ML20207B099

Interim Staff Guidance

ML20207B099
Person / Time
Issue date:	05/21/1999
From:	Brach E NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
Shared Package
ML20207B085	List: ML20207B082 ML20207B099
References
PROC-990521, NUDOCS 9905280118
Download: ML20207B099 (38)

Text

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SPENT FUEL PROJECT OFFICE OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS 1 E. William Brach, Director l 1

J INTERIM STAFF GUIDANCE l 1

1 7885288a2"3882" PDR ATTACHMENT

INTERIM STAFF GUIDANCE (ISG)

ISG-1 Damaged Fuel ,

ISG-2 Fuel Retrievability ISG-3 Post Accident Recovery and Compliance with 10 CFR 72.122(l)

ISG-4, R1 Cask Closure Weld Inspections ISG-5, R1 Normal, Off-Normal, and Hypothetical Accident Dose Estimate Calculations for the Whole Body, Thyroid, and Skin ISG-6 Establishing Minimum initial Enrichment for the Bounding Design Basis Fuel Assembly (s)

ISG-7 Potential Generic issue Concerning Cask Heat Transfer in a Transportation Accident ISG-8 Limited Burnup Credit in the Criticality Safety Analyses of i PWR Spent Fuel in Transport and Storage Casks I ISG-9 Storage of Pressurized Water Reactor (PWR) Fuel Assembly integral Components ISG-10 ASME Code Exceptions ISG-11 Storage of Spent Fuel Having Burnups in Excess of 45,000 mwd /MTU ISG-12 Buckling of Irradiated Fuel Under Bottom End Drop Conditions ATTACHMENT

Spent Fuel Project Office interim Staff Guidance -1 lasue: Damaged Fuel  !

Definition of Damaged Fuel l i

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Spent nuclear fuel with known or suspected cladding defects greater than a hairline crack or a l l pinhole leak.

1 This definition of damaged fuel applies to both spent fuel storage and transportation. I Canning of Damaged Fuel 1

l Damaged fuel, as defined in item 1 above, should be canned for storage and transportation.

i The purpose of canning is to confine gross fuel particles to a known, subcritical volume during off-normal and accident conditions, and to facilitate handling and retrievability.

This provision for canning damaged fuel applies to both spent fuel storage and transportation.

l Double Containment per 10 CFR 71.63(b)

Spent fuel, with plutonium in excess of 20 curies per package, in the form of debris, particles, loose pellets, ard fragmented rods or assemblies must be packaged in a separate inner container (second containment system) in accordance with 10 CFR 71.63(b).

This provision for double containment applies to transportation only.

l Demonstration of Fuel Condition l

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As proof that the fuel to be loaded is undamaged, the staff will accept, as a minimum, a review of the records to verify that the fuelis undamaged, followed by an extemal visual examination of I

the fuel assembly prior to loading for any obvious damage. For fuel assemblies where reactor records are not available, the level of proof will be evaluated on a case-by-case basis. The purpose of this demonstration is to provide reasonable assurance that the fuelis undamaged or that damaged fuel loaded in a storage or transportation cask is canned.

This provision for demonstrating the condition of the fuel applies to both storage and  !

transportation.

1 Recommendation:

The SRPs be revised accordingly.

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l Approved M

, i William F. Kane Date '

4 ISG-1

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Spent Fuel Project Office interim Staff Guidance -2 issue: Fuel Retrievability

$72.122(l) states:

• Retrievabl/ity. Storage systems must be designed to allow ready retneval of spent fuel or high-level radioactive waste for further processing or disposal."

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$72.236(h) states:

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'The cask must be compatible with wet or dry spent fuel loading and unloading '

facilities."

$72.236(m) states:

"To the extent practicable in the design of the storage casks, consideration should be given to compatibility with removal of the stored spent fuel from the reactor site, transportation, and ultimate disposition by the Department of Energy."

The basis of 10 CFR 72.122(1) is the Nuclear Waste Policy Act (NWPA)of 1982, $141(b)(1) (C),

(51 FR 19108 53 FR 31651). The NWPA required that Monitored Retrievable Storage (MRS) facilities be designed "to provide for the ready retrieval of such spent fuel and waste for further processing and dinosal.' in ame.nding Part 72 to permit licensing of an MRS as required by  ;

the NWPA, the Commistion determined that an independent spent fuel storage installation '

(ISFSI) must also meet the same criteria.

As utilities shut down reactors, and either plan for or actually start decommissioning, there is an increasing need to move the spent fuel from the reactor spent fuel pool to an ISFSI. These ISFSis, generally consisting of an array of spent fuel storage casks on the licensee's site, are ,

licensed or approved under the provisions of 10 CFR Part 72. In practice, the casks are loaded with spent fuel within the existing pool undet the provision of 10 CFR Part 50, then the casks are transferred from the spent fuel storage building out to the storage area. As the casks leave the spent fuel building they, and their associated operations and maintenance, transfer to the regulatory provisions of 10 CFR Part 72. After a pool has been emptied of all spent fuel, a utility may, if appropriate and in accordance with the regulations, proceed with immediate decommissioning of the spent fuel pool.

The Nuclear Regulatory Commission (NRC) (and U.S. Department of Energy (DOE))

recognized that "in the interest of reducing radiation exposures, storage casks should be designed to be compatible with transportation and DOE design criteria to the exient practicable

. . . to the extent that cask designers can avoid retum of the spent fuel from dry cask storage to reactor basins for transfer to a transport cask before moving it off site for dispe. sal" (55 FR 2g186). This, and DOE's development of a multi-purpose canister (MPC) program gave rise to dual purpose (storage and transportation) cask designs. The MPC, containing the fuel, could easily be transferred from a storage system into a transportation cask. With dual .

purpose designs, fuel no longer must be retumed to the reactor spent fuel pool for repackaging.

, Dual purpose cask designs should have the capability of being prepared for off-site transportation without having to handle individual fuel assemblies or retum to a spent fuel pool.

ISG 2

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l Insofar as a facility in question is used for interim storage, (e.g., not permanent disposal), and as long as the design of the storage system has a method to repackage into a transportation cask for shipment offsite (e.g., designed for decommissioning) for further processing or

, disposal, a facility meets the requirements of 10 CFR 72.122(l).

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Recommendation

Chapter 1 of the Standard Review Plan (NUREG-1567) should be modified to clear 1y define that compliance with 10 CFR 72.122(1) is achieved when an applicant's design is found to be in com,nliance with 10 CFR Part 72. This means all facilities that could be licensed under 10 CFR Part 72 must be designed to allow for ISFSI decommissioning and have only limited license terms. Therefore, an ISFSI considered for a license under Part 72 cannot become a " defacto" repository (i.e. the fuel is retrievable under normal conditions of operation, and therefore, the requirements of 10 CFR 72.122(l), are met).

l The staff believes that 10 CFR 72.122(l) applies to normal and off-normal design conditions and not to accidents. ISG 3 discusses the staffs recommendation for post accident recovery with regard to retrivievability.

/0fffW Approved

! William F. Kane" Date l

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l ISG.2 2

Spent Fuel Project Office interim Staff Guidance - 3 lasue: Post Accident Recovery and Compliance with 10 CFR 72.122(1)

Compliance with 10 CFR 72.122(l) has been interpreted to mean that a licensee, during any point in the storage cycle, must have a means of retrieving and repackaging individual fuel assemblies even after an accident. The stoff has reevaluated this interpretation.

Recommendation:

The staff proposes that the Standard Review Plans (SRPs) be modified to communicate the distinction between retrievability and post accident recovery. That is,10 CFR 72.122(l) applies to normal and off-normal design conditions and not to accidents. Chapter 15 and Chapter 10 of NUREG-1667 and Chapter 11 of NUREG-1536, should be modifed to focus on the identification of all credible accidents affecting public health and safety. Further, the SRPs should eliminate all references to non-credible accidents such as non-mechanistic failures of the confinement boundary. The accident analysis chapters should be rewritten to require that the staff evaluate all credible accidents and focus the review on those accidents with potential consequences resulting in the failure of the confinement boundary. Upon identification, the event shall be evaluated against the requirements of 10 CFR 72.106 and 72.122(b). Recovery methods or the need for Over-Packs or Dry Transfer Systems to maintain safe storage conditions would then not be considered and evaluated as part of the licensing process.

1 However, because a failure of the confinement boundary or other structure, system, or l component important to safety, by a means that has not been considered, is a possibility,  :

NUREG-1667, Chapter 10, Section 10.4.5

• Emergency Planning" and NUREG-1536,  ;

Chapter 11,Section V.2, " Detection of Events" should be modifed to ensure that the licensee l will have the ability to identify an accident or non-compliance situation.

Approved Yb _/ -

/0 N William F. Kane Date L

L ISG-3

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1 Spent Fuel Project Office

) interim Staff Guidance - 4, Revision 1 1

lasue: Cask Closure Weld Inspections l

Discussion:

The closure weld for the outer cover plate for austenitic stainless steel designs may be inspected using either volumetric or multiple pass dye penetrant techniques subject to the following conditions:

Dye penetrant (PT) examination may only be used in lieu of volumetric examination only on austenitic stainless steels. PT examination should be done in accordance with ASME Section V, Article 6," Liquid Penetrant Examination."

For either ultrasonic examination (UT) or PT examination, the minimum  ;

detectable flaw size must be demonstrated to be less than the critical flaw size.

