ML20155D752

Summary of 980916 Meeting with Doe,Naval Reactors (Nr) to Discuss Closure Sys Design for Naval Reactors Planned Dual Purpose Canister (DPC)

ML20155D752
Person / Time
Issue date:	10/05/1998
From:	James Shea NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:	Kane W NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
Shared Package
ML20155D749	List: ML20155D745 ML20155D752
References
TAC-L21186, NUDOCS 9811040004
Download: ML20155D752 (33)

Text

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k'l UNITED STATE 8 I

• D I NUCLEAR REGULATORY COMMISSION WASHINGTON, D.c. SneeHopi l

I October 5, 1998 MEMORANDUM TO: William F. Kane, Director .

Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards FROM: Joseph W. Shea, Project Manager _

Spent Fuel Licensing Section '

Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards

SUBJECT:

MEETING WITH DEPARTMENT OF ENERGY, NAVAL REACTORS, TO DISCUSS NAW DUAL-PURPOSE CANISTER CLOSURE SYSTEM (TAC NO. L21186)

On September 16,1998, a meeting was held between representatives of the Department of

• Energy, Naval Reactors (NR), and the Nuclear Regulatory Commission (NRC) staff to discuss the closure system design for NR's planned dual-purpose canister (DPC). In accordance with Management Directive 3.5, the meeting was not noticed as a public meeting because NR is an agency of the executive branch hot subject to NRC oversight. Attachment 1 is an at>ndance list. Attachment 2 is a copy of NR's meeting handout.

Following introductory statements, NR representatives described their ongoing efforts to design a DPC for the storage of the Navy's spent nuclear fuel. NR stated that, as part of the design process, it is aware of recent problems with closure welds on commercial spent fuel storage l casks and NRC's and industry's consideration of various closure weld inspection techniques.

l NR described the scope of its DPC usage needs which will eventually total approximately 300 l storage canisters and overpacks located at the Idaho National Engineering Laboratory. NR described that, while it does not plan to seek an NRC license under Part 71 or Part 72 for its DPC, it will submit the DPC design to the NRC for review and comment by the end of 1999. NR plans to commence loading DPCs in late 2001.

NR presented a detailed discussion of the closure system design. As depicted in the attached presentation handout, the closure system inclaoas a load bearing shear ring which transmits lifting loads and other loads from the canister lid to the canister shell. Two seal welds are applied, one each at the upper and lower edge of the shear ring. These welds are not I considered structural welds; the shear ring is analyzed for structural adequacy without structural consideration of the seal welds.

UQ, 9

9811040004 981027 PDR ORG EUSDg ENCLOSURE 1

(

W. Kane NR described weld inspection techniques. All structural welds will be radiographed in accordance with ASME Code requirements. Because the seat welds are not structural welds and because of their configuration, NR is proposing not to perform radiography or ultrasonic volumetric examination (UT) of the seal welds. NR is proposing to perform dye penetrant surface examination (PT) of the seal welds on both the root pass and the cover pass of the two pass welds.

NRC staff questioned the efficacy of revising the design to allow for UT of the closure weld. NR responded that it judged potential closure systems which allowed for UT would likely be inferior from a combination of structural, occupational dose, and cost effectiveness considerations.

NRC staff questioned NR regarding the use of mockups to better characterize the adequacy of the seal weld and PT inspection techniques. NR responded that extensive mockup welding was planned. '

NR expressed interest in obtaining written comments on its proposed seal weld design and inspection techniques. NRC staff responded by stating that the NRC Safety Evaluation Report for the NUHOMS MP-187 had been issued on September 10,1998, and contained the latest staff positions en closure weld inspection techniques. The staff also stated that it would provide written comments on NR's proposed DPC seat weld inspection techniques within 3040 days of the meeting.

After summarizing the above iss.ues, the meeting concluded.

