ML20141C298

Topical Rept Evaluation Re Rev 1 to NUH-001, Topical Rept for Nutech Horizontal Modular Storage Sys for Irradiated Nuclear Fuel. Rept Acceptable W/Listed Limitations

ML20141C298
Person / Time
Site:	07200003
Issue date:	03/28/1986
From:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
Shared Package
ML20141C270	List: ML20141C267 ML20141C298
References
REF-PROJ-M-39 NUDOCS 8604070223
Download: ML20141C298 (128)

Text

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SAFETY EVALUATION REPORT FOR NUTECH HORIZONTAL MODULAR SYSTEM FOR IRRADIATED FUEL TOPICAL REPORT U.S. Nuclear Regulatory Comission Washington, D.C. 20555 Office of Nuclear Materials Safety and Safeguards March 28, 1986 8604070223 860328 M$9 PDR

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l Table of Contents Section Eagg 1.0 GENERAL DESCRIPTION. . . . . . . . . . . . . . . . . . . . I 1.1 Introduction. . . . . . . . . . . . . . . . . . . . . 1 1.2 General Description of NUHOMS System. . . . . . . . . I 1.2.1 HSM. . . . . . . . . . . . . . . . . . . . . . 2 1.2.2 DSC. . . . . . . . . . . . . . . . . . . . . . 5 1.2.3 Handling and Transfer Equipment. . . . . . . . 5 1.2.4 Stored Materials . . . . . . . . . . . . . . . 6 1.3 Identification of Agents and Subcontractors . . . . . 7 1.4 Generic HSM Arrays. . . . . . . . . . . . . . . . . . 7 2.0 PRINCIPAL DESIGN CRITERIA. . . . . . . . . . . . . . . . . 8 2.1 Introduction. . . . . . . . . . . . . . . . . . . . . 8 2.2 Fuel to be Stored . . . . . . . . . . . . . . . . . . 8 2.3 Quality Standards . . . . . . . . . . . . . . . . . . 9 2.4 Protection Against Environmental Conditions and Natural Phenomena . . . . . . . . . . . . . . . . . 9 2.5 Protection Against Fire and Explosion . . . . . . . . 18 2.6 Confinement Barriers and Systems. . . . . . . . . . . 18 2.7 Instrumentation and Control Systems . . . . . . . . . 19 2.8 Criteria for Nuclear Criticality Safety . . . . . . . 19 2.9 Criteria for Radiological Protection. . . . . . . . . 20 2.10 Criteria for Spent Fuel and Radioactive Waste Storage and Handling. . . . . . . . . . . . . . . . . . . . 21 2.11 Criteria for Decommissioning. . . . . . . . . . . . . 22 3.0 STRUCTURAL EVALUATION. . . . . . . . . . . . . . . . . . . 23 3.1 Summary and Conclusions . . . . . . . . . . . . . . . 23 3.2 Description of Review . . . . . . . . . . . . . . . . 25 i

Table of Contents (Continued)

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Section Elgg )

3.2.1 Applicable Parts of 10 CFR 72. . . . . . . . . 26 3.2.2 Review Procedure . . . . . . . . . . . . . . . 26 3.2.2.1 Design Descriptions . . . . ... 26 3.2.2.2 Acceptance Criteria . . . . . ... 28 I 3.2.2.3 Review Method . . . . . . . . . . . . 29 3.2.2.4 Key Assumptions . . . . . . . . . . . 29 3.3 Discussion of Results . . . . . . . . . . . . . . . . 30 3.3.1 Loads. . . . . . . . . . . . . . . . . . . . . 30 3.3.1.1 Normal Operating Conditions . . . . . 30 3.3.1.2 Off-Normal Conditions . . . . . . . . 30 3.3.1.3 Accident Conditions . . . . . . . . . 31 3.3.2 Materials. . . . . . . . . . . . . . . . . . . 31 3.3.3 Stress Intensity Limits. . . . . . . . . . . . 35 3.3.4 Structural Analysis. . . . . . . . . . . . . . 35 3.3.4.1 HSM . . . . . . . . . . . . . . . . . 35 3.3.4.1.1 Normal Operating Conditions. . . 35 3.3.4.1.2 Off-Normal Conditions. . . . . . 36 3.3.4.1.3 Accident Conditions. . . . . . . 38 3.3.4.2 DSC and Internals . . . . . . . . . . 42 3.3.4.2.1 DSC Normal Operating Conditions. 42 3.3.4.2.2 DSC Off-Normal Events. . . . . . 45 3.3.4.2.3 DSC Accident Conditions. . . . . 48 3.3.4.3 DSC Support . . . . . . . . . . . . . 55 3.3.4.3.1 DSC Support Normal Operating Conditions . . . . . . . . . . 55 3.3.4.3.2 DSC Support Off-Normal Events. . 56 l 3.3.4.3.3 DSC Support Accident Conditions. 57 3.3.4.3.4 DSC Support End Connection Design . . . . . . . . . . . . 57 3.3.4.4 Transfer Cask and Vehicle . . . . . 57 ii

Table of Contents (Continued) i Section P_agg i

3.3.4.5 HSM Loading and Unloading Equipment . 58 3.3.4.5.1 Normal Operating Conditions. . . 58 3.3.4.5.2 Off-Normal Conditions. . . . . . 58 3.3.4.5.3 Accident Conditions. . . . . . . 58 3.3.4.6 Fuel Assemblies and Rods. . . . . . . 58 3.3.4.6.1 Normal Operating Conditions. . . 58 3.3.4.6.2 Accident Conditions. . . . . . . 59 4.0 THERMAL EVALUATION . . . . . . . . . . . . . . . . . . . . 60 4.1 Summary and Conclusions . . . . . . . . . . . . . . . 60 4.2 Description of Review . . . . . . . . . . . . . . . . 60 4.2.1 Applicable Parts of 10 CFR 72. . . . . . . . . 60 4.2.2 Review Procedure . . . . . . . . . . . . . . . 61 4.2.2.1 Design Description. . . . . . . . . . 61 4.2.2.2 Acceptance Criteria . . . . . . . . . 62 4.2.2.3 Review Method . . . . . . . . . . . . 62 4.2.2.4 Key Design Information and Assumptions . . . . . . . . . . . . 63 4.3 Discussion of Results . . . . . . . . . . . . . . . . 63 I

4.3.1 Analytical Methods Used by NUTECH. . . . . . . 63 4.3.2 HSH and Internals. . . . . . . . . . . . . . . 65 4.3.2.1 Normal Operating Conditions . . . . . 65 4.3.2.2 Off-Normal Conditions . . . . . . . . 66 4.3.2.3 Accident Conditions . . . . . . . . . 66 4.3.3 DSC and Internals. . . . . . . . . . . . . . . 67 4.3.3.1 Normal Operating Conditions . . . . . 67 4.3.3.2 Off-Normal Conditions . . . . . . . . 67 4.3.3.3 Accident Conditions . . . . . . . . . 67 4.3.4 Transfer Cask and Fuel . . . . . . . . . . . . 68 iii

Table of Contents (Continued)

Section Eagg 5.0 CONFINEMENT BARRIERS AND SYSTEMS . . . . . . . . . . . . . 69 5.1 Summary and Conclusion. . . . . . . . . . . . . . . . 69 5.2 Description of Review . . . . . . . . . . . . . . . . 69 5.2.1 Applicable Parts of 10 CFR 72. . . . . . . . . 69 5.2.2 Review Procedure . . . . . . . . . . . . . . . 70 5.2.2.1 Design Description. . . . . . . . . . 70 5.2.2.2 Acceptance Criteria . . . . . . . . . 70 5.2.2.3 Review Method . . . . . . . . . . . . 70

'5.2.2.4 Key Assumptions . . . . . . . . . . . 71 5.3 Discussion of Results . . . . . . . . . . . . . . . . 71 5.3.1 DSC Integrity. . . . . . . . . . . . . . . . . 71 5.3.2 Long Term Fuel Rod Behavior. . . . . . . . . . 72 6.0 SHIELDING EVALUATION . . . . . . . . . . . . . . . . . . . 73 6.1 Summary and Conclusions . . . . . . . . . . . . . . . 73 6.2 Description of Review . . . . . . . . . . . . . . . . 73 6.2.1 Applicable Parts of 10 CFR 72. . . . . . . . . 73 6.2.2 Review Procedure . . . . . . . . . . . . . . . 73 6.2.2.1 Design Description. . . . . . . . . . 73 6.2.2.2 Acceptance Criteria . . . . . . . . . 74 6.2.2.3 Review Method . . . . . . . . . . . . 74 6.2.2.4 Key Assumptions and Computer Codes. . . . . . . . . . 74 iv

Table of Contents (Continued)

Section Eilat 6.3 Discussion of Results . . . . . . . . . . . . . . . . 75 1

6.3.1 ~ Source Specification . . . . . . . . . . . . . 75 6.3.2 DSC - Cask Gap . . . . . . . . . . . . . . . . 75 6.3.3 Transfer Cask. . . . . . . . . . . . . . . . . 76 6.3.4 HSM. . . . . . . . . . . . . . . . . . . . . . 76

7.0 CRITICALITY EVALUATION

. . . . . . . . . . . . . . . . . . 78 7.1 Summary and Conclusions . . . . . . . . . . . . . . . 78 7.2 Description of Review . . . . . . . . . . . . . . . . 78 7.2.1 Applicable Parts of 10 CFR 72. . . . . . . . . 78 7.2.2 Review Procedure . . . . . . . . . . . . . . . 78 7.2.2.1 Design Description. . . . . . . . . . 78 l 7.2.2.2 Acceptance Criteria . . . . . . . . . 79 7.2.2.3 Review Method . . . . . . . . . . . . 79 7.2.2.4 Key Factors . . . . . . . . . . . . . 79 7.3 Discussion of Results . . . . . . . . . . . . . . . . 79 l

l 7.3.1 Analytical Methods . . . . . . . . . . . . . . 79 7.3.2 DSC Containing Fuel. . . . . . . . . . . . . . 80 8.0 OPERATING PROCEDURES . . . . . . . . . . . . . . . . . . . 81 8.1 Summary and Conclusions . . . . . . . . . . . . . . . 81 8.2 Description of Review . . . . . . . . . . . . . . . . 81 8.2.1 Applicable Regulations . . . . . . . . . . . . 81 v

Table of Contents (Continued)

Section hgg 8.2.2 Review Procedure . . . . . . . . . . . . . . . 82 8.2.2.1 Design Description. . . . . . . . . . 82 8.2.2.2 Acceptance Criteria . . . . . . . . . 82 8.2.2.3 Review Method . . . . . . . . . . . . 82 8.2.2.4 Key Assumptions . . . . . . . . . . . 83 8.3 Discussion of Results . . . . . . . . . . . . . . . . 83 l

9.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM . . . . . . . . . 85 9.1 Summary and Conclusions . . . . . . . . . . . . . . . 85 9.2 Description of Review . . . . . . . . . . . . . . . . 85 l

9.3 Discussion of Results . . . . . . . . . . . . . . . . 86 10.0 RADIOLOGICAL PROTECTION. . . . . . . . . . . . . . . . . . 89 10.1 On-Site. . . . . . . . . . . . . . . . . . . . . . . 89 10.1.1 Summary and Conclusions. . . . . . . . . . . 89 10.1.2 Description of Review. . . . . . . . . . . . 89 10.1.2.1 Applicable Parts of 10 CFR 72. . . 89 10.1.2.2 Review Procedure . . . . . . . . . 90 10.1.2.2.1 Design Description. . . . . . 90 10.1.2.2.2 Acceptance Criteria . . . . . 90 10.1.2.2.3 Review Method . . . . . . . . 90 10.1.2.2.4 Key Assumptions . . . . . . . 91 10.1.3 Discussion of Results . . . . . . . . . . . . 91 10.1.3.1 ALARA Considerations. . . . . . . . 91 10.1.3.2 Radiation Protection Design Featuret of DSC and HSM . . . . . 91 10.1.3.3 On-Site Dose Assessment . . . . . . 92 vi

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Table of Contents (Continued)

Section PJgg 10.2 Off-Site Radiological Protection . . . . . . . . . . 93 10.2.1 Summary and Conclusions. . . . . . . . . . . 93 10.2.2 Description of Review. . . . . . . . . . . . 93 10.2.2.1 Applicable Regulations . . . . . . 93 10.2.2.2 Review Procedure . . . . . . . . . 93 10.2.2.3 Acceptance Criteria. . . . . . . . 94 10.2.2.4 Review Method. . . . . . . . . . . 94 10.2.2.5 Key Assumptions. . . . . . . . . . 95 10.2.3 Discussion of Results. . . . . . . . . . . . 96 10.2.3.1 Normal Operating Conditions. . . . 96 10.2.3.2 Accident Conditions. . . . . . . . 97 11.0 DECOMMISSIONING. . . . . . . . . . . . . . . . . . . . . . 98 11.1 Summary and Conclusions. . . . . . . . . . . . . . . 98 11.2 Description of Review. . . . . . . . . . . . . . . . 98 11.2.1 Applicable Parts of 10 CFR 72. . . . . . . . 98 11.2.2 Review Procedure . . . . . . . . . . . . . . 99 11.2.2.1 Design Description . . . . . . . . 99 11.2.2.2 Acceptance Criteria. . . . . . . . 99 11.2.2.3 Review Method. . . . . . . . . . . 99 11.2.2.4 Key Assumptions. . . . . . . . . . 100 11.3 Discussion of Results. . . . . . . . . . . . . . . . 100 12.0 OPERATING CONTROLS AND LIMITS. . . . . . . . . . . . . . . 102 12.1 Summary and Conclusions. . . . . . . . . . . . . . . 102 vii

Table of Contents (Continued)

Section Engg 12.2 Description of Review. . . . . . . . . . . . . . . . 106 12.2.1 Applicable Parts of 10 CFR 72. . . . . . . . 106 12.2.2 Review Procedure . . . . . . . . . . . . . 106 12.3 Discussion of Results. . . . . . . . . . . . . . . . 106 12.3.1 Fuel Specification . . . . . . . . . . . . . 106 12.3.2 Limiting Conditions for Operation. . . . . . 107 12.3.3 Surveillance Requirements. . . . . . . . . . 109 12.3.4 Design Specifications. . . . . . . . . . . . 109 13.0 QUALITY ASSURANCE. . . . . . . . . . . . . . . . . . . . . 111 1

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14.0 REFERENCES

. . . . . . . . . . . . . . . . . . . . . . . . 112 APPENDIX A - Analysis of Diffusion Controlled Cavity Growth (DCCG) Damage to the Fuel Cladding in Dry Storage viii

1.0 GENERAL DESCRIPTION 1.1 Introduction This Safety Evaluation Report (SER) documents the NRC staff's review and evaluation of the Topical Report (TR) for the NUTECH Horizontal Modular Storage (NUHOMS) System for Irradiated Nuclear Fuel, NUH-001, Rev. 1,

( November, 1985 (Ref. 1). The TR was prepared by NUTECH, Inc. using the I

! format of NRC Regulatory Guide 3.48 (Ref. 2).

f l The review has been based on the proposed system's meeting the applicable requirements of 10 CFR 72, Subpart E, " Siting Evaluation Factors", Subpart F, " General Design Criteria", and Subpart G, " Quality Assurance" (Ref. 3). The review also included consideration of the appropriate parts of 10 CFR 20 for radiation protection during onsite handling, movement, and storage of spent fuel.

This review does not address either the requirements for physical protection under Subpart H, " Physical Protection" of 10 CFR 72 or those under applicable parts of 10 CFR 73, " Physical Protection of Plants and Materials". Further requirements for off-site transport of spent fuel under 10 CFR Part 71 are not addressed here since this review is limited to only on-site transfer of spent fuel at a reactor site.

1.2 General Description of NUHOMS System The NUHOMS system is an independent spent fuel storage installation (ISFSI) which provides for the horizontal, dry storage of irradiated nuclear fuel assemblies. The fuel assemblies are contained in a dry, shielded canister (DSC) made of steel which is placed inside a reinforced concrete horizontal storage module (HSM) for long term storage.

In addition to the DSC and HSM, the NUHOMS system also requires handling and transfer equipment to load the DSC with fuel, to seal the DSC, to move the loaded DSC inside a heavily shielded transfer cask from the reactor building to the HSH (elsewhere on the site), and to insert the DSC 1

into the HSM. Figures 1.1 and 1.2 show schematically the major components and operations of the NUHOMS system.

Various combinations of sizes for the DSC and HSM are possible.

However, the TR presents for review and approval a design in which the DSC holds seven irradiated pressurized water reactor (PWR) fuel assemblies and in which the HSMs are arranged in back-to-back rows of 8 HSMs in a 2x4 array.

The designs of the HSM, DSC, handling and transfer equipment, and nuclear fuel assemblies to be stored are described in more detail in the following subsections.

1.2.1 HSM Each HSM, or module, is constructed of reinforced concrete, structural steel, and stainless steel. An HSM is 5.92m long, 3.65m high and 1.7m wide.

The concrete walls and roof are of sufficient thickness to attenuate radia-tion from the stored fuel so that average surface dose rates will be less than 20 r am/hr.

The TR reference design is based on an installation of eight modules arranged in a 2x4 array on a load-bearing foundation. Each HSM can hold one DSC. The modules are arranged back-to-back so that loading of each module is accomplished through an opening in the front. The center of the opening is approximately 1.95m (6.4 feet) above the surface of the foundation. When the HSM contains one DSC, a steel door is lowered down over the front opening.

There are two steel rails inside the HSM running front-to-back which support the DSC while it is in storage. Each HSM has two air inlets on the front and two air outlets on the roof which permit natural convective air cooling of the DSC while it is in storage. Both the inlets and outlets are shielded to reduce radiation doses at the exterior of the HSM.

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1.2.2 DSC The DSC consists of a 304 stainless steel cylindrical body, two shielded end plugs, and an internal basket to hold and support seven irradiated pressurized water reactor (PWR) fuel assemblies.

The DSC body is a 0.5 inch thick stainless steel pipe. It h.as an outside diameter of 0.94m and its length will depend upon the length of the stored fuel assemblies. When welded shut, the DSC will be evacuated and backfilled with helium at 0.981 bar.

The interns 1 basket is composed of seven square cells. The structural component of the cells is stainless steel clad boral. This boral provides criticality control in the canister during wet loading operations.

Structural support of the seven cells inside the DSC is provided by circular stainless-steel spacer disks. There is one such disk under each spacer grid in the fuel assembly. Longitudinal support is provided by four support rods which run the length of the canister from one end shield to the other.

Each end of the DSC is equipped with shielded end-plugs so that when the canister is inside the transfer cask or the HSM, the radiation dose at the ends is limited. The end shields are constructed of roughly 5 inches of lead inside a stainless steel casing.

The DSC has redundant seal welds at the top and bottom. The bottom plates are shop-welded during fabrication and the top plates are welded in the field after fuel is loaded in the 05C. All connections (drain and air purge lines) are also redundantly sealed.

1.2.3 Handling and Transfer Equipment In order to support the operation of the NUHOMS system, several addi-tional components are needed for the handling of both the fuel and the DSC and for the transfer of the loaded and sealed DSC to the HSM. These items are site-specific but typically must include the following major components:

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o a welding machine o DSC evacuation and helium backfill system o a heavily shielded transfer cask o a transfer vehicle capable of moving the loaded cask across the site o a cask positioning skid to adjust the cask position at the HSM in order to allow proper alignment before the DSC is transferred to the HSM o an alignment system to align the loaded c.ask with the HSM opening o a ram to push the DSC into the HSM o similar components to reverse the process in order to retrieve fuel assemblies from the DSC, if necessary, i l The staf f has reviewed these components primarily from the point of view of feasibility. That is, these components have been reviewed only to determine if the staff believes that all operations required to support the NUHOMS system are performable by current technology, that such equipment exists, and that such a system could perform its required functions.

l 1.2.4 Stored Materials Each HSM holds one DSC and each DSC holds seven irradiated PWR fuel assenblies. The proposed system is designed to permit storage of any PWR fuel with the following characteristics:

1. initial enrichment of 3.5 percent or less

2. average assembly burnup of 33,000 mwd /MTU or less

3. post-irradiation cooling time of at least 5 years 6

4

4. fuel assembly mass between any two adjacent spacer grids of no more than 106.56 Kg, and

5. distance between fuel assembly spa:ers of 0.665m or less. -

In addition, the neutron and gamma source strengths for c single OSC are limited to 9.98x108 neutrons per second and 1.67x10 16 MeV per second, respectively.

Although not specifically stated in the Fuel Specification of subsection 10.3.1.1 of the TR, the following additional conditions must also be met by any fuel stored in NUHOMS:

6. fuel rod cladding of zircaloy, and

7. no known or suspected gross cladding failures or other structural defects prior to storage i

A fuel assembly not meeting the specified conditions must be analyzed specifically before it can be stored in the proposed NVHOMS design.

1.3 Identification of Agents and Subcontractors No agents or subcontractors for design are specifically identified in subsection 1.4 of the TR. However, in the area of nuclear criticality analysis, reference is made to work of Babcock and Wilcox Nuclear Power Division.

1.4 Generic HSM Arrays The TR is based on a 2x4 array of front-loaded HSMs. Although it was noted that other arrays and other loading methods (specifically rear-loading) are possible, none were presented for review and approval.

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2.0 FRINCIPAL DESIGN CRITERIA 2.1 IntroductJon Subpart F of 10 CFR Part 72 sets forth general design criteria for the design, fabrication, construction, testing and performance of structures, systems and components important to safety in an ISFSI. This chapter pre-sents a discussion of the applicability of these criteria to the NUHOMS tystem and the degree to which the NUTECH, Inc. TR is in compliance with these criteria. Section headings in this chapter are taken from selected sub-sections of Subpart F of Part 72.

2.2 Fuel to be Stored 1

The NUHOMS system is designed for dry, horizontal storage of irradiated PWR fuel from nuclear power stations. Acceptable fuel characteristics are presented in subsection 1.2.4 of the SER.

The DSC design is based on a hybrid set of design parameters which will accommodate standard

• fuel assembly arrays of (1) 15x15 and 17x17 designed by Babcock and Wilcox, (2) 14x14 and 16x16 designed by Combustion Engineering, and (3) 14x14, 15x15, and 17x17 designed by Westinghouse.

The criticality analysis is based on standard Westinghouse 15x15 array of fuel at an initial enrichment of 3.5 percent. The source term for shielding calculations was derived from PWR fuel with an initial enrichment of 3.2 percent, irradiated to 33,000 MWD /MTU at a specific power of 37.5 MW/MTU, and cooled for five years after irradiation before being stored in the NUHOMS system.

• Note that some special cases may require specific review, such as the extra long 17x17 Westinghouse fuel.

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1 The staff has reviewed the use of these design parmeters for the fuels in this safety evaluation report and fcund the requirements of 10 CFR Part 72 are met as ap lied to the DSC design, criticality design, and shielding design.

2.3 Quality Standards Quality standards for structures, systems and components important to l safety are required by 10 CFR Section 72.72(a). Sections 3.4 and 11.2 of the TR identifies components of the NUHOMS system which are classified as important to safety. A quality standard provides numerical criteria and/or i acceptable methods for the design, fabrication, testing, and performance of these structures, systems and components important to safety. These stan-dards should be selected or developed to provide sufficient confidence in l the capability of the structure, system, or component to perform the r.e- i quired safety function. Since quality standards are generally embodied in widely accepted codes and standards dealing with design procedures.

l materials, fabrication techniques, inspection methods, etc., judgments  ;

regarding the adequacy of the standards cited in the 1R are presented in the sections of thin report where the standards are applicable.