The critical flaw size should be calculated in accordance with ASME Section XI rnethodology; however, net section stress may be governing for austenitic stainless steels, and must not violate Section 111 requireme,its. Flaws in austenitic stainless steels are not expected to exceed the thickness of one weld j bead. i i

e if PT examination alone is used, at a minimum, it must include the root and final I layers and sufficient intermediate layers to detect critical flaws.

The inspection of the weld must be performed by qualified personnel and shall meet the acceptance requirements of ASME B&PV Code Section Ill, NB-5350 l for PT examination and NB-5332 for UT examination.

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if PT examination alone is used, a design stress-reduction factor of 0.8 must be applied to the weld design.

The results of the PT examination, irJuding all relevant indications, shall be made a permanent part of the licensee's records by video, photographic, or other means providing a retrievable record of weld integrity. Video or photographic records should be taken during the finalinterpretation period described in ASME Section V, Article 6, T-676.

Technical Basis:

Radiographic (RT) inspection is preferred for cask closure welds. However, RT may not be practical for field closure welds with fuelin the cask. UT is the next preferred inspection method but UT of stainless steel welds for the closure configurations may pose considerable difficulty and uncertainty, UT has only recently been demonstrated for carbon steel for the VSC-24 cask design. PT examination only identifies surface flaws but, if performed at sufficiently small weld depths, can provide reasonable assurance of weld integrity. The position recognizes both UT and multi-layered PT examination as acceptable methods; however UT is still preferred where practical.

Acceptable UT has not yet been demonstrated for austenitic stainless steels in the required connguration. At best, UT that would be developed may require considerable skillin execution

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iSG-4R1 2 and interpretation. Minimum detectable flaw sizes for UT will be relatively large (estimated 0.1 inch deep) and the technique is subject to false indications which may require grinding out weld ,

material unnecessarily. This additional grinding may introduce additional weld integrity l

problems and may present Al. ARA issues. Although the setup for the UT would be expensive, l  : cost could be spread over multiple containers to reduce unit cost.

PT examination only identifies surface flaws. However, the acceptance standard is no cracks or linear indications. In theory, a flaw slightly smaller than the PT examination increment size could exist, therefore, layered PT examination is necessary to assure detectable flaw sizes are j less than critical flaw site. PT examination is widely used in safety critical applications such as certain reactor pressure vessel welds and, for some applications is th's only practical technique.

, ALARA issues may arise for large welds that require multiple PT examinations.

l Austenitic stainless steels do not have a nil ductility transition temperature. Thus, the weld can sustain "large" flaws without a concern for flaw growth. This allows the use of either UT or PT

' examination although both would have limitations on detectable flaw size and both would accept '

less than critical flaws.

Finally, the Nuclear Regulatory Commission (NRC) regulates to the standard of adequate protection, not absolute assurance. Although UT is the preferred technique to PT examination,  ;

in that it is a volumetric examination, PT examination is considered to be adequate for safety, specifically for austenitic stainless steels in that it can provide reasonable assurance that flaws of interest will be identified. This position does not apply to carbon steel construction.

4 Recommendation:

1. The ISG is specifically developed for the dry storage canister top end closure weld after i the canister is loaded with spent nuclear fuel assemblies. All other dry storage canister i bottom end closure welds and shell welds should be designed, fabricated, examined, I and tested to the requirements of the appropriate subsections of the ASME Section lli i Code. I i

l 2. The top end closure welds are leak tested. No hydrostatic or pressure tests are I i required if a minimum margin of safety equal to or greater than 1.5 against design l l pressure was demonstrated by analysis. I I

( 3. The closure weld joint may be either a full thickness penetration weld or a partial I penetration groove weld. For a partial penetration groove weld, the maximum clearance I between the closure plate and the enclosure shell should not exceed 1/16 inch and the i

, minimum depth of the groove shall be equal to or larger than the enclosure shell l l thickness. The weld strength of the closure joint is based on the nominal weld area and I j the design stress intensity values for the weaker of the two materials jointed. However, I i the minimum ultimate tensile strength of the weld metal should equal or exceed the I l l base metal strength to preclude weld metal failure. I j l

4. Dry storage canisters made from austenitic stainless steels Type 304,304L,304LN, I l 316, 316L, or 316LN , the top end closure weld may be examined by either the l l ultrasonic methods (UT) or progressive PT examination as follows: l l l

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ISG-4R1 3

a. If UT examined, the UT acceptance criteria are the same as those of NB-5332 I for pre-service examination. I I

b. If PT examined, the examination should be performed progressively on the root I layer, the lesser of one half of the welded joint thickness, or % inch intervals I thereafter, and the final surface, in addition, a stress reduction factor of 0.8 shall I be applied to the weld strength of the joint. I l  ;

5. Dry storage canisters made from austenitic stainless steels other than Type 304 or 316 l  ;

listed above, may be PT examined as in 4.b above, except that the size and number of I l intermediate layers to be examined should be determined by a fracture mechanics I assessment of the weld considering the specific geometry, material properties, and I loadings. I  ;

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6. For dry storage canisters made from ferritic steels, the top end closure weld should be i examined by UT. I I

a. Based on service temperature, material properties, and critical design stress I values, determine the critical flaw size by the linear elastic fracture mechanics I i methodology specified in ASME Code,Section XI. I I

b. UT must be p'.nformed in accordance with pre-qualified procedures and l l methods. The UT examination methodology should demonstrate to be I reasonably accurate and consistently able to detect flaw size less than the l I critical flaw size determined in 6.a. The UT operators are tested and certified. l l Welding processes, weld inspection criteria, and personnel qualifications should l I be verified as being in conformance with the ASME Code. The welding process I l and technique used should be evaluated to preclude hydrogen induced cracking. I  !

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c. Progressive surface examinations, utilizing PT or magnetic particle examination I (MT), should be permitted only if unusual design and loadino conditions exist. I PT or MT must be performed after sufficiently small intarvals to ensure the I critical flaws will be detected. In addition, a stress reduction factor of 0.8 is I imposed on the weld strength of the closure joint to cover imperfections or flaws I potentially missed by progressive surface examinations. Because of brittle I fracture concems in ferritic steels, critical flaw sizes for ferritic steels are I generally small. Therefore, PT or MT must be performed on many layers of the I weld and it may become unacceptable due to ALARA concerns. The weld I  ;

design should show a sufficient safety margin and should be approved by the I NRC on a case-by-case basis. I ,

I The Standard Review Plans should be revised to clearly state the inspection criteria for spent I i fuel cask's outer cover plate closure welds. l l

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. l Approved -

E. William Brach Date

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Spent Fuel Project Office Interim Staff Guidance - 5, Revision 1 issue: Confinement Evaluation 1 Discussion: l Several changes have occurred sinco the issuance of NUREG-1536, " Standard Review Plan (SRP) for Dry Cask Storage Systems," that affect the staff's approach to confinement evaluation. The attachment to this ISG integrates the current staff approach into a revision of ISG-5. The highlights of the changes include- ,

• Reflects October 1998 revisions to 10 CFR 72.104 and 10 CFR 72.106.

• Expands and clarifies acceptance criteria associated with confinement analysis and acceptance of " leak tight" testing instead of detailed confinement analysis.

• Updates staff review guidance for design and requirements for the cask seal monitoring system and adds guidance for accident analysis of " latent" failure concerns.

• Updates source term guidance to (1) include ISG-5 recommendations, (2) include actinide activity that contributes greater than 0.01% of the design basis activity, and i (3) allow for a reduction of fines that can escape the cask (with justification by applicant).

• Deletes non-mechanistic (confinement boundary failure) accident analysis and revise staff review guidance for evaluation of normal, off-normal, and accident cases. The significant change is that the evaluated leaks are related to the as-tested leak rate.

• Updates confinement analysis section to reflect ISG-5 and describe what types of analysis should be done. Dose to lens of the eye will be addressed if skin dose and TEDE do not exceed 15 rem.

Regulatory Basis: See attachment Technical Review Guidance:

To ensure consistency in reviews, consolidate various references, and simplify the reviews, the guidance in the attachment to this ISG should be used instead of SRP Chapter 7.

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ISG-5, Rev.1 2 l Recommendation:

l SRP, Chapter 7, should be replaced with attached confinement evaluation. In addition SRP i

Chapter 11 Section V.2 should be revised regarding classificatiori ef the monitoring system to j be consistent with SRP Chapter 7. Further, SRP Chapter 2 Section V.2.b,(3)(e) should be l updated to removo reference to non-mechanistic failure of confinement boundary event.

Approved d / /~

E. William Brach 4- .N//(f 6 ate'

Attachment:

As stated 4

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1 ATTACHMENT TO ISG-5 REVISION 1 l CONFINEMENT EVALUATION i l

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l. Review Objective I l

in this portion of the dry cask storage system (DCSS) review, the U.S. Nuclear Regulatory Commission (NRC) evaluates the confinement features and capabilities of the proposed cask system. In conducting this evaluation, the NRC staff seeks to ensure that radiological releases to the environment will be within the limits established by the regulations and that the spent fuel cladding and fuel assemblies will be sufficiently protected during storage against degradation that might otherwise lead to gross ruptures. '

11. Areas of Review This chapter of the DCSS Standard Review Plan (SRP) provides guidance for use in evaluating the design and analysis of the proposed cask confinement system for normal, off-normal, and accident conditions. This evaluation includes a more detailed assessment of the confinement-related design features and criteria initially presented in Sections 1 and 2 of the applicant's safety analysis report (SAR), as well as the proposed confinement monitoring capability, if applicable. In addition, the NRC staff assesses the anticipated releases of radionuclides associated with spent fuel, by independently estimating their leakage to the environment and the subsequent impact on a hypothetical individual located beyond the controlled area boundary.

