, responsible for the storage license.
EXXON ~ Nuclear Company, Inc. Washington. Spent Fuel behavior under
~ dry storage conditions.
Fracture Control Corporation, Goleta, CA. Materials analysis.
SWANSON SERVICE Corporation, Huntington Beach, CA. Performance of stress analysis, ANSYS calculation.
e 1.4 Geheric Cask Arrays The TSAR and this review cover only a single CASTOR Ic cask. However, a
- description of the CASTOR Ic cask in three different generic arrays was provided in the TSAR:
l
- a. Single Cask
- b. Line Array
- c. Rectangular Array In a site-specific licensing case radiological aspects of an array of casks with respect to 10 CFR Part 72 requirements regarding doses received off the site of a dry cask ISFSI would have to be addressed by the license applicant. -
Although this TSAR covers only a single CASTOR Ic cask, the criticality considerations include the close packed infinite rectangular array as the most reactive.
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2.0 Principal Design Criteria e
2.1 Introduction.
Subpart F of 10 CFR Part 72 sets forth general design criteria for the design, fabrication, construction, testing and performance of structures, systems and components important to safety in an independent spent fuel storage' installation (ISFSI). In this chapter, we discuss the applicability of these criteria to the CASTOR Ic spent fuel storage cask and the degree to which the GNS TSAR is in compliance with these criteria. The headings in this
- chapter correspond to the sub-sections of Subpart F. ~
2.2 Fuel to be Stored The. CASTOR Ic cask is designed to store in a dry condition irradiated BWR fuel from nuclear power stations. The design basis fuel is UO2 with an initial enrichment of 2.4% U2ss or-less, clad in Zircalloy. The design basis fuel is assumed to have been irradiated at power loads up to 22.5 MW/MTU to an exposure of 27000 mwd /MTU and cooled sixteen months. Estimates of the radionuclide activity in spent fuel described above were made using the ORIGEN computer code.
2.3 Quality Standards Quality standards for structures, systems and components important to safety are required by 10 CFR Part 72.72(a). A quality standard provides
. numerical criteria or acceptable methods or both for the design, fabrication, testing,-and performance of these structures, systems and components.important
, to safety. These standards should be. selected or developed to provide sufficient confidence in the capability of the structure, system, or component to perform the required safety function. Since quality standards are generally
' embodied in widely accepted codes and standards dealing with design procedures, materials, fabrication techniques, inspection methods, etc., opinions regarding the-adequacy of the standards cited by the CASTOR Ic TSAR are presented in the sections of this report where the standards are applicable.
9
2.4 Protection Against Environmental Conditions and Natural Phenomena Part 72.72(b) of 10 CFR Part 72 requires the licensee to provide protection against environmental conditions and natural phenomena. Section 3.2 of the TSAR describes the structural and mechanical criteria for tornado and wind loadings, tornado missile protection, flood potential, seismic design, snow and ice loadings, structural design criteria, material properties, cask body loading due to internal pressure, loads under operating and boundary conditions, loading combinations, and thermal loads.
In this section, the discussion is limited to the adequacy of the criteria for protecting against environmental conditions and natural phenomena.
The other loading conditions are considered in the chapters of this report where they are relevant. The technical basis for accepting these criteria is defined by the regulatory requirement to consider the most severe of the natural phenomena reported for the site with appropriate margins to take into account the limitations of the data. Since the CASTOR Ic cask was not designed for a specific site, the regulatory requirement is interpreted to mean that protection against environmental conditions and natural phenomena should be provided for either by the limits specified in the TSAR or for the most severe of the natural phenomena that may occur within the boundaries of the United States.
The TSAR establishes 579 km/h (360 mph) in Section 3.2.1 as the design basis tornado wind speed. This is in accordance with Reg. Guide 1.76 (April' 1974).
A horizontal acceleration of 0.25g was established as a basis for seismic design in Section 3.2.3. This peak acceleration reflects an acceleration value recognized as acceptable in 10 CFR Part 72.66(a)(6)(ii) for ISFSI sites east of the Rockies. Therefore, the staff accepts this value as a basis for seismic design. However, the staff interpreted that the TSAR analysis showed this re-quirement as referring only to one direction. This is not appropriate. For a correct analysis, this acceleration should be combined vectorially with a com-ponent normal to this direction resulting in a maximum horizontal ground accel-eration of 0.35 g. In addition, Reg. Guide 1.60 requires that the vertical acceleration used be 2/3 of horizontal acceleration so that 0.17 g is the i
10
acceleration in the vertical direction. Therefore, the staff has accepted a horizontal acceleration of 0.25g, but has applied it in this evaluation accord-ing to appropriate NRC staff analytical procedures (see NUREG-0800, Section 3.7.1) as summarized above.
2.5 Protection Against Fire and Explosions Pursuant to 10 CFR Part 72.72(c) the licensee is required to provide pro-tection against fires and explosions. The TSAR establishes the design basis fire of 800*C (1472*F) for one half hour duration. This is a basis established
~
for Type B shipping casks under 10 CFR Part 71, Section 71.73, " Hypothetical accident conditions," Subsection 71.73(a)(3), " Thermal." As such, it consti-tutes an upper bound that is unlikely to be exceeded within a nuclear power plant site. The design basis explosion was established at 30,000 MPa (4.25 x 108 psi) for a duration of 0.2 second. A gas cloud explosion is assumed as a poten-tial mechanism. This is an unlikely event and a conservative basis for assuming an explosion on a power plant site. There are no explosive materials associated with the CASTOR Ic cask itself. The staff concurs that these are reasonable design criteria.
- 2. 6 Confinement Barriers and System Pursuant to 10 CFR Part 72.72(h), the licensee must protect the fuel cladding against degradation and gross rupture. Requirements for underwater storage of spent fuel differ from those applicable to dry cask storage and consideration of ventilation and off gas systems are site specific.
The TSAR provides analysis and cites data supporting the case that dry storage of spent fuel does not lead to degradation and gross rupture of cladding. The TSAR does not adequately address design criteria relating to the protection of fuel cladding for the maximum initial cladding temperature of 400*C (752*F). In view of this situation, the reviewers conducted an investigation directed toward determining whether the TSAR storage conditions specified were adequate. For protection to be adequate, the design of the cask should be such that degradation after a twenty year storage life should not preclude the ability of the cladding to resist gross rupture during normal 11
, o operations associated with decommissioning nor during subsequent fuel rod handling operations.
After reviewing the current research relating to spent fuel cladding dam-age mechanisms, the reviewers concluded that a diffusion controlled cavity growth (DCCG) mechanism was the only mechanism of damage for dry storage'appli-cable to the storage conditions of the fuel rods that could cause degradation and gross rupture of.the cladding. Under the influence of stress and tempera-ture, this damage mechanism progresses by the nucleation and growth of cavities along grain boundaries. This damage mechanism is serious since it can progress ,
without external evidence of damage, may not cause pin holes or through cracks to relieve the internal pressure, and manifests itself by a sudden non-ductile type of fracture. The staff has therefore paid particular attention to evaluat-ing the potential for cladding damage from this mechanism for the conditions of storage specified in this TSAR.
The only parameters that the cask designer may control to prevent cladding degradation or gross rupture in an inert environment are the maximum initial temperatures of the fuel rods and their temperature decay characteristics. Both are governed by the quantity, specific power, and age of the fuel assemblies,,
and by the heat dissipation properties of the cask. The TSAR addresses the thermal characteristics of the cask and fuel cladding integrity in Section 3.3.7.1, which notes DCCG is a potential mechanism for cladding failure,
-and Sections 5.1.3.6 to 5.1.3.8. This SER addresses the thermal evaluation in Chapter 4 and fuel cladding integrity in Section 3.4.4.9 The TSAR further established a design basis leakage rato for each double seal of 10 8 mbar 1/sec. The staff considers this leakage rate to be accept able for design.
2.7 Instrumentation and Control Systems Pursuant to 10 CFR Part 72.72(i), the licensee must provide instrumentation and control systems which monitor systems important to safety over anticipated ranges for normal and off-normal operation. The CASTOR Ic cask incorporates a 12
pressure sensitive switch monitor gauge which serves as a cask tightness surveillance system. The design criteria and description of this system appears in Section 3.3.3.2 of the TSAR and operation controls and limits in Section 10.1.2.2.
2.8 Criteria for Nuclear Criticality Safety Section 72.73 of 10 CFR Part 72 requires that spent fuel handling, transfer and storage systems be designed to be maintained subcritical. The margins of safety should be commensurate with the uncertainties in the handling, transfer c
and storage conditions, in the data and methods used in the calculations, and in the immediate environment under acc'ident conditions. Section 72.73 also requires that the design be based on either favorable geometry or permanently fixed neutron-absorbing materials. Section 3.3.4 of the CASTOR Ic TSAR addresses nuclear criticality safety criteria. Criticality analysis and prevention are reviewed in Chapter 6 of this report.
The TSAR establishes a maximum effective multiplication factor of 0.95 for all credible configurations and environments for the prevention of criticality.
This factor 's widely accepted as a criticality prevention limit, and the staff concurs with its application to CASTOR Ic cask.
2.9 Criteria for Radiological Protection Section 72.74 of 10 CFR Part 72 requires that the licensee provide adequate (a) protection systems for radiation exposure control, (b) radiological alarm systems, (c) systems for monitoring effluents and direct radiation, and (d) effluent control systems in a radiological protection program. Section 3.3.5 of the CASTOR Ic TSAR addresses radiological protection. The detailed evaluation for compliance with the regulation is discussed in Chapters 5, 7, and 10 of this SER.
The principal design features of the CASTOR Ic cask for exposure control are the inherent shielding capability of the cask and the integrity of the seals at the closure joints. Radiological alarm systems and systems for monitoring effluents and direct radiation are not applicable to the design of the storage 13
cask.~ Effluents are not a normal consequence of-the passive dry storage operation; so control systems, to provide radiological protection for this condition, are also~not applicable. Only provision (a) above is applicable to the cask with respect'to shielding capability and the possibility of leakage from seals that may degrade or suffer damage as a result of an accident.
However, it should again be noted, as in Section 2.7 above, that the double-lidded system of the cask uses a pressure measuring system as a tight-ness surveillance system. This is addressed in Section 3.3.5.3.3 of the TSAR.
A pressure of 6 bar is maintained in the space between the primary and secondary lids. A pressure drop of 3 bar will result in a signal of a leak. Thus, in
-the-unlikely event of failure of the seal on either one of the lids, personnel would be alerted by this system allowing the seal to be replaced on the lid concerned.
The shielding capability of the cask relies primarily upon the thickness and attenuation property of the ductile cast iron casting and steel closure lids which comprise the primary barriers to radiation. The cask must maintain its structural integrity under loadings associated with normal operation, accident events, natural phenomena, and environmental conditions. Of particular concern is the response of the ductile cast iron to dynamic loading conditions associated with cask drop and/or tip over. Because of the low ductility values specified for the CASTOR Ic casting, it is essential to demonstrate that its fracture toughness is sufficient to resist catastrophic brittle fracture under i
- the assumption that undetected flaws may exist at locations of maximum primary membrane or bending stress. Resistance to brittle fracture is discussed in
- . Section 8.2.1.2.4 of the TSAR, and a review of this topic is presented in Section 3.4.4.1.4 of this SER.
The TSAR also establishes the surface and two meter dose limits as 200 mrem /h.and 10 mrem /h, respectively. The staff believes that these limits are acceptable provided the site boundary for a single cask is not less than 100 meters (328 feet).
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2.10' Criteria for' Spent Fuel and Radioactive Waste Storage and Handling Pursuant to 10 CFR Part 72.75, the licensee is required to design the spent fuel storage and waste storage systems to ensure adequate safety under normal and accident conditions. These systems must be designed with (a) a capability to test and monitor components _important to safety, (b) suitable shielding for radiation protection under normal and accident conditions, (c) confinement structures and systems, (d) a heat removal capability having testability and reliablity consistent with its importance to safety and (d) means + o minimize the quantity of radioactive wastes generated.
m This section of the regulation defines the requirements for the spent fuel storage system within the context of the entire ISFSI, and compliance of the entire TSAR with this criterion needs to be documented. The TSAR addresses this requirement in summary fashion in Section 3.3.7 and refers the discussion of compliance with the details of this section of the regulation to other sections of the report.
2.11 Criteria for Decommissioning Pursuant to 10 CFR Part 72.76, the licensee is required to design the ISFSI for decommissioning. Since the dry storage cask is an integral part of the ISFSI, this requirement must be met by the cask design itself. Provision should be made to include decontamination of the cask components and removal of the radioactive spent fuel with generation of radioactive wastes and with contamination of equipment held to a minimum.