2.4 Protection Against Environmental Conditions and Natural Phenomena Section 72.72(b) of 10 CFR Part 72 requires the licensee to provide protection against environmental conditions and natural phenomena. Secticn 3.2 of the TR describes the structural and mechanical criteria for tornado and wind loadings, tornado missile protection, flood protection, seismic design, snow, ice and dead loads, pressure and thermal loads due to normal I operating conditions and accident conditions, normal and accident handling loads,. accidental drop loads, and combined loads.

This section discusses the adequacy of the selected criteria for pro-3 tecting the reinforced concrete ""' the DSC support assembly, the DSC itself and the internal basket c 4.ents against environmental conditions and natural phenomena. The above mentioned structures and component are important to safety because they contribute to the safe confinement of the

! radioactive spent fuel assemblies. The technical basis for determining the i

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adequacy of these criteria is specified by the regulatory requirement to consider the most severe of the natural phenomena reported for the site and surrounding area, with appropriate margins to take account of limitations of data. Because the NUHOMS system was not designed for a specific site, the regulatory requirement is interpreted to mean that the NUHOMS system should be reviewed against the environmental conditions and natural phenomena provided for either by the limits specified in the TR or against the most severe of the natural phenomena that may occur within the boundaries of the l United States. Table 2.1 summarizes the design criteria for normal aparating conditions. Table 2.2 summarizes the design criteria for off-normal operating cor.ditions. Table 2.3 summarizes the design criteria for the accident conditions.

As can be seen in Tables 2.1, 2.2, and 2.3, most of the design criteria for all of the safety related components are not explicitly defined by applicable codes or regulations. The vendor has applied engineering judgment to determine a performance envelope or design criteria for their system tvhich satisfies the intent of 10 CFR 72.72. Thus it is important to examine how suitable the criteria are and what, if any, restrictions or conditions should be placed on the use of the particular design, in this I

case the NUHOMS system. This section of the SER discusses these issues.

The next chapter discusses how well the NUHOMS system satisfies the l criteria, and what, if any, restrictions or conditions should be placed on the use of it by a licensee.

The normal operating loads of the NUHOMS system are dead weight loads, design basis internal pressure loads, desi0n basis thermal loads, operational handling loads and design basis live loads. The primary criteria associated with each of these loads are presented in Table 2.1 together with a descriptior, of the NUHOMS components affected by each uf the loads. The staff has revieved these criteria and considers them to be acceptable as defined in Table 2.1.

As seen from Table 2.1, a snow load criterion of 72 psf was selected l

for the design of the HSM. Although this criterion was selected from ANSt l

A58.1-1972 rather than the current edition of this code, ANSI A58.1-1982, l this criterion is acceptable to the staff. This criterion is considered

! acceptable on the basis that the NUHOMS design only considers snow loads as l

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Table 2.1. Itnal Operating Conditim Design Criteria Smary.

Applicable Replations and MC Staff Cuents Design Codes Refennced e Suitability /

[orroonent Load Tvoe Desie Parameters By MITECH hstrictims H91 DeadLoad Dead weight (including DSC wight): WA Verified by Sm.

380,132 lb.

700 Arbient Tep.

Desip Basis Maxi m inside concm te surface WA Operating taperature: 214 F is not suitable. ,

Teperatum Maxinun outside concate surface 100 F is acceptable  ;

Loads taperatum: 189 0F at arrbient taperatum: 70 0F*

Operational $fraulicRamload: 6,200 lb. WA Verified try EER.

Handling Load Snow and Ice Maxinun pmssum: 72 psf ANSI A58.11972 Not used. Included with Live toals.

Live Loads Desip load: 200 psf WA Acaptable.

DSC anc DeadLoad Weight of DSC loaded: 21,237 lbs. WA Verifiai by Sa.

Intemal Cm ponents Desip Basis Design Pmssurt: 25 psig WA Verified by SE.

Intemal Pmssure toat; Design Basis Maxinua DI Taperature: 400 0F WA VerifiedbySG.

Operathg at ex+, rime attent tap. of 1250F Terperaturv Loads Operational @!raulic Cam load: 6,?00lb. WA Verifirxl by SG, Hanf1lnj toad DSC Davi load DSC plus self might: 22,812lb. WA Verifia! by Sm.

Support Asswoly Operational $1raulic Rarn Isai: 6,200 lb. WA Verified Ly SG.

Handling Lead 0 0

• RfrE01 used 70 F in Table 3.2-1 of their TR. Itaver, they discuss tha use of 100 F as an irt)imt tsperaturu for sme condition:.

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Table 2.2. Off-Nomal Operating Condition Desip Criteria Sumary.

Applicable Regulations and MC Staff Cements Caesip Codes Refennced on Suitability /

Fiwjal_ Loari Tvoe _., Desie Parameters By MITE 04 Restrictions itN jfM Air Inlet Maxism inside concrete surface tyA Bomded by accident Bhdege teperatuti: 344 0F analysis.

Phxinn outside conotte surface taperature: 215 F at maxiam abient taperatum.: 1250F

[6C and Jared lbdraulic Ram load: 22,00?lb. .VA Verified by SER.

fntemal Condition Lor:ponents Hardling loads C',0 Jzrmed itdraulic Ran load: 22,0[0 lb. fyA Verified by SER.

Stwort Condition Assaibly Ha.d ing

. Loads t

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Table 2.3. Accident Condition Desip Criteria Sumary.

Aplicable Regulatior.s and E Staff Coments Desty Codes Referinced on Suitability /

Canoonent Load Tvoe Desim Parameters Bv PUTE01 Restrictions HSM Blockage of Maxian inside concrete surface PVA Verified by SER.

Air Inlets tenperature: 459 F S4 ject to daily and Outlets Maxian outside concrete surface inspectim for tarperature: 210 F blocked x;xnings.

At maxinn arbient tarperatum: 1250F Design Basis Maxian wind pmssum: 375 psf ACI 349-80,PEC Adeg2 ate.

Tomado Maxicun wind velocity: 360 nph Reg. Guide 1.76,

1. 61 A58.11972 NT Missile Maxinun velocity: 126 @ PUEG 0800, Verified by SER.

Types: Autonobile, 3,%7 lb. Section 3.5.1.4 8 in, dian shell, 276 lbs 1 in. solid sphere Flood Maxinn water height: 400 ft. 10CFR72.72 Adequate for limit Maxian velocity: 16.2ft/sec design. Licensee to detemine site design parameters and check against El 349-80 egiations 2, 4, 7.

Seismic Desi p horizontal acceleration: 0.25g E Reg. Guides Pmcedum for cmbining Desi m vertical acceleration: 0.179 1.60 and 1,61 ww ents not accept-10CFR72.66(a)(6) able. Otherwise acceptable.

USC and acident Drry hk wrtical &<eleration: 48g PVA No discussion of Intemal Peak horizmtal (kreieratim: 349 off-axis orientation Cmponents by NJTE01.

Accident hinun intemal pmssum: 39.7psig tVA Verified by S m.

Inton.a1 Caned by arbient tmp: 125T 0

Pressum DSC taperatum: 644 F Blcckag! cf Airinlets and Qitlets OSC arsi finod Maxinun water height: 400 ft. 10CTR72.72(b)2. See note tnier H91.

Intemal Maxinn velocity: 16.2 ft/sec Caponents Seisnic Desim horizontal acceleratim: 0.259 R % . Guides See note mder fDi.

Design vertical zaeleratim: 0.179 1.60 and 1.61 DSC Soisnic Desim horizontal acceleration: 0.259 t0C Reg. Guides See note under IGi.

I Support Design vertical acceleratim: 0.179 1.60 and 1.61 i Assmbly Blockage of Mimm DSC tmperature: 644 F IVA Ece note under HS'i.

i Air Inlets At raxima arbient imperature: 125 0F and 02tlets 13

a part of a 200 psf live load and that this total load is considered to be acceptable for design.

It should be noted that Table 2.1 lists the operating basis temperature load for the HSM as the load occurring at an ambient temperature of 700F.

The TR identifies the ambient temperature criterion as 700 F in certain parts of the TR even though some thermal calculations have been made for the 100 0F case. The staff does not consider 70 F 0 to be acceptable as the ambient temperature condition since certain locations in the United States can have steady-state temperatures over this value for extended periods of time.

However, the staff considers 100 0F to be acceptable as the ambient temperature used to determine the operating basis temperature load.

The reason that the staff considers 700 F not suitable for an ambient temperature criteria is that its use leads to a non-conservative concrete surface temperature for the HSM. The material properties of concrete 1 subjected to elevated temperatures over an extended period of time are reduced. The use of concrete in this design is discussed more fully in Section 3.3.2 of this report.

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Off-normal events are events which are expected to occur on a moderate frequency. They include blockage of HSM air inlets and a jammed canister during loading or unloading. The primary criteria associated with each of these loads are presented in Table 2.2 together with a description of the NUH0MS components affected by each of the loads. The staff has reviewed these criteria and considers them to be acceptable as defined in Table 2.2.

As stated by 10 CFR 72.72, those structures of an ISFSI that are important to safety must be able to withstand the effects of accident condi-tions due to extreme environmental conditions, natural phenomena and postu-lated accidents. The extreme environmental and natural phenomena conditions include tornado winds and tornado missiles, flood earthquakes, lightning, snow, and ice. The accident conditions include blockage of the HSM air inlets and outlets, accidental internal pressure in the DSC, and a postulated drop of the DSC (corresponding to a distance of five feet while in a IF-300 cask) resulting in a 48 g deceleration in the vertical 14 4

orientation or 34 g in the horizontal orientation.* The primary criteria associated with each of these load conditions are discussed below and are summarized in Table 2.3. The staff has reviewed these criteria and considers them to be acceptable as defined in Table 2.3 with the exceptions discussed below.

As seen from Table 2.3, only the HSM was designed to resist tornado winds and tornado missiles. This is acceptable to the staff since the HSM is shown to resist the tornado wind and missile criteria without collapse or penetration. The DSC and DSC Support Assembly are internal to the HSM and are thus unaffected by tornado winds and/or missiles. The TR establishes the design basis tornado wind velocity as 360 mph in TR Section 3.2.1. This velocity includes a rotational velocity of 290 mph and a translational velocity of 70 mph. This criteria has been taken from the most severe criteria specified in NRC Regulatory Guide 1.76 and is acceptable to the staff.

The wind pressures applied to the HSH that correspond to this tornado wind criteria have been determined using the procedure outlined in ANSI 58.1-1972. Noting that this standard has been superseded by ANSI A58.1-1982 and that significant strides have been made in recent years in understanding the effects of tornados on structures, it is recommended that any site-specific application for the NUHOMS system be checked against this new standard. It should be further noted that the staff has accepted the use of ANSI A58.1-1972 in this TR on a basis that the internal loads / stresses caused by wind pressure are very small as compared to the effects of other criteria.

• Examination of Figure 4.2-3 of the TR shows the centerline of the door in the HSM is typically 6.4 feet. Thus the DSC must be lifted to 6.4 feet before inserting into the HSM. If the IF-300 shipping cask, cited as an example in the TR, were dropped onto an unyielding surface from this height, the decelerations would exceed NUTECH's deceleration design criteria.

However, a transport cask and the impact surface (s) are site specific. The staff concludes that a drop from 6.4 feet could be readily accommodated by site specific design conditions. Therefore, these criteria are adequate.

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The tornado missiles used to design the HSM have been selected using criteria outlined in NUREG-0800, Section 3.5.1.4, III.4. The three missiles ,

and corresponding impact velocity are acceptable to the staff for the HSM '

design.

l The flood design criteria established in Section 3.2.3 of the TR is based on a determination of the flood water velocities that will cause an unanchored module to slide or overturn and of the stagnant water height that the NUHOMS system will survive without failure. Although the staff agrees with these velocity and water height calculations as being limiting design values in situations in which combined loads are not to be considered, they are not acceptable design limits when considering combined loads. For example, dead and live loads, normal operating thermal loads, and either seismic or wind loads are combined with flood loads in the load combinations specified in ACI 349-80 and used to design the HSM.

Because NUTECH did not specify a design basis flood which was included in any of their load combinations, a licensee must determine for their specific site what flood conditions exist. If any flood conditions do exist, a licensee must recalculate the appropriate load combinations in accordance with ACI 349-80 and show that the NUH0MS system is suitable for their site conditions.

A horizontal acceleration of 0.25g was established as a basis for seismic design in Section 3.2.3 of the TR. Although 10 CFR 72.66(b) does not specify a design earthquake acceleration for a canister / cask type ISFSI design, this selected acceleration is acceptable to the staff as being conservative with respect to anticipated seismic acceleration levels in most areas of the United States. This staff acceptance is further predicated on the fact that a site-specific evaluation will be required to establish geological and seismological requirements for each site-specific NUHOMS design as required by 10 CFR 72.66(b).

The vertical acceleration of 0.179 established in Section 3.2.3 of the TR is acceptable to the staff since this value is consistent with the Regulatory Guide 1.60 requirement that the vertical acceleration be 2/3 of the horizontal acceleration.

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The staff does not agree with the procedure specified in Section 3.2.3.2 of the TR for combining the effects of horizontal and vertical accelerations. The procedure used in the TR only combines the response from one horizontal direction with the response from the vertical direction. The procedure acceptable to the staff is as documented in NUREG-0800, Section 3.7.1. This procedure requires that responses are calculated from three mutually perpendicular directions (e.g., two horizontal and one vertical) and that the square root of the sum of the squares of these responses be calculated as the desired response of the structure.

Even though 10 CFR 72.72 does not define the severity of accidents or natural phenomena, a vendor or applicant must define them. Because the drop accident criteria results in the highest DSC loads, a discussion of suitability of this criteria is warranted. NUTECH has selected two deceleration values for a postulated horizontal and vertical drop of the DSC. The values were based on analytical and experimental results obtained during the development of the General Electric Company's IF-300 shipping cask. Although the IF-300 shipping cask is not required as a part of the NUHOMS design, NUTECH has stated that the deceleration values which would be imparted to the DSC when housed by an IF-300 cask represent an upper bound.

NUTECH has cited deceleration figures for three other cask configurations to support their statement. According to NUTECH, the Nuclear Assurance Corpo-ration defines 55 g as the maximum drop decelerations for the NAC S/T cask.

The Westinghouse MC-10 basket deceleration in a horizontal orientation is 55 g. The REA 2023 cask impacting on a redwood mat experiences 16.3 g and 26 g for vertical and horizontal orientations respectively (Ref. 4). Thus, NUTECH has selected two values based completely on the decelerations a OSC would experience if housed by an IF-300 shipping cask. These values are coincidentally bounded by those from other design configurations.

With regard to a possible slap-down of the cask, the TR states that the above deceleration loads "are to be used for bounding design of the transfer cask and handling system for both primary drops and any secondary slap-downs" (Ref. 1). It has not been demonstrated that a tipped-end horizontal drop of the DSC while housed by a cask (in this case, the IF-300 shipping cask) would not produce local loads in excess of the 48 g and 34 g. How-ever, the staff can accept the NUTECH criteria, because the NUTECH design 17

does not require the use of an IF-300 shipping cask. Both cask and any handling equipment are site specific. Thus it will remain the responsi-bility of the applicant to show that the two deceleration values chosen by NUTECH do bound any possible loads imposed on the DSC. Furthermore, because ASME Service Level D stress allowables are used for the accident conditions, NUTECH has specified that any drop accident will require opening of the DSC and removal of the fuel assemb'?es for inspection for possible damage. This accident condition will be discussed in greater detail in a subsequent section of this SER.

2.5 Protection Against Fire and Explosion The NUTECH TR does not specifically address protection of the NVHOMS system from potential fires or explosions. Instead, it relegates such analyses to a site-specific situation.

There are no flammable or explosive materials used in the construction and operation of the DSC or the HSM. Furthermore, the DSC has been analyzed to withstand a static external pressure of 176 psi, which would bound most explosions. Nevertheless, site specific conditions can exist with the potential for fire and explosions in or around the HSM and DSC. Therefore, any application of the NUHOMS system to a specific reactor site must analyze the consequences of fires and explosions and provide for protective and mitigative measures, as deemed necessary.

2.6 Confinement Barriers and Systems Pursuant to 10 CFR 72.72(h) the NUHOMS design must protect the fuel I cladding against degradation and gross rupture. The TR takes the position that the initial inert helium atmosphere in the DSC has no feasible way to leak out and be replaced by air and that the fuel cladding temperature will be held below levels at which damage could occur.

The staff agrees that a helium atmosphere can be maintained but believes that the mechanisms for long term cladding behavior were not j totally investigated by the applicant. After reviewing the current research l relating to spent fuel cladding damage mechanisms, the reviewers concluded I that a diffusion controlled cavity growth (DCCG) mechanism was the only I i

18

mechanism of damage for dry storage applicable to the storage conditions of the fuel rods that could cause degradation and gross rupture of the clad-ding. Under the influence of stress and temperature, this damage mechanism progresses by the nucleation and growth of cavities along grain boundaries.

This damage mechanism is serious since it can progress without external evidence of damage, may not cause pin holes or through cracks to relieve the internal pressure, and manifests itself by a sudden non-ductile type of fracture. The staff has therefore paid particular attention to evaluating the potential for cladding damage from this mechanism for the conditions of storage specified in the TR.

The only parameters that the designer may control to prevent cladding degradation or gross rupture in an inert environment are the maximum initial temperatures of the fuel rods and their temperature decay characteristics.

Both are governed by the quantity, specific power, and age of the fuel assemblies, and by the heat dissipation properties of the system. This SER addresses the thermal evaluation in Chapter 4 and fuel cladding integrity in Chapter 5 and Appendix A.

2.7 Instrumentation and Control Systems The NUTECH TR takes the position that the DSC and HSM are totally passive storage components and that monitoring instrumentation is unnecessary. It further states that system safety is demonstrated by analysis in the TR.

The staff agrees that specific instrumentation for the DSC and HSM is not needed where a program of regular surveillance is implemented as discussed in the TR. This is reviewed in Chapter 9 of this SER in conjunction with such site-specific, preoperational testing and surveillance procedures as may be set for licensed storage.

2.8 Criteria for Nuclear Criticality Safety Section 72.73 of 10 CFR Part 72 requires that spent fuel handling, transfer and storage systems be designed to be maintained in a subcritical configuration. The design safety margins should reflect design uncertainties including uncertainties in handling, transfer, and storage 19

3 conditions, data and methods used in calculations, and adverse accident environments. Section 72.73 also requires that the design be based on either favorable geometry or permanently fixed neutron absorbing materials.

Section 3.3.4 of the NUHOMS TR addresses nuclear criticality safety. The criticality analysis is reviewed in Chapter 7 of this report.

The acceptance criteria for nuclear criticality safety established in the present review was 0.95. This factor is widely accepted as a criticality prevention limit. The TR establishes that the maximum multipli-cation factor (95% confidence) for all credible configuration and environ-ments is 0.89. Thus an additional margin of safety is incorporated into the nuclear criticality safety design above what would be considered an acceptable design.

2.9 Criteria for Radiological Protection Sections 72.15, 72.67(a), and 72.68(b) set requirements for radiological protection, incorporating 10 CFR 20 by reference. Section 72.74 of 10 CFR 72 requires the provision of a) protection systems for radiation exposure control, b) a radiological alarm system, c) systems for

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monitoring effluents and direct radiation, and d) an effluent control system in a radiological protection program. Chapters 3 and 7 of the NUHOMS TR address radiological protection. The review of Radiological protection is reviewed in Chapters 6 and 10 of this SER.

The principal radiation protection design features of the NUHOMS design are both the access control and shielding, and the integrity of the double seal welds on the DSC. Radiological alarm systems and systems for monitoring effluents are not required in the NUHOMS design because of the double seal welds.

Discussion of access control in the TR is deferred to a specific site license application and is therefore not reviewed in this report.

The TR establishes shielding criteria for the NUHOMS module of an average external surface dose of less than 20 mrem /hr. In addition, criteria were established of 200 mrem /hr for the transfer cask ide and 100 mrem /hr on the DSC top lead plug. The shielding capability of the cask 20

relies primarily upon the bulk concrete shielding of the NUHOMS module and the DSC top lead plug.

Based on analyses presented in the TR, the staff concludes that the NUHOMS system meets the requirements for on-site radiological protection, including incorporation of ALARA principles. The staff also believes the NUHOMS system can be sited to meet the off-site does limits specified by Part 72.

2.10 Criteria for Spent Fuel and Radioactive Waste Storage and Handling Pursuant to 10 CFR Section 72.75, the licensee is required to design the spent fuel storage and waste storage systems to ensure adequate safety under normal and accident conditions. These systems must be designed with (1) a capability to test and monitor components important to safety, (2) suitable shielding for radiation protection under normal and accident conditions, (3) confinement structures and systems, (4) a heat removal capability having testability and reliability consistent with its importance to safety and (5) means to minimize the quantity of radioactive wastes generated.

Criteria covering items (1) through (4) have already been addressed in this SER in the preceding sections of this Chapter. Regarding item (5), the TR identifies the major contributor to generated wastes as the fuel pool water, air and helium discharged from the transfer cask, and the DSC during the loading and sealing of the DSC. No radioactive wastes requiring treat-ment are generated during the storage period either during normal or accident conditions.

Criteria for contamination limits on the exterior surface of the DSC are the same as used for off-site shipping casks, i.e. 10-4 pCi/cm 2 for beta / gamma emitters and 10-5 Ci/cm 2 for alpha emitters. Any other criteria for waste handling and treatment are referred to procedures which are site-specific.

The staff agrees to this approach for item (5).

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2.11 Criteria for Decommissioning Pursuant to 10 CFR Section 72.76, the licensee is required to design the ISFSI for decommissioning. The NUTECH TR claims that by keeping the contamination level low on the exterior of the DSC by assuring long term integrity and through controlling contamination before the DSC is placed into the HSM, then the surfaces of the HSM will have even lower contamination levels throughout its life. Under this condition the HSM can be demolished and removed for decommissioning.

The TR also postulates the disposal of the DSC in a high level nuclear waste repository as one option. Alternatively, it may be possible to decontaminate the DSC for disposal separate from the nuclear fuel, the latter to be sent to the repository. The specific decommissioning plan for the DSC must await the outcome of the U.S. high level waste disposal program and will be site-specific.

The staff concludes that adequate attention has been paid to the design of the DSC and HSM in preparation for decommissioning, considering the current state of knowledge.

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3.0 STRUCTURAL EVALUATION 1

l This chapter discusses in detail all aspects of the structural evaluation of the NUHOMS system. To give an overview of the results, a brief summary is provided. As discussed in Section 2.4, the design criteria are an integral part of the structural evaluation because 10 CFR 72.72 does not explicitly state what criteria mu:t be used. The staff summary and conclusions are therefore presented in terms of: (1) criteria suitability and any restricting conditions as they might apply to an applicant, and also (2) degree that the NUHOMS ISFSI design satisfies the criteria and any restricting conditions.

3.1 Summary and Conclusions This review has included an evaluation of all structural design criteria, analysis methodologies,. material specifications, allowable stress levels and structural analyses. The staff has reviewed the structural design of the NUHOMS system proposed by NUTECH and confirms that the design is in compliance with 10 CFR 72.72 with the exceptions outlined below.