Attachments: 1. Attendance List

2. NR Presentation Handout l

Qistribution:

NRC File Center PUBLIC NMSS R/F SFPO R/F SFLS R/F SShankman EEaston LKokajko PEng JThoma, EDO TMadden OCA WReamer. OGC SGagner. OPA NRC Attendees OFC b SFPO 4 SFPO E E SFPO E SFPO / E NAME JShes:dedM VThhrbe WHodgeh EJLeeds DATE , o /(/98 L / 0 /I/98 @ 80/98 M//198

! C = COVER E = COVER & ENCLOSURE N = NO COPY OFFlclAL RECORD COPY G:\ NAVY \099sMTG. SUM 10/1/98:dd l

)

l l I '

. e a NRC/ Naval Reactors September 16,1998 Navy Dual Purpose Canister Closure Welds ATTENDANCE LIST l blama Organization l

W. Hodges NRC/NMSS/SFPO E. Leeds NRC/NMSS/SFPO F. Sturz NRC/NMSS/SFPO i H. Lee NRC/NMSS/SFPO C. K. Battige NRC/NMSS/SFPO J. Shea NRC/NMSS/SFPO l

B. Miles DOE /NR l T. Kennedy DOE /NR E. Naples DOE /NR S. Dunn DOE /NR M. Doman Bettis B. Hallett Bettis Attachment 1

?

Review of Naval DPC Closure '

1 System ~

! E i 8

, B 4

l -

Introduction j

Description of Canister Closure System Inspection Method and Rationale f .

Resolve Questions and Comments .

i

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1 Navy Dua1 Purpose Canister System l

l Final Container System EIS published Nov 96 i for Naval spent fuel Records of Decision issued Dec 96 and Apr 97

- selected Dual Purpose Canister (DPC) System

)! - all Naval spent fuel will be loaded into DPCs and stored at the Naval Reactors Facility at the I'daho

! National Engineering and. Environmental Laboratory.

l DPC syste~m design is near completion - ,

]l procurement action starts in early 1999 j- Naval Reactors plans to submit the DPC system

!- SAR to NRC for review

l l

~ Naval Spent Fuel i

Solid metallic form - not flammable, not explosive l

Built for combat - battle shock

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- well over 50 g's

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l j Fully contains-all fission products .

l No yield of clad during storage or -

. transportation accidents

,. 3 i

l General Canister Features 1

All 316L stainless steel l

l - ductile, not susceptible to brittle fracture or i brittle crack propagation Right circular cylinder

) -212 or 187 inches long 4

l - 66.5 inches OD,1 inch thick except near .

l closure where approximately 2 inches thick

- Maximum loaded weight 49 tons

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i Design j

• Objectives l -large safety factor for lifting l - minimize closure system load during accidents

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j Components: .

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l l Closure System Seal Design i

i lnner confinement seal

-two 7/16 inch fillet welds with shear ring

-two layers of weld metal

-leak tested to 1x10-7 std cc/sec -

Outer confinement seal i .

i -two 3/8 inch groove welds with seal plate .

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-two layers of weld metal

-leak tested to at least 1x104 std cc/sec 1 . 9

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l Seal Inspection and Qualification i

Method

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For all seal welds, penetrant test of:

l - root layer

- final surface

! Inspection and qualification criteria, more restrictive of:

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- Standard Naval Reactors practice ,

! - ASME B&PVC l Section lil, NB 5350 for PT acceptance criteria l- Section IX and Section lil, NB 4360 for qualification 10 ,

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l Closure System Development i

Flat plate welding mockups l

Full diameter cylindrical welding mockups

- three completed i three with inner seal welds and shear ring .

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• one included outer seal welds and seal plate

! two being sectioned -

l - three additional complete mockups (inner and -

! outer seal) planned to be welded before the end of l

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l Closure System Development

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l Mockups have provided:

l - closure system distortion information i

- weld parameter information l - closure design improvements -

l 15 slope on vertical surfaces of inner seal to eliminate l weld undercut -

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l original three piece shear ring is now a two piece shear-j ring - eliminates tack welds and reduces number of '

manual welds .

