ML20246E113
| ML20246E113 | |
| Person / Time | |
|---|---|
| Issue date: | 07/31/1989 |
| From: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | |
| Shared Package | |
| ML20246E112 | List: |
| References | |
| REF-PROJ-M-42 NUDOCS 8907120112 | |
| Download: ML20246E113 (73) | |
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SAFETY EVALUATION REPORT RELATED TO THE TOPICAL REPORT FOR THE TN-24 STORAGE-CASK SUBMITTED BY TRANSNUCLEAR, INC.
4 U.S. Nuclear Regulatory Commission Office of Nuclear Material Safety and Safeguards July 1989 l
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8907120112 890705 l
t TABLE OF CONTENTS Page 1
GENERAL DESCRIPTION.............................................
1-1 1.1 Introduction................................................
1 1.2 General Description of the Storage Cask...........
1-2 1'
1.2.1 Cask Design Characteristics..........................
1-2 1.2.2 Operational Features.................................
1-4 1.2.3 Cask Contents........................................
1-4 1.3 Identification of Agents and Subcor, tractors.................
1-5 1.4 Generic Cask Arrays.........................................
1-5 2
PRINCIPAL DESIGN CRITERIA.......................................
2-1 2.1 Introduction................................................
2-1
- 2. 2 Fuel to be Stored...........................................
2-1 2.3 Quality Standards...........................................
2-1 2.4 Protection Against Environmental Conditions and Natural Phenomena...................................................
2-2 2.4.1 Tornado and Wind Loading..............................
2-2 2.4.2 Flood.................................................
2-3 2.4.3 Seismic...............................................
2-3 2.5 Protection Against Fire and Explosions......................
2-3 2.6 Confinement Barriers and System............................
2-4
- 2. 7 Instrumentation and Control System..........................
2-4 2.8 Criteria for Nuclear Criticality Safety.....................
2-5 2.9 Criteria for Radiological Protection........................
2-5 2.10 Criteria for Spent Fuel and Radioactive Waste Storage and Handling.....................................................
2-7 2.11 Criteria for Decommissioning................................
2-7 3
STRUCTURAL EVALUATION...........................................
3-1 i
3.1 Area of Review..............................................
3-1
- 3. 2 Acceptance Criteria.........................................
3-1
- 3. 3 Review Procedure..............
3-1 3.4 Findings and Conclusions....................................
3-2 3.4.1 Loads................................................
3-2
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3.a.1.1 Normal Operating Conditions.................
3-2 3.4.1.2 Loads Due to Environmental Conditions and Natural Phenomena.......................
3-2 l
3.4.1.3 Loads Due to Postulated Accidents...........
3-2 l
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LTABLE OF CONTENTS (Continued)
Page 3.4.2 Materials.............................................
3-3 3.4.3 Stress Intensity Limits..............................
3 3.4.4 Structural Analysis.................................
3-4 3.4.4.1 Cask Body...................................
3-4 3.4.4.1.1 Normal Operating Loads...........
3-4 3.4.4.1.2 Environmental Conditions and Natural Phenomena............
3-4 3.4.4.1.3 Accident Conditions..............
3-5 3.4.4.1.4 Fracture Toughness Evaluation....
3-6 3.4.4.1.5 Cask Thermal Stress Analysis.....
3-6 3.4.4.1.6 Tornado-Generated Missiles.......
3-7 3.4.4.2 Neutron Shie1d...............................
3-8 3.4.4.2.1 Normal Operating Loads............
3-8 3.4.4.2.2 Environmental Loads and Natural Phenomena.........................
3-9 3.4-.4.2.3 Accidents.........................
3-9 3.4.4.3 Fuel Basket..................................
3-9 3.4.4.3.1 Normal Operating Loads............
3-9 3.4.4.3.2 Environmental Loads.and Natural Phenomena.........................
3-10 3.4.4.3.3 Basket Accident Loading...........
3-10 3.4.4.3.4 Fracture Thoughness Evaluation...
3-11 3.4.4.4 Trunnions and Trunnion Bolts.................
3-11 3.4.4.4.1 Normal Operating Loads..........,.
3-11 3.4.4.5 Upper Side Impact Limiter...................
3-12 3.4.4.6 Lower Impact Limiter........................
3-12 3.4.4.7 Bolted Covers................................
3-13 3.4.4.7.1 Main Closure Lid System...........
3-13 3.4.4.7.1.1 Bolts................
3-13 3.4.4.7.1.2 Main Closure Lid.....
3-13 3.4.4.7.2 Penetrations......................
3-14 j
3.4.4.8 Fue1.........................................
3-14 3.4.4.8.1 Area of Review....................
3-14 3.4.4.8.2 Acceptance Criterion.............
3-14 3.4.4.8.3 Review Procedure..................
3-14 3.4.4.8.4 Findings and Conclusions..........
3-15 iii
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,i TABLE OF CONTENTS (Continued)
Page 4-THERMAL EVALUATION..............................................
4-1 4.1 Normal-Conditions...........................................
4-1 4.1.1 Area of Review.......................................
4-1 4.1.2 Acceptance Criteria...................................
4-1 4.1.3 Review Procedure......................................
4-1 4.1.4 Findings-and Conclusions.............................
4-2 4.2 Accident Conditions.........................................
4-2 4.2.1 Explosion............................................
4-2 4.2.2 Fire.................................................
4-3 1
4.2.2.1 Area of Review...............................
4-3 4.2.2.2 Acceptance Criteria.........................
4-3 4.2.2.3 Review Procedure............................
4-3 4.2.2.4 Findings and Conclusions....................
4-3 5
SHIELDING EVALUATION............................................
5-1 5.1 Area of Review..............................................
5-l' Acceptan'e Criteria.........................................
5-2 5.2 c
5.3 Shielding Review Procedure..................................
5-2 5.3.1 Source Specification.................................
5-2 5.3.1.1 Gamma Source................................
5-2 5.3.1.2 Neutron Source..............................
5-3 5.3.2 Model Specification...................................
5-3 5.3.2.1 Description of the Radial and Axial Shielding Configuration......................
5-4 5.3.2.2 Shield Regional Densities....................
5-4 5.3.3 Shielding Evaluation.................................
5-5 5.4 Findings and Conclusions....................................
5-5 6
CRITICALITY EVALUATION..........................................
6-1 6.1 Area of Review..............................................
6-1
- 6. 2 Acceptance Criteria.........................................
6-3 6.3 Review Procedure.........................................
6-3 6.4 Findings and Conclusions....................................
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L TABLE OF CONTENTS (Continued)
Page 7
CONFINEMENT......................................................
7-1 7.1 Area of. Review..............................................
7-1 7.2 Acceptance Criteria.............................
7-1
- 7. 3 Review Procedure............................................
7-2 7.4 Findings and Conclusions....................................
7-2 L
7.5 Confinement Requirements for the H Conditions........................ hypothetical Accident L
7-2 7.5.1 Area of Review.......................................
7-2 7.5.2 Acceptance Criteria...................................
7-3 7.5.3 Review Procedure......................................
7-3 7.5.3.1 Maximum Gaseous Activity Within the Cask....
7-3 7.5.3.2 Maximum Dose From Gaseous Activity Release..
7-3 7.5.4 Findings and Conclusions.............................
7-4 8
OPERATING PROCEDURES............................................
8-1 8.1 Area of Review..............................................
8-1 8.2 Acceptance Criteria.........................................
8-2 8.3 Review Procedure.............................................
8-2 8.4 Findings and Conclusions....................................
8-2 9
' ACCEPTANCE TESTS AND MAINTENANCE PROGRAM........................
9-1 9.1 Acceptance Tests............................................
9-1 9.2 Maintenance Program.........................................
9-1 10 RADIATION PROTECTION............................................
10-1 10.1 Area of Review..............................................
10-1 10.2 Acceptance Criteria.........................................
10-1 10.3 Review Procedure..........................................
10-2 10.3.1 Ensuring that Occupational Radiation Exposures are as low as is Reasonably Achievable (ALARA)..........
10-2 10.3.2 Radiation Protection Design Features................
10-2 10.3.3 Estimated Onsite Dose Assessment....................
10-4 10.4 Findings and Conclusions....................................
10-5 v
f TABLE OF CONTENTS (Continued)
Page 11-ACCIDENT ANALYSIS...............................................
11-1 11.1 Area of Review.............................................
11-2
- 11. 2 Accep ta nce C ri te ri a.........................................
11-2 11.3 Review Procedure.......................
11-2 11.3.1 Off-Normal Operations...............................
11-2 11.3.1.1 Event.....................................
11-2 11.3.1.2 Radiological Impact from Off-Normal Operations................................
11-2 11.3.2 Accidents...........................................
11-3 11.3.2.1 Accidents Analyzed.......................
11-3 11.4 Findings and Conclusions....................................
11-4 12 DECOMMISSIONING.................................................
12-1 12.1 Area of Review.............................................
12-1 12.2 Acceptance Criteria.........................................
12-2 12.3 Review Procedure............................................
12-2 12.3.1 Unloading of the Cask...............................
12-2 12.3.2 Decommissioning of the Cask Components.............
12-2 12.4 Findings and Conclusions...................................
12-4 13 OPERATING CONTROLS AND LIMIST...................................
13-1 13.1 Area of Review.............................................
13-1 13.2 Acceptance Criteria.........................................
13-1 13.3 Review Procedure............................................
13-1 13.4 Findings and Conclusions...................................
13-1 14 QUALITY ASSURANCE...............................................
14-1 15 REFERENCES......................................................
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GENERAL DESCRIPTION 1.1 Introduction This Safety Evaluation Report (SER) documents the staff's review and evaluation of the Topical Report (TR) for the Transnuclear TN-24 Dry Storage Cask, July 1988 L
(Reference 1).
The TR was prepared by Transnuclear Inc., using the Regulatory Guide 3.48 (Reference 2) format, as applicable. This SER utilizes the format of-Regulatory Guide 3.61 (Reference 3) with some differences in the section numbering.
The staff's review of the TR addresses the nandling, transfer, and storage of spent fuel in a TN-24 dry storage cask for an at-reactor site independent spent fuel storage installation (ISFSI).
Such storage in a ISFSI would be licensed under 10 CFR Part 72, " Licensing Requirements for the Storage of Spent Fuel in a Independent Spent Fuel Storage Installation (ISFSI)." In this TR a single dry storage cask design, the TN-24 is presented.
The staff's assessment is based on the proposed de ign's meeting the applicable requirements of 10 CFR Part 72, found under Subpart E, " Siting Evaluation i
Factors," Subpart F, " General Design Criteria," and Subpart G, " Quality Assurance," and of 10 CFR Part 20 for radiation protection for onsite receipt and storage of spent fuel in an ISFSI.
Decommissioning, to the extent that it is treated in this TR, presumes as a bounding case unloading of a TN-24 cask at the reactor site and subsequent decontamination of the cask prior to its disposition or disposal.
Use or certification of the TN-24 cask under 10 CFR Part 71, for offsite transport of spent fuel, is not a subject of this safety evaluation.
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."
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1.2 General Description of'the Storage Cask 1.2.1 Cask Design Characteristics The TN-24 dry storage cask (see Figure 1.1-1 in the TR) was developed by Transnuclear Inc., and is designed for the storage of irradiated spent fuel assemblies.
The TN-24 cask body is a right circular cylinder of SA-350, Grade LF3 forged steel 248 mm (9.75 in.) wall thickness, with a 286 mm (11.25 in.)
tnick welded closure on the bottom end made of the same material.
The upper end of the cask body is sealed by an SA-350 Grade LF3 steel-bolted closure lid which is 292 mm (11.5 in.) thick.
It has a cylindrical cavity which holds a fuel basket and is designed to accommodate 24 PWR fuel assemblies.
A protective cover is bolted to the cask body to provide weather protection for the lid penetrations.
The closure lid utilizes a double barrier seal system with two metallic 0-rings forming the seal.
The annular space between the metallic 0-rings is connected to a helium-filled tank placed between the lid and the protective cover.
Pressure in the tank is maintained above the pressure in the cask to prevent either flow of fission gases ot. or air into the cask cavity which, under normal storage conditions, is filled with helium.
The overall dimensicns of the cask are 5105 mm (201 in.) long and 2407 mm (94.75 in.) in diameter.
The unloaded cask weighs approximately 75 tonne (83 ton).
The loaded cask, including stored fuel, and contained water weighs approximately 103 tonne (113 tons).
Neutron emissions from the stored fuel are attenuated by a neutron shield located on the outside of the outer shell which contains a 137 mm (5.38 in) thickness of Dorated polymer neutron shield material encased in a shell of 19 mm (0.75 in.) thick SA 516 Gr 55 steel.
Neutron emissions from the top of the cask are attenuated during storage by a 102 u i,.0 in.) thick polypropylene neutron shield encased in SA-516 Gr 55 steel 6.4 mm (0.25 in.) thick.
This shield cap is placed on top of the cask after fuel loading.
The fuel basket has 24 cavities, each 221 mm (8.70 in) square, to hold the fuel bundles.
The fuel cavities are formed by an interlocking grid of copper plated borated (1 w/o) 304 stainless steel plates 11 mm (0.435 in.)
thick.
The spacing of the plates provide water flux traps for criticality 1-2
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The edges of the plates are guided and nested by aluminum rails that run the axial length of the cask body.
The TN-24 cask body has six attachment points for bolt-on trunnions.