The staff has reviewed all of the principal design criteria proposed by NUTECH for general applicability to NUH0MS system and has confirmed that these criteria are in compliance with 10 CFR 72.72 with the follcwing exceptions:

1. The ambient temperature for normal operating conditions is taken to be 1000 F, not 70 F0 as used by NUTECH.

2. A discussion was presented in the structural analysis portion of the TR regarding the magnitude of flood that would overload the various components of the NUHOMS system. However, the criteria presented in this discussion is considered to be too severe for use in combining the effects of flood with other appropriate load conditions. The staff accepts this omission of a " design basis" flood with the understanding that a " design basis" flood magnitude is a site-specific requirement.

For the flood condition, site-specific design parameters must be determined by a licensee. If these parameters are greater than zero, 23

they taust be combined in the load combinations specified by equations 2, 4 and 7 of Section 9.2 of the ACI Code 349-80. If they are not greater than zero, the NUTECH criteria are adequate.

3. NUTECH has chosen two values of deceleration during a postulated drop accident to bound the design of the DSC -- namely a 48 g vertical and a 34 g horizontal deceleration. The staff requires that any licensee show that the DSC will not experience any higher decelerations for any drop orientation, such as an off-axis orientation, that might be postulated for their site-specific use.

The staff has reviewed the analysis methodologies used by NUTECH in evaluating the NUHOMS structure and found them to be acceptable with the following exception:

1. The method of combining the two horizontal and one vertical compo-nent of the seismic event that was used by NUTECH was not in keeping with the conservative guidelines suggested by the NRC in NUREG 0800. However, the staff reviewed the design using the SRSS criterion as presented in NUREG-0800 and found the design to be acceptable.

The staff has reviewed the material specifications and allowable stress levels u ed by NUTECH in evaluating the NUHOMS system and confirmed that this data is in compliance with 10 CFR 72.72 with the following exceptions:

1. Since the maximum temperature on the inside of the HSM concrete module 0

has been determined to be 214 F under the effect of the design basis normal operating temperature of 100 0F and since the ACI 349-80 code only permits such a temperature to exceed 150 0 F if justified by a test program or related test data, the applicant recommends that a surveillance program be established to annually inspect the inside surfaces of the HSM concrete for the first ten years of the use of a module. These inspections may be allowed to occur at somewhat longer intervals after ten years provided that no degradation of concrete has been seen (i.e., spalling, cracking, etc.) and the original DSC is retained in the module, unless it can be demonstrated that the extent of degradation presents no safety hazard. Since the staff realizes that 24

certain locations within the HSM may be difficult to reach with inspection tools / instrumentation, the determination of the acceptability of a specific approach to performing this inspection will be the subject of a site-specific review.

2. The staff has noted an inconsistency in the modulus of elasticity data for concrete presented in the TR (Section 8.1.1 vs. Appendix D). The staff has reviewed each specific application of the original data used by NUTECH and found each use to be acceptable. However, should site-specific, parameters be outside the bounds of the TR (e.g., higher ambient temperature, higher seismic loading, etc.) it is recommended that appropriate modulus of elasticity data be used.

The staff has reviewed the structural analysis of the DSC and the DSC support and finds that:

1. The DSC and internals are in compliance with the ASME Code rules for Service Level A stresses for normal operating conditions and off-normal events.

2. The DSC and internals are in compliance with the ASME Code rules for Service Level D stresses for the drop accident conditions postulated by NUTECH.

3. The DSC support members are in compliance with the ASME Code rules for Service Level A stresses for all cases analyzed.

3.2 Description of Review This chapter evaluates the structural response of the NUHOMS dry stor-age system to loads due to normal operating conditions, off-normal condi-tions, accident canditions, and loads due to environmental conditions and natural phenomena. The review procedure discusses the assumed loads, the material properties, and the ASME or ACI code allowable stress limits. The review provides an evaluation of the structural analysis which NUTECH supplied in the TR for each of the components and systems important to safety.

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3.2.1 Applicable Parts of 10 CFR 72 The parts of 10 CFR 72 which are applicable to the review of the NUHOMS ISFSI are: 72.72(a) which deals with quality standards; 72.72(b) which requires that structures important to safety be protected against environ-mental conditions and natural phenomena, as well as appropriate combinations of effects including accident conditions; 72.72(c) which requires protection against fires and explosions; and 72.72(h) which requires protection of the fuel cladding against degradation and gross rupture.

3.2.2 Review Procedure The TR was reviewed for compliance with the applicable parts of 10 CFR 72 as outlined above. The systems comprising the NUH0MS ISFSI including the HSM, the DSC and the DSC support assembly were considered first as systems and secondly as individual parts making up complete systems. All phases of normal operating conditions, off-normal operating conditions, extreme natural phenomena, and accident conditions resulted in loading conditions which were analyzed by NUTECH and reviewed for completeness.

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3.2.2.1 Design Descriptions l

A brief description of the NUHOMS ISFSI was given in the first chapter i of this report. A more detailed description of the design is given in this l chapter which highlights aspects of the design that are important to the l structural evaluation.

The safe storage of irradiated fuel is provided by two major components l in the NUH0MS design: the canister body provides confinement of contamina-tion; the concrete module and lead end plugs of the canister provide shield-ing for biological protection. However both components contribute to the cocling of spent fuel.

The HSM is a reinforced concrete structure that provides projectile impact and weather protection for the DSC and serves as the primary biologi-cal shield for the irradiated fuel during storage.

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i The HSH is constructed of 4000 psi normal weight (145 pounds per cubic foot) concrete with Type II Portland cement meeting the requirements of ASTMC150. The aggregate will meet the specifications of the ASTMC33. The reinforcing steel is #9 ASTMA615 Grade 60.

The HSM wall thicknesset were determined to meet shielding require-ments. They will be able to protect the DSC against tornado generated missiles, which effectively bound reasonable impact accidents, as well as other environmental conditions, natural phenomena, and accidents.

The structural properties of the concrete when subje:ted to the elevated surface temperature for the long term are discussed in Section 3.3.2 of this report.

The canister is fabricated from 0.5 inch thick 304 stainless steel which has a single full-penetration weld along the entire length of the canister and double weld seals on each end. The canister encloses a basket assembly which can house seven irradiated fuel assemblies. The basket consists of seven spacer discs of 1.25 inch thick 304 stainless steel which are fixed axially by four 2.0 inch diameter 304 stainless steel rods running the length of the canister. There are seven square section boral tubes which house the spent assemblies. The primary function of the spacer discs and axial support rods is to maintain dimensional stability for the boral tubes which house the spent fuel in the event of a vertical or horizontal drop.

The end caps and the canister shell serve to confine the radioactive material by maintaining a leak-proof container as well as providing struc-tural rigidity for the drop accident. The bottom lead plug is captured between a 1.0 inch thick inner pressure plate, and a 1.25 inch thick outer cover which serves as a redundant pressure boundary. The upper lead plug has a 0.5 inch thick steei plate which when welded to the canister serves as an inner pressure boundary. A 1.25 inch thick steel top cover plate serves as an outer pressure boundary. All closure welds are multiple pass thereby reducing the likelihood of a pinhole leak.

The longitudinal weld is to be radiographed and inspected in accordance with appropriate articles of the ASME Boiler and Pressure Vessel Code Sec-27

tion III, Division 1, subsection NB. The two welds for the bottom end plates will b'e either radiographed or ultrasonically tested. Dye penetrant or ultrasonic test methods will be used for the three welds on the top lead plug. These top welds cannot be radiographed because the irradiated fuel will already be installed before these tests can be made. Additionally, a helium leak test will be made before the second seal weld for the top cover plate is completed.

The DSC rests on a fabricated support rail assembly which is bolted to corbels cast in the interior walls of the HSM. The support rails will also be welded to a cast-in-place pipe in the HSM front opening. Thermal expan-sion af the support rails is eliminated by using slotted bolt holes with suitably torqued bolts. Corrosion of the structural carbon steel will be retarded by either paint, galvanizing, or hard plating.

During loading of the DSC into the HSM, frictional loads between the DSC and the support rails in the HSM will be reduced by the use of a dry film lubricant applied to both sliding surfaces. The particular product selected by NUTECH is Everlube 812, which was designed for radiation service.

The DSC is prevented from sliding lognitudinally along the rail during a seismic event by a seismic restraining assembly. Although the details of this design are not presented in the TR, NUTECH has committed to incorporating this feature into the design (Ref. 5).

3.2.2.2 Acceptance Criteria The structural integrity of the NUHOMS HSM, DSC, and DSC support assembly will be judged adequate if it can be demonstrated that the stresses induced by the loads noted in Section 3.3.1 below are lower than the allowa-ble stress limits for the components important to safety and that all other material properties are consistent with applicable code requirements. The allowable stress limits are documented in the TR in Section 3.2.5, Tables 3.2-4, 3.2-5, and 3.2-6.

28

3.2.2.3 Review Method The method of review used to assure that the TR was in compliance with 10 CFR 72.72 involved several steps. The TR was reviewed first for com-pleteness, insuring that all areas specified by 10 CFR 72.72 were addressed, and that the standard format and content specified by Regulatory Guide 3.48 (Ref. 2) was followed. All sources cited by the TR were reviewed for applicability to the design of the NUHOMS system. Chapter 3 of the TR, which sets out the design criteria, was examined critically for appropriate-ness. All underlying assumptions stated in the TR were assessed with respect to those suggested by two of the professional societies which guide design practice for pressure vessels and reinforced concrete structures for nuclear safety related items. The societies and their respective codes are:

the American Society of Mechanical Engineers (ASME Boiler and Pressure Vessel Code for Nuclear Power Plant Components,Section III, Division 1, Subsection NB, Class 1 components 1983) and the American concrete Institute (ACI 349-80, ACI 349R-80 and 1984 Supplement to Code Requirements for Nuclear Safety Related Concrete Structures).

Secondly, Chapter 8 of the TR, which covers the analysis of the design events, was reviewed in detail. This included verifying all calculations which could be executed without resorting to computer models. The finite element computer models were verified for accuracy by examining the input and output printouts for all STARDYNE (Ref. 6) and NUTECH post processor codes. All results which were included in NUTECH's Tables 8.1-7, 8.1-8, 8.1-9, 8.1-10, 8.2-3, 8.2-4, 8.2-5, 8.2-6. 8.2-7, 8.2-9, 8.2-10, 8.2-11, 8.2-12, and Figure 8.2-14 were either verified by hand calculations or by examining the computer printouts. No independent computer analysis was performed.

3.2.2.4 Key Assumptions Assumptions made by staff reviewers in verifying NUTECH's models will be discussed on a case by case basis in later sections of this Chapter.

29

3.3 Discussion of Results ,

3.3.1 Loads The loads specified in the TR for use in designing the NUHOMS system l are described in this section together with comments by the staff regarding their acceptability. Loads are described for normal operating, off-normal operating and accident conditions.

3.3.1.1 Normal Operating Conditions Section 8.1.1 of the TR defines the normal operating conditions of the NUH0MS system. The normal operating loads of the NUH0MS system are dead weight loads, design basis internal pressure loads, design basis thermal loads, operational handling loads and design basis live loads. The primary criteria associated with each of these loads are presented in Table 2.1  !

together with the description of the NUHOMS components affected by each of I the loads. The staff has reviewed these loads and considers them to be acceptable.

An exception to the NUTECH design criteria which effects loads for normal operating conditions was the selection of the ambient temperature.

The staff has reviewed thermally induced loads as a result of the 100 F0 ambient temperature and finds the design adequate.

3.3.1.2 Off-Normal Conditions Section 8.1.2 of the TR defines the off-normal events. These are events which are expected to occur on a moderate frequency. They include blockage of HSM air inlets and a jammed canister during loading or unload-ing. The primary criteria associated with each of these loads are presented in Table 2.2 together with a description of the NUHOMS components affected by each of the loads. The staff has reviewed these loads and considers them to be acceptable.

30

3.3.1.3 Accident Conditions Section 8.2 of the TR defines the accident conditions due to extreme environmental conditions, natural phenomena and postulated accidents. The extreme environmental and natural phenomena conditions include tornado winds and tornado missiles, flood, earthquakes, lightning, and snow and ice. The primary criteria associated with each of these load conditions have previously been discussed in Section 2.4 of this SER.

The accident conditions include blockage of the HSM air inlets and outlets, accidental internal pressure in the DSC, and the postulated 48 g horizontal or 34 g vertical deceleration drop of the DSC. The primary criteria associated with each of these loads are presented in Table 2.3 together with a description of the NUHOMS components affected by each of the loads. The staff has reviewed these loads and considers them to be acceptable as defined in Table 2.3.

The exceptions to the NUTECH design criteria which effect accident condition loads have been previously discussed. They include the design basis flood for combined loads for the HSM, the off-axis orientation of the cask drop for the DSC, and method of combining seismic load components for the HSM and the DSC. The first two exceptions are acceptable to the staff on the basis that they are site-specific parameters and as such must be addressed by a licensee. The third exception has been addressed directly by the staff. The staff has made independent calculations to show that the NUH0MS system is adequately designed against the earthquake condition.

3.3.2 Materials -

The mechanical properties of all materials used in the fabrication of components important to safety are listed in Section 8.1.1, Table 8.1-2 of the TR. The source identified in these tables for properties of steel is the ASME Boiler and Pressure Vessel Code, Section III-1 Appendices. This source is an acceptable standard and is in compliance with the quality requirements of 10 CFR 72.72(a).

The sources identified in Table 8.1-2 for the structural properties of  :

lead and for borated aluminum are not recognized standards consistent with 31

i the quality requirements of 10 CFR 72.72(a). However, the material properties for lead shown in the TR were not used. The stiffness imparted by lead shielding plug was not considered in the stress analysis. Instead, steel rods were modeled as a substitute for the lead. In the case of borated. aluminum, which was the material used for fuel assembly guide tubes, the particular finite element used to model the tube was a sandwich element.

Two thicknesses of stainless steel sheet cover a borated aluminum core. The only property required for the core was the shear modulus of elasticity.

For this property, NUTECH used the shear modulus for aluminum consistent with typical aluminum alloys (Ref. 8). For these cases, the staff concurs that the quality requirements of 10 CFR 72.72(a) are met.

The source identified in Table 8.1-2 for the mechanical properties of concrete and reinforcing steel is the Handbook of Concrete Enaineerina (Ref.

9), a document that is not considered to be an acceptable standard meeting the quality requirements of 10 CFR 72.72(a). This data is supplemented by a review of concrete behavior under sustained elevated temperatures that is presented in Appendix D to the TR. The data presented in Appendix 0 is substantiated by a number of references, most of which are publications of

, the American Concrete Institute and the Portland Cement Association (PCA).

Both of these organizations publish recognized standards consistent with the quality requirements of 10 CFR 72.72(a).

It is stated in Section 1.3.1.2 that the HSM has been designed to the combined load criteria stated in the ACI 349-80 Code. This Code is recognized by the NRC staff as being a standard for use in the design of nuclear safety related concrete structures and, as such, is considered to be consistent with the quality requirements of 10 CFR 72.72(a). Appendix A to ACI 349-80 includes a discussion regarding the design of concrete structures subjected to long term high temperatures. It is stated in this Appendix that temperatures shall not exceed 1500 F except for local areas, such as around penetrations, which are allowed to have increased temperatures not to exceed 2000 F. It is further stated in this Appendix'that temperatures higher than those stated above may be allowed if: tests are provided to evaluate the expected reduction in strength; if this reduction is applied to design allowables; and if evidence is provided which verifies that the increased temperatures do not cause deterioration of the concrete either with or without load.

32

Appendix D was prepared in response to an NRC staff concern that long term temperatures in excess of the ACI 349-80 Code allowables identified above can be reasonably expected to occur over relatively long portions of the life of the HSM structure. It is shown on Table 8.1-12 of Section 8.1.3.1 of the TR that the maximum temperature is 214*F in a region on the roof of the HSM that is not a local area as defined by the Code. Temperatures as high as 459 F are

. presented in this table for accident cases.

The staff has reviewed the data presented in Section 8.1.1.5 and Appendix D of the TR_that has been presented as evidence of th e successful use of con-crete in situations in which sustained temperatures of up to 250'F can be ex-pected to occur. In addition, the staff has reviewed other data on this subject, most notably a literature review prepared by the Portland Cement Association for EPRI on the high temperature behavior of Portland Cement and refractory concretes (Ref. 10). The staff has concluded that the use of concrete in the anticipated temperature environment is acceptable from a material quality point of view pro-vided that the interior and the exterior surfaces of the HSM are given periodic inspections over the life of the installation to ensure that no visible deterio-

. ration of the concrete has taken place. The staff realizes that certain loca-tions within the HSM may be difficult to reach with inspection tools /instrumen-tation. The determination of the acceptability of a specific approach to per-forming this inspection will be the subject of a site-specific review.

These inspections could occur at periods of about one year, as the applicant recommends, during the first few years of life of the HSM with the possibility of longer periods in subsequent years provided that no significant deterioration is seen. It should be noted that one justification for this con-clusion is that the design temperatures for the HSM are expected to decrease significantly over the first few years of HSM operation and, consequently, should be less than the ACI 349-80 requirement of 150 F prior to the time of any anticipated deterioration of the concrete.

Although the staff agrees with the use of concrete at the anticipated normal operating and accident temperatures, it does not a gree with certain material properties as presented in Table 8.1-2. Specifically, the modulus of elasticity data presented in this table for 4,000 psi strength concrete 33 L

does not appear to be consistent with similar data presented in the upper portion of Fig. D.4 of Appendix D of the TR. A consistent set of data should have the following significantly reduced moduli, assuming the modulus of elasticity for unheated concrete to be 3.64x106 psi-Modulus of Elasticity (1.0x10 6psj)

Temoerature (Cearees F1 From Table 8.1-2 From Fia. D.4 100 3.64 3.46 200 3.28 2.80 300 3.09 2.11 400 2.73 1.46 It should be noted that the reduced moduli given above were for concrete heated to the specified temperatures for a period of only 2-1/2 years.

The staff has concluded that the use of the higher values of modulus of elasticity from Table 8.1-2 of the TR is conservative for those situations in which the modulus of elasticity is used in performing thermal stress analysis. However, the lower values should have been used in making calculations that involve the stiffness of the structure, such as in computing the natural frequency of the structure used for determining the seismic response of the system. The staff has checked these frequency calculations with the reduced modulus of elasticity values and has deter-mined that the increase in seismic response of the structure is small and well within the dynamic load factor of 2 conservatively used by NUTECH in computing the seismic response of the structure.

34 i

3.3.3 Stress Intensity Limits The mechanical properties of the structural materials used in the design of the NUHOMS system are listed in Table 8.1-2 of the TR. These properties, including allowable stress intensitities, are listed as a function of temperature for a variety of materials as described below.

The ASME Boiler and Pressure Vessel Code, Section III-1, was identified in Table 8.1-2 of the TR as the source for stress allowables for Type 304 stainless steel, Grade A36 carbon steel and Grade A325 steel used for bolts.

As discussed in Section 3.3.2 of this SER, these stress allowables are acceptable to the staff for use in the design of the NUHOMS system.

The Handbook for Concrete Enaineerino (Ref. 9) was identified in Table 8.1-2 of the TR as the source for stress allowables for concrete and reinforcing steel. As discussed in Section 3.3.2 of this SER, the staff does not concur in the use of Reference 9 as the authoritative source for concrete and reinforcing steel allowable stresses. However, the staff has reviewed the pertinent ACI and PCA data and concurs in the stress allowable values as presented in Table 8.1-2 of the TR for these materials.

3.3.4 Structural Analysis 3.3.4.1 HSM 3.3.4.1.1 Normal Operating Conditions The HSM concrete structure was analyzed for the effects of dead loads ,

(including effects of creep and shrinkage), live loads, and design basis temperature loads. In addition, the HSM door was analyzed for dead loads and handling loads caused by dropping the door on its bottom " zee" section, and the internal heat shield was analyzed for dead loads.

The HSM concrete structure analysis results are presented in Section 8.1.1.5 and Table 8.1-10 for each of the normal operating conditions considered. These results are presented in the form of maximum moments and shears which are compared with the ultimate moment and shear capacities of a typical cross-section of the module. These maximum moments and shears are 35 1

l l

l

1 1

summarized in Table 3.1 of this SER. It is seen from this table that these maximum moments and shears are significantly lower than the associated capacities of the module.

The staff has reviewed each of the individual normal condition cases presented in Table 3.1 and agrees that the results are reasonable and con-servative. It should be noted, however, the design basis operating thermal load case was conducted using temperature distributions across the concrete wall of the HSM of 700 F in the normal ambient thermal analysis. Since the temperature variations across the HSM walls and roof are more for this case than for the actual 100 0F design basis operating temperature care, the staff concludes that the results as presented are conservative and thus <

acceptable.

As previously noted in Section 3.3.2 of this TR, the staff finds the HSM design acceptable at an ambient temperature of 1000 F provided periodic inspection for degradation of concrete (spalling, cracking, etc.) is performed.

he staff has concluded that the structural design of the HSM for normal operating conditions has concluded that the design is acceptable.

3.3.4.1.2 Off-Normal Conditions .

The effect of increased temperatures due to the blockage of the HSM 4 inlet was the only off-normal event considered in the TR to have an effect on the HSM. A thermal analysis was performed for this event as de:cribed in Section 4.3.2.2 of this SER. The results from this thermal analysis were used to perform a structural analysis on the HSM as reported in Section 8.1.2.2 of the TR.

It is stated in Section 8.1.2.2 of the TR that this structural analysis considered the effect of a cracked cross-section when performing stiffness '

calculations. The staff agrees with the use of this cracked section analysis procedure since it is permitted as a special case in the ACI 349 80 Code. The staff has reviewed the procedure used to perform this cracked section analysis (Ref.11) together with a review of the special conditions ,

36

Table 3.1. Maximum Internal Moments and Shears in HSM.

Lead or Maximum Loading Capacities Load Comoination Description Vmax (K) Mmax finrK) Vu (K) Mu fin-K)

Individual Loads:

Dead Load (D) 6.6 614.8 50.0 3757.0 Flood Load (F) N/A N/A 50.0 3757.0 Live Load (L) 0.7 10.0 50.0 3757.0 Seismic Load (E) 11.7 499.5 50.0 3757.0 Wind Load (W) 6.5 275.8 50.0 3757.0 Design Basis Operating Thermal Load oT at 2500F 20.3 2495.2 50.0 3757.0 Air Inlet Blockage '

0 Thermal Load T a at 400 F 14.8 1818.7 47.7 3450.0 Air Inlet and Outlet Blockage Thermal Lo.ad aT ** at 400 F 25.8 3155.0 47.7 3450.0 Load Combinations:

(.2) 1.4D+1.4F+1.7L+1.7E 30.32 1718.6 50.0 3757.0 (3) 1.4D41.4F+1.7L+1.7W N/A* N/A* 50.0 3757.0 o

(4) D+F+L+To+E 39.30 3005.6 50.0 3757.0 (5) 0+F+L+T o+W N/A* N/A* 50.0 3757.0 (7a) D&F+L+Ta +1.15E (Off-Normal) 21.2 2396.8 47.7 3450.0 (7h) D+F+L+T 3 **+1.15E (Accident) 33.4 3423.1 47.7 3450.0 (10) 1.05D&l.05F+1.3t+1.3E+1.05T o 44.30 3181,0 50.0 3757.0 (11)1.050+1.05F41.3L+1.3W+1.05To N/A* N/A* 50.0 3757.0

'* Combinatior.s that include wind induced loads have been determined to be  ;

lots severe than comparable conditions that include seismic induced loads.