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3.0 : Structural Evaluation 3.1. Area of Review This' chapter evaluates the structural response of the CASTOR Ic cask to loadings under normal operating conditions, accident conditions and loads due to environmental conditions and natural phenomena. The review procedure addresses the assumed loads and material properties, the allowable' stress limits and an evaluation of the structural analysis for each of the components and systems important to safety.
3.2 Acceptance Criteria The structural integrity of the cask will be deemed adequate if it can be demonstrated that the stresses induced by the loads described in 3.1 above are lower than the allowable stress limits for the cask components important to safety.
3.3 Review Procedure The TSAR was reviewed for compliance with 10 CFR Part 72.72 (a) which refers to quality standards that govern the characterization of materials,.the establishment of stress intensity limits, and the design and analysis methods that provide confidence in the capability of the structure, system or component to perform the required safety function. The TSAR was also reviewed for com-
. pliance with 10 CFR Part 72.72 (b) which requires that protection against environmental conditions and natural phenomena be demonstrated; for compliance l
i with 10 CFR Part 72.72'(c) which requires that protection against fires and explosions be demonstrated; and for compliance with 10 CFR Part 72.72 (k) which requires that protection of fuel cladding against degradation and gross rupture be demonstrated.
3.4 Findings and conclusions i
3.4.1 Loads i
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'4
- .= .-
- 3. 4.1.1 ~ Normal Operating Conditions The TSAR specifies in Section 3.2.5.4 the design loads for normal operating conditions. They are 723,949 Pa (105 psig) internal pressure, 1.4 to 3 g
-transport and handling loads, and a 17.6 kW cask thermal load.
- 3. 4.' 1. 2 Loads Due to Environmental Conditions and Natural Phenomena
-TheLdesign basis loads due to environmental conditions and natural phenomena ~are summarized in Section 2.4 of this SER. The staff used.0.35 g horizontal acceleration plus an upward acceleration of 0.17 g to determine whether' the -cask would tip as a result of an earthquake.
'3.4.1.3 Load Due to Postulated Accidents 10 CFR Part 72.72 (b)(1) requires-that the cask be designed to accommodate the. effects of postulated accidents. The TSAR describes these postulated accidents in Chapter 8. The loads due to these accidents arise as a result of impact due to handling accidents or gas cloud explosion. -The handling accident
. assumed in the TSAR is a 1.83 m (6 foot) drop in various orientations on to a 610 mm (24 inch) thick concrete slab resting on earth with a subgrade modulus of 500 psi /in. The staff accepts six feet as a reasonable drop height provided that steps are taken to assure that the cask is not lifted to a height greater than six feet when it is moved outside the reactor for transferral to and emplacement at an-ISFSI.
i
-3.4.2 Materials I
- The standards referenced in the TSAR for most materials used in the fabrication of all components important to safety are Deutsche Industrie Normen (DIN)-standards. Where a material is not covered by a standard, such as the ductile cast iron and borated stainless steel, a specification, in American Society for Testing Materials (ASTM) format, was provided in the TSAR
, . appendices. In general, where the materials conform to the limits set by the
. specification -the DIN standards are considered by the reviewers to be quality I
- standards in compliance with 10 CFR Part 72.72 (a).
17
T .
However, the ductile cast iron in the thickness range specified for the CASTOR Ic cask is not described by any authoritative standard or specification.
In characterizing this material for design and procurement, GNS has provided, in Appendix 3 of the TSAR, a specification for the ductile cast iron in the format required by the ASTM when submitting a standard for adoption by the Society. This follows-the rules outlined in Article IV-1000 of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPV) for inclusion of materials properties in the Code. While the ASME requires the ASTM standard be published before inclusion in its Code, inclusion of the standard for this material in the ASME Code is not required by the NRC c
staff for them to conduct their review. Inclusion, while desirable, is a matter for the ASME and the applicant to consider and could delay the review process for many years. Consequently, the reviewers have chosen to evaluate the ade-quacy of the material in meeting all safety criteria of 10 CFR Part 72 where its capability is of issue on the basis of the ASTM-type specification submitted.
The justification for this procedure is that in providing an ASTM-type specification, the applicant establishes a commitment in this TSAR to provide material with stated minimum requirements and rejects material that does not meet those requirements in accordance with specified tests and test procedures.
This material specification is a part of the cask design as specified in the TSAR.
3.4.3 Stress Intensity Limits The TSAR lists in Tables 3.2-4(a) and (b) material properties and stress intensity limits for normal operating conditions, as a function of temperature, for all components important to safety. In general, where the ductility of the materials is sufficient, the stress intensity limits are in accordance with the
-standards established by the ASME BPV Code. Consequently, they conform to the quality standard requirement of 72.72 (a). On the other hand, the ductility of the CASTOR Ic cast iron, as reflected by the specified 4% minimum elongation, is not sufficient to qualify it for the design by analysis rules of Section III of the ASME BPV Codes. Neither can the safety factors recommended in Article III-2110 of the Code be used to determine the stress intensity limits.
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Thus, the use of a safety factor of four in the ultimate tensile strength spec-ified in Section 3.2.5.3 of the TSAR is prudent since it is higher than the factor of three recommended by the ASME BPV Code for ductile ferritic materials.
On the other hand, since the CASTOR Ic cask will be subjected to analysis for thermal stress, fatigue failure, and brittle fracture; and since a high level of non-destructive examination for flaws is specified in the proposed material standard; a safety factor of 5 recommended by Section VIII of the Code for duc-tile cast iron is too conservative. Furthermore, the structural design criteria for the cask body described on page 3.2-11 of the TSAR imply that under both normal operation and accident conditions, for all type of stresses, the maximum a
allowable normal stress remains well under the specified minimum yield strength.
Consequently, these criteria are acceptable for designing the CASTOR Ic cask.
The apparent lack of ductility of the CASTOR Ic cast iron requires that its response to dynamic loads be determined for the case where a flaw is present in the most critically stressed location. The design criterion for this condition is a combination of stress and flaw size that does not result in a fracture toughness stress intensity factor, K g
, greater than the critical stress IC. The level of K IC is defined in the proposed specification as intensity, K b
a minimum of 45 ksi-in at -40*C and is to be determined in accordance with ASTM E813.
3.4.4 Structural Analysis 3.4.4.1 Cask Body 3.4.4.1.1 Normal Operating Loads The cask body was analyzed for an internal pressure of 7 bar (105 psi) using a finite element code as described in Section 4.2.1.1 of the TSAR. The maximum stress was 5.36 MPa (777 psi) which is far below the allowable stress intensity limit of 54.47 MPa (7900 psi) for the cask body.
During truck transport, the cask rests on four trunnions in a horizontal position. A multiplie" of two on the cask weight is applied, reflecting the 2 g load specified in Section 3.2.5.4 of the TSAR. A simple beam analysis shows 19 .
i that the maximum stress in the cask to be 2.05 MPa (297.7 psi) which is below the stress intensity limit.
During handling by crane, the cask is supported in a vertical position on two trunnions. A finite element stress analysis, described in Section 4.2.1.1 of the TSAR, showed that the highest region of stress occurs just above the trunnion, and is 13.71 MPa (1,989 psi). The combination of pressure and handling loads is below the stress intensity limit.
3.4.4.1.2 Environmental Conditions and Natural Phenomena a
As a result of the design basis tornado wind loads and seismic loads, the staff concludes that the cask will suffer a tip over. This will give rise to dynamic loads due to impact. The TSAR addresses the tip over event in Section 8.2.1.2.3. The TSAR dismisses the necessity for performing a tip over structural analysis on the basis that the kinetic energy associated with tip over is less than that associated with the 1.83 m (6 foot) cask drcp. The staff believes that this is not necessarily true since the center of gravity of the cask will fall during tip over through a distance of 2.05 m (6.73 ft.),
and the top of the cask will fall through a distance of 5.56 m (18.24 ft.).
Furthermore, the cask may approach the balanced condition with a finite velocity which adds to the free fall energy of impact. However, the staff believes that there is sufficient margin demonstrated by the accident condition involving cask drop to absorb a higher load magnitude for tip over.
Stresses shown for the cask drop accident conditions in Section 8.2.1.2.3 of the TSAR are fractional values, in the case of the cask body, of a factor of 2.4 times the design s. tress intensity (2.45 S,), 19,000 psi, and, in the case of the primary lid, 0.7 times the ultimate stress intensity (0.7 Su), 79,200 psi; where the smaller of either 2.45 S, or 0.7 S uwas chosen in each cask for the cask body and the lid as a limit in Section 3.2.5.3 of the TSAR.
3.4.4.1.3 Accident Conditions The TSAR describes analysis of the cask body for accident conditions involving a 1.83 m (6 foot) drop in Section 8.2.1.2.3. The analysis for impact 20 t
was performed in two steps. First, the bounding force for failure of the assu.ed concrete slab was determined. This force was then applied to the cask as a static load. The staff believes that this is a valid approach. However, the model of the slab shown in Fig. 8.2-4 did not include reinforcing steel which would increase its stiffness. Consequently, the staff performed a
[ confirmatory analysis using 2% tension steel which showed that the load intensity of 34.47 MPa (5000 psi) applied to the cask for end and side drops, including tip over, is still bounding. For the corner drop, however, the confirmatory ar.alysis indicates that the maximum stress will be higher than that reported in Table 8.2-1 but will still be below the appropriate stress a
intensity limit.
3.4.4.1.4 Fracture Toughness Evaluation In this section, the fracture resistance of the ductile cast iron used for the CASTOR Ic cask is evaluated for its response to the postulated environmental and accidental conditions described above.
The fracture toughness of the ductile cast iron will be considered adequate if, for the maximum stresses applied, the minimum toughness specified in Appendix 3, and for a minimum detectable flaw size of 10 mm (0.39 in),
fracture initiation does not occur for a hypothetical flaw no less than twice
, the size of the minimum detectable flaw 20 mm (0.79 in).
A confirmatory analysis was performed using linear elastic fracture mechanics methods assuming a flaw depth of 20 mm (0.79 in.), a flaw aspect ratio of 1/6, an ideally sharp crack and a maximum stress of 108.2 MPa (15.7 ksi).
For the maximum stress level of 108.2 MPa (15.7 ksi) computed for the postulated accident conditions, the maximum anticipated flaw in the cask and the minimum fracture toughness property of 45 ksi in., failure by brittle fracture is not expected. The conservatism inherent in the analysis stems from a factor of two on the minimum detectable flaw size, the condition that the flaw coincides with the point of maximum stress in an orientation normal to applied stress, and that the flaw is an ideal sharp tipped crack. Further conservatism is indicated by the more elaborate analysis provided in the TSAR 21
based upon the work of Newman and Radju which shows that at the 1-inch depth the fracture toughness stress intensity about the crack varies only from 17.2 to 22 ksi - in b.
3.4.4.1.5 Cask Thermal Stress Analysis The thermal stress analysis for the CASTOR Ic cask was reviewed to ensure that the containment would not fail under the assumed loading conditions. The TSAR assumed that in a one year period the following thermal cases can occur:
(1) 30 cycles of cooling from 21 + -18*C (70 + 0*F) with a 96.6 km/h (60 mph)
~
wind; (2) 5 cycles of cooling from 21 + 0*C (70 + 32*F) with 127 mm (5.0 inches) of rain and a 96.6 km/h (60 mph) wind; and (3) 300 cycles ranging from 10 + 32*C (50*F to 90*F).
The requirement for structural integrity can be met if, by using ASME code methods, it is demonstrated that the maximum primary plus secondary stress is less than the 2/3 yield stress for ductile cast iron and that the fatigue usage factor due to thermal cycling is less than one. The ANSYS 3D computer code was used*by the applicant to calculate the maximum transient temperature gradient and thermal stress for each of the three thermal cases. The worst case, number 2, resulted in a primary plus secondary stress of 86.9 MPa (12,600 psi) which is more than 30% lower than the allowable stress of 127.8 MPa (18,530 psi) for ductile cast iron. The fatigue evaluation considered the three cycle cases occurring every year for a thirty year storage period. A stress concentration factor of 2.48 was used to account for the discontinuity stresses in the fin regions. The calculated fatigue usage factor of 0.174 is well below the allowable factor of one.
Based on the review of the thermal stress analysis in the TSAR, it was concluded that the cask containment will not fail. The thermal analysis in the TSAR may be referenced in a site-specific license application provided that the site environmental conditions are within the three thermal cases analyzed.