Four of these are located near the top of the cylindrical steel forging, spaced 90 degrees apart, and are used for lifting the cask.
Two trunnion. supports, 180 degrees apart, located near the bottom are used when rotating the cask to or from a horizontal position.
Four 150 mm (5.91 in) diameter lifting trunnions are attached to the upper part of the cask body with twelve 38 mm (1.50 in) diameter bolts.
Each lifting trunnion is designed to meet the requirements of NUREG-0612 for a non-redundant lifting fixture.
Two similar rotation trunnions are attached to the lower part of the cask body.
The rotation trunnions are designed to support 3.0 times the empty cask weight based on the application of a 3.0 g longitudinal load at the cask cavity center.
The TN-24 cask has three containment penetrations; one cask cavity drain, one cask cavity vent, and one interseal overpressure port.
Each of these penetra-tions is in the single lid.
The drain and vent ports utilize double metallic 0-ring seals.
The overpressure port utilizes a single metallic 0-ring seal.
The cavity drain line penetrates the closure lid and terminates at a sump relief in the bottom of the cask cavity.
This is used to drain water from the cask cavity after underwater fuel loading.
It is also used during the drying and helium back-filling of the cask cavity.
The drain valve is of the quick-disconnect type and not analyzed as part of the primary containment system.
A bolted blind flange with two concentric metallic 0-rings is provided for closure of the drain penetration.
The cavity vent penetrates the cavity through the closure lid and is provided with a bolted blind flange with two concentric metallic 0 rings for closure of the vent penetration.
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I The pressure port penetrates the closure lid to the space between the two 0-ring seals.
A bolted flange provided with a single metallic 0-ring seal connects the line from the overpressure tank to the closure lid.
- 1. 2. 2 Operational Features The TN-24 cask is designed to safely store 24 intact design basis PWR fuel
- assemblies.
Each fuel assembly is assumed to have a maximum initial enrichment i
not to exceed 3.5 w/o U-235 in Uranium.
Further assumptions limit the fuel te a maximum of 35,000 mwd /MTU burnup, a minimum decay time of 5 years after reactor discharge and a maximum decay heat load of 1 kW per assembly for a total of 24 kW for a TN-24 cask.
The heat rejection capability of the TN-24 cask maintains the maximum fuel rod clad temperature below 339 C (642 F), based on normal operating conditions with a 24 kW decay heat load, 46 C (115 F) ambient air, and full insolence.
The fuel assemblies are stored in an inert helium gas atmosphere.
The shielding features of the TN-24 cask are designed to maintain the maximum ccmbined gamma and neutron surface dose rate to less than 60 mrem /hr under
~nnrmal operations conditions.
The criticality control features of the TN-24 cask are designed to maintain the neutron multiplication factor k-effective (including uncertainties and calcula-tional bias) at less than 0.95 under all conditions.
1.2.3 Cask Contents The type of spent fuel to be stored in the TN-24 cask is LWR fuel cf the PWR type.
The 17 x 17 0FA (optimized fuel assembly) is used as a reference design in this TR.
However, for certain types nf calculations other arrays may be assumed f.or conservatism.
PWR fuel is made of shnrt cylinders (pellets) of high-fired ceramic uranium dioxide (UO ).
These pellets are 7.84 mm (0.3088 2
l in) in diameter and 12.88 mm (0.507 in) long for the 17 x 17 0FA bundles.
A 3658 mm (144 in.) long stack of 284 of these pel:ets are loaded and hermetically 1-4 L
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sealed into a zirconium alloy tubec Fuel rods are assembled into hundles in a square array,-each spaced and supported by grid structures.
The assembly has a top and bottom fitting.
The overall dimensions or a 17 x 17 0FA assembly are approximately 214.2 mm (8.434 in.) square by 4064 mm (160 in.) long. -Other typical PWR. fuel' assemblies which can be stored include 14 x 14, 15 x 15, or 17 x 17 arrays of individual rods.
- 1. 3 Identification of. Agents and Contractors Transnuclear Inc., provides the-design, engineering, analysis, and quality assurance for the TN-24 cask.
It was incorporated in the State of New York in 1965 and has offices in' Hawthorne, New York, and Aiken, South C6rolina.
Transnucica shares are privately held by Transnucleaire,S.A. of Paris, France.
The TN-24 cask may be manufactured by one or more qualified organizations.
There are no other agents or contractors involved with the TN-24 cask.
1.4 Generic Cask Arrays The ISFSI way 6 1ude cm or more TN-24 casks. The TN-24 cask shall be stored vertically on its bottom plate on a concrete pad.
The TR provides analyses of a vertical storage array consisting of.a 10 x 10 array of casks (100 casks) in Chapter 7 of the TR, A 2 x 10 array for vertical storage is also malyzed in Chapter 5 to ascertain the maximum temperature distribution in th; cask body.
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2 PRINCIPAL DESIGN CRITERIA 1
2.1 Introduction Subpart F of 10 CFR Pr.rt 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 installa-tion (ISFSI).
In this chapter, we discuss the applicability of these criteria t.o the Transnuclear TN-24 dry storage cask and the degree to which the Transnuclear TR is in compliance with these criteria.
Section headings in this chapter generally correspond to sub-sections of Subpart F of Part 72.
2.2 Fuel to be Stored The M-24 cask is designed to store in a dry condition irradiated PWR fuel from nuclea'. power stations.
The design basis fuel is zircaloy clad UO2 with an assumed initial enrichment of 3.3 percent U-235 vy weight for thermal and radiological characteristics and a maximum initial enrichment of 3.5 percent U-235 for criticality considerations.
The design basis fuel is assumed to have been irradiated to a maximum exposure of 35,000 mwd /MTU and cooled for a minimum of 5 years.
Estimates of the radionuclides activity in spent fuel described above were made using the ORIGEN 2 computer code.
- 2. 3 Quality Standards Quality standards for structures, systems, and components important to safety are required by 10 CFR Section 72.122 (a).
Section 3.4 of the TR identifies two categories of safety which are applied to all significant cask components.
Category "Important to Safety" implies critical and major impacts on safety, and would be required to be designed to accord with quality standards.
A quality standard provides numerical criteria or acceptable methods or both for the design, fabrication, testing, and performance of these structures, systems, 2-1
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and components important to safety. These standards should be selected or l
developed to provide sufficient confidence in the capability of the structure, system, or-component to perform the required safety function.
Since quality I
standards are generally embodied in widely accepted codes and standards dealing with design procedures, matarials, fabrication techniques, inspection methods, etc., judgments regarding the adequacy of the standards cited by the TN-24 TR are presented in the sections of this report where the standards are applicable.
2.4 Protection Against Environmental Conditions and Natural Phenomena Section 72.122 (b) of 10 CFR Part 72 requires the licensee to provide protection against environmental conditions and natural phenomena.
Section 3.2 of the TR describes the structural and mechanical criteria for tornado and wind loadings, flood potential, tornado missile protection, seismic design, snow and ice r
loadings, thermal loadings, combined load criteria, and structural design criteria.
In this section, the discussion is limited to the adequacy of the criteria for protecting against environmental conditions and natural phenomena.
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 TN-24 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-TR or for the most severe of the natural phenomena that may occur within the boundaries of the United States.
2.4.1 Tornado and Wind Loading The TR establishes 160.93 m/s (360 mph) in Section 3.2.1.1 as the design basis tornado wind speed.
This is in accordance with Regulatory Guide 1.76 (April 1974).
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l 2.4.2 Flood While no design basis for flood was established, the TR provides a depth limit j
_for submergence of 17.07 m (56 feet) below which no breach of containment will occur and for a. current velocity of 5.49'm/s (18 ft/s) below which no tipover will occur.
It remains for the ISFSI applicant to ensure that the specific.
. design criteria for flood are not more severe if the TR is to be used as a j
. reference.
2.4.3 Seismic
- A horizontal acceleration of 0.25 g was established as a basis for seismic design in Section 3.2.3.
This peak acceleration reflects 10 CFR Part 72.102 for ISFSI sites east of the Rockies.
The TR analysis interpreted this require-ment as referring to:only one direction.
However, the staff interpreted this requirement to mean that this acceleration should be combined vectorially with a component normal to this' direction resulting in a maximum horizontal ground acceleration of 0.35 g.
In addition, Regulatory Guide 1.60 requires that the
-vertical acceleration used be 2/3 of horizontal so that 0.17 g is the accelera-tion in the vertical direction.
2.5 Protection Against Fire and Explosions Pursuant to 10 CFR Section 72.122 (c) the licensee is required to provide protection against fires and explosions.
In Section 3.3.6 the TR establishes the design basis fire of 800 C (1475 F) for one-half hour duration.
This is a basic established for Type B shipping casks under 10 CFR Part 71, Sec-g tion 71.73, " Hypothetical Accident Conditions,." Subsection 71.73(a)(3),
" Thermal." As such, it constitutes an upper bound that is unlikely to be exceeded within a nuclear power plant site.
The design basis for explosion was established at 0.17 MPa'(25 psia) external pressure.
It remains for the ISFSI l
applicant to ensure that the site-specific criteria for explosion are not more severe if the TR is to be used as a reference.
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w' 2.6 Confinement Barriers and Systems Pursucnt to 10 CFR Section 72.122 (h)(1), the licenroe must protect the fuel cladding against degradation and gross ruptures.
The TR addresses the issue of fuel cladding degradation in Section 3.3.7.
The design criterion for the TN-24 cask requires that the maximum fuel cladding temperature limit be calculated in accordance with PNL-6189 (reference 8).
The approach used by PNL
'i is substantially in accord with the criterien adopted by the staff to assure that degradation and gross rupture does not occur over the design life of the ISFSI.
For 5 year old PWR fuel the temperature was determined to be 339 C (642 F).
A confirmatory analysis to verify the validity of this limiting temperature was performed by the staff and the results reported in Section 3.4.4.8.4 of this SER.
10 CFR 72.122(h)(3), though specifically referring to ventilation and off gas systems that are normally associated with an ISFSI, is interpreted to apply to cask storage as a requirement to confine airborne radioactive particulate materials during normal and off-normal conditions.
Consequently, closures secured by bolts or other fasteners should be designed to limit leakage to levels that do not exceed the regulatory limits of 10 CFR 72.104 and 72.106.
The TN-24 design features a single closure lid incorporating two metallic 0-ring seals.
The design criterion for c?.ch seal is a leakage rate not exceeding 10 6 atm-cc/sec of helium.
The staff considers the leakage rate to be acceptable for maintaining the cask helium atmosphere for projected storage periods of at least 20 years.
The design also provides capability to detect seal failure through pressure monitoring.
If seal failure should occur, leak tightness can be restored by either tightening the lid bolts, replacing the seals, replacing the protective cover with a thicker cover incorporating metallic 0-ring seals or welding the protective cover to the cask body.
The acceptability of the leak criterion with respect to leakage of airborne radioactive particulate and gaseous materials
'is addressed in Chapter 7 of this SER.
2.7 Instrumentation and Control Systems Pursuant to 10 CFR Section 72.122 (i), the licensee must provide instrumentation and control systems that monitor systems important to safety over anticipated 2-4 l
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<4 ranges for normal'and off-normal operation.
The TN-24 cask incorporates a
' pressure monitoring system which serves as a cask tightness surveillance system.
The design criteria and description of this system appears in Section 3.3.2 of the TR.
Considering the passive nature of cask storage, the staff finds this pressure monitoring system acceptable for monitoring cask leakage.
2.8 Criteria for Nuclear Criticality Safety Section 72.124 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, and storage conditions; in the data and methods used in the calculations; and in the immediate environment under accident conditions.
Section 72.124 also requires that the design be based on either favorable geometry or permanently fixed neutron absorbing materials.
Section 3.3.4 of the TN-24 TR addresses nuclear criticality safety criteria.
Criticality analysis and prevention are reviewed in Chapter 6 of this report.
The TR establishes a maximum effective multiplication factor of 0.95 for all credible configurations and environments for the prevention of criticality..
This factor is widely accepted as a criticality prevention limit, and the staff concurs with its application to the TN-24 cask.
2.9 Criteria for Radiological Protection Section 72.126 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 TR addresses radiological protection.
The detailed evaluation for compliance with the regulation is discussed in Chapters 5, 7, and 10 of this SER.
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The principal design features of the TN-24 cask.for exposure control are the inherent shielding capability of the cask and the integrity of the seals at the closurejoints.
Radiological alarm systems and systems for monitoring effluents and direct radiation are not applicable to the design of the storage cask.
Effluents are not a normal consequence of the passive dry storage operation; consequently control systems to provide radiological protection for this condi-tion are 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.
L However, it should again be noted, as in Section 2.7 above, that the sealing system of the cask uses a pressure monitoring device as a tightness surveillance system.
Leakage past the outer metallic seal will be manifested by a drop in interseal system pressure.
The shielding capability of the cask for gamma rays relies primarily upon the thickness and attenuation property of the steel cylinder and the steel closure lids which comprise the primary barriers to radiation.
The cask must maintain its structural integrity under loadings associated with normal op ration, accident events, natural phenomena, and environmental conditions.
Of particular concern is the response of the cask to dynamic loading conditions associated with cask drop and/or tipover.
It is essential to demonstrate that its fracture' toughness is sufficient to resist catastrophic brittle fracture under the assumption that undetected flaws may exist at locations of maximum primary membrane or bending stress.