• • Ni) TECH did not use this data in Load Combination (7) but rather used Air Inlet only blockage case data in Load Combination (7).

P i

37

~

placed on its use by the ACI 349-80 code. The staff has concluded that tF-results from this stress analysis a~re acceptable.

The only other off-normal event considered in the TR is the jamming of a canister against either the DSC support 6r HSM components during loading or unloading. Although the effect of this off-normal event on the DSC is considered in the TR, there is no consideration of the effee.t of this evsnt on the HSM, However, the staff has reviewed the effect of this off-normal event on the HSM and has concluded that the only credible effect is the direct loading of the steel liner at the cask indentation on the HSH inside the module door. The effect of this loading has been reviewed by staff and is considered to be acceptable, ,

3.3.4.1.3 Accident Conditions The HSM was analyzed for the following accident evor.ts: (1) tornado winds and tornado missiles; (2) earthquake; (3) flood; (4) blockage of air inlets and outlets; and (5) lightning.

The analyses performed to evaluate the effect of tornado winds and tornado niissiles is presented in Section 8.2.2 of the TR, The analyses performed to evaluate the effect of tornado winds include an evaluation of the possible overturning of an unanchored module and the computation of wind induced maximum moments a.nd shears in an anchored module.

The analysis performed to evaluate potential sliding and/or overturning of an unanchcred module showed that a single unanchored module wculd not overturn but would slide when subjected to the tornado wind event. It is further concluded in the TR that a 2x4 array of modules would weigh considerably more than a single module and consequently would not slide.

Finally, the TR states that, even though sliding would not occur for a 2x4 l array of modules, tie-downs between the HSM and its foundation should be used to preyent the potential risk of sliding. The NRC staff concurs with this final conclusion that tie-downs should be used.

The computation of wind induced maximum moments and shears was performed using an equivalent two-dimensional frame id6alization of the HSM and the moment distribution inethod of analysis. The maximum moments and 38

6 stears resulting from this analysis are presented in Table 8.2-3 and are less than the respective moment and shear capacities of the concrete and consequently, are acceptable.

Three analyse,s were inclyded in Section 8.2.2.2 to evaluate the effect of tornado missiles on the HSM. Two cf these analyses considered the impact of a 276 pound, 8-inch diameter blunt nosed steel object on the HSM walls / roof and on the 2-inch thick steel door at the front of the HSM. The third analysit evaluated the impact of a large missile (i.e., a 3,976 pound automcbile) on the side wall of the HSM. Although results from these analyses indicated that the HSM would not be penetrated by the missiles, it was concluded that permanent deformation would probably take place, especially considering the impact of an sutomobile. The staff has reviewed these analyses and concludes that the design is resistant to tornado missile impact.

The analyses performed to evaluate the effect of earthquakes on the HSH are presented in Section 8.2.3 of the TR. These analyses include an evaluation of the possible overturning of an unanchored module and the computation of seismically induced maximum moments and shears in an anchored '

module.

As previously discussed in Section 2.4 of this SER, the licensee selected a horizontal acceleration of 0.25 g and a vertical acceleration of 0.17 g as a basis for seismic design. Although these accelerations are considered acceptable for this TR, the procedure used tn combine the effects of these accelerations is not considered acceptable. The procedure used in the TR only combines the response from one horizontal direction with the response from the vertical direction. The procedure acceptable to the NRC staff is as documented in NUREC-0800, Section 3.7.1. This procedure requires that responses are calculated from three mutually perpendicular directions (e.g., two horizontal and one vertical) and that the square root of the sum of the squares of these responses be calculated as the desired response of the structure.

The maximum moments and shears resulting from the seismic analysis of the HSM presented'in the TSAR are shown in Table 8.2-6 of the TR and Table 3.1 of this SER. Even though these moments and shears were not computed 39

a- -m using accelerations from three mutually perpendicular directions, they are considered to be acceptable to the NRC staff. This is because the accelera-tions used in the TR calculations have been multiplied by an arbitrary dynamic load factor of 2.0 and the resulting horizontal and vertical responses have been sumed. A further justification for accepting the licensee's analysis approach for this case is that the moments and shaars in the HSM from an earthquake in a direction along the longitudinal axis of the DSC are probably very small due to the very large concrete cross-sections in this direction.

Tne analysis performed to evaluate potential sliding and/or overturning of an unanchored module showed that a single unanchored module would not either overturn or slide when subjected to the design earthquake. The staff has reviewed this analysis and concurs with its results.

The analysis performed to evaluate the effect of flood on the HSM is presented in Section 8.'2.4 of the TR. This analysis demonstrated that a single, unanchored, suteerged module would slide if the flood water velocity exceeded 16.2 ft/sec and would overturn if the flood water velocity exceeded 19.10 ft/sec. This analysis further demonstrated that a single, anchcred, submerged module would become overstressed if the flood water velocity exceeded 57.4 ft/sec. The staff has reviewed each of these analyses and agrees with the results.

The analysis performed to evaluate the effect of air inlet and outlet blockage is presented in Section 8.2.7 of the TR. Secticn 8.2.7.1 states that the structural consequences due to the weight cf debris blocking the air inlets and outlets are bounded by the structural consequences of tornado and earthquake accidents. The staff agrees with this statement.

An analysis was performed to determine the effect of high temperatures caused by the blockage of both air inlets and outlets on the structural behavior of the HSM. The results from this analysis are presented in Section 8.2.7.2 of the TR wherein it is stated that the maximum bending moment in the slab is 3,155 in-k per foot of the slab. It is further stated that this bending moment is within the moment capacity of concrete at the elevated temperature of 400 F or 3,450 in-k. Using the material properties given in fable 8.1-2 of the TR, the NRC staff has computed this moment 40

capacity to be 3,450 insk per foot of slab at the elevated temperature of 4000F. Although not presented in the TR, the NRC staff has computed the corresponding ultimate shear capacity of the HSH walls and roof to be 47.7 kips per foot of slab at a temperature of 400 0F. (

Section 8.2.7.2 of the TR states that the analysis of the maximum concrete bending moment during the air inlet and outlet blockage event was performed in the same manner as that discussed earlier for the off-normal condition of air inlet blockage. As previously noted, the cracked cross-section procedure used to perform this analysis was reviewed by the staff and found to be acceptable.

A discussion is presented in Section 8.2.6 of the TR regarding possible 2 effects from a lightning strike in the vicinity of the HSH, The section concludes that the HSM will not be damaged by either heat or mechanical forces as a result of a lightning strike. The NRC staff concurs with this conclusion.

NUTECH has evaluated a number of normal, off-normal and accident load -

combinations in Section 8.2.10 of the TR. The staff has reviewed these lead ,

combinations, which were selected from a list of load combinaticas presented in ACI 349-80, and found them to be acceptable except as discussed below:

1. NUTECH did not include the effect of flood in the various load combinations as required by ACI 349-80. As previously noted, the staff has determined that a design basis flood will be considered to be a site specific requirement and, as such, appropriate moments and shears in the concrete will have to be added to each >

load condition if the site has a flood requirement.

2. Load combination 7 as presented in Table 8.2-10 of the TR con-sidered the off-normal thermal load caused by blockage of HSM air inlets rather than the accident thermal load caused by blockage of both HSM air inlets and outlets. The staff has reviewed the effect of this accident thermal load and included it in load combination 7b of- fable 3.1 of this S~R. It should be noted that this load combination has been found to be acceptable.

41 s

s

3. No load combinations that include wind effects have been considered. This is acceptable since the effects of wind have been shown to be less severe than the effects of earthquake.

3.3.4.2 DSC and Internals 3.3.4.2.1 DSC Normal Operating Conditions The dry shielded canister was analyzed for: (1) dead weight loads, (2) design basis internal pressure, (3) design basis operating temperature loads, and (4) operation handling loads. The canister internal parts were analyzed for: (1) dead weight loads and (2) design basis operating temperature loads.

Dead Weicht Loads for DSC The dead load analysis for the DSC as presented in Section 8.1.1.1 and 8.1.1.2 of the TR considered both beam and shell bending of the canister as it is seated in the HSM. The canister is considered to be a continuously loaded beam simply supported at three longitudinal locations. The maximum primary membrane stress is 0.21 ksi, which is substantially lower than the ASME Level A stress intensity allowable of 18.7 ksi. The canister was also checked for local shell bending by considering that the total dead weight was supported uniformly by the continuous T-section rails. The resulting local bending stress was 11.35 ksi. The result must be added to the other design basis stresses and then compared to the ASME Code allowables. Table 3.2. of this SER shows that the results are below the code allowables.

Desion Basis Internal Pressure The design basis interr.al pressure of 25 psig was calculated by assum-ing 100 percent cladding failure of the fuel rods inside the canister. The resulting release of rod fill gas and 25 percent of the fission gas will 0

equilibrate at a temperature of 429 F at an ambient temperature of 125 F.

This is a conservative assumption. The primary membrane stress in the canister body was 0.91 ksi, substantially below the allowable for Service Level A of 18.7 ksi. Secondary discontinuity stresses occurring at the juncture of the shell case and the inner pressure plate junction were 42

Table 3.2 Maximum Stresses for DSC for Normal Operating Loads load Stress (ksi) ASME Type DSC Design Design i Components Stress Dead Basis Basis Operation Combined Code Type Weight Pressure Temp Handling Loads Allowable Primary 0.21 0.91 NA .11 0.21 or Sm - Level A Membrane 1.02 18.7 Canister Local Primary NA 0.91 NA NA 0.91 1.5 Sm Shell Membrane Level A 28.05 Primary Membrane 11.35 + .21 - 9.01 + .91 - 20.9 3.0 32.45 or 3 Sm

& Secondary Bend 11.55 9.16 33.06 Level A 56.10 Primary Membrane NA NA NA NA >0 Sm 18.7 8 Cover Plates Primary Membrane NA 10.03 .67 3.31 14 1.5 Sm Bending Level A 28.05 Spacer Primary 1.62 NA NA NA 1.62 Sm - l Disc Membrane Level A 18.7 Actual P = 25 psig Hydraulic Load g=1 for Tamb - Ram 0

Tamb = 125 F 700 F 6200 lbs ASME Code Allowables taken at 4000 F

1 .

calculated using equations which require slope and displacement continuity at the juncture. NUTECH used equations for slope and displacement from l Roark (Ref. 12). The TR reported a combined membrane and secondary bending stress of 9.16 ksi for the DSC shell. The staff carried out an independent calculation based on a similar method suggested by the ASME B&PV Code (Ref.

13), and essentially duplicated the numbers found in the TR. This stress was combined with the stresses from other normal operating conditions and compared with the Code allowables and found acceptable. See Table 3.2.

, The cover plates were also analyzed for bending due to internal pressure. The primary bending stress was 10.03 ksi, smaller than the Code allowable of 28.05 ksi.  !

Desian Basis Operatina Temoerature 4

NUTECH performed two analyses to determine the magnitude of thermally induced stresses in the DSC shell. The bounding case involved the interface between the DSC shell and the spacer disc. NUTECH used STARDYNE to model the differential expansion of the spacer disc at an ambient temperature of '

! 700F. They conservatively considered zero gap between the shell and the disc. The temperature profile of the disc was determined by the heat trans-fer analysis discussed in Chapter 4 of this SER. The maximum membrane and secondary bending stress intensity was found to be 20.9 ksi. The staff performed an independent calculation to verify this stress level. Using a simplified method, the staff estimated a differential expansion between the spacer disc and the shell of .0058 inches. This produced a circumferential bending stress of 15.4 ksi in the shell at the location of the spacer disc. <

When combined with other stresses as shown in Table 3.2, the NUTECH stress of 33.06 ksi was found to be under the Code allowable of 56.1 ksi.

NUTECH's calculation for the design basis thermal loads were based on

, an ambient temperature of 700 F and not the accepted temperature of 1000 F.

The staff made a simplified calculation for the 125 F case (reported in Section 3.3.4.2.3 of this SER) and concluded that the thermal stresses did not increase at the higher temperature. Because the 1000 F case is bounded by the 70 F and the 1250 F cases, and because the coefficient of thermal i expansion is a continuous function between these temperatures, the staff concludes that there would not be an increase in thermal stress for the 44

i

.100 F case. The internal pressure due to a 100 F0 ambient case would also be higher, but again, the accident case of 125 F bounds the pressure stress as  ;

shown in Table 3.3. The ASME Code allowable would be reduced so that '

combined stresses would be compared against a lower allowable. The staff concludes that the accident case for 1250F will bound the design basis thermal case of 1000F.

Operational Handlina loads for DSC The operational handling load considered by NUTECH was the push / pull force of the hydraulic ram against the DSC steel cover plate. The coeffi-cient of friction of steel against steel was conservatively taken as 0.25 causing a push / pull force of 6200 pounds in the axial direction. The resulting stresses for the canister shell and the cover plate are low, as shown in Table 3.2.

Table 3.2 shows that the maximum calculated stresses for individual load cases are lower than the Code allowables. Appropriately combined stresses are also lower than the Code allowables.

Dead Weicht and Temperature loads for DSC Internals Section 8.1.1.3 of the TR covers the stress analysis for the internal parts of the DSC for normal operating conditions. The spacer discs were analyzed for dead weight loading by means of a finite element analysis. The analysis resulted in a bearing stress of 1.62/ksi, which is less than the allowable of 18.7 ksi. The axial growth of the fuel rods is not restricted by the DSC pressure plates, therefore no thermal stresses are induced in the fuel rods or the DSC due to axial expansion. The staff concurs with the TR that adequate axial clearance exists.

3.3.4.2.2 DSC Off-Normal Events Section 8.1.2 of the TR discussed one limiting off-normal event for the DSC, namely jamming of the DSC during loading or unloading. The basis for postulating this off-normal event is that design precautions taken to pre-vent axial sticking (such as beveling of cover plates and support rails, lubrication of sliding members, etc.) fail and the leading edge of the DSC 45

Table 3.3 DSC Enveloping Load Combination DSC Design Design ASME Component Stress Dead B. Press. B. Temp Accident NUTECH Allow or Location Weight (25 psig) 0 Type (70 F) Case Comb. Comb. - Level 5' Dron*

Shell Pr. Memb. .21 .91 --

15.52 16.64 16.64 44.88-D 0 Discont. Pr.M. + Bend --

9.01 + .91 --

39.19 48.35 48.35 64.4-D Pres. 39.7 Shell Pr. Memb. .21 1.45 -- --

1.66 --

16.2-A**

@ discont. Pr.M. + Bend --

<20.9 14.55 35.45 --

48.6-A**

@ Center Pr.M. + Bend. 11.35 + .21 1.45 20.9 33.91 --

48.6-A**

- 11.56

$; Axial Bind.

Shell Pr. Memb .21 .91 -- --

1.12 --

18.7-A+

@ Discont. Pr.M. + Bend --

9.16 <20.9 --

30.06 --

56.1-A+

@ Center Pr.M. + Bend 11.56 .91 20.9 12.96 46.33 --

56.1-A+

Seismic Shell Pr. Memb .21 .91 --

3.93 5.05 --

18.7-A+

0 Center Pr.M. + Bend 11.56 .91 20.9 9.58 + 3.93 46.88 --

56.1-A+

= 13.51 Bottom 5' Dron*

! Inner Pr. Memb -- -- --

.64 .64 .64 44.88-D Pr. Plate Pr.M. + Bend --

10.3 --

51.27 61.30 61.30 64.4-D Top 5' Dron*

Lead Pr. Memb -- -- --

5.52 5.52 5.52 44.88-D Plug Pr.M. + Bend --

10.3 --

51.27 61.30 61.30 64.4-D 5' Drop

• Suppt. Ring Pr. Memb -- -- --

15.52 15.52 15.52 44.88-D Lead Plug Pr.M. + Bend -- -- --

31.99 31.99 31.99 64.4-D

Table 3.3. DSC Enveloping load Combination (continued) l l

DSC Design Design ASME Component Stress Dead B. Press. B. Temp Accident NUTECH Allow or Location Type Weight (25 psig) (70 F) Case Comb. Comb. - Level 5' Drop

• Spacer Disc Pr. Memb 1.62 -- --

40.45 42.07 40.1 44.88-D 5' Droo*

Boral Tube Pr.M. + Bend -- -- --

2.65 2.65 2.65 64.4-0 5' Droo*

2" 0 Rods Comp -- -- --

5.6 5.6 5.6 22.-D f: 5' Droo*

1/4" Fillet Primary -- -- --

11.5 11.5 11.5 22.4-D Weld Full Pen. Primary -- -- --

25.98 25.98 25.98 29.2-D All drop accident allowable stresses correspond to Service Level D at 400 0F

• The drop accident is labeled as a "5' Drop" in this table. This height corresponds to using the IF-300 cask cited as an example of a DSC transfer cask in the TR and assuming that the IF-300 cask is drcpped onto an unyielding surface. Actually, the drop is defined as one resulting in a 48 g deceleration in the vertical orientation and a 34 9 deceleration in the horizontal orientation for the DSC.

• • Allowable stresses correspond to Service Level A at 650 F for this accident case

+ Allowable stresses correspond to Service level A at 400 0F for these cases

becomes jammed against an immovable object. The force of the hydraulic ram will increase to a cut-off value equal to the full weight of DSC which is 21300 pounds. This corresponds to a coefficient of friction of 1.0 and is four times larger than the design basis coefficient. Thus a moment will be induced in the DSC. The resulting stress was shown to be insignificant.

Only 0.74 ksi resulted compared to an allowable of 18.7 ksi for Level A.

A second, more severe condition was postulated which involves sticking of the DSC on the lip of the guide sleeve of the HSM. This could cause local bending of a ring section of the DSC. The resulting stresses were shown to be 12.96 ksi. NUTECH states that this is lower than the allowable for Service Level D. When combining the local bending stress due to stick-ing with the other stresses due to dead weight, pressure and temperature, the total stress is 46.33 ksi, still below the allowable of 64.4 ksi for ASME Service Level D (see Table 3.3). In fact, this value is less than

. Service Level A, which is 56.1 ksi. Because the combined stress is less than Service Level A, no plastic deformation of the DSC body will occur, and there need be no requirement that a licensee remove and inspect the DSC for damage following an off-normal occurrence, as is required for the drop accident.

3.3.4.2.3 DSC Accident Conditions Section 8.2 of the TR defines the accident conditions associated with the NUHOMS system. The accider.. conditions which were examined for the DSC were: (1) earthquake; (2) flood; (3) blockage of air inlet and exhaust; (4) accident internal pressure; (5) cask drop corresponding to 48 g vertically and 34 g horizontally; and (6) load combinations.

Seismic Condition 1

NUTECH considered the response of the DSC to a seismic event when it is resting in the two support rails. They first performed a rigid-body )

stability analysis to show no possibility of roll-out. They then calculated the natural frequency of two modes of vibration to show that they were justified in using a static stress analysis. Finally, they calculated the stresses induced in the DSC due to the earthquake event.

48

NUTECH calculated the magnitude of a moment causing the canister to roll out of its support and compared it to a stabilizing moment causing the canister to remain in its support. As discussed in Section 3.1, the staff used 0.35 g horizontal acceleration plus an upward acceleration of 0.17 g to determine whether the cask would roll out of its support assembly. NUTECH used 0.25 g instead of combining the two horizontal components. Even when using the more conservative approach of the staff, the result showed that the DSC would not roll out and that the safety factor was 1.47 instead of 1.92.

NUTECH calculated natural frequencies of the DSC in two modes to justify using a static analysis rather than a dynamic response analysis for the seismic event. The staff checked the frequency analysis for the shell ovalling and beam bending modes and found natural frequencies of 37.0 Hz and 44.3 Hz respectively. Both figures are higher than 33 Hz, the control point A of Tables 1 and 2 of the NRC Regulatory Guide 1.60 (Ref.14). Above 33 Hz structures have a dynamic amplification factor of 1.0. The staff accepts NUTECH's equivalent static analysis rather than a dynamic analysis.

For the stress analysis, NUTECH applied a dynamic factor of 2 to account for the effects of multimode excitation. Even so, the maximum beam bending stress intensity was only 3.93 ksi. The second type of loading that NUTECH considered was partial roll-out of the canister, i.e., the full effect of the .25 g inertial force times a dynamic amplification factor of 2 was assumed to be taken out by one of the T-section guide rails. This produced a local shell bending stress of 9.58 ksi. NUTECH conservatively summed the vertical and transverse bending stresses to produce 13.51 ksi, which is considerably less than the 56.1 ksi allowable for Service Level A.

NUH0MS has provided for axial retention of the DSC in the event of an axial seismic loading. The two retaining blocks will prevent the DSC from sliding into either end of the HSM. The primary compressive stress induced in the DSC due to a .35 g horizontal resultant applied axially to the shell is only 0.13 ksi, well below the Code allowable.

49

Flood Accident Since the flood accident condition is site specific, NUTECH calculated a capacity analysis for the DSC, i.e., they determined the maximum external pressure required to achieve various code allowable stresses, both primary and secondary. The smallest external pressure was then converted into a hydrostatic head, which would represent a maximum static water column.

Section 8.2.4.2 of the TR reports the height of that column to be 407 feet.

Thus the least depth that the DSC could be submerged, without any structural damage, is 407 feet. The staff find: that this adequately satisfies 10 CFR 72.72.

Blockaae of Air Inlet and Outlet and Accident Internal Pressure The worst thermal accident postulated for the NUH0MS is complete block-age of both inlets and both outlets on the hottest design day of 125 0F. The time period for the adiabatic heat-up was assumed to be two days. This time limit is consistent with NUTECH's proposed surveillance frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This particular event is also the cause of the accident pressure case, since the higher internal temperature causes the higher internal pressure.

The DSC canister surface temperature was calculated by the one-dimen-0 sional Schmidt graphical method to be 644 F (Ref.1, p. 8.2-50). Although NUTECH did not show what the thermal stresses were for this case, the staff made a simplified calculation similar to the one carried out to check the thermal stress in Section 3.3.4.2.1 of this SER. The differential expansion between the spacer disc and the cylinder for the higher temperature did not increase, hence the stress in the DSC shell did not increase. However, because the higher temperature causes an increase in internal pressure to 39.7 psig, the stresses for the accident pressure case must be combined with the accident temperature case. Also, because the canister material heated to 6440 F has a lower allowable stress level (S, - 16.2 ksi and 3.0 S, -

48.6 ksi), the combined stress must be compared to the lower allowable for Service Level A. This result was not presented by NUTECH, but the staff performed this step and concludes that the NUTECH design is below the Code allowable, as can be seen in Table 3.3.

50

Discussion of Cask Droo The cask drop accident is by far the most critical accident which NUTECH has postulated. It is so severe that NUTECH has resorted to the use of the higher ASME Level D Service limits in order to qualify their design.

The significance of using Service Level D instead of Service Level A is that the stress limit is higher for Level D which permits gross general deformations, with some consequent loss of dimensional stability and damage requiring repair. The damage may require removal of the component from service (Ref. 15). The staff agrees with the selection of Service Level D.

NUTECH has stipulated that, in the event of a drop accident, the DSC must be opened and the fuel assemblies inspected for possible damage following any drop accident.