As prescribed in 10 OFR Part 72, the regulatory requirements for doses at and beyond the controlled area boundary include both the direct dose and that from an estimated release of radionuclides to the atmosphere (based on the tested leaktightness of the confinement). Thus, an overall assessment of the compliance of the proposed DCSS with these regulatory limits is deferred until Chapter 10," Radiation Protection," of this SRP. In addition, the performance of the cask confinement system under accident conditions, as evaluated in this section, may also be addressed in the overall accident analyses, as discussed in Chapter 11 of this SRP.

As described in Section V," Review Procedures," a comprehensive confinement evaluation may encompass the following areas of review:

1. confinement design characteristics

a. design criteria

b. design features

2. confinement monitoring capability

3. nuclides with potential for release

4. confinement analyses

a. normal conditions

b. leakage of one seal

c. accident conditions and natural phenomenon events

5. supplementalinformation l

l l 1 Attachment to ISG-5 Revision 1

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., l 111. Regulatory Requirements

1. Description of Structures, Systems, and Components important to Safety The SAR must describe the confinement structures, systems, and components (SSCs) important to safety in sufficient detail to facilitate evaluation of their effectiveness. [10 CFR 72.24(c)(3) and 10 CFR 72.24(l)]

l i 2. Protection of Spent Fuel Cladding l

l The design must adequately protect the spent fuel cladding against degradation that might otherwise lead to gross ruptures during storage, or the fuel must be confined through other means such that fuel degradation during storage will not pose operational safety problems with respect to removal of the fuel from storage. [10 CFR 72.122(h)(1))

i 3. Redundant Sealing The cask design must provide redundant sealing of the confinement boundary. [10 CFR 72.236(e))

4. Monitoring of Confinement System Storage confinement systems must allow continuous monitoring, such that the licensee will be able to determine when to take corrective action to maintain safe storage conditions. [10 CFR 72.122(h)(4) and 10 CFR 72.128(a)(1)]

5. Instrumentation The design must provide instrumentation and controls to monitor systems that are important to safety over anticipated ranges for normal and off-normal operation. In addition, the applicant must identify those control systems that must remain ope ~r ational under accident conditions.

[10 CFR 72.122(i))

6. Release of Nuclides to the Environment The applicant must estimate the quantity of radionuclides expected to be released annually to the environment. [10 CFR 72.24(l)(1)]

7. Evaluation of Confinement System The applicant must evaluate the cask and its systems important to safety, using appropriate tests or other means acceptable to the Commission, to demonstrate that they will reasunably maintain confinement of radioactive material under normal, off normal, and credible accident conditions. [10 CFR 72.236(l) and 10 CFR 72.24(d))

In addition, SSCs important to safety must be designed to withstand the effects of credible

accidents and severe natural phenomena without impairing their capability to perform safety functions. [10 CFR 72.122(b))

l l Attachment to ISG-5 Revision 1 2 l

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8. Annual Dose Limit in Effluents and Direct Radiation from an lndependent Spent Fuel Storage Installation (ISFSI)

During normal operations and anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area must not exceed 0.25 mSv (25 mrem) to the I whole body,0.75 mSv (75 mrem) to the thyroid, and 0.25 mSv (25mrom) to any other critical I organ. [10 CFR 72.104(a))

IV.- Acceptance Criteria in general, DCSS confinement evaluation seeks to ensure that the proposed design fulfills the following acceptance criteria, which the NRC staff considers to be minimally acceptable to meet the confinement requirements of 10 CFR Part 72:

1. The cask design must provide redundant sealing of the confinement boundary. Typically, I this means that field closures of the confinement boundary must either have two seal welds 1 or two metallic O-ring seals. I

2. The confinement design must be consistent with the regulatory requirements, as well as the applicant's " General Design Criteria" reviewed in Chapter 2 of this SRP. The NRC staff has accepted construction of the primary confinement barrier in conformance with Section lil, Subsections NB or NC, of the Boiler and Pressure Vessel (B&PV) Code' promulgated by the American Society of Mechanical Engineers (ASME). (This code defines the standards for all aspects of construction, including materials, design, fabrication, examination, testing, inspection, and certification required in the manufacture and installation of components.) In such instances, the staff has relied upon Section ill to define the minimum acceptable margin of safety; therefore, the applicant must fully document and completely justify any deviations from the specifications of Section Ill. In some cases after careful and deliberate consideration, the staff has made exceptions to this requirement.

3. The applicant must specify the maximum allowed leakage rates for the total primary I confinement boundary and redundant seals. Applicants frequently display this information I in tabular form, including the leakage rate of each seal. The maximum allowed leakage rate I is the "as tested" leak rate measured by the leak test performed on the cask field closure. I Generally, as discussed in items a. through d., below, the allowable leakage rate must be I evaluated for its radiological consequences and its effect on maintaining an inert I atmosphere within the cask. However, for storage casks having closure lids that are I designed and tested to be " leak tight", as defined in "American National Standard for i Leakage Tests on Packages for Shipment of Radioactive Materials, ANSI N14.5-19978, the I analyses discussed in a. through d., below, are unnecessary.' l I

a. The applicant's leakage analysis should be consistent with the methods described in 1 ANSI N14.5-1997. i l

• For casks that are demonstrated to be leak tight, the review procedures discussed in I sections V.3 and V.4 are not applicable. I 3 Attachment to ISG 5 Revision 1 1

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b. During normal operations and anticipated occurrences, dose calculations based on the I allowable leakage rate must demonstrate that the annual dose equivalent to any real I individual who is located at the boundary or outside the controlled area does not exceed I the limits given in 10 CFR 72.104(a). I  ;

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c. After a design-basis accident, dose calculations based on the allowable leakage rate l l must demonstrate that an individual at the boundary or outside the controlled area does I not receive a dose that exceeds the limits given in 10 CFR 72.106(b). I i

d. The applicant's leakage analysis must demonstrate that an inert atmosphere will be I maintained within the cask during the storage lifetime. I

4. The applicant should describe the proposed monitoring capability and/or surveillance plans for mechanical closure seals. In instances involving welded closures, the staff has l

previously accepted that no closure monitoring system is required. This practice is  !

consistent with the fact that other welded joints in the confinement system are not monitored. However, the lack of a closure monitoring system has typically been coupled with a periodic surveillance program that would enable the licensee to take timely and  !

appropriate corrective actions to maintain safe storage conditions if closure degradation I occurred. I To show compliance with 10 CFR Part 72.122(h)(4.), cask vendors have proposed, and the I staff has accepted, routine surveillance programs and active instrumentation to meet the I continuous monitoring requirements. The reviewer should r,ote that some DCSS designs l l may contain a component or feature whose continued performance over the licensing period has not been demonstrated to staff with a sufficient level of confidence. Therefore j

the staff may determine that active monitoring instrumentation is required to provide for the I detection of component degradation or failure. This particularly applies to components j whose failure immediately affects or threatens public health and safety, in some cases the '

vendor or staff in order to demonstrate compliance with 10 CFR Part 72.122(h)(4), may propose a technical specification requiring such instrumentation as part of the initial use of a cask system. After inltial use, and if warranted and approved by staff, such instrumentation may be discontinued or modified.

5. The cask must provide a non-reactive environment to protect fuel assemblies against fuel i cladding degradation, which might otherwise lead to gross rupture.8 Measures for providing  ;

a non reactive environment within the confinement cask typically include drying, evacuating air and water vapor, and backfillir.g with a non-reactive cover gas (such as helium). For dry storage conditions, experimental data have not demonstrated an acceptably low oxidation  !

rate for UO, spent fuel, over the 20-year licensing period, to permit safe storage in an air atmosphere. Therefore, to reduce the potential for fuel oxidation and subsequent cladding failure, an inert atmosphere (e.g., helium cover gas) has been used for storing UO, spent fuelin a dry environment. (See Chapter 8 of this SRP for more detailed information on the l cover gas filling process.) Note that other fuel types, such as graphite fuels for the high-temperature gas-cooled reactors (HTGRs), may not exhibit the same oxidation reactions as UO, fuels and, therefore, may not require an inert atmosphere. Applicants proposing to use atmospheres other than inert gas should discuss how the fuel and cladding will be protected from oxidation.

Attachment to ISG-5 Revision 1 4

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V. Review Procedures I

1. Confinement Design Characteristics  :

a. Design Criteria l Review the principal design criteria presented in SAR Section 2, as well as any additional detail  !

provided in SAR Chapter 7. l

b. Design Features >

Review the general description of the cask presented in SAR Section 1, as well as any additionalinformation provided in SAR Section 7. All drawings, figures, and tables describing confinement features must be sufficiently detailed to stand alone.

Verify that the applicant has clearly identified the confinement boundaries. This identification  !

should include the confinement vessel; its penetrations, valves, seals, welds, and closure  !

devices; and corresponding information concerning the redundant sealing.

Verify that the design and procedures provide for drying and evacuation of the cask interior as  ;

part of the loading operations, and that the design is acceptable for the pressures that may be experienced during these operations, i l

Verify that, on completion of cask loading, the gas till of the cask interior is at a pressure level l that is expected to maintain a ncn-reactive environment for at least the 20 year storage life of l the cask interior under both normal and off-normal conditions and events. This verification can l

- include pressure testing, seal monitoring, and maintenance for casks with seals that are not welded if these are included in chapter 12 as conditions of use. The NRC has previously accepted specification of an overpressure of approximately 14 kilopascals (-2 psig) and cask leak testing as conditions of use for satisfying this requirement, in addition,if conditions of use require routine inspection of seals by the pressure testing of the cask interior, the cask fill pressure may be linked to that activity.