A review of this topic is presented in Sec-tion 3.4.4.1.4 of this SER.
The TR also establishes in Section 3.3.5.2 (Criteria) the surface dose limit as 60 mrem /hr.
The staff believes that this limit is acceptable, provided the distance to the site boundary for a single cask is not less than 250 meters (see Sections 5.2 and 5.4 of this SER).
However, in finding these limits 1
acceptable for a 250-meter site boundary distance for a single cask, the staff notes that for site-specific analyses consideration must be given to cumulative dose rate because of reactor operations, and to individual residency time at or near the site boundary.
The nearest individual has been conservatively assumed in this evaluation to be present continuously at the site boundary.
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._o 2.10 Criteria for Spent Fuel and Radioactive Waste Storage and Handling Pursuant to 10 CFR Section 72.128, 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 reliability consistent with its importance to safety and (d) means to minimize the quantity of radioactive wastes generated.
This section of the regulations defines the requirements for the spent fuel storage system within the context of the entire ISFSI. The TR presents a summary that addresses only cladding temperature limits and nuclear criticality safety in Section 3.3.7.
Actually, the entire TR serves to demonstrate compliance with the details of this part of the regulations.
2.11 Criteria for Decommissioning Pursuant-to 10 CFR Section 72.130, the licensee is required to design the ISFSI for decommissioning.
For dry cask storage, this requirement applies to the cask design itself.
Thus, decommissioning provisions should address decontamina-tion of the cask components following removal of the radioactive spent fuel.
The quantity of radioactive wastes produced and contamination of equipment should be minimized.
The TR addresses this requirement in Section 3.5 in detail l
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3 STRUCTURAL EVALUATION 3.1. Area of. Review This chapter evaluates the structural response of the TN-24 Dry Storage 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 provided in the TR for each of the compo-nents and systems important to safety.
The structural review consists of a review for the storage requirements of 10 CFR 72 only.
No review has been made for transportation requirements.
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 noted in 3.1 above are lower than the allowable stress limits for the cask components important to safety.
The allowable stress limits are documented in the TR in Sec-tion 4.2.1.1.
Table 4.2-2 provides the structural design criteria for stress combinations.
i Information on materials is presented in Section 4A.4.
Table 4A.5-9 describes the mechanical properties used for the cask lid, cylindrical shell, and the bolts.
Table 4A.6-1 describes the mechanical properties of the basket material.
3.3 Review Procedure The TR was reviewed for compliance with 10 CFR Section 72.122(a) which refers to quality standards that govern the characterization of materials, the l
establishment of stress intensity limits, and the design and analysis methods I
that provide confidence in the capability of the structure, system, or component l
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'to perform the required safety function.
The TR was also reviewed for compliance
_ ith 10 CFR Section 72.'122(b) which requires that protection against with 10 CFR w
Section 72.122(c) which requires that protection against-fires and explosions
~be demonstrated;.and for compliance with 10 CFR Section-72.122(h) which requires that protection of fuel cladding against degradation and gross rupture be demonstrated.
3.4 Findings and Conclusions 3.4.1 Loads 3.4.1.1 Normal Operating Conditions The TR specifies-in Section 3.2.7 the norrhal operating presrure of 2.2'atm (32.3 psia).
In Section 4A.7.3 the trunnion loads are based upon NUREG-0612' H
for a redundant lifting system.
3.4.1.2 Loads Due to Environmental Conditions and Natural Phenomena The design basis loads due to environmental conditions and natural phenomena are summarized in Section 3.2 of the TR.
In accordance with Section 2.4.3 of this SER, the staff used 0.35 g horizontal acceleration plus an upward accelera-tion of 0.17 g to determine whether the-. cask would tip as a result of an earth-quake.
A maximum horizontal windspeed of 360 mph was adopted.
3.4.1.3 Loads Due to Postulated Accidents 10 CFR Section 72.122(b)(1) requires that the cask be designed to accommodate the effects of postulated accidents.
The TR describes these postulated accidents in Chapter 4 and Appendix 47.
The loads due to these accidents arise as a result of impact due to handling accidents, gas cloud explosion, or fire.
The handling accidents-assumed in the TR are an 8-foot end on drop onto a concrete pad, an 8-foot corner drop onto a concrete pad, a 7-foot side drop onto a concrete pad, a tipover.from the vertical position about the lower trunnions l ~
that are 8 feet above the ground, and, tipover from the vertical standing posi-tion.
Impact limiters are not used for any of the assumed handling accidents.
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.The staff performed' confirmatory analyses, assuming unyielding foundations, L
which indicated that the postulated accidents result in stresses in the cask Lbodythatarenot' acceptable.
It had been determined that an 18-inch drop is l
acceptable:for the cask body for the assumed handling accidents of end on drop, side drop,-corner drop, and tipover.
The integrity of the basket with fuel in o
l place at this 18-inch tipover.and side drop height, however, could not be
. established..This is discussed in more detail in Sections 3.4.4.1.3 and 3.4.4.3.3 of this SER.
3.4.2 Materials Major structural materials used for fabrication of the TN-24 dry storage cask are listed in Tables 4A.5-9 (for the cask lid, cylindrical shell and bolts) and 4A.6-1 (for the basket material).
During the course of the review some of these materials were changed.
The cask shell material was changed from'A-350 Lf1 to A-350 Lf3.
The basket material specified is to be procured in accordance with ASTM A-87 and supplemented by Transnuclear procurement specification in Appen-dix 4B of the TR'(Attachment E-10698 to letter E-10701 to M. W. Schwartz from Mike Mason dated January 30 1989).
All materials are identified by ASTM specification or ASME code designation which are related to ASTM Specifications.
These specifications are considered by the reviewers to be quality standards in accordance with 10 CFR Section 72.82(a).
3.4.3 Stress Intensity Limits The TR lists in Tables 4.2-2, 4.2-8, 4.2-10, 4.2-11, 4.2-12, and 4.2-13 stress intensity limits for primary service loads and levels A, B, C, and D service loads for all components important to safety.
In general 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 10 CFR Sec-tion 72.122(a).
The stress intensity limits for the containment vessel in Table 4.2-8a apply to A-350 Lf1 forged steel at 350 F.
The use of A-350 Lf3 will result in higher temperature are 40 and 70 ksi respectively while the corresponding minimum l
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strengths of A-350 Lf1 at the same temperature are 30 and 60 ksi.
Consequently, if the results of the analyses indicate that the limits of A-350 Lf1 are not exceeded, then the limits for A-350 Lf2 will surely not be exceeded.
In:the case of the basket material, the minimum yield and ultimate strengths are lower in the A-87 Specification than the corresponding minimum values in Appendix.48.
Where these specifications conflict the reviewer has chosen to assume the more conservative requirement.
Consequently, the stress intensity limits for the basket material should be downgraded to reflect the values in ASTM A-87.
3.4.4 Structural Analcsis 3.4.4.1 Cask Body 3.4.4.1.1 Normal Operating Loads The cask body was analyzed for an internal pressure of 250 psia using the method described in Section 4A.5.2.4 of the TR.
The maximum stress is far below the allowable stress intensity limit of 17.5 ksi for the cask body.
During truck transport, the cask rests on two trunnions (at the upper end of the cask) and a shipping skid (to support the lower end of the cask).
There is no analysis provided in the TR for this horizontal load condition.
A simple beam analysis performed by the reviewers shows that the maximum stress in the cask is well below the stress intensity limit.
During the handling by crane, the cask is supported in a vertical position by four trunnions.
A redundant four-arm yoke may be used to lift and handle the cask. The combination of pressure, bolt preload, and handling stresses is below the stress intensity limit.
3.4.4'1.2 Environmental Conditions and Natural Phenomena As a result of the design basis tornado wind loads, the staff concludes that the cask will not suffer a tipover.
The TR states in Section 3.2.3 that the 3-4
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" cask'will not.tipover as a result of the design basis earthquake.
If the horizontal components of acceleration are vectorially added (instead of L
algebraically), the combined horizontal acceleration would be 0.35 g.
Since j'
the cask is calculated to tipover at' 0.34 g's,.the cask is assumed to tipover with the'0.35 g load.
A tipover analysis is provided in the TR in Sec-tion 4A.5.7.
The staff concludes that the cask integrity'will also be maintained for snow and ice loadings, for flooding conditions and for lightning i-strikes. For tornado generated missiles see Section -3.4.4.1.6 of the SER.
3.4.4.1.3 Accident Conditions The TR describes analyses of the cask body for accident conditions in Sec-tion 4A.S.
The impact conditions considered in the TR are bottom end drop, side drop, corner drop, and tipover.
The 8-foot bottom end drop without an impact limiter is discussed in Sec-tion 4A.5.6.3.A of the TR.
A 2-D axisymmetric finite element confirmatory analysis was performed by the reviewers for this condition.
The results of the confirmatory analysis show that the maximum stress intensity in the cask body is greater than the allowable of 70 ksi (ultimate strength).
The impact target used was unyielding.
The staff has determined that stresses in the cask body are' acceptable for a bottom end impact from 18 inches.
The 7-foot side drop is discussed in Section 4A.5.6.3.B of the TR.
The results of the 3-D finite element confirmatory analysis of this handling accident show that the maximum stress intensity in the cask body is greater than the allowable of 70 ksi (ultimate strength).
The staff has determined that stresses in the cask body are acceptable for a side impact from 18 inches.
The analysis of the 8-foot corner drop (cask center of gravity over corner) accident is described in the TR in Section 4A.5.6.3C.
The confirmatory intensity in the cask body is greater than the allowable of 70 ksi (ultimate strength).
The staff has determined that stresses in the cark body are acceptable for a corner impact from 18 inches.
Severe local deformation at the point of impact will occur anr is considered acceptable.
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.s-The=tipover analysis is discussed in Section 4A.5.7 of the TR.
A confirmatory firite element analysis was performed *.)y the staff for the tipover condition.
The confirmatory analysis shows that the stresses in the cask body are accept-able.
Local yielding will occur at the point of impact.
The potential also exists that the closure seal would be lost and some bolts broken as a result'of this accident.
-i During the course of the review, the requirement for handling and storage of the loaded cask in a horizontal position was eliminated and an impact limiter was added to the top of the cask to mitigate the loads due to a tipover accident (letter E-10867 from Donald Nolan to L. Rouse dated April 20,1989).
As a result, the reviewers conclude that the cask' body is acceptable, provided drop-heights for end on and corner impacts are limited to 18 inches, the loaded cask is not subjected to handling and storage in a horizontal orientation and con-cern for local plastic deformation, and bolt failure due to a tipover accident is obviated.
3.4.4.1.4 Fracture Toughness Evaluation The TR did not provide adequate information on the fracture resistance of the A-350-Lf1 cask body material.
Durin,q the course of the review, the cask mate-rial was changed from A-350 Lf1 to A-350 Lf3.
The very low ductility at about
-120 F for A-350 Lf3 assures that the cask body is fracture resistant under dynamic loads at the lowest credible temperature that can be expected and for the thickness of the cask shell indicated.
Consequently, brittle fracture is not a credible failure mode for the cask body.
3.4.4.1.5 Cask Thermal Stress Analysis l
l The' thermal stress analysis for the TN-24 cask in Section 4A.5 of the TR was i
reviewed to ensure that-the containment would not fail under the assumed loading conditions.
The requirement for structural integrity can be met if, by using l
ASME code methods, it is demonstrated that the maximum primary plus secondary stress intensity range is less than three times the design stress, intensity (3 Sm).
The temperature gradients throughout the cask body are presented in 3-6 l
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Table 3.2-7 of'the TR.
The loading due to these gradients were incorporated into Load Case A2 which, according to Tables 4.2.8b, 4.2.8c, and 4.2.8d, indicate that the stresses are well below the allowable limit.
The staff concurs with this result.
In addition, the staff concurs with the statement on Page/4.2-6 regarding thermal stresses, since they are secondary, they need not be evaluated under accident conditions.
Clearance between the basket and the cask cavity were calculated and presented in Table 4A.6-2 of the TR.
Since there is always a positive clearance, the basket is not subject to interference loads, and no further thermal analysis is required.
Based on the review of the thermal stress analysis in the TR, it is concluded-that the cask containment will not fail.
The thermal analysis in the TR may be referenced in a site-specific license application provided that the site environmental conditions are within the thermal cases analyzed.
3.4.4.1.6 Tornado-Generated Missiles a
Tornado generated missiles that may damage the cask are described in NUREG-0800.
All missiles are assumed to impact the cask at 35 percent of the maximum wind-speed, which is defined in Regulatory Guide 1.76 to be 360 mph, thus the maximum missile velocity is 126 mph.
The point of application and orientation of the missile is that which can cause the greatest amount of damage.
NUREG-0800 also recommends that 70 percent of the postulated horizontal velocities be used, in this case 88.2 miles per hour, to assess damage caused by the vertical impact of missiles, except for the small rigid missile described below.
Types of missiles described include:
1.
A massive high kinetic energy object that is deformable tcpon 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 in.) 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.
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m In accordance with the criteria for radiological protection described in Sec-tion 2.9 of this SER, the cask must maintain its structural integrity under the impact of tornado generated missiles.
The TR addresses the subject of tornado generated missiles in Section 3.2.1.
The. analysis presented shows that the massive high kinetic energy missile will not cause cask tipover,'if the cask is hit on its side.