NUTECH has postulated both vertical and horizontal drop orientations with deceleration levels of 48 g and 34 g, respectively. However, a verti-cal or corner drop is plausible only should the overhead crane and/or lifting yoke fail while lifting the cask out of the spent fuel pool or while lifting the cask from the floor of spent fuel pool building onto the cradle of the transportation trailer. Both of these operations are likely to be governed by 10 CFR 50.

A plausible drop accident, governed by 10 CFR 72, is the dropping of the cask from the trailer onto the road bed or pad of the HSM. Such a drop would necessarily be a horizontal drop or the more general case of tipped-end drop with subsequent slap-down of the higher end. Of these two plausible accidents, NUTECH has designed for the horizontal case. However, the TR states that by designing for a vertical and a horizontal drop, the slap-down case has been bounded (Ref 1, p. 8.2-29). This statement has not been demonstrated and indeed without resorting to a dynamic analysis, it would be difficult to demonstrate since the horizontal deceleration level of 34 g was based on the plastic crushing of energy absorbers placed underneath both ends of a horizontally oriented cask. However, a cask impacting sequentially first on one end and then another end in a near-horizontal orientation would almost certainly experience higher local loading of a i spacer disc located at the DSC and subjected to the second impact.

Parameters affecting the magnitude of this second impact include the cask  ;

rigid body quantities of mass, length, orientation angle at time of drop, 51 1

1 I

l

center of gravity height at time of drop, and angular velocity at time of drop.

Horizontal Droo Section 8.2.5.2 of the TR presents the analysis for the horizontal drop

. accident. The spacer discs and the boral tubes are the elements of the DSC which are affected by a horizontal drop. Because the DSC shell is supported along its entire length by a steel liner between the shell and the transfer cask, the shell itself is not appreciably loaded in this drop orientation.

NUTECH performed a STARDYNE finite element analysis of the spacer disc. The static load on each disc was determined by applying the weight of the 26 inch segment fuel assemblies and the boral tubes to the appropriate locations on the disc. The weights of 26 inch segments of the four two-inch diameter support rods were omitted. Even though the weight of the rods represents only 5% of the 1832 pounds acting on each spacer disc, the i maximum membrane stress shown in Table 8.2-7 of the TR is fairly close to the allowable stress for Service Level D. As a conservative check, the staff arbitrarily increased the maximum stress reported by NUTECH for this case by 5%. The resulting stress, when added to the dead weight stress, is 42.03 ksi, which is less than the 44.88 ksi allowable. (See Table 3.3 of this SER.)

The staff concurs with NUTECH's position that fuel can be retrieved following a horizontal drop limited to 34 g. (See Section 3.3.4.6.2 of this SER.)

The boral tubes were also analyzed for a horizontal drop accident.

Because the seven spacer discs are located to coincide approximately with the fuel assembly grid straps, only the self weight of the boral tubes were considered. This resulted in stress well below the allowable, as shown in Table 8.2-7 of the TR.

Vertical Droo NUTECH has analyzed the canister shell, bottom inner pressure plate, top lead plug, support plate for top lead plug, two-inch diameter rods, and several welds for the vertical drop case.

52

The shell was checked for compressive membrane stress level as well as for stability under buckling due to a 48 g load. The compressive stress was 15.57 ksi compared to an allowable of 44.88 for Level D. The critical buckling stress was 14 times higher than the actual compressive stress; therefore, no problem of stability exists. Shear stresses in the weld of the bottom outer cover were calculated to be 12.64 ksi. The weld stress intensity of twice the shear stress (25.28 ksi) is lower than the Level D allowable (29.2 ksi) accounting for a 65% efficiency factor.

In order to account for the geometry of the top portion of the DSC, which includes a welded ring supporting the lead plug, NUTECH performed a STARDYNE analysis. The maximum canister shell bending stress and discon-tinuity stress is 31.52 ksi. This is less than the allowable of 64.4 ksi.

The maximum bending stress intensity of the bottom cover of the lead plug was 52.27 ksi, also lower than the 64.4 ksi allowable. The fillet weld stress was calculated to be 11.5 ksi, approximately half of the allowable 22.44 ksi.

For the bottom of the canister, another STAR 0YNE model was used.

Canister wall membrane and discontinuity stresses were found to be 39.19 ksi. The maximua bending stress intensity of the inner bottom pressure plate was 39.7 ksi. Both of these stresses are lower than the 64.4 ksi allowable for Level D.

The two inch support rods were analyzed for compressive stress and buckling due to supporting six of the seven spacer discs plus self weight times 48 g. The compressive stress was found to be 5.6 ksi, lower than the allowable of 22 ksi according to ASME Appendices XVII and F rules. Also the critical buckling stress of 97.1 ksi is significantly higher than the actual stress.

The results of these analyses show that the NUTECH design is adequate for the two postulated drop accidents. Table 8.2-7 of the TR presents all of the results.

53

l Load Combinations I Table 8.2-9 of the TR reports the load combinations for the DSC. The stress intensities in the canister at various critical locations due to l

normal operating conditions were combined with stress intensities sustained l by the canister during the accident condition. NUTECH assumed that only one accident occurs at any given time. They also took advantage of the ASME rules governing Level D stress allowables by not adding any thermal stresses to the drop accident case. In order to verify that all cases were con-sidered, the staff prepared Table 3.3 to show the individual stresses due to various load combinations for the canister internals and the canister shell.

Based on the results shown in Table 8.2-9 of the TR and Table 3.3 of this SER, the staff concurs with NUTECH's conclusion that the design is satisfactory.

DSC Fatiaue Evaluation The requirement for structural integrity of the DSC can be met by satisfying two conditions. The first condition is met if the results of calculating primary plus secondary stress intensities are within code allowables. The second condition is met if the fatigue usage factor due to the various types of stress cycles is less than unity (Ref. 17). As discussed above, NUTECH has demonstrated ASME code compliance for the first of the two conditions.

In order to demonstrate that the cumulative usage factor is less than unity, NUTECH made several very conservative assumptions. For temperature l loads, the effect of seasonal ambient fluctuations was obtained by assuming i an extreme fiber stress alternating 50 cycles from 20.9 to zero ksi, the l maximum derived thermal stress. This number of cycles corresponds to the

, 50-year design life of the system. This stress occurred in the spacer disc.

For daily temperature fluctuation, NUTECH assumed a mean stress of 20.9 ksi with an alternating stress of 5.35 ksi. The alternating stress was deter-7 mined by taking the ratio of the maximum daily temperature fluctuation of 450 F to the AT of 175 0F times the mean stress. The 1750 F AT is actually larger than the NUTECH worst case design envelope of -400 F to 125 F. The number of daily cycles was 18250 for 50 years. The internal pressure fluc-tuation caused by seasonal temperature change was evaluated by 19.4 psig at 54

-400 F and 39.7 psig for the flow blockage accident. The maximum disconti-nuity stress of the DSC shell was 14.55 ksi at 39.7 psig and 7.12 ksi at 19.4 psig. The number of cycles for this history was 50. The daily pressure fluctuation was assumed to vary from zero to 25 psig for 18250 cycles. The stress intensity associated with this was 9.16 ksi, also located in the DSC shell. One seismic event was assumed to occur during the

. life of the canister. NUTECH conservatively assumed that the load used in the equivalent static approach was repeated for 100 cycles.

NUTECH applied a fatigue stress intensification factor of four to account for geometric discontinuities. They also considered the maximum stresses for any location of all parts of the canister and internals which is much more conservative than merely considering stresses acting at one location of one part as required by the ASME Code. Section 8.2.10 and Figures 8.2-13 and 8.2-14 of the TR show that the cumulative usage factor is 0.21, which is less than the maximum allowable usage factor of 1.0.

The Staff has reviewed the structural analysis of the DSC and the DSC support and finds that:

1. The DSC and intervals are in compliance with the ASME Code rules for Service Level A stresses for normal operating conditions and off-normal events.

2. The DSC and internals are in compliance with the ASME Code rules for Service Level D stresses for the drop accident conditions postulated by NUTECH.

3. The DSC support members are in compliance with the ASME Code rules for Service Level A stresses for all cases analyzed.

3.3.4.3 DSC Support 3.3.4.3.1 DSC Support Normal Operating Conditions The support structures for the DSC were analyzed for normal operating loads including: (1) dead weight; (2) temperature loads; (3) operational handling loads.

55

Dead Weiaht Section 8.1.1.4 of the TR deals with the support assembly loads analysis for normal conditions. A linear finite element beam model was used to calculate the axial, shear and bending stress for dead weight and opera-tional handling loads. Table 8.1-8 of the TR shows that all load cases produce member stresses far below ASME Code allowable stresses for dead weight and normal operation handling loads. The allowables are taken at 3000 F and are in accordance with Service Level A.

Temoerature loads The staff concurs with NUTECH that no thermal stresses will be induced in the support system because slotted bolt holes are used and bolts are torqued in accordance with the requirements of the AISC manual. Thus friction in the bolted assembly can be overcome by the thermal expansion of the rails during normal heat up conditions.

Ooerational Handlina loads The normal operational handling load considered was a 6200 pound load applied axially to the guidt. rails. This models the normal condition of loading the DSC into the HSM with a coefficient of friction of 0.25.

3.3.4.3.2 DSC Support Off-Normal Events Section 8.1.1.4 of the TR discusses off-normal events as they relate to the support assembly. The off-normal event considered was a jammed condition where the hydraulic ram exerts a force of 21,300 pounds and at same time one half of the canister weight was assumed to act on the guide as a concen-trated force in its mid span. The analysis was performed with STARDYNE.

Table 8.1-8 of the TR reported the results, and showed that the calculated stresses are lower than the allowable stresses for Service Level A.

56

3.3.4.3.3 DSC Support Accident Conditions The support assembly was first analyzed for a seismic event to establish what the natural frequency of the support was. The lowest frequency was found to be 54.1 Hz, higher than the 33 Hz cut off for dynamic amplification. Based on this, NUTECH was justified in using a static analysis to examine the support assembly during a seismic event. NUTECH conservatively applied a non-required dynamic amplification factor of 2 for both vertical and horizontal seismic directions. For the vertical load condition, NUTECH multiplied the results of the STARDYNE static analysis by 0.34. The resulting seismic bending, axial and shear stresses were combined with dead load stresses and are significantly below code allowables as shown in Table 8.2-11 of the TR. Horizontal stresses were also analyzed using the STARDYNE dead weight analysis by applying a horizontal acceleration of 0.50g (including a dynamic factor of 2) through the center of gravity of the DSC.

The resulting force was resisted by one of the "T" rails. The maximum bending shear and axial stresses were combined with those of the dead weight analysis and were shown to be lower than the allowable in Table 8.2-11.

3.3.4.3.4 DSC Support End Connection Design The analysis of the DSC support end connection (i.e., corbel, bearing plate and bolts) is presented in Section 8.2.10 of the TR. This analysis includes consideration of dead weight, jammed condition handling loads, friction handling loads, and seismic loads. The staff review concurs with the analysis results.

3.3.4.4 Transfer Cask and Vehicle The type of cask and transfer vehicle are both site specific and should therefore be included in the applicant's license request. As discussed in Section 3.3.4.2.3 of this SER, it is incumbent on the part of a license applicant to show that the particular cask and transfer vehicle do not negate the validity of NUTECH's design criteria for a drop of the DSC. This can be shown if no accident can be postulated which exceeds a 48 g vertical deceleration and a 34 g horizontal deceleration.  ;

57

3.3.4.5 HSM Loading and Unloading Equipment 3.3.4.5.1 Normal Operatina Conditions l The most critical operational handling load which involves a site-specific hydraulic ram, transfer trailer, and optical alignment system is the sliding of the DSC into the HSM. Because the DSC and DSC support system are important to safety, they were analyzed by NUTECH and are evaluated in Sections 3.3.4.2 and 3.3.4.3 of this SER. The other equipment involved is l site specific and is not subject to evaluation in the TR.

l 3.3.4.5.2 Off-Normal Conditions ,

The off-normal events that involve HSM loading and unloading equipment are primarily site-specific. The axial sticking and/or binding of the axial body of the DSC as it is either loaded or unloaded in the HSM were discussed in sections 3.3.4.2 and 3.3.4.3 of this SER. Only the DSC and DSC support system were evaluated. No other equipment is defined as being important to safety.

3.3.4.5.3 Accident Conditions No event was identified as an accident condition which involved loading l or unloading equipment.

l l 3.3.4.6 Fuel Assemblies and Rods i 10 CFR 72.76 discusses criteria for decomissioning of the ISFSI.

Implicit in either decomissioning or in inspecting for possible damage following a drop accident is the ability of operators to remove the fuel assemblies from the DSC.

3.3.4.6.1 Normal Operatino Conditions Decomissioning is the only time when it would be necessary to remove the fuel assemblies due to normal operating conditions. A possible problem that could be postulated as a result of long term (50 years) storage of spent fuel in the horizontal rather than vertical orientation is longitudi-58

nal sagging of fuel rods due to creep, such that the fuel assemblies could not be removed from the DSC basket assembly. NUTECH did not resolve this problem either by referencing applicable literature or by analytical techni-ques (Ref. 18).

The staff examined creep as a mechanism for preventing retrieval of the

-fuel rods, and concluded that the bending stress in the rods causing creep will be self limiting should contact of the rod with the boral support tube be made. Thus the creep rate will be reduced. Furthermore there does not appear to be a creep mechanism whereby plastica 11y deformed rods could cause jamming of the fuel assembly inside the boral guide tube, considering that a 0.324 inch gap exists between the fuel rod and the spacer cell.

However, the examination performed by the staff was not exhaustive.

Should such difficulties be encountered in attempting to retrieve the fuel assemblies, the reviewers believe that the entire DSC basket assembly could be removed, permitting retrieval of the fuel assemblies.

3.3.4.6.2 Accident Conditions Should the DSC be accidentally dropped from any height, the TR speci-fies that it be removed from service and inspected for possible damage.

NUTECH addressed the potential problem of fuel assembly retrieval due to a horizontal drop in Section 8.2.5.2 of the TR. The staff concurs with NUTECH's response that plastic deformations of 0.0533 inch for the spacer disc and boral tubes could not pose a fuel retrievability problem, considering the 0.324 inch gap between the fuel and the boral tube (Ref. 19).

59

4.0 THERMAL EVALUATION 4.1 Summary and Conclusions The NUHOMS thermal design conforms to appropriate sections of 10 CFR 72 and is acceptable subject to the following condition:

1. Thermal response for the inlets blocked case is not to be used to determine loadings for any load combination.* Use of the ' inlets blocked analysis results' to determine thermal loading will neces-sitate testing to confirm adequacy of natural circulation cooling for this case.

4.2 Description of Review 4.2.1 Applicable Parts of 10 CFR 72 The thermal analysis was reviewed for conformance to 10 CFR 72 Subpart F. For normal, off-normal and accident conditions,10 CFR 72.72(h) requires that the fuel cladding be protected against degradation and gross rupture.

Sections 10 CFR 72.72(b,c) require that the system design provide protection against environmental conditions, natural phenomena and fires.

Since Section 10.3.3.1 specifies visual inspection of the HSM air inlets every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and removal of the blockage within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the inlets will not be blocked for longer than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The inlets blocked case is therefore enveloped by the accident case of all inlets and exits blocked for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Use of the accident thermal loads for HSM load combination 7 would preclude the necessity to relying on the questionable analysis with only the inlets blocked.

60

4.2.2 Review Procedure i 4.2.2.1 Design Description The NUHOMS system provides for the horizontal storage of irradiated fuel in a DSC which is placed in a concrete HSM. Decay heat is removed from the fuel by conduction and radiation within the DSC and by convection and radiation from the surface of the DSC. Natural circulation flow of air through the HSM and conduction of heat through concrete provide the mechanisms of heat removal from the HSM.

Spent fuel assemblies are loaded into the DSC while it is inside a transfer cask in the fuel pool at the reactor site. The transfer cask containing the loaded DSC is removed from the pool, dried, purged with helium and sealed. The DSC is then placed in a transfer cask and moved to the HSM. The DSC is pushed into the HSM by a horizontal hydraulic ram.

The DSC is constructed from stainless steel pipe with an OD of 37.0 inches and a wall thickness of 0.5 inch. The length of the pipe is depen-dent upon the type of fuel being stored. Within the DSC, there is a basket consisting of seven square cells of stainless steel clad boral for criti-cality control. An intact PWR spent fuel assembly is loaded into each cell, resulting in a capacity of seven assemblies per DSC. Spacer disks are used for structural support. The DSC has double seal welds at each end and rests on two steel rails when placed in the HSM.

The HSM is constructed from reinforced concrete, carbon steel and stainless steel. Passageways for air flow through the HSM are designed to minimize the escape of radiation from the HSM. Decay heat from the spent fuel assemblies within the DSC is removed from the DSC by natural draft convection and radiation. Air enters at the bottom of the HSM, flows around the canister, and exits through the flow channels in the top shield slab.

Heat is also radiated from the DSC to the inner surface of the HSM walls where again, natural convection air flow removes the heat. Some heat is also removed by conduction through the concrete.

61

The NUHOMS system utilizes a transfer cask, transporter, skid and horizontal hydraulic ram. There is no specific transfer cask that is used in the NUHOMS system. The transporter, skid and horizontal hydraulic ram are not affected by the thermal analysis. During vacuum drying of the fuel in the DSC, heat is removed by conduction through the transfer cask.

-4.2.2.2 Acceptance Criteria Potential damage mechanisms to spent fuel assemblies stored in inert gases have been addressed in Reference 20. Based on the evaluation reported in this reference, no degradation of the cladding is expected for fuel assemblies stored in inert gases at 380 0C. Therefore, a maximum temperature of 380 C is the acceptance criteria for normal operating conditions.

Meeting this acceptance criteria assures that the requirements of 10 CFR 72.72(h) are satisfied.

Reference 20 also establishes that no rods have failed in inert gas 0

exposures up to 570 C, and rods forced to failure required temperatures from 0 0 765 C to 800 C to produce ruptures. An accident temperature limit of 570 C 0 is the acceptance criteria for accidents based on the above evaluation.

ACI-349-80 specifies temperature limits for concrete structures. The thermal analysis review addresses the correctness of the reported concrete temperatures, and also the thermal input for stress analysis. Acceptability of the concrete temperatures is addressed in Section 3.3.2.

4.2.2.3 Review Method The thermal analysis was reviewed for completeness, applicability of the methods used, adequacy of the key assumptions and correct application of -

the methods. Thermal analysis was performed primarily with the HEATING-6 (Ref. 21) computer program. HEATING-6 is a part of the Oak Ridge National Laboratory SCALE package and is an industry standard code for thermal analy-sis. Representative input and output was reviewed to establish that the code use was appropriate and that the results were reasonable. Independent calculations were performed to check other portions of the analysis which did not use the HEATING-6 code, e.g. maximum internal pressure, natural circulation flow, Wooten-Epstein correlation.

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4.2.2.4 Key Design Information and Assumptions The key assumptions made in the thermal analysis are listed below.

1. The total heat generation rate for each fuel assembly is less than or equal to 1000 watts. This value is based on JRIGEN calcula- l tions and data published in the available literature. All heat is assumed to be generated in the fuel region.

2. Each dry storage canister contains seven intact PWR assemblies.

3. A factor of 1.08 to account for axial power peaking in the fuel during operation was assumed.

4.3 Discussion of Results 4.3.1 Analytical Methods Used by NUTECH In the NUHOMS topical report, thermal analyses were calculated for the ,

horizontal storage module, the dry shielded canister in the horizontal storage module and the dry shielded canister in the transfer cask. The HEATING-6 computer program was used to perform the major portion of the thermal analysis. HEATING-6 solves steady state and/or transient heat conduction problems in one, two or three dimensional Cartesian or cylindrical coordinates. It is an industry standard computer code for thermal analysis.

Air temperatures within the HSM were first established by a natural circulation cooling analysis. Steady-state natural circulation flow will occur when the buoyancy forces are balanced by friction and form loss forces. Flow areas and loss factors were designed to allow sufficient flow to maintain the desired temperature difference between the inlet and outlet air temperature.

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l i

a Thermal analysis of the horizontal storage module is performed to obtain the temperature of the inside surface of the dry, shielded canister and the temperature distribution of the module concrete, given a heat flux across the canister surface corresponding to the spent fuel heat generation  !

rate. Once this temperature is established, detailed analysis of the  ;

temperature distribution within the canister is done. A thermal analysis of the canister within a hypothetical transfer cask is done to determine the peak fuel clad temperature during the vacuum drying operation.

A two dimensional Cartesian model is used to represent the horizontal storage module in HEATING-6. The module is assumed to be infinitely long.

The plane of the model is at the location of the peak heat generation rate.

Only one-half of the module is modeled in HEATING-6, since symmetry exists about the vertical centerline. The model includes the 3.5 ft. thick concrete ceiling, the 1 ft thick concrete side walls and the floor. The external surfaces of the side walls and floor are assumed to be adiabatic.

An internal module in the 2 x 4 area is therefore represented by this model.

End modules have a 3.5 foot thick side wall. The maximum thermal gradient through the concrete will occur in the 3.5 foot thick top slab since the interior HSM wall temperature is highest for this slab. Side wall and end wall gradients will be smaller. Maximum thermal gradients are used to determine thermal loads on the HSM.

The DSC located within the module is modelled as a cylindrical shell represented as a series of 20 small rectangular slabs. The total surface area of these slabs is equal to the surface area of the canister. Heat transfer by convection and radiation is considered in the air gap between the canister and the interior surface of the module. Convection heat transfer at the outer surface of the module ceiling is included, as is solar loading on the outer surface of the module ceiling. The heat source con-sists of seven PWR assemblies, each with an assumed average heat generation rate of 1000 watts. An axial peaking factor of 1.08 is assumed. This heat generation rate is selected based on consideration of difference burnups and decay times.

Temperatures within the DSC are determined using a second HEATING 6 model. A two dimensional Cartesian model is used to represent the dry, shleided canister and the internal helium, boral clad stainless steel 64

sleeves and fuel regions. All of the heat is conservatively assumed to be generated in the fuel regions. The regions representing the DSC wall are at fixed teteperatures determined from the HSM HEATING 6 analysis. An effective thermal conductivity was determined for the fuel regions using the Wooten-Epstein correlation which has been previously used for transport cask thermal analysis. Convective heat transfer in the helium was accounted for by use of an effective thermal conductivity 60% greater than the helium thermal conductivity.

Each of the te HEATING-6 models was used to determine temperature distribution for four normal operation cases, an off-normal case of both air inlets blocked and an accident case of all inlets and exits blocked.

4.3.2 HSM and Internals 4.3.2.1 Nornal Operating Conditions A total cf four cases were considered for normal operating conditions based on the teaperature of the air at the inlet of the module. These are:

0 (1) entering air at -40 F representing " severe winter conditions,"

0 (2) entering air at 70 F representing " normal conditions," (3) entering air 0

at 100 F representing " maximum normal conditions" and (4) entering air at 1250 F representing " severe summer conditions." For design calculations at normal conditions, " maximum normal conditions" of 100 0F ambient temperature is acceptable. Use of 70 F 0 ambient temperature for normal conditions is not acceptable since the ambient temperatures at many sites can be consistently above this temperature for extended periods greater than the time constant of the HSM. For these cases the method of determining air temperature within the HSM is appropriate. While the values of loss coefficients used in the analysis appear to be somewhat low, satisfaction of the limiting condition for operation of a 100 F0 maximum air temperature rise on exit from the HSM, gives a high degree of assurance that adequate cooling is achieved.