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3.4.4.1.6 Tornado - Generated Missiles Tornado generated missiles that may damage the cask are described in NUREG-0800. All missiles are assumed to impact the cask at 35% of the maximum windspeed, which is defined in Reg. Guide 1.76 to be 360 miles per hour. Thus the maximum missile velocity is 126 miles per hour. The point of application and orientation of the missile is that which can cause the greatest amount of damage. Types of missiles described include:
- 1. A massive high kinetic energy object that is deformable upon impact with the cask. This may be represented by an 1800 kg automobile. ~
- 2. A rigid missile that tests the penetration resistance of the cask as represented by a 125 kg, 20.32 cm (8") armor piercing shell.
- 3. A small rigid object such as a solid steel sphere 2.5 cm in diameter which may pass through any openings in the protective barriers.
In accordance with the criteria for radiological protection described in Section 2.9 of this SER, the cask must maintain its structural integrity under the impact of tornado generated missiles.
The TSAR addresses the subject of tornado generated missiles in Section 8.2.1.2.2 based upon NUREG-0800 Rev. 2, " Standard Review Plan, Missiles Generated by Natural Phenomena." The safety analysis for the effect of the massive deformable object was based upon a test involving a projectile weighing 1000 kg traveling at a speed of 300 meters per sec, hitting the canister on its side. This projectile was modeled, to represent an aircraft jet engine and its kinetic energy at impact of 45 x 108 Nm which is much higher than the 28.6 x 105 NM generated by the automobile at 126 mph. The projectile consisted of a 5 mm thick, 600 mm diameter metal jacket enveloping a 255 kg steel tube. This configuration could model an automobile with the sheet metal envelope representing the body and the steel tube representing the engine block. While it is lighter than the automobile (1000 kg vs 1800 kg), the higher velocity (300 m/sec vs 56 m/sec) 23
- z. -.-
makes it a far more destructive missile. The TSAR affirms that the experi-mental tests were performed on a prototype cask (not the CASTOR Ic). with a wall thickness of 15. inches,-less-than 17.32 inches wall thickness of
-the. CASTOR Ic cask body, and that the structural integrity of the cask
.was not violated. It can be concluded, therefore, that the CASTOR Ic will
,sust'ain side impact of an automobile'at 126 mph without impairing the radiological protection afforded by the ductile cast iron body.
GNS_ reports No. GNS B 24/30, in translation from the German, describes a missile impact test. This report deals with a test which discusses, in ,
.,_ Section 2.1 of_ Appendix 8 of the TSAR, a missile impact test involving the vertical impact of a 1000 kg missile traveling at 300 m/s on the top cover plate of a CASTOR type cask. Detailed test information is included in GNS report No. GNS 24/30, translated from the German.
'It is apparent from this test that the cask seals are sensitive to extreme accident conditions. Even though there may be no apparent distortion of a lid itself,' seals could be adversely affected. It is prudent, therefore,
~to assume that there is a double seal failure under accident conditions.
The potential consequences of a double seal failure are discussed in Section 7.5.3.3 of this SER.
The TSAR does not address the effect of a rigid 125 kg missile as represented by the armor piercing shell. Consequently, an independent analysis was performed, based on work of A. C. Hagg and G. O. Sankey, which showed that at 126 mph a tornado generated missile with this configuration would not penetrate the cask or cause any significant decrease in radiological protection.
The TSAR does not address the effect of the 2.5 cm solid steel sphere specified in Section 8.2.1.2.2. However, since there are no openings in the protective barrier represented by the cask body and lids, any other effect caused by impact of this small missile would be inconsequential compared to the rigid 125 kg missile. The reviewers conclude that tornado-generated missiles will not violate the structural integrity of the cask nor the radiological protection it affords.
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3.4.4.2 Primary Lid 3.4.4.2.1 Normal Operating Loads The operating loads on the primary lid are described-in Section 3.2.5.4 of the TSAR. 'They are an internal pressure of 7 bar (102 psi), a bolt preload of 500 N/mm2 (72,498 psi), shipping and handling loads of 3 g's, and a maximum operating temperature of 100*C (212*F).
A finite element analysis'of the primary lid was made, with all of the m
. loads superimposed. Two cases were analyzed to take into account the possible load combinations. The maximum stress intensity, including pressure, acceleration, and thermal loads is 39.7 MPa (5,753 psi), well below the stress intensity limit of 231.7 MPa (33,600 psi).
' 3.4.4.2.2 Environmental Loads and Natural Phenomena While there are environmental and natural phenomena loads on the primary lid, which is interior to the cask, the staff does not expect the structural integrity of the cask to be affected by these loads.
3.4.4.2.3 Accidents The impact load resulting from a six-foot cask drop onto a concrete slab is described in Section 8.2.1.2.3 of the TSAR. The resulting stress intensity is 125.3 MPa (18,170 psi) which is well below the allowable stress intensity for the lids of 546.1 MPa (79,200 psi). This analysis in the TSAR is quite conservative in that it ignores the presence of the protective cover plate and secondary lid and assumes the impact is borne by the primary lid.
3.4.4.3 Secondary Lid 25 l.
- 3.4.4.3.1 Normal Operating Loads The operating loads on the secondary lid are described in Section 3.7.5.4 of the TSAR. They are: 7 bar (102 psi) pressure between the primary and secondary lids, 500 N/mma (72,498 psi) bolt preload, 3 g lifting and handling, and 100*C (212*F) maximum operating temperature.
The analysis for the secondary lid, as described in Section 4.2.1.3 of the TSAR, is very similar to the primary lid analysis. The pressure load applied was only 6 bars (87 psi) instead of 7. However, the margin below the e
material allowable is so large that this is not considered a problem. The maximum stress intensity is 113.3 MPa (16,430 psi); the allowable is 231.7 MPa (33,600 psi).
3.4.4.3.2 Environmental Loads and Natural Phenomena While there are environmental and natural phenomena loads on the secondary lid, the staff does not expect the structural integrity of the cask to be compromised by these loads.
3.4.4.3.3 Accidents The secondary lid serves to provide additional rigidity to the primary lid in the event of a cask drop. The loads imposed on the secondary lid during an accident were not considered in the TSAR. The staff concludes that this is reasonable.
3.4.4.4 Fuel Basket 3.4.4.4.1 Normal Operating Loads The normal operating loads on the fuel basket as defined in Section 3.2.5.4 of the TSAR are a design temperature of 400*C (752*F) and a steady state temperature distribution. A finite element analysis was used to determine the stresses and deformations as described in Section 6.2.1.4 of the TSAR. The j maximum stress was below the stress intensity limit.
26 L_.
3.4.4.4.2 Environmental Loads and Natural Phenomena The only consequence of environmental or natural phenomena on the fuel basket would result from cask tip over. The analysis for tip over is reviewed in the following section.
3.4.4.4.3 Accident Loading A drop or tip over accident was modeled by applying a 100 g loading as a static acceleration in the vertical and diagonal directions. The model C
includes lumped masses to account for the weight of the heaviest fuel expected, the 8 x 8 BWR assembly weighing 290 kg (639.3 lbm).
The results of the fuel basket analysis are presented in Section 8.2.1.2.3 of the TSAR. The fuel basket does not experience excessive plastic deformation as shown by the distorted shape plot in Fig. 8.2-24 on page 8.2-48 of the TSAR.
(Note that there is another Fig. 8.2-24 on page 8.2-66 which deals with fuel rod temperatures.) The maximum stress intensity in the fuel basket as a result of the 60 g loading is not provided by the TSAR; therefore, the staff assumes that the stress intensity limit has been exceeded. Nevertheless, the staff concludes that the ductility of the fuel basket material is sufficient to allow for small localized plastic strains.
3.4.4.5 Trunnions 3.4.4.5.1 Normal Operating Loads The operating loads on the trunnions are described in Section 3.7.5.4 of the TSAR. They are: 1.4 g lifting and 2 g transporting, with 150% of the expected contents weight.
The trunnion analysis for normal and operating loads is described in Section 4.2.1.7 of the TSAR. The heaviest load on the trunnions occurs when the cask is being carried by the crane, in which case the total load is 83.35 tonne (91.9 ton). An analysis to determine the effect of the stress risers on 27
m the trunnion, shows that the stress concentration factor is a maximum of 2.15. Using this factor, the maximum trunnion stress is 138.7 N/mm2 (20,111 psi), and the allowable is 139.0 N/mma (20,154 psi).
The staff performed a confirmatory analysis which indicated that the loads and stresses do not exceed those reported in this TSAR. Loadings due to environmental conditions and accident events were not considered relevant for the trunnions.
3.4.4.6 Primary Lid Bolts C
3.4.4.6.1 Normal Operating Loads The normal operating loads on the primary lid bolts are 3 g handling and 500 N/mm2 (72,498 psi) required contact pressure. Bolt load calculations use bolt fastening theory from VDI 2230 (VDI is the Association of German Engineers) and are described in Section 4.2.1.5 of the TSAR. The bolt stresses are summarized in Table 4.2-9, which shows the maximum stress intensity for normal operating loads to be 86.75 N/mm2 (12,578 psi), which is well below the allowable of 200 N/mm2 (29,000 psi).
3.4.4.6.2 Environmental Loads and Natural Phenomena Environmental loads or natural phenomena could result in a cask tip over.
The maximum load assumed for the tip over condition is 120 g's. The stress intensity for this case is 101.23 N/mm2 , well below the allowable.
3.4.4.6.3 Accident Conditions The accident conditions considered are a drop and an off-normal cavity pressure of 7 bar. The analysis in 3.4.4.6.2 includes these loads.
28
3.4.4.7 Secondary Lid Bolts 3.4.4.7.1 Normal Operating Loads The normal operating loads on the secondary lid bolts are 3 g handling and 500 N/mm2 (72498 psi) re'luired contact pressure. Bolt load calculations use bolt fastening theory from VOI 2230 (VDI is the Association of German Engineers) and are described in Section 4.2.1.6 of the TSAR. The bolt stresses are summarized in Table 4.2-10, which shows the maximum stress intensity for normal operating loads to be 148.51 N/mma (21533 psi), which is well below the
~
allowable of 200 N/mm2 (29,000 psi,).
3.4.4.7.2 Environmental Loads and Natural Phenomena Environmental loads or natural phenomena could result in a cask tip over.
The maximum load assumed for the tip over condition is 120 g's. The stress intensity for this case is 144.79 N/mma (20994 psi), well below the allowable.
3.4.4.7.3 Accident Conditions The accident conditions considered are a drop and an off-normal cavity pressure of 6 bar (87 psi). The analysis in 3.4.4.7.3 includes these loads.
3.4.4.8 Trunnion Bolts 3.4.4.8.1 Normal Operating Loads The normal load considered in the TSAR, Section 4.2.1.7, is a 2 g handling load on two trunnions only. This load results in a maximum stress intensity in the bolts of 242.2 N/mma (35,118 psi). The requirement in NUREG-0612 is that the stress intensity should be below the lower of 1/5 S or 1/3 S ,
y ult using a 1 g loading on the bolt. This allowable is 180 N/mma (26,099 psi),
and the stress intensity is 1/2(242.2, N/mma ), i.e., 121.1 N/mm2 (17,560 psi).
Therefore the bolt stresses meet the allowable.
29
3.4.4.8.2 Environmental Loads and Natural Phenomena No environment or natural phenomena loads are considered for the trunnion bolts. The staff accepts this assumption.
3.4.4.8.3 Accident Conditions No accident conditions are considered for the trunnion bolts. This is acceptable. A cask tip over or drop may damage any trunnions impacted and replacement of trunnions and bolts may be necessary.
3.4.4.9 Fuel Rods 3.4.4.9.1 Area of Review In this section, the integrity of the fuel rod cladding is evaluated for compliance with the requirement of 10 CFR Part 72.72(h). The TSAR addresses this requirement by establishing a maximum cladding temperature of 400*C (752*F) i as a design criterion and demonstrating that the design of the cask is such that this temperature limit will not be exceeded for the assumed design basis fuel and environmental conditions. However, the relationship between maximum cladding temperature and the state of damage of the fuel , rod cladding during its storage life is not fully addressed. Of particular concern, is the ability of the clad-ding to maintain its integrity during normal handling at the end of storage life as well as under normal conditions prevailing during its storage life.
The system to be reviewed consists of pressurized Zircalloy cylinders in an inert helium atmosphere. The thermal environment is characterized by an initial temperature of 400'C (752*F) which decays over time in accordance with the curve shown in Fig. 5.1-13 of the TSAR.
l 3.4.4.9.2 Acceptance Criterion The requirements of 10 CFR Part 72.72(h) will be met if it can be demonstrated that, for the design configuration of CASTOR Ic cask, damage accumulation is negligible at the end of storage life.