The reviewers agree with this conclusion.
There may be some damage to the neutron shield due to the impact of the massive missile on the cask; neutron shield damage is addressed in Section 3.4.4.2 of this SER.
The staff determined that the rigid 125 kg (2n. T 1rmor piercing shell posed the. greatest damage potential to the cask.
The TR addresses this missile in Section 3.2.1.3b.
The analysis shows that the cask will not tipover due to an impact with the missile on the side, and that the cask will not be penetrated by the missile.
The analysis also shows that a missile impacting the top of the cask would penetrate through the weather protective cover, but not the cask lid.
Some local plastic deformation would occur at the point of impact of the outer cask wall or lid during an impact with this missile.
The TR addresses the effect of the 2.5 cm solid steel sphere in Section 3.2.1.3c.
The analysis shows that this missile does not cause tipover and does not pene-trate the cask.
The analysis shows that it might cause some plastic deformation on the cask surface or weather protective cover.
Since there are no openings in the protective barrier represented by the cask body and lids, this small missile will not cause any other damage to the cask.
3.4.4.2 Neutron shield 3.4.4.2.1 Normal Operating Loads The analyses of the neutron shield shell are provided in Section 4A.8 of the TR.
The neutron shield shell supports borated polyester resin contained in aluminum shells.
The neutron shield shell is comprised of a 0.75-inch thick steel cylinder 94.75 inches 0.D. and two steel end disks 1 inch thick.
The steel material is SA 516 Grade 55.
The neutron shield shell is analyzed for a 3-8
I pressure of 25 psig inside the shell and outside the shell in the TR.
The shell is also analyzed for handling loads of 3 g's in the vertical direction'and 5 g's in the horizontal. direction.
The staff has reviewed these analyses and found L
the' stresses resulting from these loads to be less than the allowable stresses
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for the steel material.
-3.4.4.2.2 Environmental Loads and Natural Phenomena.
While'there are environmental and natural phenomena loads on the neutron' shield, the staff does not expect the structural integrity of the neutron shield to be affected by these loads, with the exception of the tornado missiles.
The tor-nado missile loads on the neutron shield are discussed in Section 3.2.1.3 in the TR. There is a good chance that part or all of the neutron shield will be damaged by a tornado missile; however, the cask integrity will not be affected, as discussed in Section 3.4.4.1.6, of this SER.
The shielding analysis for this condition is discussed in Section 11.3.2.1 of this SER.
3.4.4.2.3 Accidents The analysis of the response of the neutron shield to end drop impact loads is not addressed in the TR.
The staff assumed the neutron shield will fail under this accident.
The radiological consequences of this failure are discussed in Section 11.3.2.1 of this SER.
' Most of the energy from a tipover condition will be absorbed by the upper impact limiter for the cask.
It is possible that a portion of the lower neutron shield will be damaged by this condition.
This will not affect the overall cask integrity. A shielding analysis for the damage neutron shield condition is discussed in Section 11.3.2.1 of this SER.
3.4.4.3 Fuel Basket 3.4.4.3.1 Normal Operating Loads The TR addresses the arealysis of the fuel basket in sections 4.2.1.2 and 4A.6.
Since the loaded TN-24 cask is limited to a vertical orientation for handling 3-9
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and normal operating conditions, the basket members are loaded only their own weight in this orientation.
An analysis is provided in the TR for the normal handling loads.
3.4.4.3.2 Environmental Loads and Natural Phenomena The only consequence of environmental or natural phenomena on the fuel basket.
would result from' cask tipover.
The analysis for tipover is reviewed in the following section.
1 3.4.4.3.3. Basket Accident Loading l
The staff performed confirmatory 3-D finite element analyses on the basket with fuel in place for an 18-inch horizontal (side) drop.
Several impact targets were used and stresses and plate buckling were compared to allowables.
Several orientations of the basket plates relative to the impact surface were also considered.
At the 18-inch drop height onto an unyielding target, it was determined that stresses in the basket were not acceptable and that basket plates were likely to buckle.
The basket was also analyzed using a 3-foot thick layer of concrete resting on an unyielding surface as the impact target.
The concrete properties used were those specified in the TR, and the concrete was modeled using a psuedo-tensor concrete / geologic model.
Again, stresses were not acceptable, and there was a potential for buckling.
When the effect of the unyielding surface below the concrete was relaxed, stresses and buckling were still not acceptable.
The staff, therefore, concluded that the integrity l
i of the basket with fuel in place at this 18-inch drop height could not be established.
J A confirmatory analysis of the basket subject to a corner drop was performed.
It was determined that the stresses induced in the basket due to this accident were less than the allowable stress.
During the course of the review the design of the basket was revised by extending j
basket plate L5 to the cask wall, where it is supported along with plate L2 in j
a double rail, and by increasing the basket plate thickness from 8 mm to 11 mm j
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(letter Donald Nolan to L. Rouse E-10795 dated March 16,1989).
In addition, as mentioned before, the requirement for horizontal handling and storage was eliminated, and an impact limiter was added to the top of the cask to mitigate the loads due to a tipover accident.
A confirmatory analysis was performed using these ' revised conditions with the-result that the stresses induced into the basket due to tipover were less than the allowable stress for the basket material.
Basket orientations of 0, 15, 30, and 45 degrees were considered in this analysis.
3.4.4.3.4 Fracture Toughness Evaluation The resistance to brittle fracture of the basket under postulated accidental loading conditions is assured provided the material is procured in accordance with paragraphs 4.2.2.7, 4.2.2.8, and 4.2.2.9 of Appendix 4B of the TR.
3.4.4.4 Trunnions and Trunnion Bolts 3.4.4.4.1 Normal Operating Loads The TN-24 h.as six trunnions:
four trunnions are spaced at 90 degree intervals near the top of the cask; two trunnions are spaced 180 degrees apart near the bottom of the cask.
The trunnions consist of an outer shoulder (small diameter) used for lifting, and an inner shoulder (large diameter) used for rotation of the cask between horizontal and vertical and for support during transfer.
The trunnions are designed to support a vertical lift on the outer shoulders equivalent to three times and five times the weight of a fully loaded (including water) cask.
For a load three times the weight, the maximum tensile stresses siiall not exceed the minimum yield strength of the trunnion material.
For a load five times the weight, the maximum tensile stresses shall not exceed the minimum ultimate tensile strength of the trunnion material.
For tilting, th<
trunnions are evaluated for a 3 g load on the inner shoulder.
The stresses are limited to the minimum specified yield strength of the trunnion material.
For transfer, the trunnions are evaluated for the combined inertia g loads of 2 g vertical, 2 g longitudinal, and 1 g lateral.
The stresses are limited to the minimum specified yield strength of the trunnion material.
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4 The analysis for these conditions is contained in Section 4A.7 of the TR.
The analysis uses a cask design weight of 225,481 lbs.
In the calculation of the stresses in the trunnion shoulders, the formula used to calculate the maximum shear stress (V/A) is actually the average shear stress across the section.
Assuming the maximum shear stress to be 2V/A for a thin-walled cylinder (slightly conservative), the maximum shear stresses listed in Table 4A.7-1 would double in magnitude.
The resulting stress intensities also increase.
Comparison of the stress' intensities based on 2V/A with the allowable stresse:: indicate that the trunnion shoulders are acceptable.
3.4.4.5 Upper Side Impact Limiter The impact limiter is discussed in document E-10843 that is attached to letter E-10867 dated April 29, 1989.
The impact limiter is provided on the top of the cask to mitigate. inertial loadings in the event of cask tipover.
The limiter is a simple annular shaped unit comprised of redwood, supported by radial steel gussets sealed in a carbon steel shell.
The maximum deceleration in the cask was determined to be 43 g using this limiter.
More information on the upper impact limiter is given in letter E-10929 from Donald Nolan to L. Rouse dated May 25, 1989, and in letter E-10995 from Donald Nolan to William Lloyd dated June 7, 1989.
A confirmatory analysis was performed on the basket using an inertial loading of 50 g.
The difference between 43 g and 50 g is due to a dynamic amplification factor of 1.15 between the cask and the basket.
The orientation of the basket relative to the ground was varied from 0 to 45 degrees.
It was determined that 4
the stresses in the basket were acceptable for the tipover accident.
3.4.4.6 Lower Impact Limiter Detailed information was not provided in the TR for a lower impact limiter.
The cask shall not be lifted more than 18 inches vertically without a lower impact limiter.
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3.4.4.7 Bolted Covers
.The TR addresses the analysis of the bolted covers in Section 4.2.1.1, while.
m the results of the analysis of the bolts, lid, and cask are given'in Table 4.2-8.
An independent' analysis of the main closure lid system as well as the penetra-
. tion. cover systems'was performed to confirm that the stresses in these systems do not exceed the allowable design stress of the materials used.
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.3.4.4.7.1 Main Closure Lid System 3.4.4.7.1.1 Bolts-The' forty eight, 1.5-inch diameter lid bolts are made of SA-320 grade L43 steel.
The bolt analysis described in Section 4A.5.3 of the.TR specifies a preload sufficient to seat the metallic seals and_ maintain the seal under any loading
' conditions.
The gasket load for seating is 4,474 lbs per lineal inch which requires a preload per bolt of at least 19,346 lbs.
However, a torque specified
-such that a bolt stress equal, at least, to 25,000 psi is generated.
This corresponds'to a bolt preload of about 37,000 lbs.
The force due to the maximum internal: pressure of 250 psi is 17,864 lbs per bolt.
Thus, the total bolt load due to seal seating and pressure is 37,210 lbs which indicates that the preload is sufficient to maintain the seal.
Since the thermal coefficient of expansion of the bolts is slightly larger than that of the lid, a small reduction of bolt tension results when the lid is heated above the room temperature'at which the bolts were torqued.
This is compensated by applying a somewhat higher torque at room temperature.
The TR also considers the effect of the rotations at the junction of the lid and the cylindrical shell that ' induce bending stresses in the bolt.
The combination of tensile and bending stress of 47,200 psi is below the maximum allowable of 3 Sm of 93,600 psi.
3.4.4.7.1.2 Main Closure tid The primary stresses in the lid are from the pressure load and the gasket load.
The secondary stresses are thermal stresses from the temperature gradients in the radial direction of the lid and across the lid thickness.
The sum of the 3-13
primary loads with a pressure of 250 psi and an estimate of the thermal stresses is less than the design stress intensity for the SA-350 Grade Lf1 lid material.
For a pressure of 250 psi. from an accident, the sum of the primary and secondary stresses is less than 1.5 times the design stress intensity limit.
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.3.4.4.7.2 Penetrations l
l A confirmatory analysis of the bolts used to attach the penetration covers to the main closure lid of the storage cask was performed.
This analysis confirms that the tensile stress in the bolts from an. accident pressure of 250 psi does not exceed the yield strength of the bolt material.
3.4.4.8 Fuel 3.4.4.8.1 Area of Review In this section, the integrity of the fuel rod cladding is evaluated for compliance with the requirement of 10 CFR Section 72.122(h).
The system to be reviewed consists of pressurized Zircaloy cylinders in an inert helium atmosphere.
The fuel rod temperature limit of 339 C (716 F) is
-specified in the TR.
3.4.4.8.2 Acceptance Criterion The requirements of 10 CFR Section 72.122(h) will be met if it can be demon-strated that, for the design configuration of TN-24 cask, damage accumulation is negligible at the end of storage life.
3.4.4.8.3 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.
There are
- 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, 3-14
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g 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 reference 10).
-3.4.4.8.4 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 temperature manifests itself by accelerating the rate of degradation mechanisms activated by both stress and corrosion.
Stress corrosion cracking (SCC) occurs as a result of synergistic combination 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, in part, to the large external pressure on the fuel rods.
The only known cause of cladding l
failure due to SCC occurred in a reactor during a ramp-up.
No other failures 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 accumula-tion 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 condi-tions.
The possibility exists, however, that cracks may be present that were 3-15 l
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' 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 r
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 Zircaloy have been know to cause cracking by embrittling the cladding.
Terminal solubilities of hydrogen'in Zircaloy increase with temper-1:
ature.
If the temperature subsequently decreases, hydrides wil.1 precipitate in l'
an orientation determined by the stress level.
Normally the hydride precipitates l
in a circumferential direction and is not a problem even at hydrogen concentra-tions 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 (725 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 (725 F) could be tolerated for creep rupture lives well i
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 tnan those that prevail under dry storage conditions.
)
The creep damage mechanisms in the high temperature, high stress regime are 3-16
__=
l
- ..+
l 1
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.
The staff examined this matter to determine potential mechanisms for significant creep damage under dry storage conditions applicable to the case of the TN-24 cask.
The only mechanism for any of the failure modes considered above that the staff found which represented a possible potential for cladding degradation and grocs rupture 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 reference 10).
l l
A confirmatory thermal analysis revealed that the maximum temperature _for the fuel rods could be as high as 339 C (642 f) with a temperature decay curve as shown in Figure 3-1.
Assuming a maximum room temperature fill gas pressure of 475 psig (490 psia) as specified on page 2A-321 of reference 9 and computing the pressure at the temperature limit of 339 C., the progress of damage based l
upon the DCCG analysis indicated the area of decohesion after 20 years of 1
storage would be less than 10 percent.
Consequently, an initial storage temperature not exceeding the design limit of 339 C (642 F) is acceptable for meeting the requirements of 10 CFR 72 Section 72.122(h).