The HEATING-6 input and output for the " severe summer condition" of 0

125 F entering air was reviewed in detail. No problems were found in the input and the results are reasonable in terms of order of magnitude and trends. Values used for solar radiation on the top shield block, axial peaking, convection heat transfer coefficients and material properties can 65

be characterized as best estimate values. Approximation of the cylindrical DSC as a series of rectangular regions is such that the key parameter, namely, surface heat transfer area, is preserved. Trends in the results among the various cases were also reviewed, and found to be reasonable.

Temperatures of HSM internals such as bolts embedded in concrete and DSC support rails were determined by hand calculations from the HEATING-6 results. The values presented are reasonabic. Use of these results to determine the magnitude of thermal cycles for the fatigue analysis is also acceptable.

4.3.2.2 Off-Normal Conditions The off-normal condition considered was the total blockage of both air inlets. In this case, cooling air must enter through one of the exits and leave through the other exit. There is no definitive mechanism which operates to select one or the other of the exits as the one which will support inflow. A slight shifting of the heat flux peak and the presence of the internal shield block at one end do not appear sufficient to assure a stable steady state air flow. Determination of loss factors is at best an ,

inexact science for this ccmplex geometry, and the assumptions regarding flow direction and steady-flow for the inlets blocked case are open to question. Therefore, thermal load analysis using input from this case is unacceptable.

• Since the inlets will be inspected for blockage every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> according to the surveillance requirements in Section 10.3.3.1, the heating period can therefore not exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> plus up to another 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to ,

clear the inlets. Thus the results of this case are bounded by the accident condition discussed in the next section.

4.3.2.3 Accident Conditions The total blockage of all air inlets and exits was analyzed as the -

accident case. Adiabatic heatup of the various components was assumed, with the HSM providing the slowest heatup rate. Adiabatic heating starting at '

the 125 F inlet temperature condition is an appropriate assumption for this 66 1

case, and the resulting temperatures are reasonable and acceptable for use in the thermal loads analysis (see Table 8.1-12, page 8.1-76 of the TR). The Schmidt graphical method used for the analysis is an appropriate tool for cal-culating the adiabatic heatup rate. Since it is assumed that the blockage will be cleared within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, heatup was calculated over this period.

4.3.3 DSC and Internals 4.3.3.1 Normal Operating Conditions The four normal operating conditions were also analyzed for the DSC and internals, namely, 1) entering air at -40"F representing " severe winter condi-tions", 2) entering air at 70 f representing " normal conditions", 3) entering air at 100 F representing " maximum normal conditions" and 4) entering air at 125*F representing " severe summer conditions". HEALING-6 input and output for the 125'F ambient air case was examined in detail and found to be correct.

Trends and magnitude of the resulting temperature distribut. ions are also reason-able. ' Maximum fuel cladding temperatures were calculated for the four cases and found to be less than the 380 C limit in all cases.

Results of these calculations were used as input to another HEATING-6 cal-

'culation of the temperature distribution in the spacer disc. The resulting spacer disc temperatures are acceptable for use in thermal stress analysis, as are temperatures obtained from the DSC HEATING-6 analysis.

4.3.3.2 Off-Normal Conditions As discussed in Section 4.3.2.2, the natural circulation cooling analysis for this case is not accepted. Since the results of the DSC analysis depend upon results of the HSM calculation, their use is not accepted. This case is bounded

-by the accident case discussed below.

4.3.3.3 Accident Conditions Temperature distribution within the DSC was determined for the case of all air inlets and exits blocked for a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period. A steady-state 67

temperature distribution was assumed within the DSC, since its heatup rate is faster than that of the HSM. The resulting temperature distribution is acceptable for use in determining thermal loads. Maximum fuel cladding temperature was calculated to be below the 5700 C accident Ilmit.

4.3.4 Transfer Cask and Fuel In this case, the inside surface temperature of the transfer cask was determined by calculating the steady-state temperature distribution through the cask which was modeled as a series of cylindrical annular regions. The surface temperature of the DSC was then determined from the conduction, convection and radiation heat losses from the canister to the cask. The most limiting condition of vacuum drying of the DSC was analyzed. This case is limiting due to the lower effective thermal conductivity within the evacuated DSC. In this case, the maximum fuel cladding temperature was just below the 3800 C limit for normal operations. When the DSC is not evacuated, the maximum temperature will be significantly Icwer due to the higher effective thermal conductivity within the DSC. Off normal and accident thermal conditions were not considered for the transfer cask where the fuel will be resident for only a relatively short period.

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! 5.0 CONFINEMENT BARRIERS AND SYSTEMS 5.1 Sumary and Conclusion s

The staff has reviewed the features of the NUH0MS design which provide confinemcat of radioactive material and, specifically, protection of the fuel rod cladding. The review was directed at two aspects of the design:

(1) the mechanical integrity of the DSC and (2) the long term behavior of cladding in an inert atmosphere.

As a result of this review the staff concludes that the NUHOMS design conforms to applicable parts of 10 CFR 72.72(h). Confinement is assured by '

a radiographic inspection of the longitudinal full penetration weld, ultra ,

sonic inspection of the two welds for the bottom plug, and helium leak testing and dye penetrant testing of the welds for the top lead plug and top plate, respectively. The acceptance leak rate.for helium leak testing is less th.an 10-6 atm - cc/sec. The less rigorous dye test procedure used for it.e top end plate can be considered acceptable due to the helium leak testing of the inner weld, and due to the fact that two seals are used instead of one, as for the longitudinal weld. Radiographic inspection of the top plug welds is not feasible due to the fact that irradiated fuel will a? ready be installed before the tests can be made.

5.2 Description of Review 5.2.1 Applicable Parts of 10 CFR 72 Paragraph (1) of Section 72.72(h) of 10 CFR is pertinent to storage of spent fual in NUH9MS. It requires that "the fuel cladding shall be pectorted against degradation and gross ruptures." The remaining paragraphs of that section, which relate to storage of fuel in water and to off-gas and ventilation systec, are not applicable to this review.

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5.2.2 Review Procedure 5.2.2.1 Design Description The NUHOMS design provides protection of the fuel cladding by storing fuel assemblies in an inert atmosphere of helium. The helium atmosphere is first established after the fuel is loaded into the DSC. The loaded DSC is welded closed, hydrostatically tested at 50 psig for 10 minutes, drained of water, and evacuated. A vacuum is held until no more than an estimated one gram of water remains. Then the DSC is back-filled with helium to a pressure of 1.5 atmospheres for purposes of helium leak-testing of the primary weld.

After the end weld is checked for leaks, the DSC is again evacuated and backfilled with helium at 14.5 psia. The evacuation lines are sealed and the top end cover is welded to the DSC. The field welds and the shop welds on the bottom and along the longitudinal seam are expected to maintain the internal helium atmosphere intact for the full time of storage of the DSC in the HSM. No device (e.g. gauge) is made part of the system for verifying the maintenance of the helium atmosphere.

5.2.2.2 Acceptance Criteria The design will be found acceptable if the TR shows that (1) there is a high likelihood that the DSC internal helium atmosphere will remain intact and (2) there is no long term cladding degradation mechanism in a helium atmosphere which could cause significant degradation or gross ruptures.

5.2.2.3 Review Method The review of the TR was directed at two aspects of the design: (1) the mechanical integrity of the DSC and (2) the long term behavior of cladding in an inert atmosphere.

The staff reviawed DSC integrity from the point of view of weld quality and inspections, adequacy of leak check methods on welds, other leakage paths, and long term helium migration. Reviewers also checked the calculated stresses in the DSC under normal, off-normal and accident 70

conditions in order to verify that they were in the acceptable range.

Cyclic fatigue of the DSC was also analyzed.

The staff reviewed cladding degradation by reviewing literature in order to identify known and postulated nechanisms of gross failure of fuel in inert atmospheres. Based on the literature search, calculations were performed of postulated failures by the mechanism of diffusion controlled cavity growth using a conservative set of assumptions. This was the only failure mechanism considered likely under the NUHOMS storage conditions, 5.2.2.4 Key Assumptions Assumptions made in review of the TR regarding confinement systems are: ,

1 1

1. The diffusion rate of helium through the DSC is no greater than 10-8 g-moles / year at nominal design conditions and as much as 10-5 g-moles / year at accident conditions, as stated by the applicant.

1

2. The values used for various properties of the zircaloy cladding in the analysis of diffusion controlled cavity growth (DCCG) and the DCCG mathematical model lead to a very conservative estimate of degradation (see Appendix A).

3. The fuel cladding is protected at steady state temperatures of up i to 3800 C and on short term transients up to 570 0 C if in an inert atmosphere.

5.3 Discussion of Results 5.3.1 DSC Integrity In the review of the structural analysis of the DSC, the staff has found that the design is acceptable.

The commitment to design and fabricate the DSC's bottom and longitudinal welds to the ASME Boiler and Pressure Vessel Codes provides reasonable assurance of leak-tightness at these locations.

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The top end plate welds are made in the field. The TR states that end plate welds are either radiographed, ultrasonically tested, or tested by dye penetrant "to the same standards". The dye penetrant method of testing reveals information about the weld surface only, hence a weld tested by this method does not meet the same standards as the radiographic method.

Furthermore, the top end welds cannot be radiographed because irradiated fuel will be in place before these tests can be performed. However, the staff finds that this procedure is acceptable because the primary welds are first leak-tested by a helium detector and because the two top end plates represent a double seal.

Since the DSC contains no penetrations for sampling or gauges, there are no diffusion or leakage paths for helium other than the welds and the primary metal. Presuming the weld integrity to be equivalent to that of the parent metal, the staff also concludes that diffusion is not a potential mechanism to permit escape of helium and ingress of oxygen.

The staff concludes that DSC design and fabrication methods will result in a high likelihood that the internal DSC helium atmosphere will remain intact over its storage lifetime.

5.3.2 Long Term Fuel Rod Behavior Predictions of the DCCG phenomena are made for fuel stored in NUHOMS at normal operating conditions in Appendix A to this SER. As shown in Appendix A, fuel rod degradation postulated by this mechanism is insignificant. For this analysis the constant ambient temperature of 700F (21 0C) used by the applicant was accepted. While ambient temperatures above and below 700F (210 'C) would be experienced over the years of storage, this ambient temperature is a reasonably conservative value for the assessing of long-term damage to the fuel cladding attributable to DCCG for most of the continental United States.

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6.0 SHIELDING EVALUATION 6.1 Summary and Conclusions The NUHOMS shielding design conforms to the ALARA requirements of 10 CFR 72 and to acceptable shielding methods ana practices. The staff concludes, based on the analysis presented in the Topicai Report, that the shielding is designed to ensure that surface dose rates satisfy the criteria established in the Topical Report subject to the following conditions:

1. No more than seven (7) fuel assemblies meeting the specifications discussed in Chapter 12 of this SER are contained in a DSC, or

2. The maximum neutron source strength per canister is 9.98x10 8 neutrons per second, and the maximum gamma-ray source strength is 1.67x10 16 MeV per second.

6.2 Description of Review 6.2.1 Applicable Parts of 10 CFR 72 The applicable part of 10 CFR 72 regarding the shielding evaluation of the NUTECH Horizontal Modular Storage System is the requirement of 10 CFR Part 72.3 related to ensuring that occupational exposures to radiation are as low as reasonably achievable.

6.2.2 Review Procedure  ;

6.2.2.1 Design Description The principal design criterion for the NUHOMS module is to provide an I average external surface dose (gamma and neutron) of less than 20 mrem /hr.

In addition, the design criteria at the side of the cask and at the DSC top lead plug are 200 mrem /hr and 100 mrem /hr, respectively.

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6.2.2.2 Acceptance Criteria The shielding design is acceptable if the shielding evaluation results provide reasonable assurance that these design criteria are satisfied in the NUH0MS design.

-6.2.2.3 Review Method The shielding analysis in the Topical Report was reviewed. Independent or confirmatory calculations were not performed. Rather, an assessment of the appropriateness of the shielding methods was made and checks of the computer input data were performed. The results were also evaluated for sel f-consistency.

6.2.2.4 Key Assumptions and Computer Codes The major assumption is the specification for fuel to be stored, which is discussed in Section 6.3.1 below.

Three computer codes were used in the shieiding analysis reported in the Topical Report. ANISN, a one-dimensional discrete ordinates code, was used to estimate neutron and gamma-ray dose rates at the outer HSM wall, centerline of the DSC shield plug, and outside the loaded transfer cask.

DOT-IV, a two-dimensional discrete ordinates code, was used to estimate the neutron and gamma-ray dose rates for the DSC end plugs, the HSM roof, the HSM air outlet duct, and the DSC-transfer cask annular gap. QADMOD-G, a three-dimensional point kernel code, was used to determine the primary gamma-ray dose rates for the outer HSM wall, the DSC end plugs, the DSC-cask annular gap, and the HSM air vent penetrations. Manual albedo calculations were used in conjunction with the QADMOD and ANISN results to estimate the upper bound for the reflected gamma-ray dose rates for primary gamma-ray streaming through the air vent penetrations and the DSC-cask annular gap.  ;

The cross sections used in the shielding analysis were the Bugle-80 l multigroup library which contains 47 neutron groups and 20 gamma-ray groups.

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6.3 Discussion of Results 6.3.1 Source Specification The neutron and gama radiation sources include the design basis irradiated fuel and activated portions of the fuel assemblies. The shield-ing analysis includes both primary neutrons and gama-rays from these sources, as well as secondary gamma-rays from interactions of neutrons with the DSC and shielding materials.

The shielding is designed for a neutron source strength of 9.98x10 8 neutrons per second and a gama source strength of 1.67x10 16 MeV per second.

Any combination of fuel irradiation time, burnup, specific power, enrich-ment, and postirradiation time which gives neutron and gama source strengths no greater than these design sources are acceptable for storage.

The design basis is derived from a burnup analysis of PWR fuel subjected to 880 full power days at an average specific power of 37.5 MW/MTHM resulting in an average fuel burnup of 33,000 mwd /MTHM. The analysis considered an initial enrichment of 3.2 weight percent 2350 , and a postirradiation time of 5 years. Fuel which meets these criteria are bounded by the neutron and gamma-ray sources used in the shielding analysis. Any other combination of irradiation time, burnup, specific power, enrichment, and postirradiation time is also acceptable from a radiological protection standpoint provided the gamma and neutron source strength criteria are satisfied.

The neutron and gamma source energy spectrum used for the shielding analysis were derived from the ORIGEN burnup analysis and reported in Tables 7.2-1 and 7.2-2 of the Topical Report, respectively.

6.3.2 D5C - Cask Gap The dose rate in the DSC-Cask annular gap was evaluated with a combination of DOT-IV and a manual albedo method described in the Topical Report. The size of the annular gap depends on the specific transfer cask ,

that is used and is therefore site-specific. The analysis presented in the Topical Report assumed a 4-inch thick shell of depleted uranium with a 0.5 inch steel inner lining and a 1.5-inch steel outer shell. The annular gap was assumed to be 0.25 in. The dose rate reported for the DSC-cask gap is 75

282.64 mrem /hr. This dose rate assumes that the gap is filled with water.

Without water the dose rate is estimated to be 5 rem /hr; however, no person-nel access at the gap location is required during this condition.

Because the transfer cask design and consequently the size of the annular gap is site-specific, a determination will be necessary on a site-

. specific basis that the gap analysis presented in the Topical Report is bounding.

6.3.3 Transfer Cask The dose rate reported for the transfer cask surface is 5.6 mrem /hr.

This dose rate is based on a one-dimensional discrete ordinates model which has been determined to be a suitable approximation of the actual geometry.

However, the shielding effectiveness of the transfer cask is the dominant factor for the surface dose. Because the selection of the transfer cask is site-specific, a determination will be necessary on a site-specific basis to determine whether the transfer cask shielding analysis presented in the Topical Report is a bounding analysis.

6.3.4 HSM The dose rates reported for the HSM are based on a combinat s' of one-and two-dimensional discrete ordinates analyses. A manual albedt ;reatment of the streaming through the penetrations was also part of the analysis method. For most of the important dose points, dose rates have been esti-mated by more than one method. The comparisons are generally self-consis-  ;

tent, leading to an acceptable assessment of the adequacy of the reported  ;

results. l The dose rate reported for the HSM walls or roof is 2.84 mrem /hr. The dose rate reported at the air outlet is 11.61 mrem /hr with the shield cap in j place and 496.64 mrem /hr with the shield cap removed. The dose rate at the air inlets is reported as 51.72 mrem /hr.

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The dose rates at the HSM loading door are reported to be 2.29 mrem /hr at the center of the surface of the steel door, and 5.63 mres/hr with the steel door removed. The dose rate at a position 4.5 ft in front of the door is reported as 0.69 mrem /hr.

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7.0 CRITICALITY EVALUATION

7.1 Summary and Conclusions The largest effective multiplication factor (k-effective) reported in the TR is 0.8884. This value was determined for a bounding configuration of the NUHOMS design. Thus, it is concluded, based on the analysis presented in the TR, that the NUH0MS system is designed to remain in a subcritical configuration and to prevent a criticality accident. The NUHOMS system is in compliance with 10 CFR 72.73 as long as the following conditions are met:

1. a maximum fuel enrichment of 3.5%

2. all guide sleeves contain a minimum concentration of B-10 of 0.02 gm/cm 2

3. the fuel assemblies are no more reactive than 15x15 rod arrays.

7.2 Description of Review 7.2.1 Applicable Parts of 10 CFR 72 The applicable part of 10 CFR 72 regarding nuclear criticality safety is the requirement of 10 CFR 72.73.

7.2.2 hview Procedure 7.2.2.1 Design Description The NUHOMS DSC internals are designed to provide nuclear criticality safety during wet loading operations. Fuel loading in an unborated pool was determined to be the worst case nuclear criticality configuration. Analyses were performed to confirm an adequate criticality safety margin for this worst case configuration. After fuel loading and DSC drying, the spent fuel assemblies are not moderated, assuring subcriticality during subsequent operations and configurations.

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7.2.2.2 Acceptanc'- Jriteria The requirement of 10 CFR 72.73 can be met if it is demonstrated that the effective multiplication factor for the NUHOMS design is less than 0.95 for all credible configurations and environments.

-7.2.2.3 Review Method The criticality analysis presented in the TR was reviewed. Independent or confirmatory calculations were not performed. Rather, checks of the computer inputs were performed and the results of the analysis presented were compared for consistency with similar calculations reported in the literature.

7.2.2.4 Key Factors The key factors in the criticality analysis were: (1) the maximum fuel enrichment is 3.5%; (2) the 15x15 rod array is conservative regarding nuclear criticality compared to other fuel assemblies (i.e.14x14,16x16, and 17x17); (3) the guide sleeve material is stainless steel clad boral with a thickness of 0.125 in; and (4) the fuel rod / water array can be represented by a suitably weighted homogeneous mixture.

7.3 Discussion of Results 7.3.1 Analytical Methods The criticality analysis presented in the TR was performed using the KEN 0 IV computer code. The cross sections used in the analysis were a 123 energy group data set. An estimate of the KEN 0 bias for the criticality analysis was estimated by comparing the results from KEN 0 with those of 21 critical experiments. For a 95% confidence level, a bias of - 0.01745 ok was obtained from the analysis of the critical experiments using the method-ology and data employed in the analysis of the NUH0MS design and includes code operation, cross section generation procedure, spacing variation, and statistical precision of the KEN 0 results.

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The computational procedure involved the following steps: performing the resonance self-shielding with the NITAWL code; weighting of the homogen-ized fuel cross sections using XSDRNPM; final mixing of the cross sections for KENO using NITAWL; and finally performing the KEN 0 analysis for the case of interest. This i,s an accepted procedure in the SCALE system for nuclear criticality analysis.

The validity of the homogenization procedure was demonstrated by comparing the results of an infinite array consisting of a fuel asseinbly within a boral shroud. The array was water reflected on the top and bottom.

Comparisor of the KEN 0 analysis using homogenized fuel / water region with discrete modeling of the fuel rods demonstrated a slight conservative bias of +0.0027 ak (although no credit was taken for this bias).

7.3.2 DSC Containing Fuel Several cases were analyzed to evaluate the effect of fabrication tolerances, water temperature and density. The largest k-effective value reported in the TR is 0.8884. This value includes 2 sigma as well as the bias estimated from the verification analysis.

On the basis of the analysis presented in the TR, it is concluded that the NUHOMS is designed to be maintained in a subcritical configuration and to prevent a nuclear criticality accident in compliance with 10 CFR 72.73.

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8.0 OPERATING PROCEDURES 8.1 Summary and Conclusions The staff has reviewed proposed operations described in Chapters 5, "0PERATION SYSTEM", and 9*, " CONDUCT OF 0FERATIONS", of the TR. Portions of Chapters 3 (3.1.2) and 4 (4.7) contain summaries of operation and were also reviewed.

Operations described in the TR are intended to serve as an example only and are not submitted for approval. Therefore, the review was limited to evaluating the feasibility of accomplishing the various activities.

Approval of operations by the staff must await submittal of a site-specific application.

The staff concludes from its review that the operating sequence and steps proposed in the TR are feasible. If a site-specific applicant develops his own detailed operating procedures from the TR descriptions, there is no reason to believe they could not be made to meet the NRC's regulatory requirements. However, since NUH0MS is a new system that has not been built and tested, approval of site-specific procedures will be contingent upon successful demonstration of most "first-of-a-kind" features.

8.2 Description of Review 8.2.1 Applicable Regulations l

The regulations used in the review of the TR included appropriate parts of 10 CFR 20 under the heading of " PERMISSIBLE DOSES, LEVELS, AND CONCENTRATIONS", and those paragraphs of Subparts E and F of 10 CFR 72 related to potential operational accidents (e.g. cask drop), off-normal events, and radiological doses.

Section 9.6, " Decommissioning Plan", of the TR is reviewed in Chapter 11 of this SER.

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l l

8.2.2 Review Procedure 8.2.2.1 Design Description Chapter 5 of the TR presents a generic description of the handling, l transfer and storage operations for NUH0MS. The operations considered i unique to this system include:

o DSC evacuation and helium backfill o DSC top cover welding and weld inspection o transfer of fuel across the site on a specially designed vehicle in a cask which is either built specifically for this use or is a modified version of an existing cask, which is able to unload the DSC at the HSM o positioning and aligning the cask with the HSM opening while it sits on the transfer vehicle o pushing the DSC into the HSM from the cask cavity o reversing the order of loading the DSC into the HSM in order to be able either (1) to retrieve the spent fuel from the DSC on-site or (2) to ship the loaded DSC off-site.

8.2.2.2 Acceptance Criteria Since all operations are generic and no approval is sought, acceptance criteria are not applicable to this review.

8.2.2.3 Review Method The sequence of operations and the step-by-step procedures proposed in the TR for the handling, transfer and storage of spent fuel were reviewed to determine if any portion of the proposed system might not function as planned. The reviewers used engineering judgment and past experience in a review of all proposed steps to reach a determination of feasibility. For those situations in which accidents might occur, a judgment was made of 82

whether the results reported in the TR were reasonable or, lacking results, whether mitigating measures were available which could be implemented on a site-specific basis.

8.2.2.4 Key Assumptions It is assumed that approval of operating procedures will be given only

-on a site-specific basis.

8.3 Discussion of Results In the review of NUH0MS operations, special attention was given to the following issues:

1. Are inspection procedures and records normally available to determine the characteristics and the mechanical and stru.ctural integrity of fuel assemblies prior to loading them into a DSC?