30 ,
O @
3.4.4.9.3 Review Procedure The integrity of the fuel rods under dry storage conditions was evaluated with reference to the damage mechanisms that are likely to be effective. The TSAR lists several potential mechanisms for fuel cladding failure which include fracture as the terminal event of stable or unstable crack propagation, stress corrosion cracking induced by fission products, hydriding, stress rupture due to creep, oxidation and diffusion controlled cavity growth. Since the cask is designed to maintain an inert gas (helium) environment for the fuel rods, oxidation is precluded and need not be considered further as a potential damage mechanism. The effect of the remaining damage mechanisms were assessed based upon a review of the available data and conclusions of researchers involved in
)
cladding integrity studies (See also Appendix A).
3.4.4.9.4 Findings and Conclusions Three fundamental agents contribute to fuel cladding degradation under dry storage conditions: stress, temperature, and an aggressive environment.
Under normal conditions the stress in the cladding is due to internal gas pressure in the fuel rod. The major component of this gas is helium which is introduced into the free fuel to moderate the effects of the external pressure while in the reactor core. In the course of time, fission products accumulate in the fuel rod cavities. Besides contributing to the internal pressure of the fuel rod, the fission products may also attack the inner surface of the cladding. The effect of terrperature manifests itself by accelerating the rate of degradation mechanisms activated by both stress and corrosion.
Stress corrosion cracking (SCC) occurs as a result of a synergistic combi-nation of a susceptible material, an aggressive environment and high stress.
The corrosive environment associated with SCC of fuel rods has been attributed to fission products generated during irradiation. While the specific agent has not yet been identified, iodine, cesium, and cadmium are considered the most likely agents. SCC may also be related to pellet cladding interaction (PCI) but this has only been observed during reactor operation due, inpart, to the large external pressure on the fuel rods. The only known cause of cladding failure due to SCC cccurred in a reactor during a ramp-up. No other failures 31 .
from this cause are known to have occurred either during pool storage or under dry storage conditions. One explanation may be the pellet temperatures during dry storage are much lower than those in a reactor. Consequently, the generation of fresh fission products at the cladding is slowly reduced during dry storage.
Furthermore, the activation of SCC requires stress levels substantially above those that can reasonably be expected to prevail under dry storage conditions.
The possibility exists, however, that cracks may be present that were initiated during reactor operation. Under these conditions, the stresses generated at the crack tips may be large enough to cause crack extension. However, should such a crack penetrate the cladding it is likely that the internal pressure will be relieved and, as a consequence, effectively terminate the progress of the SCC damage mechanism. The staff concludes, therefore, the SCC is not a damage mechanism that can lead to gross rupture of the fuel rod cladding.
Hydrides in Zircalloy have been known to cause cracking by embrittling the cladding. Terminal solubilities of hydrogen in Zircalloy increase with temperature. If the temperature subsequently decreases, hydrides will precipitate in an orientation determined by the stress level. Normally the <
hydride precipitates in a circumferential direction and is not a problem even at hydrogen concentrations up to 400 ppm. At hoop stress levels of 90 to 95 MPa the hydride will precipitate in a radial direction which can encourage crack penetration. At 400*C (752*F) the hydrogen concentration could be as high as 200 ppm. Brittleness may be induced as the fuel rods decrease in temperature during dry storage. However, the hoop stresses in the cladding are not expected to be high enough to cause a radial orientation of the hydride and consequent crack initiation. It is remotely possible that pre-existing cracks under stress can induce the diffusion of hydrogen to the crack tips where substantially higher concentrations could precipitate hydride in a manner that would encourage crack extension. However, as is the case of SCC, crack penetration would result in a loss of fuel rod internal pressure and termination of the damage mechanism. The staff concludes, therefore, the delayed hydriding is not a damage mechanism that can lead to gross rupture of the fuel rod cladding.
Creep rupture is a potential failure mode under dry storage conditions.
Researchers have demonstrated that using a Larson-Miller approach temperature limits from 380*C (716*F) to 400*C (752*F) could be tolerated for creep rupture 32
_ _ __~
lives well beyond that required for interim storage of spent fuel. The Larson-Miller approach, however, is somewhat empirical since it depends upon the existence of experimental data to establish the appropriate parameter.
Practicality limits the duration of creep rupture tests, which are usually conducted at stress levels and temperatures far higher than those that prevail under dry storage conditions. The creep damage mechanisms in the.high temperature, high stress regime are different from those that occur at lower temperatures and stresses. Consequently, predictions based on a Larson-Miller mode are clouded with sufficient uncertainty to warrant a more fundamental approach to cladding degradation under creep conditions.
c The staff examined this matter to determine potential mechanisms for significant creep damage under dry storage conditions applicable to the case of the CASTOR Ic cask. The only creep damage mechanism (in fact the only mechanism for any of the failure modes considered above) that the staff found which represented a possible potential for cladding degradation and gross rup-ture was diffusion controlled cavity growth (DCCG), which is most applicable to the conditions of dry storage. Damage is manifested by the nucleation and growth of cavities at the grain boundaries which, in effect, reduces the area of material available to resist loads. The measure of damage is the fraction of the grain boundary area that undergoes decohesion. The reviewers developed a method to determine the level of damage as a function of time (See Appendix A).
The progress of damage based upon the applied methods indicated the area of decohesion at the end of twenty year storage life to be less than 8%. Based upon the degree of conservatism maintained throughout the analysis, it can be concluded that this level of damage is insignificant and would not be exceeded.
Consequently, an initial storage temperature not exceeding 400*C (752*F) for the design basis fuel is acceptable for meeting the requirements of 10 CFR 72 Part 72.72(h).
Staff evaluation of potential damage mechanisms to spent fuel assemblies stored under conditions specified in the TSAR, leads to the conclusion that for storage at 400*C (752*F) or less in a helium atmosphere, there is no potential for significant deterioration or degradation of the cladding.
33 ,
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'40 Thermal Evaluation 4.1 Normal Conditions
~
4.1.1 Area of Review -
~
. ~
J The thermal analysis presented in the TSAR Was reviewed to evaluate the protection provided to prevent fuel cladding degr' adation and gross ruptures in -
compliance with 10 CFR Part 72.72(h). TheCASTORIccaskbasdescribedin '
Section 1.2 of the TSAR basically consists of.A thick-walled ductile iron casting with a stainless steel basket separating the 16 BWR fuel assemblies. .
There is a 13 mm nominal clearance between the' fuel' basket and the cask wall.
Th TSAR provides a thermal analysis for loading, transporting and storing
.8x8 and/or 7x7 BWR fuel. The maximum host output of any specific assembly is 1.2 kW just prior to loading into the cask.
4.1.2 Acceptance Criteria The requirements of 10 CFR Part 72.73 can be met if it is demonstrated that, for the CASTOR Ic cask design and operation, the maximum. fuel cladding temperature is less than 400*C (752*F) to prevent cladding rupture as established in Section 3.4.4.9 of this SER. .
I 4.1.3 Review Procedure The thermal analysis in the TSAR was reviewed and confirmatory calculations were performed to ensure that the fuel rod clattding temperature is below 400*C a.,
(752*F). The thermal analysis in the TSAR was performed with a two-dimensional finite difference code developed and benchmarked against test data by the applicant. The two-dimensional heat transfer model was based on an infinitely long heat source with a homogenous power distribution. A power peaking factor of 1.10 was used in the hottest region of the casP basket. Radiation e,nd conduction heat transfer modes within the basket regions wera consider'd. e The radiation heat transfer included shadowing and non-uniform radiation effects between the rods with an average emissivity of 0.8.
34
The confirmatory analysis was performed using a simplified two-dimensional model for the cask and basket and a modified version of the Wooton-Epstien Correlation to calculate the maximum temperature of the fuel cladding. A single fuel bundle was fir.311y analyzed in the center region of the basket with an average basket wall temperature of 340 C (644*F). Sensitivity calculation were made to evaluate rod emissivities in the range of 0.7 - 0.9 and basket wall average temperatures of 320*-360*C (600*F-680 F).
4.1.4 Findings and Conclusions
~
The maximum cladding temperature calculated in the TSAR is 400*C (752*F) for the cask in the upright position with a 54*C (129*F) ambient condition and an average heated output of 1.1 kW per assembly. Both the experimental data t
measured by the applicant and the confirmatory analysis were within the 400*C (752*F) maximum temperature. On the basis of the independent analysis and the TSAR evaluation including the experimental data, it is concluded that the fuel cladding will remain below 400*C (752*F) during storage to prevent cladding degradation and gross rupture in compliance with 10 CFR Part 72.72(h).
The thermal analysis in the TSAR is acceptable for referencing provided that the maximum heat output of any single assembly is less than 1.2 KW and the total heat content stored within the basket is below 17.6 KW.
We also note that operations during cask loading would occur within a reactor spent fuel pool area and does not come under this review. However, this process is briefly described in Section 5.1.1 of the TSAR including spent fuel loading, lifting of the cask to the pool surface, primary lid closure and seal testing with drying of the cask cavity region by a vacuum system and pressurization to a subatmospheric pressure (0.8 bar) with helium. Subsequently, the secondary lid closure and sealing is performed with an interlid helium pressure of 6 bar established. The commitment to maintain the maximum fuel pin cladding temperature at a limit of 400*C (752*F) is made as an operatism limit .
in the TSAR.
35
,1
1 Cask unloading is also briefly described for unloading at a reactor spent fuel pool in Section 3.5.1.2 of the TSAR with prevention of release of gaseous activity during this process discussed.
Procedures for cask loading, unloading, and decontamination, as adapted for site-specific conditions and use, will be described and committed to in detail by a license applicant.
4.2 Accident Conditions 4.2.1 Explosion The gas explosion analysis in the TSAR was reviewed to assure that the cask containment does not fail and that the radiation doses in 10 CFR Part 72.67 are not exceeded. The only damage possible to the cask due to a gas explosion is that which could result from tip over which had been addressed in Section 8.8.1.
It is concluded that the tip over analysis findings and conclusions in Section 3.4.4.1.2 apply to the gas explosion event.
4.2.2 Fire 4.2.2.1 Area of Review The thermal analysis for an accidental fire was reviewed in the TSAR to determine if any radioactive release could occur in violation of 10 CFR 72.68.
The TSAR assumed that the cask is exposed to a 800*C (1472*F) engulfing fire for 1/2 hour. The cask was analyzed for a heat load of 27 kW which is significar.tly greater than the specified licensing condition of 17.6 kW.
4.2.2.2 Acceptance Criteria
.The requirements of 10 CFR Part 72.23 can be met if it is demonstrated that the pressure in the cask remains below 7 bar (103 psi), the closure seal remains intact and the shielding is unimpaired during and following the fire, l
l l
e L
4.2.2.3 Review Procedure The TSAR references experimental testing in which a quarter-length cask was exposed.to 800*C (1472*F), for 1/2 hour. The test demonstrated that the
-seal would remain intact and that the shield material was unimpaired. Calcula-tions were performed to demonstrate that even if there were 100% failure of the cladding the cavity pressure would be less than the design pressure of 7 bar
.(105 psi). The thermal analysis results showed the maximum temperatures to be 363*C (685*F) on the outside, and less than 200*C (392*F) on the inside walls of the cask for the fire event. The maximum fuel clad temperature calculated a
was 410*C (770*F).
4.2.2.4 Findings and Conclusions Based on the experimental testing and the thermal analysis it was concluded that-the cask containment will not fail and the requirements in 10 CFR Part 72.68 as discussed in Section 11.0 of this SER will be met. The fire evaluation in the TSAR may be cited as a reference in a site-specific license application except for the maximum cavity pressure of 1.62 bar (23.5 psi) which is not pro-perly substantiated. A confirmatory calculation indicated that the cavity pres-sure would be 5 bar, (72.5 psi) which is still below the design pressure of 7 bar (105 psi).
37 .
5.0 Shielding' Evaluation 5.1 Area of Review
.Part-72.67(a) of 10 CFR Part 72 requires that during normal operations and anticipated occurrences, the annual dose equivalent to any.real individual-
-located beyond the controlled area shall not exceed 25 mrem to the whole body, 75 mres to the thyroid and 25 mrem to any other organ as a result of exposure to (1) planned discharges of radioactive materials, radon and its daughters excepted, to the general environment, (2) direct radiation from ISFS1 operations and (3) any other radiation from aranium fuel cycle operations within the region.
As_ stated, the. maximum annual dose equivalent allowed to the whole body from all sources at the ISFSI site is 25 mrem.