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4 THERMAL EVALUATION 4.1 Normal Conditions 1
4.1.1 ' Area Of Review The thermal analysis e esented in the TR was reviewed to evaluate the protection provided to prevent fuel cladding degradation and gross ruptures in compliance with 10 CFR Section 72.122(h).
The cask body is a right circular cylinder of SA-350', Grade LF3 forged steel that has a 248 mm (9.75 in.) thick wall and provides a cavity that is 1600.2 mm (63 in.) in diameter and 4146.55 mm (163.25 in.) long.
The TR provided a thermal analysis for transporting and storing 17 x 17 0FA (optimized. fuel assembly) PWR fuel. The maximum heat output of any intact fuel rod. bundle is 1.0 kW just prior to loading into the cask.
4.1.2 Acceptance Criteria The requirements of 10 CFR Section 72.122-(h) can be met if it is demonstrated that, for the TN-24 Storage cask, the maximum fuel cladding temperature does not exceed a limit specified in Section 3.4.4.8.4 of the SER that would ensure negligible cladding damage over the design storage life of 20 years.
4.1.3 Review Procedure The thermal analysis in the TR was reviewed and confirmatory calculations were performed to ensure that the fuel rod cladding temperature does not exceed that specified in the TR.
The steady-state thermal analysis in the TR was performed with the finite difference codes HEATING 5 and SCOPE, assuming a solar absorptivity
- of 0.3 for the cask surface.
The staff did an independent confirmatory transient analysis with the finite element code TOPAZ 20.
In this confirmatory analysis,
- a power peaking factor of 1.1 was assumed in the hottest region of the cask basket, the solar absorptivity of the cask surface was assumed to be 0.3, a time 4-1
p.
t-averaged value of insolation was used, and a modified Wooton-Epstein Correlation was used to calculate the maximum temperature of the fuel cladding.
4.1.4 Findings and Conclusions The maximum cladding temperature specif"ed in the TR in Table 5.1-3 is 339 C l
under normal conditions. The confirmatory analysis performed by the reviewers revealed that for a peak pow r output of 1.0 kW per assembly, the maximum clauding temperature would be 341 C (645 F) using a basket emmissivity of 0.7 and a Westinghouse 17 x 17 0FA fuel bundle.
Since the modified Wooton-Epstein l
correlation is a conservative method for estimating the temperature rise across a fuel bundle, it is concluded that the fuel cladding will remain below 341 C (645 F) during storage to prevent cladding degradation and gross rupture in compliance with 10 CFR Section 72.122(h).
The thermal analysis in the TR is acceptable for referencing provided that the maximum heat output of any single assembly does not exceed 1.0 kW, the total heat content stored within the basket does not exceed 24 kW, the surface absorptivity of the outer shell of the cask does not exceed 0.30, the basket emmissivity is greater than 0.70, and the storage cask is back-filled with helium.
Operations during cask loading that occur within the reactor spent fuel pool area are described in Tables 5.1-1, 5.1-2 and Section 5.1.2 of the TR, including characterization of the fuel, 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 a pressurization with helium.
Procedure for cask loading, unloading (including sampling and fuel cool-down),
and decontamination, as adapted for site-specific conditions and use, will be described iri detail by a license applicant.
4.2 Accident Conditions 4.2.1 Explosion 10 CFR Part 71 requires an evaluation of 20 psia over pressure for transportation casks. The TR includes a calculation to show that an external pressure of 104 psia 4-2
f is required for the initiation.of the yielding of the cask outside surface.
It
-is concluded-that'the TN-24 cask is structurally adequate to withstand any j
credible explosive overpressure.
4.2.2 Fire i
4.2.2.1 Area Of Review The thermal analysis for an accidental fire was reviewed in the TR to determine if'any radioactive release could occur in violation of 10 CFR Section 72.106.
l The TR assumed that the cask is exposed to a 800 C (1472 F) engulfing fire for 1/2 hour.
4.2.2.2 Acceptance Criteria The requirements of 10.CFR Section 72.122(c) can be met if it is demonstrated that the fuel rod cladding temperature remains below 380 C (716 F).
4.2.2.3 Review Procedure The confirmatory analysis on the fire accident was performed with a two-dimensional finite element code, and the fuel rod cladding temperature was calculated with the modified Wooton-Epstein Correlation.
The analysis assumed loss of the neutron shield prior to the fire, and an external cask absorptivity i
i of 0.8.
4.2.2.4 Finding and Conclusions The confirmatory analysis for the fire indicated a maximum fuel temperature of 346 C (654 F) about 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after initiation of i.he fire using an absorptivity of 0.8.
The maximum fuel rod temperature during and after a fire will be below 380 C.
The cack design is structurally adequate '
meet the requirements of 10 CFR Sections 72.122(c) and 72.122(h).
i 4-3
P-l l
4 5 SHIELDING EVALUATION
' 5.1 Area of Review Section 72.104(a) of 10 CFR Part 72 requires that during normal operations and anticipated occurrences, the annual dose equivalent to any individual 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 operations, and (3) any other radiation from uranium fuel cycle operations within the region.
Section 72.106(a) of 10 CFR Part 72 requires that for each ISFSI site, a controlled area shall be established.
Since the TR is generic in nature, no specific controlled area has been established.
However, the TR has provided a dose rate calculation (see Figure 7.3-6 of the TR), which for a distance of 250 meters _(820 feet) from a single cask, yields an annual dose of less than 25 mrem to an individual assumed to be continuously present at that distance.
In addition to the above, TN addresses the shielding design criteria in TR.
Sections 1.2.2 (Principal Design Criteria), 1.2.5 (Structural Features),
3.3.5.2 (Shielding), 7.1.2 (Design Considerations), and supplemental letter E-10703 dated January 31, 1989.
The revised maximum design dose rate (neutron
+ gamma) is 125 mrem /hr at any cask surface.
TN calculated values are given in i
Table 7.3-4 (TN-24 Dose Rates at Short Distances) of the TR.
l Approximation of measured dose rates are derivable from information supplied in supplemental letters E-10697 dated January 26, 1989, and E-10703 dated January 31,
.1989.
Maximum values are 11.9, 121.9, and 49.9 mrem /hr at the surface of the cask top, side, and bottom, respectively, and represent approximated measured dose rates.
The maximum dose rate at the cask side occurs below the neutron shield and will be accessible.
5-1
y
- 5. 2 : Acceptance Criteria k
Since the TR must be generic in its approach and cannot address site-specific conditions'for a license applicant's given array configuration and size, the case of a single cask is examined to evaluate cask shielding design adequacy.
Arbitrarily, we have set the minimum distance to the site boundary at 250 meters (820 feet).
For the case of a single loaded TN-24 cask and the conservative assumption that an individual is continuously present, cask shielding is accept-able if it can be shown that the annual dose to an individual at the site boundary does not exceed 25 mrem.
5.3 Shielding Review Procedure 5.3.1 Source Specification 5.3.1.1 Gamma Source The TR addresses the gamma source for the active fuel region of the spent fuel elements in Sections 3.1.1 (Materials to be Stored) and 7.2.1 (Characterization of Sources).
Supplemental information is also provided in TN letters E-10697 dated January 26,1989, and E-10703 dated January 31, 1989.
No axial source distribution is described.
The gamma source strength is determined from an ORIGEN2 calculation using the Westinghouse 17 x 17 array fuel assembly described in Tables 3.1-1 (Physical Characteristics of Westinghouse fuel Assemblies) and 3.1-2 (Thermal, Gamma and Neutron Sources for the Design Basis 17 x 17 PWR Fuel Assembly).
In this calculation, the average burnup is 35,000 MWD /MTU, the specific power is 37.5 MW/MTV, the initial fuel enrichment is 3.3 percent (E-10697), and 461 Kg (1,014 lb) per assembly of heavy metal is considered.
j The irradiation time and interim shutdown periods used are not stated in the TR.
The cooling time for the spent fuel used in the shielding evaluation is 5 years.
Activation of the hardware in the active fuel region is included in the gamma source strength (E-10697); activation of the cladding material is not.
L The gamma sources for the top and bottom end fittings and the gas plenum regions are addressed in TR Section 7.2.1 (Characterization of Sources) and TN letter E-10697 dated January 26, 1989, The gamma source strengths are determined from 5-2 m
ORIGEN2 calculations using the spatially and spectrally adjusted neutron flux
_' generated by ORIGEN2 for the active fuel region.
5.3.1.2 Neutron Source l
I The TR addresses the neutron source for tie active fuel length of the spent fuel q
elements in Sections 3.1.1 (Materials to be Stored) and 7.2.1 (Characterization of Sources).
No axial source distribution is described.
The neutror. source I
strength is determined from an ORIGEN2 calculation using the Westinghouse 17 x 17 array fuel assembly described in Tables 3.1-1 (Physical Characteristics of Westinghouse Fuel Assemblies) and 3.1-2 (Thermal, Gamma and Neutron Sources for the Design Basis 17 x 17 PWR Fuel Assembly).
The major input parameters for this calculation are described in Section 5.3.1.1 (Gamma Sources) of this SER.
Suberitical multiplication is automatically included with the neutron transport calculations.
5.3.2 Model Specification The shielding model is addressed in TR Section 7.3.2.2 (Shielding Analysis) and TN letter E-10697 dated January 26, 1989.
TN assumes one and three dimensional geometries with a homogenized circularized spent fuel array.
Basket material densities are reduced by 75 percent in the top and bottom end fitting and plenum 3
zones to account for the reduced shielding effectiveness of the basket in the axial direction.
Vent, drain, and pnrt penetrations into the closure lid, and streaming through the aluminum boxes in the radial neutron shield are not specifically modeled.
In the staff evaluation, the TN spent fuel array was transformed into an equivalent cylindrical surface source for the radial analyses and an equivalent disk surface source for the axial analyses.
Shield model dimensions were those supplied by TN.
s 5-3
- I The effectivetshield thicknesses in centimeters for'the TN-24 cask were
)
i Stainless Neutron Location Iron Steel Shield Top 29.21 1.0 11.83 i
~ Bottom 28.50 0
11.50 Side 24.00 2.0 13.50 Points at the' top, bottom, and side were evaluated both'for normal and accident conditions.
Neutron dose, neutron induced gamma dose, and gamma dose were assessed at the most vulnerable points.
Streaming effects were considered during loading conditions.
I 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 (Shielding Analysis) and TN letter E-10697 dated January 26, 1989.
Dose point locations are at the axial midplane for the radial values and on the cask centerline for the axial values.
Figure; presented for the radial e.nd axial shielding configurations appear adequate for the TN model.
-5.3.2.2 Shield Regional Densit,ies Material densities (gm/cc) and atom number densities (atoms / barn-cm) are addressed in Section 7.3<2.2 (Shielding Analysis) of the TR and TN letter E-10697 dated January 26, 1989.
1 Material densities for 0 in the fuel / basket zone, H in the polypropylene/ steel and resin / aluminum zones, and B in the resin /alumf. um zone have been omitted from the TN QAD-CGGP analyses, i
The. atom number densities for U-235 and U-238 in the fuel / basket zone used in by TN in the XSDRNPM analyses reflect an enrichment of 3.7 percent instead of the 3.3 percent used in the ORIGEN2 source term calculation.
5-4
l Elemental and material density data used in the staff shielding evaluation were those supplied by TN.
l 5.3.3 Shielding Evaluation l
The TR addresses the shielding evaluation in Sections 3.3.5.2 (Shielding) and I,
7.3.2.2 (Shielding Analyses).
Additional information is also provided in letters E-10697 dated January 26, 1989, and E-10703 dated January 31, 1989.
l The TR shielding calculations are performed with the QAD-CGGP and XSDRNPM codes; flux-to-dose rate conversion factors are those of ANSI /ANS-6.1.1.
Effects of the fire and loss of neutron shield accidents are calculated from the model.
The SER shielding calculations are performed with COG (reference 4).
COG uses the Monte Carlo method to transport both neutrons and gamma rays.
The neutron and gamma dose at the side surface of the cask were determined as was the average gamma dose at the top and the bottom of the cask.
5.4 Findings and Conclusions TN's approximated raeasured total (neutron + gamma) maximum surface dose rates for normal conditions are 11.9, 121.9, and 49.9 mrem /hr at the top, side, and bottom, respectively, and are less than 125 mrem /hr at any accessible surface.
Staff calculations confirm the TR results.
For computation of the annual dose. commitment, TN and the staff have assumed the dose rate at the active fuel midplane as representative of the cask average.
Annual dose commitment at 250 meters from a single cask to an individual, conservatively assumed to be continuously present, is calculated by the staff to confirm the TN value of 16.2 mrem /yr which is less than the 25 mrem /yr allowed under Section 72.104(a).
For arrays involving more than one cask, a license applicant will have to assess the conditions for the site concerned and the total number of casks to arrive at a suitable distance.
l 5-5 4
6 CRITICALITY EVALUATION 6.1 Area of Review In compliance with 10 CFR Section 72.124, the criticality analysis presented in the TR was reviewed to determine if the TN-24 cask is designed to be sub-critical and to prevent a nuclear criticality accident.
The TN-24 cask with its fuel basket is described in Section 1.2 of the TR.
The cask body is a right circular cylinder of SA-350, Grade LF3 forged steel that has a 248 mm (9.75 in.) thick wall and provides a cavity that is 1600.2 mm (63 in.) in diameter and 4146.55 mm (163.25 in.) long.