Coordinate with the structural reviewer (Chapter 3 of this SRP) to ensure that the applicant has provided proper specifications for all welds and, if applicable, that the bolt torque for closure  ;

devices is adequate and properly specified.

If applicable, assess the seals used to provide closure. Because of the performance requirements over the 20-year license period, evaluate the potential for deterioration. The NRC staff has previously accepted only metallic seals for the primary confinement. Coordinate with the thermal reviewers (Chapter 4 of this SRP) to ensure that the operational temperature range for the seals,'specified by the manufacturer, will not be exceeded.

2. Confinement Monitoring Capability

. The NRC staff has found that casks closed entirely by welding do not require seal monitoring.

However, for casks with bolted closures, the staff has found that a seal monitoring system has been needed in order to adequately demonstrate that seals can function and maintain a helium 5 Attachment to ISG-5 Revision 1

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atmosphere in the cask for the 20-year license period. A seal monitoring system combined with periodic surveillance enables the licensee to determine when to take corrective action to maintain safe storage conditions. (Note that some fuel designs may not require an inert atmosphere in the cask. In such designs, a periodic surveillance program to check seal leak tightness may be appropriate.)

Although the details of the monitoring system may vary, the general design approach has been to pressurize the region between the redundant seals, with a non-reactive gas, to a pressure greater than that of the cask cavity and the atmosphere. The monitoring system is leakage I tested to the same leak rate as the confinement boundary. Installed instrumentation is routine;y I checked per surveillance requirements. A decrease in pressure between these seals indicates l ,

that the non-reactive gas is leaking either into the cask cavity or out to the atmosphere. For I normal operations, radioactive material should not be able to leak to the atmosphere; hence this I design allows for detecting a faulty seal without radiological consequence. Note that the i volume between the redundant seals should be pressurized using a non-reactive gas, thereby preventing contamination of the interior cover gas.

The staff has accepted monitoring systems as not important to safety and classified as l Category B under the guidelines of NUREG/CR-6407d. Although its function is to monitor I confinement sealintegrity, failure of the monitoring system alone does not result in a gross I release of radioactive material. Consequently, the monitoring system for bolted closures need not be designed to the same requirements as the confinement boundary (i.e., ASME Section Ill, Subsections NB or NC).

Dependant on the monitoring system design, there could be a lag time before the monitoring i system indicates a postulated degraded sealleakage condition. Degraded sealleakage is I leakage greater than the tested rate that is not identified within a few monitoring system I surveillance cycles. The occurrence of a degraded seal without detection is considered a I

" latent" condition and should be presumed to exist concurrently with other off-normal and I design-basis events (see SRP section 2, paragraph V.2.b.). Note that once the degraded seal I condition is detected, the cask user will initiate corrective actions. I For the off normal case, the monitoring system boundary remains intact and this condition I would be bounded by the off normal analysis, if the monitoring system would not maintain I integrity under design-basis accident conditions, additional safety analysis may be necessary. I The staff recognizes that the possibility of a degraded seal condition is small and that the I possibility of a degraded seal condition concurrent with (A design basis event that breaches the i monitoring system pressure boundary is very remote. However, these probabilities have not i been quantified. To address this concern, the staff accepts a demonstration that the probability I of occurrence of a latent, degra91 seal, condition concurrent with a design basis event that I 4

breaches the monitoring system coundary is acceptably low (e.g. less than 1 X 10 per year ). I Alternatively, the staff accepts a demonstration that the dose consequences of this event are I within the limits of 10 CFR 72.106(b). I Examine the specified pressure of the gas in the monitored region to verify that it is higher than both the cask cavity and the atmosphere. Coordinate with the structural and thermal reviewers (Chapters 3 and 4 of this SRP) to verify the pressure in the cask cavity.

Attachment to ISG-5 Revision 1 6

Review the applicant's analysis to verify that the total volume of gas in'the seal monitoring system is such that normal seal leakage will not cause all of the gas to escape over the lifetime of the cask. In determining the proposed maximum leakage rate, the applicant should consider the volume between the redundant seats of the confinement cask, the minimum pressure to be maintained, and the length of the proposed routine recharge cycle. The applicant should then specify the leakage rate as an acceptance test criterion in SAR Section 9, even though the actual leakage rate of the seals is expected to be significantly lower.

For redundant seal welded closures, ensure that the applicant has provided adequate justification that the seal welds have been sufficiently tested and inspected to ensure that the weld will behave similarly to the adjacent parent material of the cask. Any inert gas should not leak or diffuse through the weld and cask material in excess of the design leak rate.

Verify that any leakage test, monitoring, or surveillance conditions are apprcpriately specified in SAR Sections 9 and 11, the license, and/or the Certificate of Compliance.

3. Nuclides with Potential for Release The NRC staff has determined that, as a minimum, the fractions of radioactive materials I available for release from spent fuel, provided in Table 7-1 for pressurized-water reactor (PWR) I fuel and boiling-water reactor (BWR) fuel for normal, anticipated occurrences (off-normal), and I accident conditions, should be used in the confinement analysis to demonstrate compliance I with 10 CFR Part 72. These fractions account for radionuclides trapped in the fuel matrix and I radionuclides that exist in a chemical or physical form that is not releasable to the environment I under credible normal, off-normal, and accident conditions. Other release fractions may be i used in the analysis provided the applicant properly justifies the basis for their usage. For I example, the staff has accepted, with adequate justification, reduction of the mass fraction of I fuel fines that can be released from the cask. I I

The staff has accepted the following rod breakage fractions for the confinement evaluations: I I

1% for normal conditions i 10% for off normal conditions I 100% for design basis accident and extreme natural phenomena l i

For the source term, the NRC staff has accepted, as a minimum for the analysis, the activity I from the Co" in the crud, the activity from iodine, fission products that contribute greater taan 1 0.1% of design basis fuel activity, and actinide activity that contributes greater than 0.01% of I the design basis activity. In some cases, the applicant may have to consider additional I radioactive nuclides depending upon the specific analysis. The total activity of the design basis I fuel should be based on the cask design loading that yields the bounding radionuclide inventory I (considering initial enrichment, bumup, and cool time). I 1

l l

l 7 Attachment to ISG 5 Revision 1

1 l

l

\ .

l Table 7.1* I Fractions Available for Release ~ l PWR AND BWR FUEL l

Norrnal and Off- Hypothetical Accident normal Conditions C.onditions i Fraction of gases released due t a cladding breach, fat 0'3 0.3 Fraction of volatiles released due to a cladding breach, fut 2 X 10 4 2 X 10 4 Mass fraction of fuel released as fines due to cladding breach, f, 3 X 10 5 3 X 105 Fraction of crud that spalls off cladding, fe 0.15' 1.0" Values in this table are taken from NUREG/CR-6487 5. 1 I

Except for"Co, only failed fuel rods contribute significantly to the release. Total fraction of I radionuclides available for release must be multiplied by the fraction of fuel rods assumed to have l 1 failed. I I

t in accordance with NUREG/CR-6487, gases species include H-3,1-129, Kr-81, Kr-85, and Xe-127; I volatile species include Cs-134, Cs-135, Cs-137, Ru-103, Ru-106, Sr-89, and Sr-90. I I

1. The source of radioactivity in crud is "Co on fuel rods. At the time of discharge from the reactor, I the specific activity, S,, is estimated to be 140 pCi/cm' for PWRs and 1254 pCi/cm2 for BWRs. I Total"Co activity is this estimate times the total surface area of all rods in the cask'. Decay of I "Co to determine activity at the minimum time before loading is acceptable. l I

The quantities of radioactive nuclides are often presented in SAR Section 5, since they are i generally determined during the evaluation of gamma and neutron source terms in the shielding I analysis. Coordinate with the shielding review (Chapter 5 of this SRP) to verify that the I applicant has adequately developed the source tenn. I I

It is important to recognize that design basis normal or accident conditions resulting in I confinement boundary failure are not acceptable. Preservation of the confinement boundary I during design basis conditions is confirmed by the structural analysis. The confinement I analyses demonstrate that, at the measured leakage rates, and assumed nominal I meteorological conditions, the requirements of 10 CFR 72.104(a) and 10 CFR 72.106(b) can be I met. Each ISFSI, wether it is a site specific or a general license, is also required to have a site I specific confinement analysis and dose assessment to demonstrate compliance with these I regulations. I Attachment to ISG 5 Revision 1 8

4 4

4. Confinement Analysis Review the applicant's confinement analysis and the resulting doses for the normal, off-normal, I and accident conditions at the controlled area boundary. I I

The analysis typically includes the following common elements: I I

. 8 calculation of the specific activity (e.g. Ci/cm ) for each radioactive isotope in the casi cerity I based on rod breakage fractions, release fractions, isotopic inventory, and cavity free volume -

1 I

using the tested leak rate and conditions during testing as input parameters, calculation of I the adjusted maximum seal leakage rates (Cm8/sec) under normal, off-normal, and I hypothetical accident conditions (e.g. temperatures and pressures) I l

calculation of isotope specific leak rates (Q,- Ci/sec) by multiplying the isotope specific I activity by the maximum seal leakage rates for normal, off-normal, and accident conditions i I

determination of doses to the whole body, thyroid, other critical organs, lens of the eye, and I skin from inhalation and immersion exposures at the controlled area boundary (considering I atmospheric dispersion factors - X/Q) I l