Part 72.68 (a) of 10 CFR Part 72 requires that for each ISFSI site, a controlled area shall be established. Since the TSAR is generic in nature no specific controlled area has been established. For the purposes of this review, the minimum distance of 100 meters (328 ft.), as given in Part 72.68 (b) of-10 CFR Part 72, is assumed.
In addition to the above, the TSAR addresses-the GNS shielding design criteria in Section 1.2.2 (Principal Design Criteria), Section 3.3.5.2
. (Shielding), and Section 10.1.2.1 (Cask Surface /Off-Surface Dose Rate Limit).
As stated, the maximum dose rate (neutron + gamma) is 200 mrem /h at the cask surface.and 10 mrem /h'at 2 meters (6.56 ft.) from the cask, i
l L 5.2 Acceptance Criteria L
l Cask shielding is deemed acceptable if it can be shown that the annual dose equivalent at 100 meters (328 ft.) from a single. cask is less than 25 mrem including reactor background. Satisfaction of the GNS design criteria-
[- .is also of interest but is not the basis upon which acceptance will be
- evaluated.
I' q
l l
38 .
L-
5.3- Shielding Review Procedure 5.3.1 Source Specification 5.3.1.1 Gamma Source The TSAR addresses the gamma source for the active length of the spent fuel elements (designated as Source Region No. 3) in Section 3.1.1.2.2 (Activity Inventory), Section 3.1.1.2.4 (Photon Emission), Section 3.1.1.2.6 (Summary of Spent Fuel Characteristics), Section 7.2.1 (Characterization of Sources), and Appendix 2 (ORIGEN Burnup Calculation for BWR Fuel). Supplemental information is also provided by GNS under separate letters dated 17 October and 14 November 1984. The gamma source strength is determined from an ORIGEF calculation using the TYPE 1 7x7 array design basis fuel described in Section 3.1.1.1 (Description of the Fuel Assemblies). In this calculation, the average burnup is 27,000 mwd /MTU, the specific power is 22.5 MW/MTU, the initial fuel enrichment is 2.4%, the irradiation time is four 300-day cycles with interim shutdown of 30 days, and 3,120 Kg (6,878 lbm) of heavy metal are considered. The cooling time for the spent fuel used in the shielding evaluation is one year. Activation of the cladding material is not included in the. gamma source strength result.
The gamma source for the head and foot pieces of the spent fuel elements (designated as Source Region Nos. 1 and 4, respectively) is addressed in Section 7.2.1 (Characterization of Sources) and Section 7.3.2.2.1 (Shielding Calculation) of the TSAR. The gamma source strength is determined from an ORIGEN activation calculation using a mean neutron flux of 1.5 x 1012 neutron
!- s/cm2 -sec, an irradiation time of 1,200 days, and a maximum cobalt content of 2000 ppm. Neither tables nor text are specific in their identification of the sources for the discrete energy gamma rays emitted from the head and foot piece regions. The 0.83 MeV gamma undoubtedly arises from the s4Fe (n, p) s4Mn
. reaction; the 1.17 and 1.33 MeV gammas may arise from either the s9Co (n, Y) 80Co or SONi (n, p) 80Co reactions.
i o
39 i
5.3.1.2 Neutron Source The TSAR addresses the neutron source for the active length of the spent fuel elements (designated as Source Region No. 3) in Section 3.1.1.2.2 (Activity
-Inventory), Section 3.1.1.2.5 (Emission of Neutrons), Section 3.1.1.2.6 (Summary of Spent Fuel Characteristics), Section 7.2.1 (Characterization of Sources),
and Fppendix 2 (ORIGEN Burnup Calculation for BWR Fuel). Supplemental information is also provided by GNS under separate letter dated 17 October 1984.
The neutron source strength is determined from an ORIGEN calculation using the Type 1 7x7 array design basis fuel described in Section 3.1.1.1 (Description of ,
the Fuel Assemblies). The major input parameters for this calculation are described in Section 5.3.1.1 (Gamma Sources) of this SER.
5.3.2 Model Specification The TSAR addresses the shielding model in Section 7.3.2.2.1 (Shielding Calculation). For the cask wall, GNS assumes a cylindrical geometry with homogenized source region and an effective layer thickness for the moderator rods and fin area. In the case of the bottom and covers, it is assumed that GNS used a cylindrical source with slab shield geometry. In our evaluation we have assumed the same source and shield geometry as GNS. We have examined the gamma dose using cylindrical volume source with slab shield at side and end geometries. We have also examined the gamma dose and coupled neutron and gamma dose using a three dimensional finite cylinder analysis.
5.3.2.1 Description of the Radial and Axial Shielding Configuration The radial and axial shielding configurations are addressed in Section 7.3.2.2.1 (Shielding Calculation). Supplemental information is also provided by GNS under separate letter dated 17 December 1984. Dose point locations for surface and one meter distances are at the middle of the cask side wall, the middle of the bottom, and the middle of the covers. At two meters (6.56 feet) distance, only the dose at the middle of the cask side wall is computed.
40
v Figures presented for the radial and axial shielding configurations appear adequate for the GNS model. However. for the verification analyses they are incomplete. Figures presented for t5e bottom and cover calculations are incompletely labeled in the various source "egions; the presence of boron is not indicated. The effect of these missing cata is not significant.
5.3.2.2 Shield Regional Densities Material densities (g/cm3) are addressed in Section 3.2.5.3 (Material Properties), Section 7.3.2.1 (Shielding Design Features), and Section 7.3.2.2.1 (Shielding Calculation) of the TSAR. Atom number densities (atoms / barn-cm) are ~
addressed in Section 7.3.2.2.1 (Shielding Calculation); only Source Region No. 1 (head piece), Source Region No. 2 (upper gas plenum), Source Region No. 3 (active fuel length), and Source Region No. 4 (foot piece) are included. Atom number densities for the shield materials are not included. Supplemental information is also provided by GNS for the protection plate and other shield materials under separate letters dated 14 November 1984 and 17 December 1984.
Material densities given in Section 3.2.5.3 (Material Properties) and Section 7.3.2.1 (Shielding Design Features) differ in the values for the steel in the bottom and covers. GNS has used the higher density (7.85 vs.
3 7.70 g/cm ) in their shielding calculations.
Atom number densities given in Section 7.3.2.2.1 (Shielding Calculation)
'for the iron and bcron in Source Region Nos. 1 through 4 are incorrectly calculated. They are, however, conservative in that they indicate a density that is less than the combined density for iron and boron. The effect of these missing data is not significant.
5.3.3 Shielding Evaluation The TSAR addresses the s.'.ielding evaluation in Section 3.3.5.2 (Shielding),
Section 7.3.2.2.1 (Shielding Calculation), and Section 7.3.2.2.2 (Experimental Verification). GNS shielding calculations are performed with ANISN and DOT codes with 100 neutron and 20 gamma groups; flux-to-dose rate conversion factors are those of ANSI /ANS-6.1.1. Analytical results are then compared with experimental 41
.. e results from a Caster Ic loaded with representative BWR fuel. In our evaluation of the GNS calculations we use the SHIELD and MORSE-L codes and consider the experimental results as well.
. SHIELD is an unbenchmarked code for gamma dose from simple source geometries and is' based upon analytical solutions to simple integration kernels. Buildup factors employed in this use of SHIELD are for a point isotropic source in iron.
MORSE-L (Ref. 4 and 5) is a special version of MORSE which uses the Monte Carlo method to' transport both neutrons and gamma rays. With SHIELD we determined the gamma dose at all GNS dose point locations using cylindrical volume source with slab shield at side and end geometries. With MORSE-L we determined the gamma dose and coupled neutron and gamma dose at the surface of the bottom, top, and side wall using a three dimensional finite cylinder analysis. Flux-to-dose rate conversion factors are also those of ANSI /ANS-6.1.1. Gamma sources as tabulated by GNS are used in the SHIELD code computations. For MORSE-L, the 100 neutron
-and 20 gamma groups are collapsed to 30 and 15, respectively.
5.4 Findings and Conclusions Our calculations verify the GNS computational results. Further validation of their computations.is provided via the experimental results-for the CASTOR Ic.in Section 7.3.2.2.of the TSAR. Annual dose commitment at 100 meters, the minimum permissible distance to a controlled area' boundary under Part 72.68(b),-
to a real individual, conservatively assumed to be continuously present, from a single cask is calculated to be less than 40 mrem. This is greater than
, 25 mrem / year allowed under Part 72.67(a). Thus, for the assumption of an individual continuously present at the controlled area boundary, the distance
'to such boundary for a single cask should be 150 meters or more. A lower residency time could reduce this distance requirement. For arrays involving more than one cask, a license applicant would have to assess the conditions for the site concerned to arrive at a suitable distance.
4 Cask shielding is acceptable.
l 1
42
=,
6.0 Criticality Evaluation 6.1 Area of Review.
The criticality analysis presented in the TSAR was reviewed to determine if the CASTOR Ic cask is designed to be subcritical and to prevent a nuclear criticality accident in compliance with 10 CFR Part 72.73. The CASTOR Ic cask with its fuel basket is described in Section 1.2 of the TSAR. The cask basically consists of a thick-walled cylinder ductile iron casting. A fuel basket with a minimum of 0.9% baron in the stainless steel material is used to separate the fuel and to prevent criticality. ~
The TSAR provides a criticality analysis for loading, transporting and storing 8 x 8 and/or 7 x 7 BWR fuel. The fuel to be stored is assumed (for criticality purposes only) to have a maximum enrichment of 3.2% and a maximum density of 96%.
6.2 Acceptance Criteria The requirement of 10 CFR Part 72.73 can be met if it is demonstrated that, for the CASTOR Ic cask design, the effective multiplication factor is less than 0.95 for all credible configurations and environments.
6.3 Review Procedure The criticality analysis in the TSAR was reviewed and verification calculations were performed for comparison to ensure that the CASTOR Ic design is subcritical at all times. The criticality analysis presented in the TSAR was performed with the KENO IV code and was based on the following assumptions:
(1) the fuel was enriched 3.2%; (2) the fuel was unirradiated; (3) the stainless steel basket had a boron content of .7%; and (4) the fuel and water inside each assembly formed a homogenous mixture at 20*C (68*F). The calculations were performed with the cask in the upright position with the fuel centered and also with the cask in the horizontal position with the fuel to one side of the storage space. Calculations were performed using two different neutron 43
cross'section libraries, the 27-Group Westfall Library and 123-Group LLC-16 Library. The benchmark calculations for the KENO IV code presented in the TSAR were reviewed.
Our confirmatory analysis was independently performed with the KENO Va Code and was based on the following assumptions: (1) the fuel enrichment was
~3.2%; (2) the fuel was unirradiated; (3) the stainless steel basket had a boron content of .7% ; and (4) discrete pins were modeled inside each assembly inside the cask with the water at 20*C (68*F). Calculations were performed with the fuel centered, to the outside and to the inside of the basket space using the 27-Group Westfall Library and the 123 group DLC-16 library.
6.4 Findings and Conclusions The largest k-effective value reported in the TSAR is 0.896 0. 008 .for
-the cask in the upright position in an infinite array of closely spaced casks containing 8 x 8 fuel assemblies in 20*C (68*F) water. The largest k-effective value calculated in the confirmatory analysis was 0.838 1 0.004 for the cask in the upright position. Both the TSAR and verification cases bound the actual fuel and basket configurations and materials and result in k-effective values below .95 which is the maximum acceptable design limit. On the basis aof the independent analysis.and the TSAR evaluation, the staff concludes that the CASTOR Ic cask is designed to be maintained subcritical and to prevent a nuclear critical accident in compliance with 10 CFR Part 72.73.
44 ,
7.0 Confinement 7.1 Area of Review The confinement analysis presented in the TSAR was evaluated to ensure that the annual doses specified in 10 CFR Part 72.67 (a) are not exceeded during normal operations and anticipated occurrences. The cask basically consists of 14-inch thick ductile cast iron walls with a double closure on the top.
The primary closure is made of stainless steel with filling and flushing connections. The secondary closure is also made of stainless steel and has a pressure gauge installed on it. Both closures and all their openings are "
sealed with flexible metal seals. Elastomer seals are used only to assist in leak checking the closure systems after the cask is filled with helium. The space between-the primary and secondary closures is used as a gas barrier and pressure monitoring space. Each of the barriers is designed to have a leak rate < 10 8 mbar 1/sec.
7.2 Acceptance Criteria The requirements of 10 CFR Part 72.67 (a) can be met if it is demonstrated that the annual doses are within regulatory requirements even if one assumes 1%,10% or 100% fuel rod failures occurring and both seals remaining leak tight.