2. Is the DSC able to withstand some reasonable combinations of drop height, cask design (with impact limiters), and transfer vehicles while still maintaining its mechanical integrity, including retrievability of the fuel?

3. Are the dose rates, distances, and worker residence times during the DSC top welding operations reasonable and do they result in acceptably low exposures?

4. Are the dose rates, distances, and worker residence times for loading the DSC into the HSM reasonable and do they result in acceptably low exposures?

5. Are the dose rates, distance of personnel from the DSC in the HSM, and personnel residence time during normal operation, off-normal events and accidents reasonable and do they result in dose rates below levels specified by regulations?

6. Are the alignment dimensional tolerances between the HSM and the transfer cask achievable and can the DSC be easily retrieved from the HSM after 20-40 years of storage?

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The staff believes that NUHOMS can be operated to safely handle, transfer and. store spent nuclear fuel based on a review of the issues listed above. However, approval of operations will require review of detailed site-specific information.

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9.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM  !

l 9.1 Summary and Conclusions The staff has reviewed the proposed acceptance tests and maintenance programs for the storage of spent fuel in NUHOMS. Most of these activities are site specific. The TR does specify the following generic requirements which must be met by the system:

1. The dose rates at the end of the DSC shield plug and at the surface of the HSM after the DSC is first inserted are restricted to specified values consistent with ALARA principles.

2. The maximum rise in air temperature from the HSM inlet to the HSH outlet after the DSC is initially loaded into the HSM is limited to a predetermined figure of 100 0F. At this value the maximum fuel cladding temperature is predicted to remain below 380 00.

3. Daily inspections (surveillance) of the HSM air inlets and outlets is required to ensure that airflow is not interrupted. An annual inspection of the HSM interaals is also recommended to identify potential airflow blockage and material degradation. The results of such inspections may require corrective action, which could be classified in the category of maintenance.

The staff finds that these generic activities, when augmented by a complete set of site-specific activities, will provide for safe operation of the DSC in the HSM when applied to a site-specific situation.

9.2 Description of Review The review was performed by grouping the proposed test and maintenance activities into the following phases:

1. Design, procurement and fabrication of components

2. Construction and installation of the system leading to start-up 85 I

3. Initial loading of spent fuel

4. Long-term passive storage The tests and maintenance activities proposed in the TR during each phase were identified and evaluated for completeness. Those activities which will be the subject of a site-specific application were not re';iewed

-in detail. Those proposed generically were reviewed to determine whether they provide for safe operation of the components which are important to safety.

9.3 Discussion of Results Generic pre-operational acceptance tests of the total NUHOMS system are not proposed. The TR takes the position that such tests are site specific.

Section 9.2 of the TR does, however, recognize the uniqueness of the first NUHOMS installation and outlines a set of pre-operational tests in Table 9.2-1.

Although acceptance tests during spent fuel handling, transfer and storage are for the most part also considered to be site specific, subsection 10.3.2 of the TR does establish limiting conditions on certain critical parameters prior to the time that passive storage begins. Dose rates at the end of the DSC shield plug and at the surface of the HSH after the DSC is first inserted are restricted to specified values consistent with ALARA principles. The maximum rise in air temperature from the HSH inlet to the HSH outlet after the DSC is initially loaded into the HSM is also limited to a predetermined figure of 100 0F. At this value, the maximum fuel cladding temperature is predicted to remain below 380 0C.

Numerous acceptance tests are required in the fabrication of the DSC and its internals. Section 11.2 of the TR identifies the HSM as important to safety, which implies that its design and fabrication will also be controlled. The details of these tests are contained in the commitments to the ASME Boiler and Pressure Vessel Code, the ACI 349 Code, the NUTECH j quality assurance program, the applicant's site-specific quality assurance program, and various procurement specifications. The last two items will be l reviewed with a site specific application.

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l Maintenance of NUHOMS is addressed in the TR in the following ways:

1. Maintenance of the system in order to assure continuous operation is not required since the system is totally passive once the spent fuel is in long term storage. However, daily inspection (surveillance) of the HSM air inlets and outlets is required to ensure that air flow is not interrupted. An annual inspection of the HSM internals is also recommended to identify potential airflow blockage and material degradation. The detailed procedures to be used during such inspections, which must address criteria for determining the effect of degradation, are site-specific.

The results of such inspections may require corrective action, which could be classified in the category of caintenance.

2. Maintenance of the fuel handling and transfer equipment is site-specific. The major components involved are the transfer cask, transfer vehicle, alignment system, and hydraulic ram. (Note:

the devices used for lifting heavy loads while the DSC is in the reactor or spent fuel pool building are assumed to be covered under a Part 50 license).

In summary, the TR treats acceptance testing and maintenance in the following ways:

1. Pre-operational acceptance testing of the system is site-specific.

2. Acceptance testing of components which are important-to-safety (the DSC, DSC internals and HSM) is subject to industry codes and standards, NUTECH's quality assurance program (as applicable), a site-specific applicant's quality assurance program (as applica-ble) and various procurement specifications, the last two items being site-specific.

3. Generic limiting conditions for operation are applied in the TR which, if not met, require corrective action.

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4. Surveillance of the HSM during the passive storage phase is required, which may result in maintenance activitica if the NVHOMS ;

performance is jeopardized. As noted above, detailed surveillance i procedures are site-specific.

5. Maintenance of equipment used in handling and transfer of spent fuel is a site-specific requirement.

The staff finds that this treatment is acceptable and that the generic activities, when augmented by a complete set of site-specific items, will provide for safe operation of the DSC in the HSM when applied to a site-specific situation. Special attention needs to be given to establishing criteria which define when corrective actions are required.

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10.0 RADIOLOGICAL PROTECTION 10.1 On-Site 10.1.1 Summary and Conclusions The shielding, confinement, and handling design features of the NUHOMS system conform to the on-site radiological protection requirements of 10 CFR 20, and are considered acceptable for the set of conditions assumed in this review. The NUHOMS design and operational procedures are also consistent with the objective of maintaining occupational exposures as low as reasonably achievable. Detailed discussions of access control, surveillance, and other operational aspocts affecting on-site exposure are deferred to the site-specific license application.

10.1.2 Description of Review 10.1.2.1 Applicable Parts of 10 CFR 72 Part 72.15 of 10 CFR requi h s the licensee to provide the means for controlling and limiting occupation 11 radiation exposures within the limits given in 10 CFR Part 20, and for meeting the objective of maintaining exposures ALARA.

Part 72.74(a) of 10 CFR requires that radiation protection systems shall be provided for all areas and operations where on-site personnel may be exposed to radiation or airborne radioactive materials.

Part 20.101(a) of 10 CFR .20 states that any individual in a restricted area shall not receive in any period of one calendar quarter from radioactive material and other sources of radiation a total occupational dose in excess of 1.25 rems to the whole body. Part 20.101(b) states that, under certain conditions, the quarterly dose limit to the whole body is 3 rems in any calendar quart.er.

Guidance for ALARA considerations is also provided in Regulatcry Guides 8.8 and 8.10.

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10.1.2.2 Review Procedure 10.1.2.2.1 Design Description.

The radiation protection features of the NUHOMS design include both access control, radiation shielding, and containment of radioactive materials. Access to the NVHOMS installation site will be controlled by a periphery fence so as to meet 10 CFR 72 requirements. The details of the access control features which would be described in the applicant's site license application are site-specific and not part of the TR, The shielding features of the NUHOMS module have been discussed in Chapter 6.0 of this SER. Shielding has been designed to provide an average external surface dose on the module of less than 20 mrem /hr. The contain- '

ment features of the DSC control the release of gaseous radionuclides and are described in Section 1.3.1.1 of the TR.

An assessment of the expected operational dose incurred by site personnel during fuel handling, transfer, and emplacement activities is presented in Section 7.4 of the TR, The cumulative dose is calculated by estimating the number of individuals performing each task and the amount of time associated with the operation. The resulting man-hours are multiplied by the appropriate estimated dose rates for the location of the activity.

The dose rates were estimated from the results of the shielding analysis.

10.1.2.2.2 Acceptance Criteria.

Radiation protection for on-site personnel is considered acceptable if it can be shown that: there are non-site-specific considerations for maintaining occupational radiation exposures at levels which ALARA: that the protection is in compliance with appropriate guidance and/or regulations; -

and that the dose from associated activities to any individual does not exceed the limits of 10 CFR 20.

10.1.2.2.3 Review Method.

Independent or confirmatory calculations of the on-site doses were not made. Rather, the calculational methods and results presented in the TR were reviewed for correctness and self-consistency.

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10.1.2.2.4 Key Assumptions.

The on-site dose assessment was based on loading the HSM with the design basis fuel which is cooled for five years after irradiation. The assessment also assumes that operational procedures and associated doses will be as reported in Table 7.4-1 of the TR.

10.1.3 Discussion of Results 10.1.3.1 ALARA Considerations The design of the DSC and HSM exhibit several features that are speci-fically directed toward ensuring that personnel doses are ALARA.

These include:

bulk shielding in the HSM to reduce the surface dose to below an average of 20 mrem /hr.

lead shield plug on the ends of the DSC to reduce the dose to workers performing drying and sealing operations.

use of a shielded transfer cask for transporting the DSC from the fuel handling building to the NUHOMS installation.

- doyble seal welds on each end of the DSC to provide redundant radioactive tnaterial containment.

filling of the cask and DSC with demitieralized water and sealing the DSC-cask gap to reduce contamination of the DSC exterior during loading.

10.1.3.2 Radiation Protection Design Features of DSC tad HSM There are several radiation protection oesign features of the DSC and HSM. Detailed drawings and material specifications of the radiaticn protec-tion design features are provided in Section 7.3 cf the TR. The DSC body is a section of 0.5-inch thick, 36-inen inside diameter stainless steel pipe.

Two lead-filled end plugs and three steel plates provide neutron and gamma 91

shielding at the ends of the DSC. During transfer operations, shielding in the radial direction is provided by a transfer cask.

The HSM provides shielding in both the radial and axial directions during the storage phase. Bulk shielding consisting of concrete of nominal 42-inch thickness provides neutron and gamma-ray shielding. The front of the module is covered by a two-inch thick steel plate.

Four penetrations in the module are necessary to provide a convective air cooling path. These penetrations are designed to minimize radiation streaming. Openings to the HSM interior are placed above the end shield regions and not directly over the active fuel region. The outlet air pene- ,

trations hav.e two 90 degree bends and are covered with precast concrete shielding end caps.

r 10.1.3.3 On-Site Dose Assessment Table 7.4-1 of the TR provides a summary of the operational procedures which result in radiation exposure to personnel. The cumulative dose 1s calculated by estimating the number of individuals peforming each task and the amount of time associated with the operation. The resulting man-hour figures are multiplied by appropriate dose rates near the transfer cask surface, the exposed DSC top end, or the HSM front.

The total estimated on-site dose for necessary operations associated with the fuel handling and transfer activities for one NUH0MS module is 259.6 mrem. The highest individual dose for these activities is 125.6 mrem.

The largest contribution (155.4 mrem total and 77.7 mrem individual) to the on-site exposure is from the operation to set up the automatic welder to seal weld the lead plug to the canister. All other operations are estimated to result in doses on the order of 10 mrem or less per person. Site workers will also be exposed to direct and air-scattered (skyshine) radiation from filled modules. Examples of activities involving such exposure are surveil-lance of the modules and site operations which are not associated with stored fuel but which are performed in the general vicinity of the storage area. Major factors influencing the magnitude of these exposures are the occupancy times and spatial distribution of workers, and the intensity of the radiation field. An assessment of such exposures is deferred to site-specific applications.

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10.2 Off-Site Radiological Protection 10.2.1 Summary and Conclusions The shielding and confinement design features of the NUHOMS system conform to the off-site radiological protection requirements of 10 CFR 72 and are considered acceptable for the set of conditions assumed in this review. Site-specific factors such as the capacity of the storage array, the distance and direction of the nearest boundary of the controlled zone, and the contribution of reactor plant effluents to the off-site dose must be considered in the compliance evaluation for a specific proposed ISFSI.

10.2.2 Description of Review 10.2.2.1 Applicable Regulations Section 72.15(a)(13) of 10 CFR requires, in part, that a safety assessment be performed on the potential dose or dose commitment to an individual located outside the controlled area as a result of radioactivity releases caused by accidents or natural phenomena events.

Section 72.67(a) of 10 CFR requires that during normal operations and anticipated occurrences, the annual dose equivalent to any real individual located beyond the controlled area shall not exceed 25 mrem to the whole body, 75 mrem to the thyroid, and 25 mrem to any other organ as a result of exposure to (1) planned discharges of radioactive materials (except for radon and its daughter products) to the general environmental, (2) direct radiation from ISFSI operations, and (3) any other radiation from uranium fuel cycle operations within the region.

Section 72.68(b) requires that any individual located on or near the closest 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.

10.2.2.2 Review Procedure The two principal design features which limit off-site exposures during normal operations are the containment capabilities of the double-seal welded 93

DSC, and the radiation shielding of both the DSC and the HSM. The containment features of the DSC control the release of gaseous radionuclides and are described in Section 1.3.1.1 of the TR. The radiation shielding i design features limit the direct radiation exposure rate and are described in Section 7.3.2.1 of the TR, and an analysis of the shield design is '

provided in Section 7.3.2.2. Additionally, Section 7.4.1 describes the construction of a dose-versus-distance curve from the shield analysis results.

Postulated accidents that could result in the loss of shielding or the release of radionuclides are analyzed in Section 8.2. In particular, an accident resulting in the loss of both air outlet shielding blocks is analyzed in Section 8.2.1, while an instantaneous loss of 30% of fission gas inventory from a DSC is assessed in Section 8.2.8. Other accidents are assessed in Section 8.2, but the TR concludes that none of these accidents represent credible sources of off-site dose consequences.

This evaluation focuses on the off-site doses resulting from normal operations and the two postulated accident events. These doses are assessed for compliance with 10 CFR 72. The minimum distance selected for the evaluation of compliance with this section is 200 m, which is a reasonable assumption for the minimum distance to the nearest boundary of the l controlled area. (10 CFR 72.68 requires a minimum distance of at least 100 m to the nearest boundary of the uncontrolled area.) Dose results are also evaluated at 300 m.

10.2.2.3 Acceptance Criteria l Off-site radiological protection features of the NUHOMS system are deemed acceptable if it can be shown that the non-site-specific considerations result in off-site dose levels which are in compliance with the applicable sections of 10 CFR 72, and that these doses to off-site individuals are ALARA.

10.2.2.4 Review Method The review for off-site radiological protection mainly involved a detailed evaluation of the methods applied and the results obtained in the applicable TR sections, supplemented by additional information provided by 94

NUTECH on these methods and results. For the case of off-site doses from direct and scattered (or "skyshine") radiation, an evaluation was performed on the application of the DOT, QADMOD, and ANISN radiation transport codes for calculating gamma-ray and neutron dose equivalent rates at various locations in and a' ound r the HSM, and in the construction of a dose-versus-distance curve using the SKYSHINE-II model. The dose rates predicted by this curve for off-site distances of 200 and 300 m were used to evaluate ,

compliance with 10 CFR 72.67(a).

The accident analyses provided in Section 8.2 of the TR wr.re evaluated for technical soundness, and the results of the DSC leakage event were compared to results calculated independently.

10.2.2.5 Key Assumptions The assessment of off-site dose from normal operations assumes the following:

(1) The recipient of the dose resides at a distance of 200 or 300 m from the face of a filled,184-module array of NUHOMS modules.

t (2) An occupancy factor of unity is assumed, and no credit is taken for attenuation in building materials.

(3) The dose rate as a function of distance from a filled array is as illustrated in Figure 7.4-1 of the IR.

The consequences of the loss of shielding f>1ocks event assumes the following:

(1) The air outlets remain unshielded for a period of 7 days (although  ;

normal recovery should be accomplished in less than 30 minutes).

(2) The resultant dose rate at the air outlet is 495 mrem /hr, and the dose rate decreases with distance according to the values listed in Table 8.2-2.

(3) The recipient of the off-site dose consequences is present for the  ;

entire duration of the recovery at a distance of 200 or 300 m.

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The consequence assessment of the DSC leakage event assumes the following: -

, (1) The fraction of the noble gas (assumed to constit entirely of Kr-

85) and I-129 inventory which is released is 0.3, as recommended by Regulatory Guide 1.25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage facility for Boiling and Pressurized Water Reactors".

(2) The release is assumed to last for a period of I hour or I week.

(3) Short-term atmospheric dispersion factors were obtained from Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequence of a loss of teolant Accident for Pressurized Water Reactors",

(4) Inhalation rates and external dose conversion factors were ob-tained from Regulatory Guide 1.109, " Calculation of Ahnual Doses to Man from Rcutine Releases of Reactor Effluents fcr the Purpcse of Evaluating Compliance with 10 CFR Part 50, Appendix I".

Inhalation dose conversion facters for I-129 were obtained from NUREG-0172, " Age-Specific Radiation Dose Commitment Factors for a One vear Chronic latake".

(5) The distance from the release point to the receptor is 200 or 300 m.

10.2.3 Discussion of Results 1G.2.3.1 Normal Operating Conditions .

The dose to an aff-site individual residing at a distance of 200 or 300 m fron a filled array of 184 NUHOMS taodules is computed as approximately <

66 mrem /yr or 25 mrem /yr, respectively. Since the assessment methodology ,

ccnservatively assumed full-time occupancy in the direction of maximum nff- l site dose, as well as no attenuation by inuitding materials, it is likely that off-site doses to e "real" individual wculd be lower. Although nearest i I

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off-site residence, fuel storage capacity, contribution of off-site dose N from reactor plant effluents, etc. must be. carefully considered, it is likely that normal operation of a NUHOMS array would comply with the requirements of 10 CFR 72. .

10.2.3.2 Accident Conditions x

The dose to an off-site individual at a distance of 200 or 300 m as a ,

result of a loss of shielding block accident is computed as 0.02 mrem or 0.006 mrem, respectively. These doses are well below the limits prescribed by 10 CFR 72.

The dose to an off-site individual at a distance of 200 or 300 m from a ruptured DSC is computed as follows:

.s Dose Eouivalent frem) 1-Hr Release 1-Wk Release Oraan 299_m 300 m 200 m 300 m Whole Body 0.034 0.017 0.004 0.002 Thyroid 0.43 0.22 0.052 0.025 Skin 2.0 1,4 0.34 0.16 These doses are all within the limits prescribed by 10 CFR 72. Note that the calculated results assume that all of the fuel cladding has been breached as a result of sor,e accideht. It should also'be noted that, as it.dicated in the TR, no crsdible conditions have been identified which could breach the canister body or fail the double real welds at each end of the DSC. Thus, these dose results are only presented to demonstrate the inherent safety of the NUHOMS design.

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4 11.0 DECOMMISSIONING 11.1 Summary and Conclusions The applicant has discussed decommissioning considerations for the DSC and HSM. The TR takes the position that decommissioning of the DSC. and of all handling and transfer equipment and a decommissioning plan are site-specific issues. However, the TR proposes that the HSM could be demolished and removed from a site by conventional methods. This position is based on two assumptions:

1. The initial contamination levels on the exterior of the DSC are limited to the following values:

Beta / Gamma Emitters: 10-4 pCi/cm 2 Alpha Emitters: 10-5 pCi/cm 2

2. There is no credible situation in which the DSC could fail and release radioactivity while it is inside the HSM.

The staff finds that the proposed design and procedures are in conformance with the intent of 10 CFR 72.76, but withholds formal approval pending review of a site-specific case. -

11.2 Description of Review 11.2.1 Applicable Parts of 10 CFR 72 10 CFR 72.76 provides criteria for decommissioning. It requires that considerations for decommissioning be included in the design of an ISFSI and that provisions be incorporated to (1) decontaminate structures and equip-ment, (2) minimize the quantity of waste and contaminated equipment, and (3) facilitate removal of radioactive waste and contaminated materials at the time of decommissioning.

Although 10 CFR 72.18 defines the need for a decommissioning plan which includes financing. This type of plan is appropriate only to a site-specific situation. Therefore, 10 CFR 72.18 was not used in this review.

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11.2.2 Review Procedure 11.2.2.1 Design Description j l

The two primary components reviewed against 10 CFR 72.76 are the DSC and the HSM. The TR takes the position that decommissioning of the former

-is site-specific but the latter could be demolished and removed by conven-tional means if the contamination levels of the DSC outer surface are held to specified levels.

Based on the proposed procedures described in Sections 3.3.2.1, 4.4.1 and 5.1.1, the contamination levels of the DSC will be determined by taking surface swipes of the upper one foot of the exterior of the DSC while it is  !

in the transfer cask prior to making the first closure weld. This swipe is used as a representative sample of the DSC body. If the specified limits are not met, the annular space between the DSC and the transfer cask will be flushed with demineralized water until they are met. By minimizing DSC contamination the potential for HSM contamination is held to an absolute minimum.

i 11.2.2.2 Acceptance Criteria 10 CFR 72.76 does not provide specific criteria for acceptance.

Therefore, the designs will be reviewed against good nuclear engineering practices which include (1) a design that is initially decontaminated, and (2) a system that controls the spread of contamination.

11.2.2.3 Review Method Decommissioning is discussed in a general way in Section 3.5 of the TR.

Other descriptions in the TR which relate to the two criteria identified l above include initial contamination limits (Section 3.3.7.1) and contamina-l tion control (Sections 3.1.2.3, 3.3.2.1, 4.4.1, 5.1.1, and 7.1.2). These l sections of the TR were reviewed to assess adequacy of the proposed design in meeting the acceptance criteria.

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11.2.2.4 Key Assumptions The reviewers have assumed that:

1. The contamination limits proposed for the exterior of the DSC can be demonstrated, and

2. There is no credible situation in which the DSC could fail and release radioactivity while it is inside the HSM.

11.3 Discussion of Results The material presented in Section 3.5 of the TR addresses decommis-sioning of the HSM and the DSC. Other sections identified in Section a 11.2.2.3 of this report amplify the discussion in Section 3.5. Decommis-sioning of other components is site-specific.

The TR takes the position that the decommissionino of the DSC is to be addressed on a site-specific basis. It identifies uncertainties in the DSC's internal contamination levels and in the waste disposal policies of the United States when the DSC is ready for decommissioning, presumably a l few decades after the DSC is initially loaded.

However, the TR does address radioactivity levels on the outside sur-faces of the DSC. Section 3.3.7.1 commits to limiting initial contamination levels on the outside surface of the DSC to the following values:

Beta / Gammas emitters: 10-4 pCi/cm 2 Alpha emitters: 10-5 pCi/cm2 These levels can be converted to a total activity level on the DSC surface of roughly 15 microcuries of beta / gamma emitters and 1.5 microcuries of alpha emitters.

The reason for requiring a clean exterior surface of the DSC is to reduce potential contamination of the HSM. If these levels are met initially by the DSC, the contamination in the HSM from this source will be much lower than these values.

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The applicant has claimed that failure of the DSC and release of radionuclides is not feasible under normal, off-normal or accident conditions. Therefore, the possible contamination levels of the HSM ere limited to be much less than initial levels on the exterior of the DSC.

This will allow the HSM to be demolished and disposed of using conventional methods.