The applicant should specify maximum allowable *as tested" seal leakage rates as a Technical l Specification, as discussed in Chapter 12. Guidance on the calculations of the specific activity I for each isotope in the cask and the maximum allowable helium seal leakage rates for normal, I off-normal, and accident conditions can be found in NUREG/CR-6487 and ANSI N14.5-1997. l The minimum distance between the casks and the controlled area boundary is generally also a - I design criterion; however,10 CFR Part 72 requires this distance to be at least 100 meters from I the ISFSI. l l

For the dose calculations, the staff has accepted the use of either an adult breathing rate (BR) I of I d 8 7 2.5x10dm /s, as specified in Regulatory Guide 1.109 , or a worker breathing rate of I 8

3.3x10 m /s, as specified in EPA Guidance Report No.11.8 The dose conversion factors I (DCF) in EPA Guidance Report No.11 for the whole body, critical organs, and thyroid doses I from inhalation should be used in the calculation. The bounding DCFs from EPA Report No.11 I should be used for each isotope unless the applicant justifies an attemate value. No weighting i or normalization of the dose conversion factors is accepted by the staff. For each isotope, the I committed effective dose equivalent (CEDE,- for the intemal whole body dose) or the I committed dose equivalent (CDE,- for the intemal organ dose) are calculated as follows: l l

CEDE, or CDE,(in mrem per year for normal /offnormal or mrom per accident) l

= Q,

• DCF,

• X/ O

• B-Rate

• Duration

• conversion factor

• I l

Dhe conversion factor, it required. converts the input units into the desired form. e.g. mrem / year. I 9 Attachment to ISG 5 Revision 1

n

, For the contributions to the whole body, thyroid, critical organs, and skin doses from immersion i (extemal) exposure, the DCFs in EPA Guidance Report No.12'should be used. Again, no I

j. weighting or normalization of the dose conversion factors is accepted by the staff. I

! l l The deep dose equivalent (DDE,- for the external whole body) and the shallow dose equivalent l l (SDE,- for the skin dose) are calculated as follows: l l

DDE, or SDE, (in mrom per year for normal /offnormal or mrem per accident) I

= Q,

• DCF,

• X/ Q

• Duration

• conversion factor

• l l

l The total effective dose equivalent, TEDE = E CEDE, + E DDE, I i

l For a given organ, the total organ dose equivalent, TODE = E CDE, + E DDE, I I

The total skin dose equivalent SDE = E SDE, I I

Compliance with the lens dose equivalent (LDE) limit is achieved if the sum of the SDE and the i

, TEDE do not exceed 0.15 Sv (15 rem). This approach is consistent with guidance in ICRP- l 26".

I l I in general, the staff evaluates analyses for normal, off-normal, and accident conditions. l l-

a. Normal Conditions l I

For normal conditions, a bounding exposure duration assumes that an individual is present I at the controlled area boundary for one full year (8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />). An alternative exposure I duration may be considered by the staff if the applicant provides justification. I I

Because any potential release, resulting from seal leakage, would typically occur over a i substantial pedod of time, the staff accepts (for applications for certificates) calculation of I the atmospheric dispersion factors (X/ Q) according to Regulatory Guide 1.145" assuming I D-stability diffusion and a wind speed of 5 m/s. I l

For the likely case of an ISFSI with multiple casks, the doses need to be assessed for a I hypothetical array of casks during normal conditions. Therefore, the staff anticipates that I

' the resulting doses from a single cask will be a small fraction of the limits prescribed in I 10 CFR 72.104(a) to accommodate the array and the external direct dose. I 1

Note: If the region between redundant, confinement boundary, mechanical seals is I maintained at a pressure greater than the cask cavity, the monitoring system boundaries I are tested to a leakage rate equal to the confinement boundary, and the pressure is 1 routinely checked and the instrumentation is verified to be operable in accordance with a l Technical Specification Surveillance Requirement, the staff has accepted that no discemible I leakage is credible. Therefore, calculations of dose to the whole body, thyroid, and critical I organs at the controlled area boundary from atmospheric releases during normal conditions I would not be required for normal conditions. I I

I Attachment to ISG 5 Revision 1 10

! =

l

b. Off-normal Conditions l I I

For off-normal conditions, the bounding exposure duration and atmospheric dispersion I factors (X/ Q) are the same as those discussed above for normal conditions. I

, I i To demonstrate compliance with 10 CFR 72.104(a), the staff accepts whole body, thyroid, I and critical organ dose calculations for releases from a single cask. However, the dose I contribution from cask leakage should also be a fraction of the limits specified in l l 10 CFR 72.104(a) since the doses from other radiation sources are added to this I contribution. I I

l c. Accident Conditions l I

For hypothetical accident conditions, the duration of the release is assumed to be 30 days 1 (720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />). A bounding exposure duration assumes that an individualis also present at l l the controlled area boundary for 30 days. This time period is the same as that used to I demonstrate comp,iance with 10 CFR 100 for reactor facilities licensed per 10 CFR 50 and I provides good defense in depth since recovery actions to limit releases are not expected to I exceed 30 days. I I

For hypothetical accidents conditions, the staff has accepted calculation of the atmospheric I dispersion factors (X/ Q) of Regulatory Guide 1.145 or Regulatory Guide 1.25s2 on the basis I of F-stability diffusion, and a wind speed of 1 m/s. l l

To demonstrate compliance with 10 CFR 72.106(b), the staff accepts whole body, thyroid, I critical organ, and skin dose calculations for releases of radionuclides from a single cask. l l

S. SupplementalInformation Ensure that all supportive information or documentation has been provided or is readily available. This includes, but is not limited to, justification of assumptions or analytical procedures, test results, photographs, computer program descriptions, input and output, and applicable pages from referenced documents. Reviewers should request any additional information needed to complete the review.

VI. Evaluation Findings Review the 10 CFR Part 72 acceptance criteria and provide a summary statement for each.

These statements should be similar to the following model:

e Section(s) - of the SAR describe (s) confinement structures, systems, and components (SSCs) important to safety in sufficient detail in to permit evaluation of their effectiveness.

e The design of the [ cask designation) adequately protects the spent fuel cladding against degradation that might otherwise lead to gross ruptures. Section 4 of the safety evaluation report (SER) discusses the relevant temperature considerations.

11 Attachment to ISG-5 Revision 1 l

l l

e The design of the [ cask designation) provides redundant sealing of the confinement system closure joints by .

e The confinement system is monitored with a monitoring system as discussed above (if applicable). No instrumentation is required to remain operational under accident conditions.

e The quantity of radioactive nuclides postulated to be released to the environment has been assessed as discussed above, in Section 10 of the SER, the dose from these releases will be added to the direct dose to show that the [ cask designation) satisfies the l regulatory requirements of 10 CFR 72.104(a) and 10 CFR 72.106(b). l o The cask confinement system has been evaluated [by appropriate tests or by other i means acceptable to the Commission) to demonstrate that it will reasonably maintain confinement of radioactive material under normal, off-normal, and credible accident  ;

conditions. I e The staff concludes that the design of the confinement system of the [ cask designation) is in compliance with 10 CFR Part 72 and that the applicable design and acceptance criteria have been satisfied. The evaluation of the confinement system design provides reasonable assurance that the [ cask designation) will allow safe storage of spent fuel.

This finding is reached on the basis of a review that considered the regulation itself,  !

appropriate regulatory guides, applicable codes and star.dards, the applicant's analysis and the staff's confirmatory analysis, and accepted engineering practices.

Vll. References

1. American Society of Mechanical Engineers,"ASME Boiler and Pressure Vessel Code,"

Section lil, Subsections NB and NC.

2.- American National Standards Institute, institute for Nuclear Materials Management, "American National Standard for Leakage Tests on Packages for Shipment of Radioactive Materials," ANSI N14.5,1997. l

3. Pacific Northwest Laboratory, Evaluation of Cover Gas impurities and Their Effects on the Dry Storage of LWR Spent Fuel," PNL-6365, November 1987.

4. Laboratory " Classification of Transportation Packa ing and Idahgpent Dry Fuel StoragNational tem C Engineeri p nents According to importance to Safe NUREG/CR-6407, INEL /0551, February 1996.

5. U.S. Nuclear Regulatory Commission," Containment Analysis for Type B Packages i Used to Transport Vanous Contents," NUREG/CR-6487, November 1996. I

6. R.P. Sandoval, et al., Sandia National Laboratories. " Estimate of CRUD Contribution to I Shipping Cask Containment Requirements," SAND 88-1358, TTC-0811, UC-71, 1 January 1991. l l Attachment to ISG 5 Revision 1 12

7. U.S. Nuclear Regulatory Commission, " Calculations of Annual' Doses to Man from i Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with I 10 CFR Part 50, Appendix 1," Regulatory Guide 1.109, October 1977. I

8. U.S. Environmental Protection Agency, Federal Guidance Report No.11, " Limiting i Values of Radionuclide intake and Air Concentration and Dose Conversion Factors for I i Inhalation, Submersion, and Ingestion," DE89-011065,1988. I

9. U.S. Environmental Protection Agency," Federal Guidance Report No.12: External i Exposure to Radiouclides in Air, Water, and Soil," EPA 402-R-93-081, September 1993. I

10. International Commission on Radiation Protection, " Statement from the1980 Meeting of I the ICRP," ICRP Publication 26, Pergammon Press, New York, NewYork,1980. I j

11. U.S. Nuclear Regulatory Commistion, " Atmospheric Dispersion Models for Potential I Accident Consequence Assessments at Nuclear Power Plants," Regulatory Guide 1 1.145, February 1989. I !