7.3 Review Procedure The confinement analysis in the TSAR was reviewed and confirmatory calculations were performed to ensure that the regulatory dose limits are not exceeded. Only BWR fuel assemblies with intact cladding will be loaded into a CASTOR Ic cask. While cladding failures are assumed by the TSAR, they may not
.necessarily occur. The TSAR analysis assumed that fuel clad failures of 1%,
10% and 100% occur under unspecified conditions. The radioactive gases and vapors considered for release are tritium, krypton, iodine and cesium with the release fractions into the cask cavity obtained from Regulatory Guide 1.25 and NUREG-0069 for 27,000 mwd /MTV. The analysis is based on molecular diffusion mechanisms only, since a 6 bar (87 psi) gas barrier exists between the first and second closure.
45
A confirmatory analysis was performed with'similar assumptions except that choked flow conditions were assumed to exist between the second closure and the outside environment. The internal gas pressure was also calculated to confirm the flow mechanism between the first and second barriers.
7.4 Findings and Conclusions From the TSAR, the maximum release of radioactive materials from a single cask with 100% cladding failure, both seals intact, and released during fifty years, the dose is 2.8 x 10.s mrem / year (TSAR Tab. 8.1-3) which is only a small fraction of the 25 mrem / year allowed for normal operations in Part 72.67(a).
Confirmatory calculation results show releases approximately two times higher.
Assuming 100% cladding failure, maximum activity release and both seals remain-ing intact, the dose from the double seal system in one year at 100 m is an integrated dose of only 9.2 x 10 4 mrem whole to the whole body and 1.0 x 10 3 mrem thyroid. The confirmatory calculations further show that the maximum pres-sure in the cask is < 4 bars (56 psi) with 100% fuel rod failure and an average gas temperature of 300*C (572*F), which confirms that molecular diffusion flow prevails between the first and second barriers. If one of the seals fails, an alarm activates that alerts the ISFSI personnel to check and replace the seal.
Compliance with 10 CFR Part 72.67 is site dependent and depends on the number of' casks being stored. Thus, a license applicant must assess conditions for the' cask array proposed for his specific site.
7.5 Confinement Requirements for the Hypothetical Accident Conditions 7.5.1 Area of Review Part 72.15 (a)(13) of 10 CFR Part 72 requires, in part, an analysis of the potential, dose or dose commitment to an individual outside the controlled area from accidents or natural phenomena events that result in the release of radioactive material to the environmental or direct radiation from the ISFSI. ,
Part 72.68 (b) of 10 CFR Part 72 requires that any individual located on or near the nearest boundary of the controlled area shall not receive a dose 46 i 1
greater than 5 rem to the whole body or any organ from any design basis accident. The minimum boundary distance allowed is 100 meters.
Our review focuses on the release of radioactive material to the environ-ment.
7.5.2 Acceptance Criteria Cask confinement of radioactive material is deemed acceptable if it can be shown that the release of material subsequent to an accident shall not deliver a dose of 5 rem to any individual within 100 meters (328 feet) of a single cask. ~
7.5.3 Review Procedure The review consists of consideration of: (1) the maximum gaseous activity within the cask, (2) the maximum dose from gaseous activity release, and (3) the maximum dose from gaseous and particulate activity release.
7.5.3.1 Maximum Gaseous Activity Within the Cask The TSAR addresses the maximum activity expected to be found within the cask in Section 3.3.2.2 (Activity Release). Only the nuclides 3H, asKr, 1291, 134Cs, and 137Cs are identified. Cladding tube failures of 1, 10, and 100 percent were addressed.
7.5.3.2 Maximum Dose From Gaseous Activity Release The maximum dose expected from gaseous activity release is addressed in Section 8.1.1.3.1 (Seal Malfunction) and Section 8.2.1.2.7 (Thermal Loads) of the TSAR.
Overall, Section 8.2 (Accidents) is lacking in its discussion of the dose consequences due to seal malfunction resulting from the accidents analyzed.
However, assuming the confinement will not be breached, that is, at least one cask lid remains sealed, and that the doce consequences discussed in Section 8.2.1.2.7 (Thermal Loads) are a maximum, then information provided is sufficient for NRC staff to complete our review.
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7.5.3.3 Maximum Dose from Gaseous and Particulate Activity Release As the staff noted in Section 3.4.4.1.6, Tornado-Generated Missiles, a test of a missile impact on the too of a CASTOR type cask shows a failure of metal seals for the primary and secondary lids can occur, at least under extreme conditions. While this may result in no significant release of either gaseous or particulate radioactive material, the staff has performed an analysis using a hypothetical worst-case accident based on transportation scenarios and additionally has arbitrarily assumed 100% fuel cladding failure to provide an upper bound to consequences of a seal failure for both the secondary and primary lids of the cask. _
The CASTOR Ic cask was designed for storage and transportation of irradia-ted spent fuel assemblies, and its design fulfills the IAEA international speci-fications for type B(U) packaging. Although storage of spent fuel is the only use evaluated in this report, a hypothetical worst-case accident based on trans-portation accident scenarios is being evaluated to establish an upper bound accident impact for storage applications. The transportation accident scenario is not considered credible for storage situations. It has been chosen merely to determine estimates for release of radionuclides from the spent fuel to the cask cavity and then to the environment rather that arbitrarily assuming a non-mechanistic accident release.
The release fractions used in this analysis were based on ref. 7 for scenario 5 (a worst-case for air-cooled casks). This scenario considers all release mech-anisms that are credible for air-cooled casks. The mechanism for release of radioactivity considered appropriate for this evaluation was an impact rupture which somehow caused mechanical disruption of t;ie c. adding and subsequent depres-surization of 10 percent of the fuel rods. The fraction (20%) of the spent fuel inventory of noble gases generated in the reactor that are in the gap is released to the cask cavity. Because of the low temperatures, the remainder of fission products released are assumed to be particulates that are swept out of the rods as they depressurize after rupture. The spent fuel inventory frac-tion that is swept out as particulates is 2 x 10 8 48
c
. , e.
Once radionuclides have been released from the fuel rods they must,then find.a path out of the cask. The result of accident damage is not expected to provide a pathway from the cask cavity to the environment with a large cross section. Only a small section of failed cask seal would be the most likely release pathway. (In the case of the CASTOR Ic, failure of a least two seals would be required, .such as the metallic seals for the secondary and primary lids, with no credit being taker, for any elastomer seal present.) Before the radionuclides are released to the' environment, they must pass many places that
-are relatively cool and through small passages. As a result, radionuclides can. condense, plate out, or be filtered out before escaping the cask. -For gas-cooled casks (in this case helium cooled) 60% of the noble gases in ~
the cask cavity are assumed to be released and 5% of the particulates.
After the radioactive material escapes the cask, there are two factors important in determining whether the particules reach people; the fraction that becomes
. suspended in air and the fraction that is respirable (less than 10 microns aero-dynamic diameter). Five percent of the particulates were assumed to be smaller than 10 microns and remain-as an aerosol.
The resultant dose to an individual located at 100 meters from a cask would be about 19 mrem to the whole body and 16 mrem to the thyroid. These doses are much less than the maximum 5 rem dose from a design basis accident to an individual at the minimum controlled area boundary as required in 10 CFR 72.68(b). The dose is also much less than the Environmental Protection Agency's protection
. action guides. For a very conservative assumption of 100% failure of fuel cladding, the dose would be 190 mrem to the whole body and 160 mrem to the thyroid at 100 meters, again much less than the 5 rem limit established in 10 CFR 72.68(b).
7.5.4 Findings and Conclusions The dose consequence due to gaseous activity release from a single cask following an accident-in which one of the lid metal seals malfunction is less than 0.4 mrem in one week at 100 meters. For a bounding case assuming failure of seals on both the primary and secondary lids and 100% failure of fuel 49
cladding, the whole body dose.is 190 mrem from gaseous and particulate radio-active material release. The doses are far less than the 5 rem specified in 10 CFR Part 72.68(b).
t 1
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8.0 Operating Procedures In Chapter 9, " Conduct of Operations," of the GNS TSAR Rev. 3, GNS states:
"This~ chapter addresses items related to a specific license application and
'thus is not relevant from a topical report standpoint." Operating procedures were therefore not subject to review by the staff.
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9.0 Acceptance Tests and Maintenance Program 9.1 Acceptance Tests The GNS-TSAR Rev. 3 does not address the subject of acceptance tests.
As noted in our Introduction (Section 1.1), the TSAR was generated in the format of Regulatory Guide 3.48. In Regulatory Guide 3.48, Chapter 9, " Conduct of Operations," covers such tests under Section 9.2, "Preoperational Testing and Operation." The TSAR treats this as a site-specific matter. This is acceptable to the staff. We note, however, that test procedures are committed to under the quality assurance program for the CASTOR Ic cask cited in Appendix 9, " Quality Assurance Plan for CASTOR Dry Spent Fuel Storage / Transport Cask," of the TSAR under " Matrix of Quality Assurance Requirements" on page 2.
9.2 Maintenance Program Maintenance is addressed only briefly in the GNS TSAR Rev. 3. In Sec-tion 5.1.3.5 GNS states: " Maintenance tasks associated with the storage cask are not foreseen. There are no periodic tests necessary for the cask monitoring system. Visual inspection of the cask during storage may be done but this is not an operation in the sense of maintenance," and in Section 4.5, GNS states:
"Due to the specific design of CASTOR casks, repair is not foreseen at the ISFSI.
Maintenance is restricted to minor procedures such as elimination of defects of the outer decontamination coating and routine inspections if requested. In the case of an AR storage, restoration of double confinement in the improbable case of failure of one of the sealing systems will be executed inside the power plant building at the spent fuel pool..."
This description of maintenance is acceptable for the TSAR. However, for a license applicant proposing to use an array of casks at a reactor site, a detailed description of site-specific maintenance activities and procedures will be required.
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10.0 Radiation Protection 10.1 Area of Review 10 CFR Part 15(a)(5) requires the licensee to provide the means for
' controlling and limiting occupational radiation exposures within the limits '
given in 10 CFR Part 20, and for meeting the objective of exposures as low as is reasonably achievable.
10 CFR Part 72.74(a) states, in part, that radiation protection systems shall be provided for all areas and operations where onsite personnel may be exposed to radiation or airborne radioactive materials.
Guidance is also provided in Regulatory Guide 8.8, "Information Relevant To Ensuring That Occupational Radiation Exposures At Nuclear Power Stations Will Be As Low As Is Reasonably Achievable," and Regulatory Guide 8.10,
" Operating Philosophy.For Maintaining Occupational Exposures As Low As Is Reasonably Achievable."
Our review focuses on those policy, design and operational considerations associated with occupational exposures as low as is reasonably achievable that are not site specific. In this regard our review is limited. A second area of our review focuses on the estimated onsite dose from direct radiation and gaseous activity release during normal operations.
10.2 Acceptance Criteria l Radiation protection is deemed acceptable if it can be shown that the non-site-specific considerations for occupational radiation exposures as low as is reasonably achievable are in compliance with appropriate guidance and/or regulations, and that the dose from the handling, surveillance and maintenance f of casks are not in excess of Part 20 limits.
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-10.3-' Review Procedure The review is divided into three main parts: (1) ensuring that occupational radiation exposures are as low as is reasonably achievable,-(2) radiation protection design features, and (3) estimated onsite dose assessment.
10.3.1 Ensuring that Occupational Radiation Exposures are as Low as is Reasonably Achievably (ALARA)
Non site-specific policy, design, and operational considerations are addressed in Section 7.1.1, 7.1 2, and 7.1.3, respectively, in the TSAR.
The TSAR sections cited above describe how the CASTOR Ic cask is designed to meet the ALARA requirements. These requirements are met through the massive shielding, the passive nature of the system,.the ruggedness of design and the double confinement system utilized.
The objectives of Regulatory Guide 8.8 with regard to access control, shielding, decontamination and monitoring are also met by the design features.
The staff evaluated the non-site-specific information provided by GNS in comparison with the guidance and/or regulations cited in Section 10.1 of the SER.
10.3.2 Radiation Protection Design Features Installation design features are addressed in Sections 1.2, 1.3, 1.4, 3.1.1, 3.3.2, 3.3.5, 5.1.2, 6.5, 7.1.2, and 8.1.2 of the TSAR.
TSAR Sections 1.2, 1.3, 1.4 provide a physical description of the design of the cask. Included in this description are the features pertaining to shielding, the gas containment system and other features pertaining to radiation protection.
Section 5.1 describes the operations involved with fuel loading, closing up the cask, monitoring, and setting the cask in its storage position.
54
Sections 6.5 and 7.1 describe the radiation protection involved with normal operation of the cask. ,
Section 8.1 describes the radiation protection aspects of off-normal operations with the cask.