The staff finds that the proposed design and procodures are in conformance with the intent of 10 CFR 72.76. It will be necessary to review each site-specific application before determining whether demolition and removal of the HSM can be done by conventional measures. The staff notes that decommissioning of the DSC and of other equipment and a decommissioning plan are aho matters properly addressed in a site-specific application.

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12.0 OPERATING CONTROLS AND LIMITS 12.1 Summary and Conclusions Although operating controls and limits are normally reviewed as part of an application for a site-specific license, the staff has reviewed the set of generic operating controls and limits found in Chapter 10 of the TR.

These controls and limits are summarized in Table 12-1. The proposed operating controls and limits, with the exception of the fuel specifica-tions, are found acceptable.

In addition to the fuel specifications, another limiting condition of operation was identified regarding the dropping of a loaded DSC. The differences in the fuel specifications and the additional limiting condition of operation are stated as follows:

1. In 10.3.1.1 Fuel Specifications,

a. for the shielding design, specifications should be added for specific power (s37.5 MWT/MTHM) and initial enrichment (3.2%

U-235).

b. the alternative specification for gamma source strength should be restated as 1.67x1016 Mev/sec per canister (or 5.4x10 16 photons /sec/ canister).

c. only zircaloy-clad fuel is acceptable for storage in NUH0MS.

d. assemblics with known or suspected gross cladding failures or structural defects will not be stored in NUH0MS.

2. In 10.3.2, Limiting Conditions for Operation, an additional specification is needed which requires that any loaded DSC which is dropped must be opened and the fuel assemblies removed and inspected for damage.

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Table 12-1 Summary of Specifications from Chapter 10 of NUHOMS TR ToDic SDecification TR Reference Fuel Specifications Type Any PWR Fuel 10.3.1.1 Burnup 5 33,000 mwd /MT Post Irradiation Time > 5 years Initial (Beginning of 1 3.5% U-235 Life) Enrichment Weight Per Distance 1 106.56 kg Between Any Adjacent Spacers, Per Assembly Distance Between Spacers s 0.665 m Any fuel not specifically filling the above requirements may still be stored in the NUHOMS system, if all the following requirements are met:

Decay Power Per 1 1 kw Assembly Neutron Source Per 1.43 x 100 n/sec/

Canister assembly Gamma Source Per 7.76 x 1015 Canister photons /sec/ canister

• With spectrum bounded by that shown in Table 3.1-4 End of Life 0.8% U-235 Fissile Content 0.5% Pu-239 0.1% Pu-241 DSC Vacuum Pressure Vacuum Pressure I torr 10.3.2.1 During Drying Time at Pressure: Not less than I hour DSC Helium Backfill Helium 0.0 psig 0.5 10.3.2.2 Pressure psig backfill pressure (stable for 30 minutes after filling) 103

Table 12-1 Summary of Specifications from Chapter 10 of NUHOMS TR (cont'd)

-Topic Soecification TR Reference DSC Helium. Leakage Leakage 10.3.2.3 Rate Test of weld 10'gateofprimary atm - cc/sec Primary Weld DSC Dye Penetrant Test Acceptance standards for 10.3.2.4 of Secondary Weld liquid penetrant examination ASME Boiler and Pressure Vessel Code Section III, Division 1, Subsection NB-5350 (1983)

Liquid Penetrant Acceptance Standards Dose Rate at End of Dose Rates at the following 10.3.2.5 DSC Lead Shield locations:

Plug Center of Lead Shield Plug 200 mrem /hr Center Between Siphon and 100 mrem /hr Vent Tube Penetrations Cask-Canister Gap (at 4 300 mrem /hr locations 900apart)

Center of Steel Cover Plate 50 mrem /hr Edge of Steel Cover Plate 100 mrem /hr Surface Dose Rates on Surface dose rates at the 10.3.2.6 the HSM While the followirg locations:

DSC is in Storage

1) Outside of HSM door on 10 mrem /hr centerline of DSC

2) Center of air inlets 200 mrem /hr

3) Center of air outlets 200 mrem /hr Average Dose rates for the following surface:

1) Roof 50 mrem /hr

2) Front /Back 50 mrem /hr

3) Side 50 mrem /hr Dose rates one meter from the center of the following faces of a unit of modules:

1) Front /Back 20 mrem /hr

2) Side 20 mrem /hr 104

Table 12-1 Summary of Specifications from Chapter 10 of NUHOMS TR (cont'd)

Topic Soecification TR Reference Maximum Air Tempera- Maximum air temperature rise 1000F 10.3.2.7 0

ture Rise from HSM (55.6 C)

Inlet to Outlet Alignment of Cask and The cask must be aligned with respect to 10.3.2.8 HSM for DSC Transfer the HSM so that the longitudinal Operation centerline of the DSC in the cask is within 1/16" of its true position when the DSC rests on the HSM.

Surveillance of the Normal visual inspection 10.3.3.1 HSM Air Inlets frequency Every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Accident visual inspection Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after accident Surveillance of the Visual inspection of the inlets air 10.3.3.2 HSM Inside Cavity pathways, the outlets air pathway, and the concrete surface,- and embedded bolts.

Material Specification All components comprising the DSC 10.3.4.1 of Pressure- pressure boundary will be provided from retaining Components ASME SA 240 Type 304 stainless steel or equivalent Boron Content of DSC The sleeves shall contain a minimum 10.3.4.2 Guide Sleeves effective B-10 loading of 0.02 g/cm2 over the length of the active fuel.

105

l 12.2 Description of Review 12.2.1 Applicable Parts of 10 CFR 72 10 CFR 72.33 defines the requirements for operating limits and contecls. That section only applies to specific licenses, not to reviews and approvals of topical reports. However, to the extent that operating controls and limits in a topical report are referenced in an application for a license, they require approval by the NRC.

12.2.2 Review Procedure The staff has reviewed Chapters 3, 7, 8, and 10 of the TR with special attention given to those parts which form the basis for a set of generic operating controls and limits. The criteria for and results of the safety analyses provided in the first three above-mentioned chapters were used to review the limiting conditions proposed in Chapter 10.

12.3 Discussion of Results Section 10.3 of the TR presents one fuel specification, seven limiting conditions for operation, two surveillance requirements, and two design (actual and material) specifications. The TR identifies all these controls and limits as being generic to NUHOMS and necessary for safe operation.

Reference is also made to the need for two design specifications for the HSM, but no detailed requirements were provided.

The requirements which were provided comprise a set of controls and limits for use with the proposed NUH0MS design. They will have to be augmented by additional specifications or revised to accommodate site-specific issues, but they do serve as a basis for review as a minimum set of requirements.

12.3.1 Fuel Specification The fuel specification of section 10.3.1.1 of the TR restricts the type of fuel acceptable for storage in the proposed NUHOMS design to ensure that peak fuel rod temperatures, radiation source terms, neutron multiplication 106

factor, and stress on the DSC and its internals are below specified design limits.

As noted in Chapter 6 of this SER, the specific shielding analysis was based on fuel irradiated at a specific power of 37.5 MWT/MTHM for 33,000 MWD /MTHM with an initial enrichment of 3.2 percent U-235. These limits should be incorporated into the fuel specification. Furthermore, it is noted that both shielding and nuclear criticality design are acceptable at enrichments at i 3.2 percent, whereas up to 3.5 percent criticality design is acceptable but shielding design must be verified by comparing neutron and gamma source terms with those listed in the alternative specification given in Chapter 6.

The decay power listed of 1 kw and the neutron source strength and its spectra are consistent with the analyses of the TR. The gamma source strength is not consistent with the value given in Tables 1.2-2 or 3.1-1 unless it is restated as 1.67x1016 Mev/sec or 5.4x10 16 photons /sec/ canister.

The alternate set of parameters is not complete because:

1. No mention is made of weights or distances required between spacer disks with the alternative set of requirements.

2. The initial enrichments corresponding to the quoted end-of-life values are not provided anywhere in the criticality analysis section of the TR.

A revised statement of the fuel specification incorporating these differences is given in Table 12-2. With these modifications the fuel speci-fication is acceptable.

12.3.2 Limiting Conditions for Operation The seven limiting conditions for operation (LCO) are acceptable as proposed (see Section 10.3.2 of the TR). The eighth LC0 is an example only and will be the subject of a site-specific application.

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Table 12-2 Fuel Specification for NUH0MS System Type Only PWR Fuels Listed in Table 3.1-2 of TR Fuel Cladding Zircaloy-clad fuel with no known or suspected gross cladding failures Burnup 1 33,000 MWD /MTHM at a specific power of s 37.5 MW/MTHM Post Irradiation Time > 5 years Initial (Beginning of 1 3.2% U-235 Life) Enrichment -

Weight per Distance 1 106.56 kg Between Any Adjacent Spacers, Per Assembly Distance Between Spacers s 0.665 m Any fuel not specifically meeting the above burnup and specific power requirements, and with an initial enrichment of s 3.5% U-235, may be stored in the NUHOMS system if it can be shown that the following requirements are met:

Decay Power per 1 1 kw Assembly Neutron Source 1 9.98 x 108 n/sec per Cannister Gamma Source 1 1.67 x 1 Mev/sec or per Cannister 15.4x10(6 photon /sec/ canister with spectrum bounded by that in Table 3.1-4 of TR 108

The staff notes, however, that Section 3.2.5.2 of the TR imposes a condition on the loaded DSC that in the event it is dropped, it must be opened and the fuel assemblies removed and inspected for damage. The reason for this requirement has to do with the use by the applicant of Service Level D for the drop accident, which allows for deformation of materials.

In this case the staff requires as an additional LC0 that retrieval, of the

-fuel assemblies from the DSC following a loaded cask drop from a height exceeding 15 inches is required. Further, the maximum loads produced on the DSC shall not exceed those bounded by the 48 g vertical and 34 g horizontal drop orientation decelerations postulated for a DSC in a transfer cask in Section 8.2.5 of the TR.

12.3.3 Surveillance Requirements The two surveillance requirements are acceptable, except that the statement that "... analysis showed that blockage of the air inlet alone did not result in unacceptable temperatures." cannot be supported (refer to the discussions in Section 4.3.2.2 of this SER). Nevertheless, since the "all inlets and outlets blocked" case envelopes the " inlets only blocked" case and since the proposed surveillance protects against this more severe case, the staff finds that the balance of the specification and its intent are acceptable.

12.3.4 Design Specifications The two design specifications are acceptable. They address the DSC.

Since the HSM is important-to-safety the staff has reviewed the need for design specification (s) for the HSM described in the TR.

The primary safety functions of the HSM are shielding and physical protection of the DSC. An additional concern of the staff is the assurance of DSC retrievability from the HSM.

The issue of shielding is handled by the LC0 on HSM surface dose rates.

The physical protection of the DSC is assured by the use of reinforcing bars in the HSM concrete and the wall thicknesses of the HSM design.

109

Retrievability will require structural integrity of the support system, which will be periodically verified by the HSM internal surveillance specification.

In conclusion the staff does not see a need for additional generic design specification (s) for the HSM.

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13.0 QUALITY ASSURANCE 13.1 Conclusions Chapter 11, " Quality Assurance," Revision 2, of the TR references NUTECH's Quality Assurance Manual, final draft of Revision 7 (these were docketed March 10, 1986 along with additional response information under Project M-39).

The NRC staff has reviewed all this information and finds it acceptable.

Chapter 11 of the TR, with its reference to NUTECH's Quality Assurance Manual, describes an acceptable quality assurance (QA) program for the design and manufacture of dry storage cask systems for the long-term interim storage of irradiated fuel assemblies. This includes the following revision, as committed to by NUTECH. The scope of the NUTECH QA program, given in Section 11.2 of the TR, will be revised as follows:

NUTECH will apply its QA program to those activities related to the DSC and HSM for which NUTECH has responsibility.

Additional system components such as the transporter, skid, hydraulic ram, and consumables (including dry film lubricant) shall not be considered important-to-safety.

Thus, when properly implemented, the QA program described by NUTECH meets the requirements of 10 CFR 72.80.

l 111 l

,. _ - -. - -1

14.0 REFERENCES

1. NUTECH engineers, Inc., " Topical Report for the NUTECH Horizontal Modular Storage (NUHOMS) System for Irradiated Nuclear Fuel," NUH-001, Rev. 1, November, 1985.

2. U.S. Nuclear Regulatory Commission, " Standard Format and Content for the Safety Analysis Report for an Independent Spent Fuel Storage Installation (Dry Storage)," Regulatory Guide 3.48 (1981).

3. U.S. Government, " Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation," Title 10 Code of Federal Regulations, Part 72, Office of the Federal Register, Washington, DC, (1984).

4. Letter from John Massey, Ph.d., NUHOMS Program Manager, NUTECH, Inc., to J. P. Roberts, Project Manager, Advanced Fuel and Spent Fuel Licensing Branch, USNRC, Attachment 3, dated January 24, 1986. Available from the NRC Public Document Room, 1717 H Street, NW., Washington, DC, 20555, docketed under Project No. M-39.

5. Letter from John Massey, Ph.d., NUHOMS Program Manager, NUTECH, Inc., to J. P. Roberts, Project Manager, Advanced Fuel and Spent Fuel Licensing Branch, USNRC, Detailed Comments on NUTECH Topical Report, NUH-001, Question No. 92, dated May 29, 1985. Available from the NRC Public Docu-ment Room, 1717 H Street, NW., Washington, DC, 20555, docketed under Project No. M-39.

6. Control Data Corporation, "STARDYNE User Information Manual" (1984).

7. CRC Handbook of Tables for Applied Engineering Science, 2nd Edition, pp. 111 and 118, (1973).

8. U.S. Department of Defense, Metallic Materials and Elements for Aerospace Vehicle Structures, MIL-HDBK-5A (1966).

112

9. Fintel, M., Handbook of Concrete Engineering, Van Nortrand Rembold Co.,

New York, New York (1974).

10. Normal and Refractory Concretes for LMFBR Applications, Volumes 1 and 2, EPRI-ND-2437, (June, 1982).

11. Kar, A. K., " Thermal Effects in Concrete Members," Transaction of the 4th International Conference on Structural Mechanics in Reactor Technology, (1977).

12. Roark, R. and Young, W., Formulas for Stress and Strain, Fifth Ed.,

McGraw-Hill Book Company, (1982).

13. ASME, Boiler and Pressure Vessel Code, Sect!on III, Division I, Class 1, Subsection NB, Appendix A, Articles A-2000, A-5000, and A-6000 (1983).

14. U.S. Nuclear Regulatory Commission, " Design Response Spectra for Seismic Design of Nuclear Power Plants," Regulatory Guide 1.60, Rev. 1, (1973).

15. ASME, Op. Cit., Articles NCA-2142.1.

16. ASME, Op. Cit., Article F-1120.

17 ASME, Op. Cit., NB-3222.

18. Letter from John V. Massey Ph.d., NUHOMS Program Manager, NUTECH, Inc.,

to J. P. Roberts, Project Manager, Advanced Fuel and Spent Fuel Licensing Branch, USNRC, Detailed Comments on NUTECH Topical Report, NUH-001, Question No. 55, dated May 29, 1985. Available from the NRC Public Document Room, 1717 H Street, NW., Washington, DC, 20555, docketed under Project No. M-39.

19. Letter from John V. Massey Ph.d., NUHOMS Program Manager, NUTECH, Inc.,

to J. P. Roberts, Project Manager, Advanced Fuel and Spent Fuel Licensing Branch, USNRC, Detailed Comments on NUTECH Topical Report, NUH-001, 113

i Question No. 15, dated May 29, 1985. Available from the NRC Public Document Room, 1717 H Street, NW., Washington, DC, 20555, docketed under Project No. M-39.

20. A. B. Johnson, Jr. , and E. R. Gilbert, " Technical Basis for Storage of Zircaloy-Clad Spent Fuel in Inert Gases." PNL-4835, September 1983.

21. D. C. Elrod, et. al., " HEATING-6: A Multidimensional Heat Conduction Analysis with the Finite-Difference Formulation," NUREG/CR-0200, Vol. 2, Section F10, ORNL/NUREG/CSD-2/V2, October 1981.

114

-l .

APPENDIX A ANALYSIS OF DIFFUSION CONTROLLED CAVITY GROWTH (DCCG) DAMAGE TO FUEL CLADDING IN DRY STORAGE

1.0 INTRODUCTION

The only damage mechanism that the staff found with a possible potential for cladding degradation and gross rupture was DCCG. The staff has examined this potential and based on available information has developed a method to determine the level of damage which could occur under dry storage conditions for the NUHOMS module as a function of spent fuel time in storage.

2.0 REVIEW PROCEDURE

-The area fraction of decohesion at at any time can be ascertained by satisfying the following equation A t f 7 f

f )

=

f G(t)dt (2-1)

Aj o where A j is the initial area fraction of decohesion due to the nucleation of stable cavities and A7 is the area fraction of decohesion that occurs over the period of time t .

7 Furthermore, Ag (1 p sin a) (1 - A) f(A) = 1 1 3 A (2-2)

A(7 En g g - A(1 z))

32 F

B (a) 060, Dgb(t)

( )

• 3n T(t) (2-3)

Fy (a) g3 1

The terms of expression (2-2) and (2-3).are defined as follows:

a = grain boundary cavity dihedral angle Q = atomic volume-6 = grain boundary thickness

~

o ,= stress on the cladding k = Boltzman's constant A = average cavity spacing D

gb

= grain boundary diffusion rate T = absolute temperature FB (a) = nsin za Fy (a) = 2n/3 (2 - 3cosa - coszy)

Some of the foregoing terms may be further defined by 1

a = cos- (TB )

Dgb = Dogb exp[-Q/RT(t)]

Dogb = grain boundary diffusion coefficient .

Q = activation energy for grain boundary self-diffusion R = gas constant 1

T =_ free surface energy i TB= grain boundary surface energy Much of the review effort focused on establishing the values of the param-eters in the above expressions. Where there was wide divergence in reported values, the value that led to the most conservative result was selected.

2.1 Grain Boundary Cavity Dihedral Angle, a 1

For clean surfaces in pure metals Raj and Ashby suggest that TB= 2 so that a is computed to be about 75*. To account for non-ideal condi-tions, a value for a of 50* was used in the analysis.

t 4

2

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

2.2 Atomic Volume, O The atomic volume can be estimated from

^

O = Np where A is the atomic weight = 91 N is Avogadro's number = 6.02 x 1023 p is the specific gravity = 6.55 gms/cc which gives a value for 0 of 2.31 x 10 29 m3 / atom. This agrees Closely with a value of 2.37 x 10 29 ms / atom reported by Lloyd.2 However, Chin, et al.3 used the cube of the Burgers vector, b = 3.23 x 10 8 m, which gives an atomic volume of 3.37 x 10 20 m3 / atom. For the sake of conservatism, the value for 0 = 3.37 x 10 29 .3/ atom was selected for the analysis.

2.3 Grain Boundary Thickness, 6 The grain boundary thickness defines the area through which grain boundary vacancies migrate to the cavity. The disorder that characterizes the structure at the grain boundary is only a few atoms thick. Since grain boundary diffusion rates are many orders of magnitude greater than volume diffusion rates, a grain boundary thickness of 3 Burger's vectors is ionsidered adequate. Consequently, a value of 6 = 3(3.23 x 10 10) = 9.69 x 10 10 m was selected for the analysis.

2.4 Stress on the Cladding, o, The cladding stress is due to the fuel rod internal pressure at the storage temperature. There is considerable uncertainty regarding the level of pressure in the fuel rod, either from rod pressurization, fission gas release, or volume increase due to creep strain. Blackburn4 states con- l servatively that the fuel rod internal pressure is no greater than the external

]

reactor system pressure, which is 2250 psi for a PWR, at operating temperature.

This results in a conservative stress level in the fuel rods of 17830 psi, (12.3 MPa) which was used for the analysis.

3 l

I o

2.5 Average Cavity Spacing, A The value of this parameter has been particularly difficult to establish.

Cavity spacing depends upon the density of nucleation sites and will vary with

'the type of nucleation mechanism. Experimental work conducted at Cornell 5

University indicated a spacing in unirradiated Zircalloy-2 of from 10 to 20 x 10 3 m. This experimental work further established that grain boundary cavities do form at 350*C especially at stresses over 100 MPa. The cavity density appeared to reach a saturation level after about 10 days suggesting a ,

limited number of nucleation sites in the material. Consequently, it is not ,

likely that the intercavity spacing, A, will decrease during dry storage as a result of further nucleation. Conservatism dictated the use of the lower value i

of 10 x 10 6 for the analysis,

/

2.6 Grain Boundary Diffusion Rate, D gb There are many reported values of volume diffusion rate for a-Zirconium but few with respect to grain boundary diffusion rate. The two values spe-cific for grain boundary diffusion are 6 x 1010 exp (112/RT) reported by Chin 3 and 5.9 x 10 8 exp (131/RT) reported by Garde, et al.6 The latter is the more conservative value by about two orders of magnitude and was, consequently, used for the analysis.

2.7 Temperature, T The temperature dependence of grain boundary decohesion was established

, using the temperature decay curve provided on Figure 4-3 in a submittal by NUTECH7 supplementing its topical report. Since the data as reflected by measured values terminates at approximately ten years from beginning of storage, it was conservatively assumed that the temperature would remain constant thereafter.

3.0 FINDINGS AND CONCLUSIONS The progress of damage based upon the methodology and assumed values for the parameters previously described indicates that the area of decohesion at 4

the end of twenty year storage life to be less than 4 percent. Sased upon the degree of conservatism maintained throughout the analysisi it can be concluded that this level of damage is insignificant and would not be exceeded, Conse-quently, based on the assteed constant ambient temperature of 70*F (21*C) ovet- ,

the twenty years of storage, which would correspond at storage initiation to a temperature not exceeding 322 C for the cladding of the design' basis fuel in a caaister emplaced in a NUHOMS concrete module, the requirements of 10 CFR /2, Section 72.72.(h), are met.

l 5

p L

REFERENCES

1. R. Raj and M. F. Ashby, "Intergranular Fracture at Elevated Temperature,"

Acta Met. , Vol. 23, p. 653 (1975).

2. L.1. Lloyd, " Thermal Expansion of Alpha-Zirconium Single Crystals,"

ANL-6591, Argonne National Laboratory (1963).

3. 8. A. Chin, N. H. Madsen and M. A. Khan, " Application of Zircalloy Deforma-tion and Fracture Maps to Predicting Dry Spent Fuel Storage Conditions,"

Department of Mechanical Engineering, Auburn University, Auburn, A1.

t h ...

• c i

4. L. D. Blackburn et al.,' " Maximum Allowable Temperat.ure for Storage of t

i ,f -..,

i j-Spent Nuclear'. Fue1,"h 3 'HEDL-TME78.37,.VC70,May1580.

4 ,

( ,f,'

k t .

(

5. R. L. Keusseyan, " Grain Bounddrp;.Slibi*ng and Related Phenomena," Doctoral Dissertation, Cornell [ University,1985.. ,

I .,,

l- _ ... .

i

6. A. M. Garde, H. M. Chung, and.T.g F . Kaisner, "Micrograin Superplacticity 5 ,; ,

f inZircalloyat850*C,l"ActsMet.,Vo.'26,p.153(1978).

7. Letter from John V. Massey, Ph.D., NUHOMS Program Manager, NUTECH, Inc.

to John Roberts, Project Manager, Advanced Fuel and Spent Fuel Licensing i- Branch, USNRC, dated January 17, 1986. Available from the NRC Public Document Room, 1717 H Street NW., Washington, DC 20 5, docketed under Project No. M-39.

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