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i 10CFR72 Requirements l

10CFR72.236:

j - (d) Confinement features must be provided such I

_ that dose limits are met - dose limits will be met for l both normal and design basis accident at l controIIed area boundary i

i .

- (e) Ca~sk rnust be designed to provide redundant -

sealing of confinement system - inner and outer l seals provide redundant sealing of confinement i~ system i- .

14 f.

e i

l l 10CFR72 Requirements ij) Cask must be inspected to ascertain no defects i which could significantly reduce confinement -

effectiveness - liquid penetrant inspection and leak l testing will be used to show confinement effectiveness -

l (I) Cask and systems important to safety must be~ .

l evaluated to demonstrate that they maintain .

l confinement under all credible conditions - SAR

!. drop and non-mechanistic tipover analyses will I.

i show confinementis maintained i, 15 i

l. -

l i

j NUREG-1536 Guidance .

l NUREG does not address Naval DPC .

l System Closure (closure different than typical j commercial closure)

! - NUREG says use ASME B&PVC.Section Ill, .

! Subsections NB or NC

- NUREG focus on Category C welds l

• Category C welds require RT and either MT or PT -

} For welds where RT is not feasible, can do UT i

j. -

NRC has accepted PT, helium leak test, and redundant seals l.

4 16

-_r - , - - - - , _

i Recommendation l t Use ASME B&PVC, Section lil, subsection NB criteria for " Specially Designed Welded

Seals"

- NB 3227.7 for design ~

! - NB 4360 for fabricstion

- NB 5271 for inspection .

PT only .

Obtain NRC concurrence that PT of Naval

~

i l- DPC system closure welds is acceptable j, 17

l

.]

i

! SPENT FUEL ~ PROJECT OFFICE t l OFFICE OF NUCLEAR MATERIAL SAFETY  !

AND SAFEGUARDS i

William F. Kane, Director l 4 1 i

INTERIM STAFF GUIDANCI:

4 4

4 i

}

i l l 4

ENCLOSURE 2

" - ~ ~ "

~' - ~ ' , ~ " -

~

. .. 1 INTERIM STAFF GUIDANCE (ISG)  ;

ISG-1 Damaged Fuel i .

ISG-2 Fuel Retrievablity '

l ISG-3 Post Accident Recovery and Compliance with 10 CFR 72.122(l) .

ISG 4 Cask Closure Weld Inspections -

ISG-5 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 i

9 9

l l

e <

l

Spent Fuel Project Office Interim Staff Guidance -1 Issue: Damaged Fuel Definition of Damaged Fuel Spent nuclear fuel with known orsuspected cladding defects greater than a hairline crack or a pinhole leak.

This definition of damaged fuel applies to both spent fuel storage and transportation.

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

The purpose of canning is to confine gross fuel particles to a known, suberitical 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.

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, and 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.

Demonstration of Fuel Condition 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 fuel is undamaged, followed by an external visual examination of 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.

Recommendation:

The SRPs be revised accordingly.

Approved /

William F. Kane ' Date ISG-1 m m -- - - - ,

i . ..

Spent Fuel Project Office

. inteir m Staff Guidance -2 I

i lasue: Fuel Retrievability -

%72.122(l) states: -

"Retrievability. Storage systems must be designed to allow ready retrieval of spent fuel or high-level radioactive waste for further processing or disposal."

$72.236(h) states:

"The cask must be compatible with wet or dry spent fuelloading 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."R 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 disposal." In amending Part 72 to permit licensing of an MRS as required by the NWPA, the Commission determined that an independent spent fuel storage installation i

(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 ISFSts, 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 under 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 extent 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 disposal" (55 FR 29186). 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 4

+

e -

_ .7 .

d . . .

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).

Recommendation:

l Chapter 1 of the Standard Review Plan (NUREG-1567) should be modified to clearly define that

) compliance with 10 CFR 72.122(l) is achieved when an applicant's design is found to be in compliance 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).

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 staff's recommendation for post accident recovery with regard to retrivievability.

l Approved #

William F. Kane" Date l'

l l

l lSG 2 2 l _ _

~ .

l e . . .