The upper end of the cask body is sealed with an SA-350, Grade LF3 steel bolted closure lid which is 292 mm (11.50 in.) thick. A protective cover is bolted to the cask body to provide weather protection for the lid penetrations.
The closure lid utilizes a double barrier seal system with two metallic 0-rings forming the seal.
The annular space between the metallic 0 rings is connected to a helium filled tank placed between the lid and the protective cover.
Pressure in the tank is maintained above pressure in the cask to prevent either flow of fission gasses out or air into the cask cavity which, under normal conditions, is filled with helium.
The lower end of the cask is welded to the sides of the cask and is made of SA-350, Grade LF3 steel that is 286 mm (11.25 in.) thick.
Neutron shielding is provided by a borated polyester resin compound surrounding the cask body.
The resin compound is cast into long, sleno'er aluminum containers.
The array of resin-filled containers provides a 136.65 mm (5.38 in.) thick shield and is enclosed within a 19.05 mm (0.75 in.) smooth outer shell constructed of SA 516, Grade 55 steel in two half cylinders.
A disk of poly-propylene 101.6 mm (4 in.) thick, encased in a 6.35 mm (0.25 in.) SA 516, Grade 55 steel shell is attached to the lid after fuel loading to provide neutron shielding.
The spent fuel assemblies are supported by a basket of proprietary design.
This design provides for support of 24 intact spent fuel bundles.
The fuel 6-1
- e
- i assembly analyzed'in'the TR is the Westinghouse 17 x 17.0FA (optimized fuel assembly).
This fuel, assembly design is stated in the TR to yield higher L
k-effective values with' respect i similar analyses using a 14 x-14 or a 15 x;15 PWR design.
Each spent.ael bundle location has an inside-dimension of.
about 22.098 cm x 22.098 cm (8.7 inches x 8.7 inches) square.
Each fuel assembly may have a maximum initial enrichment of 3.5 w/o U-235 (See Section 6.4 of this SER) and a maximum of 35,000 MWD /MTU burnup.
Boron-stainless steel neutron L
absorbing material ne.ar the fuel is required for criticality control during underwater fuel loading and unloading.
The criticality analysis presented in the TR was performed with AMPX cross-section processing codes that included BONAMI and NITAWL, and the Monte Carlo criticality code KENO-Va combined with the 27 group cross-section set based on the Evaluated Nuclear File, version B-IV data.
These codes and cross sections are on the SCALE system available from the Reactor Shielding Information Center, Oak Ridge, Tennessee.
The criticality analysis in the TR is based on the following assumptions:
(1) the fuel is enriched to 3.7 w/o U-235 in uranium; (2.) the fuel is unieradiated;-(3) the boron content in the boron-stainless steel plates is 1.0 l
w/c Boron with about 18.5 w/o B-10 in the Boron; (4) the fuel pins, basket, and
{
cask were modeled discretely at 20 C (68 F); and (5) a two-dimensional model is I
used such that the cask is assumed infinite along the vertical axis (the lid and cask bottom were not modeled).
The staff found k-effective to be greater i
than 0.95 for the off-center placement of the fuel bundles in the basket when modeled in three dimensions.
Ste Section 6.4 of this SER.
The calculation method and cross-section values which were used in the criticality analysis in the TR were verified by comparison with critical experiment data j
for assemblies similar to those for which the cask was designed.
Five critical experiments were analyzed.
These experiments considered water moderated, oxide fuel arcays separated by various materials (Boral, borated steel, and water, for example) that simulate Light Water Reactor (LWR) fuel storage conditions.
K effective results were calculated for the models of each of these critical i
experiments.
From these k effective results, the bias of the computational tool was determined.
See Section 3.3.4.3 of the TR.
6-2
g.
- 4,..e 4
.j u,
- 6. 2 Acceptance Crite'ria u
The requirement of 10 CFR Section.72.124 can be met if it is demonstrated that, for the TN-24 cask design, the effective multiplication factor is less than 0.95 (Keff < 0.95) for all credible configurations and environments.
6.3 Review Procedure
'The critic #,ty analysis in the TR was reviewed and verification calculations l
were performed for comparison to ensure that the TN-24 design is subcritical at l
j all times during transport'and storage.
The criticality review for the 24 fuel bundle basket configuration was performed with the KENO-Va code combined with the 123GROUPMTH cross-section set in refer-i ence 5.
Criticality calculations using these computational tools were performed on an IBM 3033 mainframe computer at the Oak Ridge National Laboratory (ORNL).
The criticality review was based on the following assumptions:
(1) the fuel was enriched to 3.5 w/o U-235 in uranium; (2) the fuel was unirradiated; (3) the boros content'in the boron-stainless steel plates is 1.0 w/o Boron with about 18.3 w/o B-10 in the Boron; and (4) the fuel pins, basket, and cask were modeled discretely at 20 C (68 F).
The cask and its contents were modeled in three dimensions.
The fuel; assembly analyzed in the SER was the Westinghouse 17 x 17 0FA (optimized fuel assembly).
The fuel rods, basket, and cask were modeled discretely in three dimensions.
Both centered and off-centered loading of the fuel bundles in each basket location were calculated.
The off-centered loading had the fuel bundles placed as closely as possible near the corner of each hole in the basket nearest the center of the cask.
The calculation method and cross-section values used in the criticality review were verified by comparison with critical experiment data for assemblies similar to those'for which the cask was designed.
Four critical experiments were analyzed.
6-3 2x__
i
. j I
i
' l These' experiments included water moderated, oxide fuel arrays separated by
. borated stainless steel plates on two sides of a linear three fuel bundle array for two different UO enrichments (references 6 and 7) at near optimum water 2
moderation.
From these k-effective results, the bias of the computational tools used in the criticality review were determined.
l 6.4 Findings and Conclusions The largest'k effective value reported in the TR is 0.941 for the two-dimensional a
model with each fuel bundle centered in each hole in the basket.
Thie is un upper limit value at 95 percent confidence.
This stems from a calculated k-effective value of 0.9206 + 0.0044.
The correction to 0.9206 is due to the application of the bias for the computational tool from calculated results of the five critical experiments.
A multiplier of 2 on the one-sigma values was used for the upper limit at 95 percent confidence, as described in the TR.
The largest k effective value found in the cor.firmatory analysis was 0.95 for
.off-centered loading of the fuel bundles in each basket location.
The off-centered. loading had the fuel bundles placed cs closely as possible near the corner of each hole in the baskre nearest the center of the cask.
This is an upper. limit value at 95 percent confidence with biases applied Mist calculations were also performed in the confirmatory analysis for-watsr densities between 0.001 and 1 grams /cc in the cask.
No k-effective values higher than 0.95 were found.
The confirmatory analysis models of the TN-24 cask bound the actual fuel and basket configurations and materials.
The calculated results for these models show the peak k-effective to be less than the maximum acceptable design limit of 0.95 for W-17 x 17 0FA fuel enriched to 3.5 w/o U-235 in uranium in the UO 2 for 24 fuel bundles in the basket.
On the basis of the TR evaluation and the confirmatory analysis, the staff concludes that the TN-24 cask is designed to be maintained subtritical and to prevent a nuclear criticality accident in l
compliance with 10 CFR Section 72.124.
6-4
_m.-_.
mm___
mm
b 7 CONFINEMENT 7.1 Area of Review The confinement analysis presented in the TR was evaluated to ensure that the annual doses specified in 10 CFR Section 72.104(a) are not exceeded during normal operations and anticipated occurrences.
The TN-24 cask with its fuel basket is described in Section 1.2 of the TR.
The cask body is a right circular cylinder of SA-350, Grade LF3 forged steel that has a 248 mm (9.75 in.) thick wall and provides a cavity that is 1600.2 mm (63 in.) in diameter and 4146.55 mm l
(163.25 in.) long.
The upper end of the cask body is sealed with an SA-350, Grade LF3 steel-bolted closure lid which is 292 mm (11.50 in.) thick.
A I
protective cover is bolted to the cask body to provide weather protection for the lid penetrations.
The closure lid utilizes a double barrier seal system with two metallic 0-rings forming the seal.
The annular space between the metallic 0-rings is connected to a helium-filled tank placed between the lid and the protective cover.
Pressure in the tank is maintained above pressure in the cask to prevent either flow of fission gases out or air into the cask cavity which, under normal conditions, is filled with helium.
The lower end of the cask is welded to the sides of the cask and is made of SA-350, Grade LF3 steel that is 286 mm (11.25 in.) thick Neutron shielding is provided by a
- borated polyester resin compound surrounding the cask body.
The resin compound is cast into long slender aluminum containers.
The array of resin-filled containers provides a 136.65 mm (5.38 in.) thick shield and is enclosed within a 19.05 mm (0.75 in.) smooth outer shell constructed of SA 516, Grade 55 steel in two half cylinders.
A disk of polypropylene 101.6 mm (4 in.) thick, encased in a 6.35 mm (0.25 in.) SA 516, Grade 55 steel shell is attached to the lid after fuel loading to provide neutron shielding.
7.2 Acceptance Criteria The requirements of 10 CFR Section 72.104(a) can be met if it is demonstrated that the annual doses are within regulatory limits.
l 7-1 IL
.,. 4 I
7.3 Review Procedure The confinement. analysis in the TR was reviewed and confirmatory calculations 1
were performed to ensure that the regulatory dose limits are not exceeded.
The cask is loaded with spent fuel in the storage pool.
The cask is then removed
[
from the pool, drained and helium dried.
The cask is filled with helium and l
1eak-checked to 10 6 atm-cc/sec at the primary seal.
Because of decay heat from the fuel, the pressure in the cask can increase to 2.2 atm under normal storage j
conditions.
The TR analysis assumes that fuel clad failures are 100 percent l
under accident conditions.
Under accident conditions the cask pressure can
{
increase to 3.89 atm because of the release of helium and r;oicactive gases and vapors from the assumed failed fuel rods.
For this analysis, the TR assumes i
a Westinghouse 15 x 15 PWR assembly, because it has the greatest free gas volume.
The radioactive gases considered for release are tritium and krypton with the release fractions into the cask cavity obtained from Regulatory Guide 1.25.
An independent analysis was performed to confirm the pressure in the cask for normal and accident conditions. The leakage rates for normal conditions were not determined due to the unique design of the pressurization system for this cask.
For the accident conditions, we have assumed the instantaneous release of H-3 and Kr-85.
7.4 Findings and Conclusions For the accident conditions of 100 percent fuel rod cladding failure and instantaneous release, the expected H-3 and Kr-85 releases are 712 Ci and 23,5SO Ci, respectively.
The dose consequences of these activities are discussed in the following section.
7.5 Confinement Requirements for the Hypothetical Accident Conditions 7.5.1 Area of Review Section 72.24(m) of 10 CFR Part 72 requires, in part, analysis of the potential dose or dose commitment to an individual outside the controlled area from accidtnts i
l or natural phenomena event.s that result in the release of radioactive material 7-2
{
+-
L L
l
'to the environmental or direct radiation from the ISFSI.
Section 72.106(b) of 10 CFR Part 72 requires that any individual located on or near the nearest L
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 chosen is 100 meters to conform with the minimum allowable controlled area boundary distance required in sections 72.106(b).
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 t.? any individual a dose of 5 rem outside the controlled area.
7.5.3 Review Procedure The review consists of consideration of:
(1) the maximum gaseous activity within the cask, (2) the max @um dose from gaseous activity release.
7.5.3.1 Maximum Gaseous Activity Within the Cask The TR addresses the maximum H-3 and Kr-85 gaseous activities expected to be found within the cask in sections 7.2.2 (Airborne Radioactive Material Sources) and 8.1.2 (Radiological Impact from Off-Normal Operations).
Volatile isotopes with limited availability such as Cs-134 and Cs-137 are also identified and quantified.
Cladding tube failures of 100 percent are assumed.
7.5.3.2 Maximum Dose From Gaseous Activity Release The maximum dose expected from gaseous activity release is adoressed in Sec-tion 8.1.2 (Radiological Impact from Off-Normal Operations) of the TR.
TN assumes the available gaseous inventories of H-3 and Kr-85 are released to the environment.
Site boundaries are set at 100 and 500 meters.
The maximum dose to an individual at the minimum site boundary (100 meters) following an accident in which the available gaseous inventories of H-3 and 7-3
n
' Kr-85 are released has been calculated by-the staff.
In computing'the doses due to' gaseous activity release, the staff has assumed the following:
(1) 100 percent cladding tube failure; (2) the release fractions of Regulatory Guide 1.25;
,(3) a shielding factor of 1.0 for Kr-85; (4) the population weighted inhalation rate of Regulatory Guide 1.109; (5) the inhalation dose and whole body dose factors of Reguletory Guide 1.109; and (6) F-stability atmospheric diffusion with a windspeed of 1 meter /sec with plume meander.
p.
7.5.4 Findings and Conclusions i
The dose consequence due to gaseous activity release from a single cask following
.an accident.in which the available gaseous inventories of H-3 and Kr-85 are released is less than 0.516 rem to the whole body at a site boundary of 100 meters.
Accident consequences are less than the 5 rem established in 10 CFR Section 72.106(b),
Compliance with 10 CFR Section 72.106(b) 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 site.
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'7-4
8 OPERATING PROCEDURES Operating procedures for the'TN-24 cask are described in Chapter 5 (Operation Sy. stems) and Section 7.1 (ALARA) of the TR.