10.3.3 Estimated Onsite Dose Assessment Information important to the estimate of the onsite collective dose is found in Sections 3.3.2, 3.3.5, 6.5, 7.3.2, and 7.4 of the TSAR.
TSAR Section 3.3 provides information on the radioactivity (gaseous) release from the cask and the dose rate predicted for the cask.
Sections 6.5 and 7.3 provide information on the radiation protection aspects pertaining to the estimated dose.
The dose to an employee was computed by the staff for normal operations from the sum of the dose from gaseous activity release and the dose due to direct radiation. One cask arrival and one routine maintenance operation per year is assumed. Exposure doses due to direct radiation used in the staff review are those appearing in Tab. 3.3-11 of the TSAR. In computing the dose due to gaseous activity release, the staff has' assumed the following: (1) 100%
cladding tube failure; (2) both lid seals are intact; (3) activity release is at a maximum for each gaseous radionuclide; (4) the occupational inhalation rate of Regulation Guide 1.25; (5) the exposure times and distances in Tab.
3.3-10 of the TSAR; (6) the inhalation dose and exposure dose factors of Regulation Guide 1.109 and NUREG-0172 (Ref. 6); and (7) F-stability atmospheric diffusion.
4 10.4 Findings and Conclusions 7 Non-site-specific policy, design, and operational considerations are in compliance with appropriate guidance and/or regulations, and the dose from a single cask to any individual from direct radiation and gaseous activity l
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.o-
. release during normal operations is estimated to be less than 15 mrem per
' year to the whole body.
Radiation protection is acceptable.
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11.0 Accident Analysis 11.1 Area of Review 10 CFR Part 72.15(a)(5) requires, in part, an analysis of the potential dose or dose commitment to an individual outside the controlled area from acci-dents or natural phenomena events that result in the release of radioactive material to the environment or direct radiation from the ISFSI.
10 CFR Part 72.67(a) requires that during normal operations and anticipated occurrences the annual dose equivalent to any real individual who is located "
beyond the controlled area shall not exceed 25 mrem to the whole body, 75 mrem to the thyroid and 25 mrem to any other organ as a result of exposure to (1) planned discharges of radioactive materials, radon and its daughters excepted, to the general environment, (2) direct radiation from ISFSI opera-tions and (3) any other radiation from uranium fuel cycle operations within J
the region.
10 CFR Part 72.68(b) requires that any individual located on or near the nearest boundary of the controlled area shall not receive a dose greater than 5 rem to the whole body or any organ from any design basis accident. The minimum distance established is 100 meters.
Our review focuses on the dose from direct radiation and gaseous activity release associated with postulated off-normal and accident events. In the context of this review, off normal events are anticipated occurrences.
11.2 Acceptance Criteria Cask safety in the event of postulated off-normal and accident events is deemed acceptable if it can be shown that the dose from a single cask to any individual from direct radiation and gaseous activity release is not in excess of the applicable values given in Section 11.1 above.
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11.3 Review Procedure The review is divided into two main parts: (1) off-normal operations and (2) accident events.
11.3.1 Off-Normal Operations 11.3.1.1 Event Postulated events for off-normal operations are addressed in Section 8.1.1.
Postulated causes for the events are found in Section 8.1.1.1. The means of detecting the various events are discussed in Section 8.1.1.2. Analysis of the effects and consequences and the proposed corrective actions appear in Sections 8.1.1.3 and 8.1.1.4, respectively.
11.3.1.2 Radiological Impact from Off-Normal Operations The radiological impact from off-normal operations is addressed in Sec-tion 8.1.2 of the TSAR. Dose due to direct radiation beyond the controlled area is discussed in Section 7.6.
The radiological impact from off-normal operations is computed for an indi-vidual outside the controlled area. In computing the doses due to gaseous activity release, the staff has assumed the following: (1) 100% cladding tube failure; (2) failure of the primary seal; (3) activity release is at a maximum for each radionuclide; (4) the population weighted inhalation rate of Regulation Guide 1.109 for the offsite individual; (5) an inhalation exposure time of one week for the offsite individual; (6) the inhalation dose and exposure dose fac-tors of Regulation Guide 1.109 and NUREG-0172 (Ref. 6); (7) a direct radiation dose at the controlled area boundary of 100.0 meters based upon the ratio of the surface and 100 meter cylindrical volume source SHIELD results; (8) the side wall total dose rate in Tab. 7.3-7 of the TSAR; and (9) F-stability atmospheric diffusion. The staff recognizes that a failed primary seal is to be replaced in the reactor building. Nevertheless, we assumed this to be the worst case for release of gaseous activity and for dose consequences to an individual replacing a failed secondary seal.
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11.3.2 Accidents 11.3.2.1 Accidents Analyzed i The accidents analyzed, cause of the accidents, and the accident analysis are presented in Sections 8.2.1, 8.2.1.1, and 8.2.1.2, respectively, of the TSAR. With one exception, there is no discussion of the radiological conse-quences or accident dose calculations results. Reiterating our position from Section 7.5.3.2 of the SER, assuming the confinement will not be breached, that is, at least one cask lid remains sealed, and that the dose consequences discussed in Section 8.2.1.2.7 of the TSAR are a maximum, the information provided is sufficient for NRC staff to. complete our review.
In the staff review of the radiological impact of the accidents presented, we followed the same procedure and made the same assumptions as those discussed in our review of off-normal operations. Additionally, we assumed the dose-consequence discussed in Section 8.2.1.2.7, namely, that the release of cesium
-was doubled and therefore the dose from 134Cs and 137Cs were double.
However, as discussed in Section 3.4.4.1.6 of the SER, failure of the seals for both the primary and secondary lids may be possible under extreme accident conditions. Based on this, staff has calculated upper bound conse-quences of such failure arbitrarily assuming 100% percent fuel cladding failure in Section 7.5.3.3 of the SER.
11.4 Findings and Conclusions The dose consequence due to gaseous activity release from a single cask following postulated off-normal and accident events is less than 0.4 mrem in one week to the whole body and thyroid outside the controlled area, whose boundary is assumed to be at the minimum allowable distance of 100 meters, if cask confinement is assumed not to be breacted.
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Assuming failure of seals for both the primary and secondary lids and
- arbitrarily' assuming a very conservative 100% failure of the fuel cladding, the doses would be 190 mrem to the whole body and 160 mrem to the thyroid at a distance of 100 meters. Accident consequences are much less than the 5 rem limit established in 10 CFR Part 72.68(b).
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1 12.0 Decommissioning 12.1L Decommissioning Plan A decommissioning plan for a site-specific ISFSI as required by 10 CFR 72.18lis not applicable. However, in Section 3.~5.1.2 of the TSAR, it is noted that decommissioning could consist simply of the unloading, decontamination and disposal-of the cask.
l 12.2' Decommissioning Considerations
.e 12.2.1! Cask Body Activation
.The cask is slightly activated by the neutron flux emanating from the spent fuel. ORIGEN calculations were performed to determine the extent of activation products in the cask body. Results of these calculations are presented in Sec-
. tion 3 5.1.1 of the TSAR. Four activation nuclides were considered; Mn-54, Fe-55, Fe-59, and Co-60.
Only the isotope Co-60 is of importance. This isotope originates from nickel which is limited to 1.3% in the cask body material. Therefore, the maximum cobalt content-in the cask body material is 100 ppm.
i For the inner region, approximately 23.2 metric tonnes of the CASTOR Ic cask are taken into consideration which give a total Co-60 activity of CASTOR
'Ic. cask of 8.5 x 10 5 Ci after 10 years storage and 2.8 x 10.s after 50 years storage. For the outer region, 14.1 metric tonnes of the CASTOR Ic cask are
! taken into consideration, which give a total Co-60 activity of 4.9 x 10 8Ci f after 10 years, and 1.6 x 10 s Ci after 50 years storage. A total Co-60 activ-l ity for the whole CASTOR Ic cask of 9.0 x 10 5 Ci is calculated after a storage
. period of 10 years, and of 3.0 x 10 5 Ci for a storage period of 50 years.
L .
These concentrations of Co-60 present are well below the exempt concen-l tration limits' set in 10 CFR 30.70. Therefore, the decontaminated cask could be i
~given an unrestricted release.
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7, e 12.2;2 Unioading the Cask
.The decommissioning of the cask consists of the unloading of the fuel'at a reactor pool, which is briefly described in Section 3.5.1.2 of the TSAR.
Prior to unloading the cask a closed vessel will be connected first to the inter-lid space and then to the cask cavity to contain potential gaseous activity and present radioactivity release. For the unloading, the cask cavity is flushed by pumping water through the internal cavity by means of the flushing connection. A special cool-down installation is used to lower the fuel rod temperatures, if necessary, so that the unloading can be done in the spent fuel pool at the reactor. e 12.2.3 Findings and Conclusions The cask design is consistent with the requirements of 10 CFR Part 72.76 that an ISFSI be designed for decommissioning. The actions involved in cask unloading, decontamination and disposal are also consistent with the require-ments of 10 CFR Part 72.18 as feasible elements of a site specific decommission-ing plan.
For a site-specific license application, the applicant would be expected to' develop and commit to detailed procedures for use in unloading the cask.
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. :o 13.0 Operating Controls and Limits 13.1 . Area of Review *
.Each license issued under 10 CFR Part 72 shall include license conditions pursuant to 10 CFR Part 72.33. In addition to the conditions pursuant to 10 CFR Part 72.33(b), each application for a-license under 10 CFR Part'72 shall include proposed technical specifications pursuant to 10 CFR Part 72.16 and consistent
~with 10 CFR Part 72.33(c). The final approved technical specifications will be made-part of the operating license.
The technical specifications of a license define certain features, char-acteristics and conditions governing operation of an installation. Technical specifications cannot be changed without approval of the NRC.
13.2 Acceptance Criteria .
Consistent with 10 CFR Part 72.33(c) the operating controls and limits established in Chapter 10 of the TSAR will be deemed acceptable if they cover, for the cask, all required safety limits, limiting conditions for operation surveillance requirements and design features.
13.3 Review Procedure Operating controls and limits which may serve as a basis for licensing conditions are derived from the analyses and evaluation included in the TSAR.
-13.4 Findings and Conclusions l
The staff reviewed the specific operating limits summarized on page 10.1-4 of the TSAR. The limits established for lifting height, dose rate, leakage, specific power, cladding temperature, helium pressure and water content reflect the design criteria upon which the safety analyses were based and are acceptable.
With regard to the fuel characteristic limit' 4 scribed in Section 10.1.2.3 of the TSAR, the maximum enrichment is limi+4A :( l . Z4. Consequently, though the design basis fuel specified in the TSlo wqn aview was 2.4% enrichment, the
- staff concurs that the operating limit for the spent fuel be set at 3.2%.
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_ , l A license applicant would review parameters covered in the TSAR and develop appropriate proposed technical specifications and license conditions l
'for his site-specific conditions. '
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4 64
14.0 Quality Assurance Chapter 11 " Quality Assurance," of the TSAR references Appendix 9, " Quality Assurance Handbook for Dry Spent Fuel Storage Cask (QAH)," Revision 1 of the QAH
-(docketed July 25, 1984). However, the QAH has been superseded by submittal by General Nuclear Systems, Inc. (now the applicant) of the " Quality Assurance Plan for the CASTOR Dry Spent Fuel Storage / Transport Cask" (docketed March 5, 1985, under Project Nos. M-34 and M-37 and 71-0510), which includes Revision 2 of the QAH and which responds to NRC staff comments sent to Gesellschaft fur Nuklear Service mbH (formerly the applicant) by letter dated November 13, 1984, for Revision 1 of the QAH.
- The quality assurance program established by General Nuclear Systems, Inc., which is based on the criteria of Appendix B to 10 CFR Part 50 and which references ANS/ASME NQA-1-1979, " Quality Assurance Program Requirements for Nuclear Power Plants," is acceptable for referencing without further review in a license application to receive and store spent fuel under 10 CFR Part 72, provided that the license applicant has an NRC approved Quality Assurance Program which meets the requirements of Appendix B to 10 CFR Part 50 and which is applied to this activity.
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. o 15.0 References
- 1. Gesellschaft fur Nuklear-Service mbH (GNS), " Topical Safety Analysis Report-(TSAR) for CASTOR Ic Cask Independent Spent Fuel Storage Installation (Dry Storage)" Revision 3, September 1984.
- 2. U.S. Nuclear Regulatory Commission Regulatory Guide 3.48, " Standard Format and Content for the Safety Analysis Report for an Independent Spent Fuel Storage Installation (Dry Storage)," October 1981.