Spent Fuel Project Office Interim Staff Guidance -3 Issue: Post Accident Recovery and Compliance with 10 CFR 72.122(l)

Compliance with 10 CFR 72.122(I) 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 staff 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(1) 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 modified 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.

However, because a failure of the confinement boundary or other structure, system, or 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 modified to ensure that the licensee will have the ability to identify an accident or non-compliance situation.

Approved

. William F. Kane Date ISG 3

s . . . -

) -

Spent Fuel Project Office Interim Staff Guidance - 4 lasue: Cask Closure Wold inspections Position: 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:

1

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

l

' 2) For either ultrasonic examination (UT) or PT, the minimum detectable flaw size must l o uld be be demonstrated to be less than the critical flaw size. The critical flaw size sh' I calculated in accordance with ASME Section XI methodology; however, not section l stress may be goveming for austenitic stainless steels, and must not violate Section 111 requirements. Flaws in austenitic stainless steels are not expected to exceed the thickness of one weld bead.

l

3) If PT alone is used, at a minimum, it must include the root and final layers and sufficient intermediate layers to detect critical flaws.

a

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

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

l

6) The results of the PT examination, including 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.

, Basis: Radiographic (RT) inspection is preferred for cask closure welds. However, RT is not practical for field closure welds with fuel in 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 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 as acceptable methods; however UT is still preferred where practical.

9 j

Acceptable UT has not yet been demonstrated for austenitic stainless steels in the required

configuration. At best, UT that would be developed may require considerable skillin execution and interpretation. Minimum detectable flaw sizes for UT will be relatively large (estimated O'.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

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

ISG 4 1

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l J-l problems and may present ALARA issues. Although the setup for the UT would be expensive, j cost could be spread over multiple containers to reduce unit cost.

1 PT only identifies surface flaws. However, the acceptance standard is no cracks or linear l Indications. In theory, a flaw slightly srnaller than the PT examination increment size could exist, therefore, layered PTis necessary to assure detectable flaw sizes are less than critical flaw size. PT is widely used in safety critical applications such as certain reactor pressure l

l vessel welds and, for some applications is the only practical technique. ALARA issues may arise for large welds that require multiple PT examinations.

i I Austenitic stainless steels do not have a nil ductility transition temperature. Thus, the weld can l sustain 'large" flaws without a concem for flaw growth. This allows the use of either UT or PT j although both would have limitations on detectable flaw size and both would accept less than

. critical flaws.

l l Finally, the Nuclear Regulatory Commission regulates to the standard of adequate protection, not absolute assurance. Although UT is the preferred technique to PT, in that it is a volumsine examination, PT 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. l l

Recommendation:

The SRPs be revised accordingly.

I Approved "W411am F. Kano ' Date

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l ISG.4 2

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3 Spent Fuel Project Office l

Interim Staff Guidance - 5

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issue: Normal, off-normal, and hypothetical accident dose estimate calculations i for the who!c body, thyroid, and skin.

While the staff considers there will be no effluent from a storage cask, the applicant must demonstrate compliance with 10 CFR 72.104(a),10 CFR 72.106(b), and 10 CFR 72.126(d). It is the staffs assumption that leakage of gases, volatiles, fuel fines, and crud is credible and should be addressed. The analysis should be based on a source term that includes all radionuclides that are greater than 0.1% of the total activity present in the fuel plus iodine (this would result in approximately 95% of the dose if the totalinventory were included). The guidance given in Table 4-1 of NUREG-1617 (DRAFT) and technical bases given in '

NUREG/CR-6487 provide release fractions relative to gases (fa), volatiles (fy), fuel fines (f,),

and crud (fe). NUREG/CR-6487 also provides the basis for determining the quantities of radionuclides for inclusion in the source term. The use of a computer code such as SAS2H to generate this source term or the shielding source term is acceptable to the staff.