12. U.S. Nuclear Regulatory Commission,"Assum tions Used for Evaluating the Potential I Radiological Consequences of a Fuel Handli Accident in the Fuel Handling and i Storage Facility for Boiling and Pressurized ter Reactors," Regulatory Guide 1.25, 1 March 1972. I 13 Attachment to ISG-5 Revision 1

Spent Fuel Project Office Interim Staff Guidance -6 issue: Estsblishing minimum initial enrichment for the bounding design basis fuel assembly (s).

The Standard Review Plan, NUREG-1536, Chapter 5,Section V,2 recommends that "the applicant calculate the source term on the basis of the fuel that will actually provide the bounding source term," and states that the applicant should, "either specify the minimum initial enrichment or establish the specific source terms as operating controls and limits for cask use."

A specified source term is difficult for most cask users to determine and for inspectors to verify.

The specification of a minimum initial enrichment is a more straightforward basis for defining the allowed contents. The specification should bound all assemblies proposed for the casks in the application. Specific limits are needed for inclusion in the Certificate of Compliance.

Lower enriched fuel irradiated to the same bumup as higher enriched fuel produces a higher ,

neutron source. Sometimes fuel assemblies are driven to bumups beyond the value normally i expected for the given enrichment. According to the U.S. Department of Energy's Characteristic Data Base, the lower enrichment for fuel bumed to 45,000 mwd /MTU is about 3.3%. The neutron source for an initial enrichment of 3.3% is expected to be 70% higher than the neutron source for 4.05% enriched fuel.  !

Recommendation:

Rewrite the last sentence of paragraph 1 in Chapter 5,Section V,2 (page 5-3) to read

" Consequently, the SAR should specify the minimum initial enrichment as an operating control and limit for cask use, orjustify the use of a neutron source term, in the shielding analysis, that

.specifically pounds the neutron sources for fuel assemblies to be placed in the cask. Absent adequate justification acceptable to the staff, the SAR should not attempt to establish specific source terms as operating controls and limits for cask use."

l Approved William F. Kane Date l .

ISG4 i

k

Spent Fuel Project O# ice interim Sta# Guidance -7 lasue: Potential Generic issue Concerning Cask Heat Transfer in a Transportation Accident Staff raised two major ist.ues conoeming the adverse effects of fission gases to the gas-mixture thermal conductivity in a spent fuel canister in a post accident environment. The two major concorr.s were: (1) the reduction of the thermal conductivity of the canister gas by the mixing of fission gases expelled from failed fuel pins and (2) the resultant temperature and pressure rise within the can5ter. Since the fission gas is typically of a lower conductivity than the cover gas, its mixing with the cover gas tends to lessen the thermal performance of the mixture.

Furthermore, since additional gas is introduced into the canister, the internal pressure will increase as will the bulk temperature of the gas. The combination of these phenomena, if they are great enough, would pose a containment issue if the design basis pressure is exceeded.

The first step in resolving this issue involved reviewing NUREG/CR-5273, Vol. 4, describing a suitable method to predict the change in gas conductivities as a function of increasing fission gas concentrations. Although the fission gases are a collection ofiodine (I), krypton (Kr), and xenon (Xe), as a conservatism, all fission gases generated were assumed to be the heaviest gas present, which is usually xenon. This reduced the problem to a binary mixture and simplified the calculation of the gas mixture properties. The practicability of this assumption is that the heavier gases exhibit a lower thermal conductivity for standard temperature and pressure conditions. Two separate methods were used to verify the reduction in the thermal conductivity with respect to temperature and concentration. Of the methods selected, the mixture properties were determined using the mole fraction of the mixture and a complex function of viscosities and molecular weights. These estimation techniques are based on the kinetic theory of gases.

The VSC-24 was selected for modeling as being a general cask representative of those in operation today. The sealed VSC-24 canister is backfilled with 1 atm of helium (He). In accordance with the Nuclear Regulatory Commission-accepted practice of assuming that all fuel rods fail during an accident with 30% of the fission gases entering the canister, the appropriate thermal properties of the He-Xe mixture were evaluated as a function of bulk temperature at standard pressure. These thermal conductivities were compared with thoso of pure He at the same standard pressure. This revealed a general percentage reduction in the mixture conductivites over a wide range of temperatures. In this study, the percentage reduction was a sizable 70%. Applying this reduction to the cask model, a parametric study of the cask was performed to determine the impact of the changing conductivity on the inner-canister component temperatures. The bulk temperature of the gaseous region was also evaluated to gauge the pressure increase within the canister. The results of this analysis are presented in the table below.

c ISG 7

e 4

e

• Temperature Temperature w/70%

Modeled Case tuoriginalThermal Reduced Therr.d Temperature Laosten ConductMey[R] C - M ai(R]

12 hr Memmum s47 EBB Suk Gas Therme! Lead Trennent 1236 1272 Peak Clad Vsc h h 973 sS7 a d Gas Tm cook 1348 1371 Peak Cted As can be seen from the table, the large decrease in gas conductivity elevated the peak cladding temperature by less than 3% and increased the bulk gas temperature by less than 3%.

The resultant pressure increase accounting for the higher gas tempercture was a maximum of 4 psi which is less than a 10% increase in the reported value. This pressure increase is a very conservative estimate based on the assumption that the initial gas temperature was that of the canister shell.

Recommendation:

Change the Standard Review Plan for Transportation Packages for Spent Nuclear Fuel, NUREG 1617, and the Standard Review Plan for Dry Cask Storage Systems, NUREG-1536 as follows:

Under the conditions where any of the cask component temperatures are close (within 5%) to their limiting values during an accident or the MNOP is within 10%

of its design basis pressure, or any other special conditions, the applicant should consider, by analysis, the potential impact of the fission gas in the canister to the cask component temperature limits and the cask intamal pressurization.

Approved William F. Kane Date t

ISG 7 2

4 Spent Fuel Project Office interim Staff Guidance-8 issue: Limited Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks Discussion:

When fuel is irradiated in a reactor, the reactivity of the fuel decreases. This reduction of reactivity with burnup is caused by the change in fbsile content of the fuel (i.e., bumup of U-235 and production of Pu-239 and other fissile actinides), the production of actinide neutron absorbers, and the production of fission-product neutron absorbers. Until now, criticality safety analyses for spent fuel casks, including storage, transport, and dual-purpose casks, were performed under the assumption that the fuel was unirradiated. This " fresh fuel" assumption was used as a bounding condition because of unresolved issues over the technical basis and m0thods for including credit for fuel bumup in the criticality analysis of spent-fuel casks.

The U.S. Department of Energy (DOE) has been working on the development of a topical report that proposes e method for taking bumup credit in casks for transporting and storing spent fuel from pressurizM watu reactors (PWRs). DOE's proposal has been submitted to the U.S.

Nuclear Regulebry Commission (NRC) and has gone through two cycles of revisions based on NRC's review ar d comment. Based on the technical information provided in DOE's topical report, with its supporting technice.! reports, and information available from other sources, both foreign and domestic, the staff has now found sufficient basis to approve a ljmited level and scope of bumup credit.while it pursues the development of a more complete basis for more comprehensive bumup credit.

As just' tied through the review of additional supporting data and analysis, the staff will issue revised interim guidance to reflect the evolution of approved methods for greater levels of burnup credit. Future revisions of the Standard Review Plans (NUREG-1617, NUREG-1536, and NUREG-1537) will incorporate or reference the current guidance, as appropriate.

Background:

Existina NRC Uses of Burnuo Credit in Soent Fuel Storaae The Office of Nuclear Reactor Regulation (NRR) has long allowed the use of bumup credit in the borated spent fuel storage pools at PWR plants. This is based in part on the established ability of licensees to predict the core burnup behavior over hundreds of reactor years of operation. Additional safety assuranc6 is based on application of the double contingency principle as defined in ANSI /ANS-8.1-1983, and in Title 10, Code of Federal Regulations (10 CFR), Section 72.124(a), which re yires two unlittely, independent, concurrent events to produce a criticality accident. For ext v.ple, if soluble boron is normally present in the spent fuel pool wa4r, the tess of soluble boron ; considered as one accident condition and a second concurrent acc' dent need not be ass Jmed. Alternatively, credit for the presence of soluble boron in PWR pools may be assumed in evaluating other accident conditions such as the misloading of fresh fuel assemblies into racks restricted to irradiated fuel. Typically, there is sufficient soluble boron in PWR poots to maintain at least a 5% suberiticality margin even if an entire bumup-dependent storage rack were misloaded with fresh fuel assemblies.

4 ISG-8 2

- As noted by DOE and others, burnup credit calculations can also be found in the applicants' safety analysis reports (SARs) for two approved single-purpose dry storage casks for PWR spent fuel (i.e., NUHOMS-24P and VSC-24). There, the applicants performed burnup credit calculations in evaluating hypothetical underboration events during wet loading or unloading of the dry storage casks. However, the staff's safety evaluation reports for those cases used the

" fresh fuel" analysis assumption in combination with credit for boron in the water, Boron credit was made possible by creating in the license or certificate a Technical Specification requiring two independent verification controls on soluble boron concentration during wet loading and unloading operations. This satisfied the double-contingency criterion of 10 CFR 72.124(a) while obviating consideration of loss-of-boron events in the review under 10 CFR Part 72.

Although triple contingencies are not directly considered in the staff's evaluations for 10 CFR Part 72 reviews, applicants have chosen to retain the burnup credit calculations in t5eir SARs in order to help address plant-specific requirements under 10 CFR 50.59 for use of the casks at reactor spent-fuel pools. Those burnup credit calculations are acknowledged by NRC's Office of Nuclear Material Ssafety and Safeguards as illustrating an additional safety margin, of uncertain magnitude, that goes beyond the regulatory requirements of 10 CFR Part 72.