This SER review is limited to the procedures as presented by Transnuclear in this TR.
The staff does not, at this time, make any prior judgment on the operating procedures that must be included as part of the license application for an ISFSI storage facility using the'TN-24 cask.
8.1 Area of Review 10 CFR Section 72.24(h) requires the applicant to submit "a plan for the conduct of operations including the planned managerial anr' controls system, and the applicant's organization, and program for training of personnel...." While this provision applies, primarily to the ISFSI, the operations involved in loading, transporting, and storing of the spent fuel' are closely associated with the design of the cask to the extent that design features.are incorporated to facilitate the conduct of these operations.
Consequently, the review of the operating procedures is limited to the specific operations of handling the cask from the time it is loaded in the storage pool until it is placed upon the storage pad.
Managerial and administrative controls would only be relevant if the cask design were such that only administrative controls could ensure that the spent fuel could be safely handled and stored under conditions that would not pose a hazard to operating personnel or the public.
10 CFR Part 20 covers the standards for protection against radiation that must be met during the operation of a ISFS1.
Regulatory Guides 8.8 and 8.10 provide guidance to ensure that occupational radiation exposures will be "as Low as is Reasonably Achievable" (ALARA).
I The TN-24 Cask TR addresses the cask receipt, loading, and some onsite i
transportation procedures at the ISFSI.
Procedures for unloading the cask l
8-1
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S,k j
> y..
are not covered in the TR, even'though th'is' operating procedure is inseparable from decommissioning.
Detailed operating procedures.for replacing lid seals are also not covered.
This_ review only covers the inspections, tests, and special preparations of the cask for loading spent fuel.
Section 5.1.1 of the TR addresses the loading procedures while Section 7.1 of the TR addresses the issue of ensuring that the occupational radiation doses 'are ALARA.
8.2 -Acceptance Criteria i
The TR should provide a detailed description of the procedures for loading, draining and drying.the cask, creating an inert environment for the spent fuel, assuring the effectiveness of the seals at the bolted closure joints, trans-porting the loaded cask to the storage pad, and assuring that occupational radiation exposures are maintained ALARA as required by 10 CFR Section 72.24(e).
i I
8.3 Review Procedure I
TR Section 5.1.1 provides a general description of the operational procedures i
for loading the cask and preparing it for storage.
More detailed procedures L
describing the receipt and loading of the TN-24 cask at the ISFSI are described in flowsheet form in Fig. 5.1-1.
Inspections and tests are described as part of the preparation for loading.
l TR Section 7.1 describes the general procedures to be followed to meet the requirements of 10 CFR Part 20, Regulatory Guide 8.8, and Regulatory Guide 8.10.
These general procedures for radiation protection and meeting ALARA limits for
.l occupational exposure apply to the cask loading procedure.
8.4 Findings and Conclusions The operational procedures for loading the TN-24 cask at the ISFSI are in compliance with the appropriate guidance and/or regulations.
These procedures must be incorporated into the operational procedures for the ISFSI.
The proce-dures for onsite transportation not covered in the TR and for unloading must be added to the operational procedures for the ISFSI.
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9 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM 9.1 Acceptance Tests j
The TN-24 TR does not address the subject of acceptance tests and maintenance.
in any detail.
Specific mention of tests relating to decontamination, cavity pressure, surface temperature, radiation levels, and instrument function appear in Table 5.1-1.
As noted in Section 1.1 (Introduction) of this SER, the TN-24 TR 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
" Pre operational Testing and Operation." With the exception of the limited examples cited above, the TR treats this as a site specific matter in its Chapter 9.
This is acceptable to the staff for this TR. We note, however, that test procedures are required under 10 CFR 50, Appendix B.
A complete set of inspection and test procedures will be required in the license application for the ISFSI.
9.2 Maintenance Program Maintenance is addressed only briefly in the TN-24 TR.
In Section 5.1.3.5 (Maintenance Techniques) the TR states " Maintenance during normal storage is expected to be minimal.
Other than visual inspection and possible instrument calibration and paint touch-up, no maintenance is anticipated." This treatmint of maintenance is acceptsble for the TR.
However,'for a license applicant proposing to use an array of casks at an ISFSI, a detailed description of site-specific maintenance activities and procedures will be required.
L 1
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e ia 10 RADIATION PROTECTION 10.1 Area of Review 10 CFR Section 72.24(e) 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 Section 72.126(a) dates, in part, that radiation protection systems shall be provided for alI 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 i
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 Occupations 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 I
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-J specific considerations for occupational radiation exposures as low as is l
' tr reasonably achievable are in compliance with appropriate guidance and/or regulations, and that the dose from the transporting, storage, and repair of casks are not in excess of Part 20 limits.
1 I
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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 prot'ection design features, and (3) estimated onsite dose assessment.
l 10.3.1 Ensuring that Occupational Radiation Exposures are_ as Low as is Reasonably Achievable (ALARA)
Non-site-specific policy, design, and operational considerations are addressed 1
l in Sections 7.1.1 (Policy Considerations), 7.1.2 (Design ~ Considerations), and I
7.1.3 (Operational Considerations), respectively, in the TR.
The TR sections cited above describe how the TN-24 cask is designed to meet the ALARA requirements.
These requirements are met through the massive shielding, the passive nature cf 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 Transnuclear in comparison with the guidance and/or regulations cited in Section 10.1 of this SER.
10.3.2 Radiation Protection Design features Installation design features are addressed in Sections 1.1.2 (Principal Design Features of Installation), 1.2 (General Description of Installation), 1.3 (General Systems Description), 3.1.1 (Materials to be Stored), 3.3.2 (Protec-tion by Multiple Confinement Barriers and Systems), 3.3.3 (Protection by Equipment and Instrumentation), 3.3.5 (Radiological Protection), 3.5 (Decom-missioning Considerations), 4.5 (Shipping Cask Repair and Maintenance), 5 1.1 (General Description), 5.1.2 (Flow Sheets), 5.1.3.5 (Maintenance Techniques),
5.3 (Other Operating Systems), 5.4 (Operation Support Systems), 6.0 (Waste Confinement and Management), 7.1.2 (Design Considerations), 7.2 (Radiation l
10-2
Sources),'and 7.3 (Radiation Protection Design Features).
Supplemental informa-tion was also provided in Transnuclear letters E-10697 dated January 26, 1989, E-10703 dated January 31, 1989, E-10724 dated February 13, 1989, and a letter dated February 16, 1989.
TR Sections 1.1.2'(Principal Design Features of Installation), 1 2 (General Description of Installation), and 1.3 (General Systems Description) 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.
Sections 3.1.1 (Materials to be Stored), 3.3.2 (Protection by Multiple Confinement Barriers and Systems), 3.3.3 (Protection by Equipment and Instrumentation), 3.3.5 (Radiological Protection), and 3.5 (Decommissioning 1
Considerations) provide information basic to the principal design of the cask.
Included are descriptions of the spent fuel characteristics and major source terms; the confinement barriers and seals; the lid tightness monitoring system; the direct dose from a single cask associated with cask loading, transport, and emplacement; and neutron activation of the cask materials over the storage period.
. Section 4.5 (Shipping Cask Repair and Maintenance) identifies typical periodic maintenance procedures and describes the repair procedures associated with a leaking cask seal.
Sections 5.1.1 (General Description), 5.1.2 (Flow Sheets), and 5.1.3.5 (Maintenance Techniques) describe the handling operations in the cask loading, decontamination, and storage areas.
Included in this description are the estimated number of personnel and their associated exposure periods and loca-tions.
Section 5.3 (Other Operating Systems) identifies the component /
equipment spares and addresses the installation of the containment cover in the event of double seal failure.
Section 5.4 (Operation Support Systems) provides a general description of the lid tightness monitoring system.
Section 6.0 (Waste Confinement and Management) reiterates that there will be no out-leakage of gas from the cask cavity during normal operations.
10-3
Sections 7.1.2'(Design Considerations), 7.2 (Radiation Sources), and 7.3 (Radiation Protection Design Features) provide information' basic to radiation protection and. shielding.. Included are discussions of. design considerations, the source terms for the spent fuel and airborne radioactive material, and the shielding design features and analyses.
10.3.3' Estimated Onsite Dose Assessment Information important to the estimate of the onsite collective dose is found in Sections 3.1.1 (Materials to be Stored), 3.3.5 (Radiological Protection),-
5,.1.2 (Flow Sheets), 7.2 (Radiation Sources), 7.3 (Radiation Protection Design Features), and 7.4 (Estimated Onsite Collective Dose Assessment).
Supplemental information was also provided by Transnuclear in letters E-10697 dated January 26,1989, E-10703 dated January 31, 1989, and a letter dated February 16, 1989.
With the exception of Section 7.4 (Estimated On-Site Collective Dose Assessment),
the general centents of the above pertinent sections are described in the previous section of this SER.
The description provided in section 7.4 includes a reference to the section containing an estimate of the direct radiation collective dose associated with various loading, transporting, and emplacement operations of a single cask; an estimate of the direct radiation collective dose associated with maintenance and repair operations of a single cask; the dose rate as a function of distance from a single cask and 100 cask array; and the direct collective dose at a site boundary of 250 meters associated with a l
generic 100 cask array.
The dose from a single cask to any individual from direct radiation during transporting, storage, maintenance, and repair operations was computed by the staff.
Spacific operations considered are those grouped under transfer to storage area, storage area, periodic maintenance, and major maintenance.
One transport to storage, monthly visual surveillance, semi-annual function testing of instrumentation, one surface defects repair operation, one replacement of seals, and one containment cover installation are assumed during the year.
10-4 I
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C Dose ~ from direct radiation is based upon the times, distances, and dose rates of-the TR.
Times and distances required to perform the *tarious operations are t'aken from Table 5.1-2 (Anticipated Time and Personnel Requirements for Cask
'. Handling' Operations).
Dose rates are taken from either Table 7.3-4 (TN-24 Dose Rates at Short Distances) or Table 7.4-1 (Maintenance and Repair Operaticis
' Annual Exposure), with the most conservative being used in all instances.
For additional conservatism, beginning of life is assumed for all operations.
Dose form gaseous activity release is not evaluated.
Gaseous activity release is not considered credible under. normal storage conditions for the TN cask-
-configuration.
At 100 meters from a single cask, the dose to any individual from direct radiation during normal operations (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> /wk and 50 weeks /yr period of exposure) was also evaluated.
2 Dose from direct radiation is estimated from the single cask curve of Figure 7.3-6 (Dose Rates at Long Distances).
Dose form gaseous activity release is not evaluated.
Gaseous activity release is not considered credible under normal storage conditions for the TN cask configuration.
10.4 Findings and Conclusions 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 during normal operations is estimated to be less than 531 mrem /yr to the whole body from the transport, storage, maintenance, and repair operations.
At 100 meters from a single cask, the dose to any indi-vidual from direct radiation for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> /wk and 50 weeks /yr is less than 24.1 mrem /yr.
I i
The staff concludes that the radiation protection is acceptable.
I 10-5
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L11 ACCIDENT ANALYSIS
.'11.1 Area of Review-L10 CFR Section 72.24(m) 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 environment'or direct radiation from the ISFSI.
- 10'CFR Section 72.104(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) l planned discharges of radioactive materials, radon and its daughters excepted, to the' general environment, (2) direct radiation from ISFSI operations and (3)
Lany other radiation from uranium fuel cycle operations within the region.
.10 CFR Section 72.106(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.
Two
> distances are considered:
(1).100 meters, the minimum allowable controlled area boundary distance required in Section 72.68(b); and (2) 250 meters, the minimum distance assumed in the shielding evaluation (see Sections 5.2 and 5.4
- of this SER).
Our review focuses on the dose from direct radiation and activity release associated with. postulated off-normal and accident events.
In the context of this review, off-normal events.are anticipated occurrences.
As such, the minimum distance chosen is 250 meters to conform with the minimum distance
- assumed in the shielding evaluation (see Sections 5.2 and 5.4 of this SER).
11-1 l
m.
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 indi-vidual from direct. radiation and activity release is not.in excess of the applicable values given in section 11.1 above.
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 Two events are identified for off-normal operations:
(1) seal failure on a cask, and (2) malfunction of the cask pressure monitoring system.
Specific causes of the events are not addressed in Section 8.1.1.1 (Postulated Cause of the Event).
The means of detecting the events are discussed in Section 8.1.1.2 (Detection of Event).
Analysis of the effects and consequences, and the proposed corrective actions in the case of these two events appear in Sec-tions 8.1.1.3 (Analysis of Effects and Consequences) and 8.1.1.4 (Corrective Action),respectively.
11.~3.1.2 Radiological Impact from Off-Normal Operations Section 8.1.2 (Radiological Impact from Off-Normal Operations) presents no estimate of the collective doses at 250 meters due to gaseous activity release following the off-normal events of seal failure on a cask and malfunction of the cask pressure monitoring system.
Section 7.4 (Estimated Of f-Site Collective Dose Assessment) of the TR describes the direct radiation collective dose versus distance for a single cask and 100 cask array.
Supplemental information was also provided in Transnuclear letter E-10697 dated January 26, 1989.
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The radiological impact from off-normal operations only involves gaseous activ-ity release and is computed for an individual outside the 250 meter controlled l-area.