An assumed minimum leakage duration for normal and off-normal conditions of one year and a minimum leakage duration for hypothetical accident conditions of 30 days are considered appropriate. The analyses should use the leak rate from tested conditions as adjusted to account for bounding conditions of storage for normal, off normal, and accident conditions (temperatures and pressures) as an assumed leakage rate for the normal, off-normal, and accident conditions.

For determination of X/Q values, absent site-specific data, use of Class D stability and wind velocity of five m/s meteorological conditions may be used for long term conditions (normal,-

off normal). For short-term conditions (a hypothetical accident), use of Class F stability and wind speed of one m/s meteorological condition is considered bounding.

The staff accepts dose calculations using Dose Conversion Factors from EPA Federal Guidance Reports 11 and 12 on an isotope specific basis. No weighting or normalization of the dose cenversion factors is acceptable since this may not be conservative. The resultant dose for normal conditions of storage must be a small fraction of the 25 mrem /yr limit of 10 CFR 72.104(a) to accommodate an array of casks and extemal direct dose. For off-normal conditions, only one cask need be considered, but the resultant dose must be a fraction of the

' dose limit so that when added to the dose from normal conditions it still meets 10 CFR 72.104(a).

. O ISG 6

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• Recommendation:

Update NUREG-1536 Chapter 7, Sections 3 and 4, (pages 7 7-7); Chapter 10,Section V, 3, (page 10-3); and Chapter 11,Section V,3 ( page 11-3) to reflect this new position. NUREG-1567 should also be updated accordingly.

Approved N //

William F. Kane Date .

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ISG-5 2

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Spent Fuel Project Office interim Staff Guidance - 6 issue: Establishing min!murn 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 4

! bounding source term," and states that the applicant should, *either specify the minimum initial l l

enrichment or establish the specific source terms as operating controls and limits for cask use." l A specified source term is difficutt 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. -

l Lower enriched fuel irradiated to the same bumup as higher e.nriched fuel produces a higher l neutron source. Sometimes fuel assemblies are driven to bumups beyond the value normally l 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 l 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 bounds 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."

Approved William F. Kane Date 1

i I~ ISG.6 l

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4 Spent Fuel Project Office interim Staff Guidance -7 lasue: Potential Generic laaue Concerning Cask Heat Transfer in a Transportation  !

Accident 1 l

Staff raised two major issues concoming the adverse effects of fission gases to the gas-mixture l thermal conductivity in a spent fuel canister in a post accident environment. The two major concems 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 canister. 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 intemal pressure will ,

increase as will the bulk temperature of the gas. The combination of these phenomena,if they I 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 l 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 l xenon (Xe), as a conservatism, all fission gases generated were assumed to be the heaviest j gas present, which is usually xenon. This reduced the problem to a binary mixture and

- l 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 l 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 evalua.ted as a function of bulk  ;

temperature at standard pressure. These thermal conductivities were compared with those of pure He at the same standard pressure. This revealed a general percentage reduction in the i mixture conductivites over a wide range of temperatures. In this sidy, the percentage  ;

reduction was a sizable 70%. Applying this reduction to the cask model, a parametric study of l the cask was performed to determine the impact of the changing conductivity on the inner-1 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.

4 9 0

, ISG-7

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' Temperature Temperature #10%

Modeled case WoriginalThermal Reduced Thermal . Temperature Lacetion conduceMiy[R) ConductMiy[R] l 12 hr Mazhum s47 868 ThermalLead aukGee )

i Trench.t 1236 1272 Peak clad j vac centenerm s73 as7 aumone Treneler cask 3343 1371 peak clad l

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 temperature 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 i canister shell.

Recommendation:

t Change the Standard Review Plan for Transportation Packages for Sper t Nuclear Fuel,

!. NUREG-1617, and the Standard Review Plan for Dry Cask Storage Systerns, NUREG-1536 as follows:

! Under the conditions where any of the cask component temperatures are close l (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 potentialimpact of the fission gas in the canister to the j cask component temperature limits and the cask intamal pressurization.

l Approved #

l William F. Kane Data r

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