Therefore, the burnup credit analyses for wet loading and unloading of dry storage casks can be viewed as technically consistent with NRR's applications of bumup credit at PWR spent-fuel storage pools in that both are used, on a risk informed basis, only in addressing the criticality safety margins for extreme hypothetical events that are considered extremely unlikely or incredible.

The need for considering bumup credit after drying and closure of casks has generally been avoided in 10 CFR Part 72 storage applications by showing that fresh-water ingress into sealed dry storage casks is not credible. Specifically, the double-contingency criterion is satisfied by showing that water ingress into a storage cask would require both a flooding event and an accident that would cause seal failure. On the other hand, transportation regulations under 10 CFR Part 71 include explicit requirements for assuming fresh-water inleakage in the criticality analysis of transport packages for fissile materials. Sections 6.5.4 and 6.5.5 in NUREG-1617, " Standard Review Plan for Transportation Packages for Spent Nuclear Fuel,"

further discuss the water-inleakage considerations for spent-fuel evaluations under 10 CFR Part 71.

Bumuo Credit in Other Countries Several regulatory bodies outside the U.S. have allowed various uses of burnup credit in wet storage and handling operations, and also in reprocessing. However, transportation uses of bumup credit have been granted to-date only in France. The French reprocessing program has developed an extensive set of proprietary validation data to support the limited credit needed for shipping modem PWR fuels with higher initial enrichments in the existing fleet of casks. Safety authorities in the United Kingdom and Japan are now working toward similar uses of burnup

- credit in transport packages. As noted above, validation benchmarks provided from French and other foreign or proprietary sources may be considered as part of the expanded technical basis needed for future NRC approval of greater levels of bumup credit.

l

l l

ISG-8 3

! Recommendation:

When performing criticality safety analyses for spent-fuel casks, limitW oartial credit for the reactivity effects of fuel burnup may be taken as follows:

Method:

Use the method described in DOE's report entitled, " Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages," DOE /RW-0472, Rev. 2, except as otherwise specified below.

Assume fuel burnup is 50% of the verified and adjusted burnup level from plant records.

Scope:

Applies to intact commercial PWR fuel only.

Includes actinide effects only (change in fissile content and actinide neutron absorbers).

Ranae of Aeolication:

Covers UO, fuel with nominal initial enrichments up to 4.0 weight percent U-235.

Covers assembly-average burnup levels up to 45 GWD/MTU.

Establishina the Burnuo Value:

Use the reactor-record assembly burnup, as adjusted, when confirmed by a direct assembly measurement, performed in the storage pool or loading facility, that is calibrated to the reactor records for a representative set of measured assemblies.

Measurement confirmation must be within a 95% confidence interval based on the measurement uncertainty. Adjust the burnup-record value by reducing the record value by the combined uncertainties in the records and the measurement. Bumup measurement may be based on gamma emissions of the Cs-137 isotope. The requirement for bumup verification measurements is consistent with Regulatory Guide 3.71 and is justified in part by numerous licensee-reported events involving failure of administrative controls on plant spent-fuel records and the selection, handling, and placement of spent fuelin the storage pools at reactors.

Approved aw JI6

l. E. William BracM Date j

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Spent Fuel Project Office Interim Staff Guidance - 9 lssue: Storage of Pressurized Water Reactor (PWR) Fuel Assembly Integral Components Discussion:

The Standard Review Plan does not provide explicit guidance on the storage of rod cluster control assemblies, burnable poison (rod) assemblies, thimble plugging assemblies, and primary and secondary source assemblies as materials associated with the storage of spent fuel assemblies. While control rods are mentioned in the Standard Review Plan as possible contents, specific information and guidance is lacking. l Regulatory Basis:

Title 10, Code of Federal Regulations (10 CFR), Section 72.3, " Definitions," states, " Spent Nuclear Fuel or Spent Fuelmeans fuel that has been withdrawn from a nuclear reactor l following irradiation, has undergone at least one year's decay since being used as a source of energy in a power reactor, and has not been chemically separated into its constituent elements by reprocessing. Spent fuelincludes the special nuclear material, byproduct material, source i material, and other radioactive materials associated with fuel assemblies."

Technical Review Guidance:

Standr d Review Plan for Dry Cask Storage Systems, NUREG-1536, Chapter 2," Principal Desig' Oriteria,"Section IV.2.a states,"The applicant should define the range and types of spent it el or other radioactive materials that the DCSS [ dry cask storage system] is designed to store . . . For DCSSs that will be used to store radioactive materials other than spent fuel, that is, activated components associated with a spent fuel assembly (e.g., control rods, BWR fuel channels), the applicant should specify the types and amounts of radionuclides, heat generation, and the relevant source strengths and radiation energy spectra permitted for storage in the DCSS" [page 2-4].

Recommendation:

Rod cluster control assemblies are materials that may be associated with the storage of spent fuel. When used, burnable poison (rod) assemblies, thimble plugging assemblies, and neutron source assemblies, are integral components of a PWR fuel assembly and are also associated with the storage of spent fuel. These components may be approved for storage in a DCSS if the applicant submits information and the safety / technical justification for the proposed DCSS contents for staff review and approval. The staff should incorporate this information as proposed contents in the license, certificate of compliance, or technical specification.'

'It should be noted that if a heense, certificate of comphance, or technical specification has already been issued and does not specifically allow storage of these components, there is no other regulatory rehef to allow new contents other than an amendment. Therefore, the apphcant should seek to amend its hcense, certificate of comphance, or technical specification.

L,*L ISG-9 2 Specifically, the technical review staff should consider the following in its review:

• The design bases source term (radiological and thermal) should be based on a saturation value for activation of cobalt impurities or on cobalt activation from a specified maximum burn-up and minimum cool time. The reviewer should consider other activation products, as appropriate.

• The effects of gas generation must be considered in the design pressure for the cask, including (1) the release of gas from additional components and (2) the volume occupied by additional components on the cask internal pressure.

• Additional weight and length of the proposed material must be considered in the -

structural and stability analyses.

• . The thermal analysis must consider (1) the added heat from these components and (2) the effects of heat transfer within and to/from the fuel assembly by the addition or absence of these components. This would ultimately affect the maximum predicted cladding temperature.

• In terms of a criticality evaluation, absent direct physical measurements, no credit

should be assumed for any negative reactivity from residual neutron absorbing material remaining in the control components. A bounding analysis would assume that no control components are present. Credit for water displacement may be taken provided adequate structural integrity and placement under accident conditions is demonstrated.

Also, the reviewer may need to consider the effects of displacing borated water, if applicable.

To the degree that comparable reactor technologies have similar attributes (burnable poisons, bypass flow restricting devices, and sources), the reviewer should similarly accept that the material may be stored in a DCSS as noted above.

The Standard Review Plans (NUREG-1536 at citation noted above, and NUREG-1567, at

' Section 4.4.1.1) should be revised to clearly state those rod cluster control assemblies,

- burnable poison (rod) assemblies, thimble plugging assemblies and primary and secondary source assemblies may be stored in a DCSS, evaluated appropriately, and the proposed contents included in the license, certificate of compliance, or technical specification.

Approved MYf E. William Brach Date'

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4 Spent Fuel Project Office interim Staff Guidance-10 issue: ASME Code Exceptions There is no existing American Society of Mechanical Engineers (ASME) Code for the design and fabrication of spent fuel dry storage casks. Therefore, ASME Code Section ill,is referenced by NUREG-1536, " Standard Review Plan for Dry Cask Storage Systems," as an acceptable standard for the design and fabrication of dry storage casks. However, since dry storage casks are not pressure vessels, ASME Code Section lil, cannot be implemented without allowing some exceptions to its requirements.

Discussion:

Title 10, Code of Regulations (10 CFR) Part 72, was established to provide requirements and criteria for the issuance of licenses to receive, transfer, and possess power reactor spent fuel and other radioactive materials associated with spent fuel storage. However, to date, no industry code or standard exists for the design and fabrication of dry cask storage systems.

Therefore, the industry adopted, and NRC accepted, the use of ASME Code Section 111.

ASME Code Section til was developed to provide guidance to design and fabricate pressure vessels. Since spent fuel dry storage systems are not required to be pressure vessels in every aspect (e.g., they do not require relief valves) not all of the requirements of the code apply or are practical. Therefore, in the past, NRC has allowed specific exceptions to the code for those requirements that were not applicable or practical to implement for fuel dry storage cask systems.

Early spent fuel dry cask storage licenses and certificates of compliance were issued without documenting commitments as to which specific exceptions to ASME Code Section 111, were approved. Poor quality assurance practices during design and fabrication led to additional, and in some cases unacceptable, deviations from the Code without appropriate certificate holder design review or NRC review and approval.

Recommendation':

Commitments to ASME Code Section Ill, with approved exceptions, should be documented in the application and in 10 CFR Part 72 licenses, certificates of compliance, or technical specifications issued by NRC. In addition, to ensure that problems similar to those identified with the use of ASME Code Section lil, do not exist in other areas important to safety, all codes and standards applied to components important to safety and associated deviations as committed to by the applicant, should be included in the license, certificate of compliance, or technical specification, in the event that during fabrication, deviations to codes are required and the deviations do not impact the quality or safety of the component, an exception to the requirements of the license, certificate of compliance, or technical specification may be granted with approval of the NRC.