In computing the dose due to H-3 and Kr-85 gaseous activity release, the staff has assumed the following:
(1) 10 percent cladding tube failure; (2) the release fractions of Regulation Guide 1.25; (3) a leakage rate of 1.1 x 10 5 atm-cc/sec from Section 3.3.2.2 (Analysis of Cask Pressures and Leakage Rates);
(4) a shielding factor of 1.0 for Kr-85; (5) the population weighted inhalation rate of Regulation Guide 1.109 for the offsite individual; (6) an exposure period of 1 year; (7) the inhalation dose and whole body dose factors of Regulation Guide 1.109; and (8) F-stablilty atmospheric diffusion with a l
windspeed of 1 meter /sec with plume meander.
11.3.0 Accidents 11.3.2.1 Accidents Analyzed TR Section 8.1.2 (Radiological Impact from Off-Normal Operations) addresses the worst case situation of the release of the available gaseous inventories of H and Kr-85 from the cask cavity.
Other volatile isotopes with limited availabil-ity such as Cs-134, and Cs-137 are not evaluated.
Supplemental information supplied in Transnuclear letter E-10697 dated January 26, 1989, addresses the worst case situation of the complete loss of the neutron shield and outer shell from the radial surface.
No other accidents described in the TR have consequences worse than the above.
In the staff review of the radiological impact of the instantaneous release of the gaseous contents, we followed the same procedures and made the same assumptions as those discussed in our review of the maximuir, dose from gaseous activity release (see Section 7.5.3.2 of this SER).
Dose from direct radiation is derived from the loss of neutron shieli eccident and is computed from the TN maximum side surface gamma and neutron dose rates presented in Transnuclear letter E-10697 dated January 26, 1989, and the TN predicted dose versus distance curve.
A worst case condition of no corrective actions is assumed for the period of one year.
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._-__-__---_-__---_________-_-_-_-------------------A
-i 11.4 Findings and Conclusions The dose consequences due to gaseous activity release from a single cask following off-normal and accident events are less than 2.4 x 10 4 mrem and 0.516 rem, respectively.
The dose consequence due to direct radiation from a single cask following the accidental loss of neutron shield and axial lead slump is 1.9 rem /yr at 100 meters and less than 0.183 rem /yr at 250 meters.
For the accident events, the total dose from gaseous activity release and direct radiation is less than 1.71 rem at 100 meters and less than 0.270 rem at 250 meters.
Clearly, accident consequences are less than the 5 rem limit established i
in 10 CFR Section 72.106(b) at a minimum site boundary of 100 meters.
Compliance with 10 CFR Section 72.106(b) 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 site.
11-4 l
.e 12 DECOMMISSIONING 12.1 -Area of Review 10 CFR Section 72.30 provides requirements for a site-specific decommissioning plan, including financing.
Among the items to be addressed is the disposal of residual radioactive materials after all spent fuel has been removed.
10 CFR Section 72.130 provides requirements for decommissioning and states, in part, that the ISFSI shall be designed for decommissioning.
Among the items to be addressed under this part are the provisions to facilitate decontamination of equipment, the provisions to minimize the quantity of radioactive wastes and contaminated equipment, and the provisions to facilitate the removal of radio-active wastes and the materials at the time of permanent decommissioning.
49 CFR Sections 173.421, 173.423, and 173.435 provide information on the radionuclides activities that may be transported as limited quantity materials.
10 CFR Sections 30.14 and 30.70 address radionuclides concentrations that are exempt from licensing requirements.
I 10 CFR Sections 30.18 and 30.71 address radionuclides quantities that are exempt from licensing requirements.
10 CFR Sections 61.55 and 61.56 address radionuclides concentrations for Class A wastes and the characte:ristics of such waste.
A decommissioning plan for a site-specific ISFSI as required by 10 CFR Part 72.30 is not applicable for a topical report.
Therefore our review focuses on the non-site-specific elements of decommissioning and in particular the decommissioning of a single cask.
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.12.2 Acceptance Criteria Cask-decommissioning is deemed acceptable if'it can be shown that regulations cited in Section 12.1 above have been followed as appropriate, and where limits can be applied, these have not been exceeded.
12.3 Review Procedure The review is divided into two main parts:
(1) unloading of the casks; and.
.(2) decommissioning of the cask components.
12.3.1 Unloading of the Cask A brief description of cask unloading is presented in a supplemental letter E-10724 dated February 13, 1989.
Only wet unloading (reactor pool) is mentioned.
After removal of the protective cover and disconnection of the over pressure system, the facility.off gas system intake will be connected to the cask-drain
~
valve and the cask cavity gas sampled and/or vented.
The cask cavity is then-filled with water, the lid bolts removed, and the cask lowered into the pool
.for unloading.
Subsequent to the unloading, the ISFSI site may elect to remove internal cask cavity surface contamination.
For a site-specific license application, the applicant would be expected to develop and commit to detailed procedures for use in unloading the cask.
12.3.2 Decommissioning of the Cask Components Activation of the fuel basket, cask body and lid, neutron shield, and neutron shield shell and protective cover are discussed in Section 3.5 (Decommissioning Considerations) of the TR.
Neutron fluxes obtained from the XSDRNPM shielding calculations were used in conjunction with ORIGEN-2 to calculate the activities of C-14, Cr-51, Mn-54, Fe-55, Fe-59, Co-58, Co-60, Ni-59, Ni-63, Nb-94, and Tc-99 in the cask body 12-2
[
.4 and lid; C-14, Cr-51, Mn-54, Fe-55, and Fe-59. in the fuel basket;. H-3, C-14, and Zn-65 in the neutron shield; and C-14 and Fe-55 in the neutron shield shell i
l and protective cover at 30 days subsequent to unloading..Scme of these nuclides emit no gamma rays (H-3, C-14, and Ni-63).
Others (Fe-55 and Ni-59) emit gamma rays with energies of less than 8 kev.
In evaluating the activation products, the staff has assumed a minimum decay period of 60 days for the fuel basket and 30 da)s for the cask body and lid, l
neutron shield, and neutron shield shell and protective cover.
At 60 days, 2.43x10 12 Ci of C-14, 1.37x10 3 Ci of Cr-51, 3.11x10 4 Ci of Mn-54, 3.57x10 3 Ci of Fe-55, 4.22x10 5 Ci of Fe-59, 4.50x10 4 Ci of Co-58, 3.14x10 3 Ci of Co-60, 3.25x10 6 Ci of Ni-59, 4.26x10 4 Ci of Ni-63, 3.06x10 12 Ci of Nb-94, and 8.55x10 12 Ci of Tc-99 remain in the fuel basket.
At 30 days, 8.83x10 12 Ci of C-14, 8.35x10 5 Ci of Cr-51, 3.66x10 3 Ci of Mn-54, 3.97x10 2 Ci of Fe-55, and 7.42x10 4 Ci of Fe-59 remain in the cask body and lid; 3.49x10 20 Ci of H-3, 8.45x10 20 Ci of C-14, and 1.98x10 5 Ci of Zn-65 remain in the neutron shield; and 2.48x10 14 Ci of C-14 and 1.12x10 5 Ci of Fe-55 remain in the neutron shield shell and protective cover.
Materials quantities used by Transnuclear in the activation calculations are representative of those in the cask itself.
Weights for the fuel basket, cask body and lid, neutron shield, and neutron shield shell and protective cover are 7,643 kg (16,815 lb), 61,184 kg (134,605 lb), 5,747 kg (12,643 lb), and 5,333 kg (11,733 lb), respectively; volumes for the fuel basket, cask body and lid, 3
neutron shield, and neutron shield shell and protective cover are 0.94 m,
3 3
3 7.78 m, 3.71 m, and 0.69 m, respectively.
With respect to decommissioning of the cask components, their activation product concentrations are such that the cask components at the assumed decay periods subsequent to unloading (60 days for the fuel basket and 30 days for the cask body and lid, neutron shield, and neutron shield shell and protective cover) contain license exempt concentrations of H-3, C-14, Cr-51, Mn-54, Fe-55, Fe-59, Co-58, Co-60, and 2n-65.
Furthermore, the activities or concentrations are such that the cask components may be classed as limited quantity materials for offsite transportation and may be disposed of as Class A waste.
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'12.4 Findings and Conclusions The cask. design-is consistent with the requirements of 10 CFR Section 72.130 that an ISFSI be designed for decommissioning.
The actions involved in cask unloading are also consistent with the requirements of 10 CFR Section 72.30 as feasible elements of a site-specific decommissioning plan.
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I 13 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 Section 72.44.
In addition to the conditions pursuant to 10 CFR Section 72.44(b), each application for a license under 10 CFR Part 72 shall include. proposed technical specifications pursuant to 10 CFR Part 72.26 and consistent with 10 CFR Section 72.44(c).
The final approved technical specifications will be made part of the license.
The technical specifications of a license define certain features, characteristics 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 Section 72.44(c), the operating controls and limits established in Chapter 10 of the TR 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 l
Operating controls and limits which may serve as a basis for licensing conditions are derived from the analyses and evaluation included in the TR.
13.4 Findings and Conclusions The staff reviewed the specific operating limits summarized in Chapter 10 of the TR.
The limits established for these parameters reflect the design criteria upon which the safety analyses were based and are acceptable.
With regard to the fuel characteristic limits cascribed in Section 10.1.2.4 of the TR, the 13-1
l maximum initial enrichment is limited to 2.3 percent.
Fuel assembly burnup shall not exceed 35,000 MWD /MTV.
Storage of intact fuel bundles for the Westinghouse 17 x 17 0FA is used as the design basis in the TR.
A maximum handling height of 18 inches without impact limiters should be included as an operating limit.
Horizontal storage is not permitted.
The TN-24 cask must be stored with the upper side impact limiters attached.
l The licensee must assure that the redwood impact limiter does not deteriorate l
over the design life of the ISFSI.
l The cask basket must not be inadvertently moved while inserting or removing a fuel bundle with other fuel in place.
This is a criticality concern.
The surface absorptivity coefficient for solar radiation of the cask must be less than 0.30.
The total decay heat for the cask must be limited to 24 kW and the decay heat for each fuel bundle must be limited to 1.0 kW.
The maximum fuel rod temperature during vacuum drying must not exceed 380 C (716 F).
The maximum fuel rod temperature during storage should not exceed the design limit of 339 C (642 F).
A license applicant for an ISFSI must review parameters covered in the TR and develop appropriate proposed technical specifications and license conditions for the site-specific conditions.
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14 QUALITY ASSURANCE In Chapter 11, " Quality Assurance Program," of Revision 1 of the TR, TN has q
committed to implement its " Quality Assurance Program for Design, Fabrication, Inspection and Testing of Storage Systems for Spent Fuel and Associated Radioactive Materials" (TN document E-9213) to all TN-24 activities that are important to safety.
Chapter 11 refers to TR Table 3.4-1, " Classification of Structures, Components and Systems," which shows the items important to safety and the items not important to safety.
The staff has reviewed TN's commitments for quality assurance given and referenced in Chapter 11 of the TR.
The staff finds that the TN commitments q
for quality assurance meet the requirements of subpart G of 10 CFR Part 72 for the TN-24 cask and are, therefore, acceptable.
The TR can be referenced without further quality assurance review in a license application to receive and store spent fuel under 10 CFR Part 72, provided that the applicant applies its NRC approved quality assurance program that meets the requirements of Appendix B to 10 CFR Part 50 to the design, construction, and use of the spent fuel storage installation.
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15 REFERENCES 1.
Transnuclear, Inc., "TN-24 Dry Storage Cask Topical Report," E-7107, L
Transnuclear, Inc., Two Skyline Drive, Hawthorne, New York, July 25, 1988.
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 3.61 " Standard Format and Content For A Topical Safety Analysis Report For A Dry Spent Fuel Storage Cask," February 1989.
4 4.
T. Wilcox and E. Lent, "C0G: A Particle Transport Code Designed to Solve the Boltzmann Equation for Deep-Penetration (Shielding) Problems - Volume I, Users Manual," UCRL - Draft Copy, Lawrence Livermore National Laboratory, Livermore, California (October 1986).
5.
SCALIAS-77, " Selected FORTRAN 77 Modules for SCALE-3.1," CCC-475, Radiation Shielding Information Center (RSIC), September 1986.
6.
S. R., Bierman et al., " Critical Separation Between Subcritical Clusters Of 4.29 Wt % 23s0 Enriched U0 Rods In Water With Fixed Neutron Poisons,"
2 NUREG/CR-0073, May 1978.
7.
S. R., Bierman et al., " Critical Separation Between Subcritical Clusters Of 2.35 Wt % 23su Enriched U0 Rods In Water With Fixed Neutron Poisons,"
2 PNL-2438, October 1977.
8.
I.S., Levy et al., " Recommended Temperature Limits for Dry Storage of Spent Light Water Reactor Zircaloy-Clad Fuel Rods in Inert Gas," PNL 6189, May 1987.
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l 9.
" Characteristics of Spent Fuel, High Level Waste, and Other Radioactive Wastes Which May Require Long-Term Isolation, Appendix 2A. Physical Description of LWR Fuel Assemblies," DOE /RW-0184, Vol. 3 of 6, U.S. Department of Energy, OCRWM, December 1987.
10.
M. W. Schwartz and M. C. Witte, " Spent Fuel Cladding Integrity During Dry Storage," UCID 21181, Lawrence Livermore National Laboratory, l
Livermore, California, September 1987.
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