. 3. U.S. Nuclear Regulatory Commission Regulatory Guide 3. (CE-306-4),
" Standard Format and Content for a Topical Safety Analysis Report for a Dry Spent Fuel Storage Cask," August 23, 1984.
- 4. T. Wilcox, " MORSE-L, A Special Version of the MORSE Program Designed to Solve Neutron, Gamma and Coupled Neutron-Gamma Penetration Problems,"
UCID-16680, Lawrence Livermore National Laboratory, Livermore, California (September 1972).
- 5. T. Wilcox, " MORSE Cross Section Library Tapes," UCID-16683, Lawrence Livermore National Laboratory, Livermore, California (July 1973).
- 6. G. R.'Hoenes and J. K. Soldat, " Age-Specific Radiation Dose Commitment Factors for a One-Year Chronic Intake," NUREG-0172, Battelle Pacific Northwest Laboratories, Richland, Washington (November 1977).
i l
- 7. E. L. Wilmont, Transportation Accident Scenarios for Commercial Spent Fuel, SAND-80-2124, (Albuquerque, NM.: Sandia National Laboratories, February 1981).
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APPENDIX A-ANALYSIS OF DIFFUSION CONTROLLED CAVITY GROWTH (DCCG) DAMAGE TO FUEL CLADDING IN DRY STORAGE
- 1. 0 INTRODUCTION The only damage mechanism that the staff found with a possible potential for cladding degradation and gross rupture was DCCG. The staff has examined
' this potential and based on available information has developed a method to determine the level of damage which could occur under dry storage conditions for the CASTOR Ic cask as a function of sptat fuel time in storage.
2.0. REVIEW PROCEDURE The area fraction of decohesion at grain boundaries in Zircaloy cladding at any time can be ascertained by satisfying.the following equation A t I I dA l_ f f(4)
=
f G(t)dt (2-1)
A o 9
where A 9 is the initial area fraction of decohesion due to the nucleation of I
stable cavities and Af is the area fraction of decohesion that occurs over the period of time t .
f Furthermore, (1 - (A9 /A)b sin a) (1 - A) f(A) * (2-2) 1 1 3 A A b (2 2n A A(1 - I)) ,
32 F
B (a) 06a, Ogb(t)
G(t) = 3n 3 T(t) (~)
l Fv (a) u l
l l
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The terms of expression (2-2) and (2-3) are defined with the aid of Figure 2-1 as follows:
a = grain boundary cavity dihedral angle O = atomic value 6 = grain boundary thickness '
a ,= stress in the cladding k = Boltzman's constant A = average cavity spacing -
D gb = grain boundary diffusion rate T = absolute temperature FB (a) = nsin 2a Fy (a) = 2n/3 (2 - 3cosa - cos2 a)
Some of the foregoing terms may be further defined by a = cos 1(TB )
g exp[-Q/RT(t)]
D gb
=D D
gbo
= grain boundary diffusion coefficient Q = activation energy for grain boundary self-diffusion R = gas constant T = free surface energy TB= grain boundary surface energy Much of the review effort focused on establishing the values of the param-eters in the above expressions. Where there was wide divergence in reported values, the value that led to the most conservative result was selected.
2.1- Grain Boundary Cavity Dihedral Angle, a For clean surfaces in pure metals Raj and Ashby (Ref. 1) suggest that Tg = T/2 so that a is computed to be about 75 . To account for non-ideal condi-
.tions, a value for a of 50* was used in the analysis.
2 .
- o. ~u t
1All slII x
. a . GRAIN 6 BOUNDARY
.. + O 5 .- , .' __
. J .
CAVITY
[ .' .
uvN g
- Tg I
l
\
l V C/
I FIGURE 2-1
- 3
2.2 ' Atomic Volume, O The atomic volume can be estimated from O*N where A is the atomic weight = 91 N is Avogadro's number = 6.02 x 102s p is the specific gravity = 6.55 gms/cc which gives a value for O of 2.31 x 10 29 m3 / atom. This agrees closely with a
.value of 2.37 x 10 29 3/ atom reported by Lloyd (Ref. 2). However, Chin, et al.,
(Ref. 3) used the cube of the Burgers vector, b = 3.23 x 10 8 m, which gives an atomic volume of 3.37 x 10 29 m3 / atom. For the sake of conservatism, the value for 0 = 3.37 x 10 29 m3 / atom was selected for the analysis.
2.3 Grain Boundary Thickness, 6 The grain boundary thickness defines the area through which grain boundary vacancies migrate to the cavity. The disorder that characterizes the structure at the grain boundary is only a few atoms thick. Since grain boundary diffusion rates are many orders of magnitude greater than volume diffusion rates, a grain boundary thickness of 3 Burger's vectors is considered adequate. Consequently, a value of 6 = 3(3.23 x 10 10) = 9.69 x 10 10 m was selected for the analysis.
2.4 Stress on the Cladding, 6, The cladding stress is due to the fuel rod internal pressure at the stor-age temperature. There is considerable uncertainty regarding the level of pres-sure in the fuel rod, either from rod prepressurization, fission gas release or volume increase due to creep strain. Johnson (Ref. 4) estimates that 30 psi is a typical internal pressure for BWR fuel rods at 25"C while 75 psi is reached only occasionally. Based upon the assumption that the internal pressure is equivalent to external pressure in a reactor, Blackburn (Ref. 5) computes a pressure of 550 psi. This results in a conservative stress level of 8800 psi (60.7 MPa) at 400*C which was used for the analysis.
4
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- 2. 5 Average Cavity Spacing, A The value of this parameter has been particularly difficult to establish.
Cavity spacing depends upon the density of nucleation sites and will vary with the type of nucleation mechanism. Experimental .ork conducted at Cornell University (Ref. 6) indicated a spacing in~unirradiated Zircalloy-2 of from 10 to 20 x 10 8 m. This experimental work further established that grain boundary cavities do form at 350*C especially at stresses over 100 MPa. The cavity density appeared to reach a saturation level after about 10 days suggesting a limited number of nucleation sites in the material. Consequently, it is not
'likely that the intercavity spacing, A, will decrease during dry storage as a result of further nucleation. Conservatism dictated the use of the lower value of 10 x 10 8 for the analysis.
2.6 Grain Boundary Diffusion Rate, D ab There are many reported values of volume diffusion rate for a-Zirconium but few.with respect to grain boundary. diffusion rate. The two values spe-cific for grain boundary diffusion are 6 x 10 18 exp (112/RT) reported by Chin (Ref. 3) and 5.9 x 10 8 exp (131/RT) reported by Garde, et al., (Ref. 7). The latter is the more conservative value by about two orders of magnitude and was, consequently, used for the analysis.
2.7 Temperature, T The temperature dependence of grain boundary decohesion was established using the temperature decay curve provided on Page 5.1-39 of the TSAR. Since the data as reflected by measured values terminates at approximately two years from beginning of storage, it was assumed that the temperature would remain constant thereafter.
3.0 Findinas and Conclusions The progress of damage based upon the methodology and assumed va10es for the parameters previously described indicates that the area of decohesion at the end of twenty year storage life to be less than 8%. Based upon the degree 5 ,
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of conservatism maintained throughout the analysis, it can be concluded that this level of damage is insignificant and would not be exceeded. Consequently, an-initial storage temperature not excaeding 400*C for the design basis fuel in a CASTOR Ic cask is acceptable for meeting _the requirements of 10 CFR-72 Section 72.72.(k).
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a v REFERENCES
- 1. R. Raj and M. F. Ashby, "Intergranular Fracture at Elevated Temperature,"
Acta Met., Vol. 23, p. 653 (1975).
- 2. L. T. Lloyd, " Thermal Expansion of. Alpha-Zirconium Single Crystals,"
ANL-6591, Argonne National Laboratory (1963).
- 3. B..A. Chin, N. H. Madsen and M. A. Khan, " Application of Zircalloy Deformation and Fracture Maps to Predicting Dry Spent Fuel Storage Conditions," Department of Mechanical Engineering, Auburn University, Auburn, A1.
- 4. A. B. Johnson, Jr. , " Behavior of Spent Nuclear Fuel in Water Pool Storage," BNWL-2256, September 1977.
- 5. L. D. Blackburn, et al., " Maximum Allowable Temperature for-Storage of Spent Nuclear Reactor Fuel," HEDL-TME 78-37, VC 70, May 1980.
- 6. A. M. Garde, H. M. Chung, and T. F. Kaisner, "Micrograin Superplacticity in Zircalloy at 350 C," Acta Met., Vo. 26, p. 153 (1978).
- 7. 'R. L. Keusseyan, " Grain Boundary Sliding and Related Phenomena," Doctoral dissertation, Cornell University,1985.
l 8 '
L l
ENCLOSURE 2 LIMITATIONS AND OPERATING CONDITIONS The following limitations and operating conditions are to be considered in developing site-specific technical specifications for the use of the CASTOR Ic cask for receipt, handling and storage of spent fuel at an ISFSI located at a reactor site. Additional technical specifications and limitations, as appropriate, will be necessary with respect to the issuance of any site-specific license under 10 CFR Part 72, but these are not addressed in this review. (However, some areas requiring consideration are listed for information under Cask Related Activities and are cross-referenced to corresponding items under Cask Operating Conditions).
Limitations
- 1. Basic Components:
The Basic Components of the Model CASTOR Ic cask that are important to safety are listed in Table 3.4-1 of the TSAR.
- 2. Fill gas: Helium
-3. Fuel Stored
- a. Types of Boiling Water Reactor fuel found acceptable are those listed in Section 3.1.1.1 of the TSAR.
- b. Maximum number of BWR fuel assemblies stored per cask shall be < 16.
- c. Assemblies to be contained in the CASTOR Ic fuel basket shown in GNS Drawing No. E528.13-15.2 Rev. O in App.1 of the TSAR.
- d. Initial fuel enrichment shall not e'xceed 3.2 percent U 235 by weight.
- e. Fuel Assembly Burnup shall not exceed 27,000 MWD /MTU at not more than 22.5 MW/MTU specific power.
- f. Spent fuel assemblies known or suspected either to have gross cladding defects or to have structural defects sufficiently severe to adversely affect fuel handling and transfer capability shall not be loaded into the cask for storage. Pa rtial assemblies, that is, a ssemblies from which fuel pins are missing must not be stored unless dummy fuel pins are used to displace an amount of water equal to that displaced by the original pins.
Cask Operating Conditions Operating Limit
- 1) Max. Lifting Height 6 feet during delivery to and emplacement at the storage site or movement therefrom (Movement of the cask within the reactor buildings is subject to heavy loads limitations of the ,
reactor operating license.)
- 2) Dose Rate:
a) at 2 m Distance from the cask surface 1 10 mrem /h b) at Surface of the cask 1200 mrem /h
- 3) Cask Tightness:
(Standard He-Leak Rate) a) Primary Lid Seal 110-6 mbar 1/s b) Secondary Lid Seal 110~ mbar 1/s
- 4) Max. Specific Power of One 1.1.1 kW Fuel Assembly As Calculated with ORIGEN-2 (For spent fuel with the maximum specified burnup, this specific power value assures that fuel must have decayed 16.4 months after discharge from the reactor core).
- 5) Max. Calculated Cladding Temperature 400 C including during loading operations
- 6) Initial Helium Pressure (Inter 11d) 6 ba r
- 7) Initial Helium Pressure Limit (Cavity) 8001100 mbar
- 8) Partial Pressure of Air (Cavity) 3 mbar (holding for attained during cask evaccuation 10 min.
and drying.
- 9) Water Content in Cavity 110 grarc, attained during cask evaccuation a nd d ryi ng. ,
4
- 10) Siting Limitations a) 10 CFR Pa rt 72.67 (a) b) 10 CFR Part 72.68
Technical Specifications Listing Cask Monitoring
~
Pressure gage: A functional check during preparation of cask for placement in storage (as described in Section 3.3.3.2 ofJ the TSAR) shall be performed to ensure the gage is functioning.
Cask Related Activities Not Covered Activities associated with the receipt, storage and unloading of the cask for which limitations and operating conditions have not been identified include, but are not necessarily limited to the following items. These have not been included either because they are site specific or because they would be covered under a 10 CFR Part 50 Operating License of a reactor.
- 1) Cask handling and loading operations.
- 2) Cask Decontamination.
- 3) Cask sealing and drying (See also Cask Operating Conditions, items 3 and 5-9).
- 4) Cask transfer to ISFSI.
- 5) Cask emplacement at ISFSI (See also Cask Operating Conditions, item 1).
- 6) ISFSI siting (See also Cask Operating Conditions, item 10).
- 7) Cask return to reactor.
8). Cask handling and unloading operations including cavity sampling and cask cooling.
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