' Current license or certificate of compliance holders, may request that this process be added to their license or certificate of compliance by seeking an amendment.

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ISG-10 2 Therefore, the staff recommends that applicants implement a condition to a license, certificate I

of compliance, or technical specification to Section 4, " Design Features," which describes the codes committed to and acceptable deviations. The condition or technical specification should describe a process to address one time deviations from the ASME Code that may occur during fabrication. The following is an example:

4.3 Codes and Standards, l

The American Society of Mechanical Engineers (ASME) Boiler anc, Pressure Vesse!

Code, Section 111,1992 Edition with Addenda through 1994, is the governing Code for the storage system.  !

4.3.1 Design Exceptions to Codes, Standards, and Criteria Table 4-1 lists all approved exceptions for the design of the ISFSt.

l 4.3.2 Construction / Fabrication Exceptions to Codes, Standards, and Criteria i Proposed alternatives to ASME Code Section 111,1992 Edition with Addenda through 1994 including exceptions allowed by Section 4.3.1 may be used when i authorized by the Director of the Office of Nuclear Material Safety and Safeguards or designee. The applicant should demonstrate that: ,

l

1. The proposed alternatives would provide an acceptable level of quality and safety, or

2. Compliance with the specified requirements of ASME Code, Section 111,  !

1992 Edition with Addenda through 1994, would result in hardship or  !

unusual difficulty without a compensating increase in the level of quality and safety.

Request for exceptions should be submitted in accordance with 10 CFR 72.4.

LIST OF ASME CODE EXCEPTIONS FOR PLANT ISFSI Table 4-1 Component Reference ASME Code Exception, Justification &

Code Section/ Article Requirement Compensatory Measures Cask specific data to be added as applicable.

Approved E. William Brach Date L _

F Spent Fuel Project Office interim Staff Guidance-11 issue: Storage of Spent Fuel Having Burnups in Excess of 45,000 mwd /MTU Discussion:

l The Standard Review Plan (NUREG-1536) does not address storage of spent fuel having burnups in excess of 45,000 mwd /MTU. For spent fuel having bumups less than 45,000 mwd /MTU, there is sufficient experimental data to support the long-term and short-term temperature limits identified in NUREG-1536. Thus, the staff has generally accepted storage of spent fuel with burnup up to 45,000 mwd /MTU. However, there is limited data to show that the l cladding of spent fuel with burnups greater than 45,000 mwd /MTU will remain undamaged l during the licensing period. Limited information suggests increased cladding oxidation, l increased hoop stresses and changes to fuel pellet integrity with increasing burnup up to about i 60,000 mwd /MTU. These burnup dependent effects could potentially lead to failure of the cladding and dispersal of the fuel during transfer and handling operations.

Regulatory Basis

L In accordance with 10 CFR 72.122 (h)(1), the spent fuel cladding must be protected from degradation that leads to gross ruptures, or the fuel must otherwise be confined so that degradation of the cladding will not impose operational safety problems. Further,10 CFR 72.122(l) requires that the storage system be designed to allow ready retrieval of the spent fuel from the storage system.

l Technical Review Guidance:

! The reviewer should verify that spent fuel having bumups greater than 45,000 mwd /MTU will l remain intact for the licensing period by reviewing the following information supplied by the i

applicant:

• Methodology used to derive the maximum cladding temperature limit. The NUREG-1536 endorsement of the diffusion-controlled cavity growth (DCCG) method to calculate the

maximum cladding temperature limit during dry storage is overly restrictive and relatively i

inflexible. Recently developed literature does not support the use of this model for i zirconium-based materials. Verify that the use of alternative methods are based on l phenomena expected to be encountered under dry storage conditions. Further, verify that l these models are validated with experimental data and that the associated modeling i uncertainties are addressed. An example of an acceptable modelis the Commercial Spent ,

Fuel Management Program model described in PNL-6189 and PNL-6364. I I

Experimentally derived creep data (e.g., time to creep rupture, strain rate under storage  ;

temperature and pressure conditions, etc.). This data will ensure that creep strains are well l below those that would result in cladding damage or excessive deformation. Verify that the i tests were performed using high bumup fuel, or comparable cladding material specimens, under conditions (i.e., temperature, stress and strain rate) that approximate those expected for dry storage. Accelerated tests are acceptable in the event that long duration tests are impractical. However, the effects of creep resulting from different creep and/or deformation j l

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ISG-11 2 f mechanisms, which are likely to occur over different temperature and stress regimes, .

should be considered and evaluated.

A calculation, or measurement, of the cladding hoop stress. This information will aid in establishing both the parameters of the accelerated creep tests, outlined above, and the accuracy of the cladding life prediction. Verify that the stress calculation includes the effects of (1) a reduction of thickness due to cladding oxidation, (2) the initial fuel rod backfill gas pressure, (3) the buildup of fission products in the fuel rod, and (4) the generation of other gases (e.g., helium, etc.) due to effects caused by the irradiation of any internal cladding coatings. Experimental data should be used, as necessary, to verify any assumed values for the oxide thickness or the increase in pressure caused by the buildup of gases.

An estimation of the amount of hydrogen absorbed by the cladding during reactor operation.

This information will ensure that the concentration levels associated with hydride embrittled zirconium alloys are well below those that could significantly reduce the ductility, or overall integrity, of the cladding.

Information about the integrity of the fuel pellets (i.e., post-reactor operation pellet size, estimated size and quantity of pellet fragments, etc.). This will support criticality analyses of  !

potentially reconfigured fuel.

Recommendation:

The staff proposes that Chapter 4 of NUREG-1536 be modified to permit the storage of high burnup fuel assemblies provided that the information described above demonstrates that the cladding will remain intact for the license period. Further, NUREG-1536 should be modified to

, endorse technically defensible methodologies for estimating the maximum cladding temperature limits under dry storage conditions. However, absent this data, high burnup fuel assemblies could be enclosed by approved baskets to (1) confine the fuel such that degradation of the fuel during storage will not pose operational problems with respect to its transportation or removal from storage, and (2) maintain suberiticality based on optimum 3 moderation conditions and no potential for buckling and failure of fuel rods, grid spacers, and end fittings under the hypothetical accident conditions.

Approved l  ;

! E. William Brach Date  ;

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Spent Fuel Project Office interim Staff Guidance-12 issue: Buckling of irradiated Fuel Under Bottom End Drop Conditions Discussion:

Fuel rod buckling analyses under bottom end drop conditions have traditionally been performed to demonstrate integrity of the fuel following a cask drop accident. The methodology described by Lawrence Livermore National Laboratory (LLNL) to analyze the buckling of irradiated spent fuel assembly under a bottom end drop in their report UCID-21246 is a simplified approach. It l assumed that buckling occurred when the fuel rod segment between the bottom two spacer grids reached the Euler buckling limit. The weight of fuel pellets was neglected in the analysis; only the weight of the cladding was considered. Material properties for unirradiated cladding were used. The buckling analysis also neglected the stiffness of the pellets which could have i

been fused or locked to the cladding. It assumed the total weight of the cladding to be on top of the fuel rod segment between the bottom two spacer grids. In addition, it also assumed that the fuel rod segment between the bottom two spacer grids was pin connected. The restraint and lateral support of the fuel basket structure to the fuel assemblies were ignored in the analysis.

The weight of pellets and irradiated material properties should be included in any end drop analysis. With these changes, the simplistic method of UCID-21246 may not yield acceptable results. For example, the staff conducted calculations using the same methodology as LLNL report UCID-21246 except irradiated material properties for the clad, and the weight of fuel pellets are included in the calculations. The most vulnerable fuel assembly in the LLNL report, a 17x17 Westinghouse fuel assembly, was chosen for this exercise. Euler buckling loads for the clad were calculated using the following formula:

! P, = n 2El/L8 where

! E = 10.47 x 105 psi l l

l 1 = 1/4n x (r,'- r,') = 1/4n x (0.187'- 0.1645') = 3.85 x 10" in' L = 24 inches The results indicate that P, = 69 lb l

Since the weight of cladding and pellets for the 144 inch-long fuel rod is about 4.98 lb, the -

buckling load in terms of gravitational acceleration (g) is P,/W = 69/4.98 = 13.86 g l

This is considerably smaller than the 82 g reported in the LLNL report UCID-21246. However, there are several bounding assumptions in this approach which make the results unrealistically low for predicting cladding failure.

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ISG-12 2

Conclusion:

Analyses of fuel rod buckling performed to demonstrate fuelintegrity following a cask drop accident yield results which contain a large margin to actual failure. The calculated onset of buckling does not imply fuel or cladding failure. Where such analyses yield unacceptable results, more realistic analyses of dynamic fuel behavior are appropriate and acceptable if the cladding stress remains below yield strength, the fuel integrity is assured.

Recommendation: .

l 1

if the analytical approach described in the LLNL report UCID 21246 for axial buckling is used to assess fuel integrity for the cask drop accident, the analysis should use the irradiated material properties and should include the weight of fuel pellets.

Alternately, an analysis of fuel integrity which considers the dynamic nature of the drop accident and any restraints on fuel movement resulting from cask design is acceptable if it demor.v 'tes that the cladding stress remains below yield. If a finite element analysis is performed, the analysis model may consider the entire fuel rod length with intermediate supports at each gud ,

support (spacer). Irradiated material properties and weight of fuel pellets should be included in the analysis.

The appropriate section of Standard Review Plan, NUREG-1536, should be revised to clearly Reflect analytical approach for fuel rod bucking analyses.

Approved tOt ** d 7!#

E. William Brach (fate L_