ML20154N212

To Application for Amend of Coc 1004 for Dry Spent Fuel Storage Casks.Submittal Provided in Listed Format to Facilitate Staff Review.Proprietary Encl Withheld

ML20154N212
Person / Time
Site:	07201004
Issue date:	10/09/1998
From:	Grenier R External (Affiliation Not Assigned)
To:	Mcginty T NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
Shared Package
ML20138L335	List: ML20154N212 ML20154N217 ML20154N223
References
NUH03-98-1413, NUH3-98-1413, NUDOCS 9810210210
Download: ML20154N212 (27)

Text

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~72-l 0 O l TRANSNUCLEAR WEST October 9,1998 NUH03-98-1413 Mr. Tim McGinty Spent Fuel Project Office, NMSS U. S. Nuclear Regulatory Commission i1555 Rockville Pike Rockville, MD 20852

Subject:

Application for Amendment of Certificate of Compliance No.1004 for Dry Spent Fuel Storage Casks, Revision 2

References:

1.

Certificate of Compliance No.1004 (effective January 23,1995) 2.

NRC's Request for Additional Information (TAC No. L22252) dated March 17,1997

Dear Mr. McGinty:

e In response to the Reference 2 request for additional information, Transnuclear West is herewith submitting the subject revision to our license amendment application for our Standardized NUHOMS* Horizontal Modular Storage System for Irradiated Nuclear Fuel, NUH003.12002. As in the case of the previous revisions to this amendment application, this submittal is provided in the following format to facilitate your staff's review:

Attachment A: Summary of License Amendment Changes Attachment B: Questions and Answers Attachment C: Changed CSAR Pages Attachment D: Suggested Changes to Certificate of Compliance Pages Attachment E: Calculation Packages (Proprietary Information)

Attachment F: Affidavit for Proprietary Information

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,.m, 9810210210 981009 PDR ADOCK 07201004 C

PDR T nsnuclear West Inc.

39300 Civic Centor Drive, Suite 280, Fremont, CA 94538 k

Phone: 510-795-9800 + Fax: 510-744-6002

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Mr. Tim McGinty NUH03-98-1413 Spent Fuel Project Office, NMSS October 9,1998 1

Should you or your staff require additional information to support review of this application, i

please do not hesitate to contact Mr. Walter Bak (510-744-6018) or me (510-744-6020).

i Sincerely,'

7 k

Old, Robert M. Grenier

' President and Chief Operating Officer

- cc:.

Mr. Joe Shea U.S. NRC Public Document Room, U.S. NRC RMG-98-017

Enclosure:

Nineteen (19) copies of Revision 2 (15 of which are under separate cover) j 1

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2

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Attaciunent F Page 1 of 2 AFFIDAVIT PURSUANT TO 10 CFR 2.790 Transnuclear West Inc.

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i State of California

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SS.

County ofAlameda

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I, Robert M. Grenier, depose and say that I am President and Chief Operating Officer of Transnuclear. West Inc., duly authorized to make this affidavit, and have reviewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph immediately below. I am submitting this affidavit in conformance with the provisions of 10 CFR 2.790 of the Commission's regulations for withholding this information.

%e information for which proprietary treatment is sought is contained in the following document:

Application for Amendment of Certificate of Compliance No.1004 for Dry Spent Fuel Storage Casks (At*=chment E)

His hment has been appropriately designated as proprietary.

I have personal knowledge of the criteria and procedures utilized by Transnuclear West Inc. in designating information as a trade secret, privileged or as confidential commercial or financial information.

Pursuant to the provisions of paragraph (b) (4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure, included in the above referenced document, should be withheld.

1)

The information sought to be withheld from public disclosure are licensing drawings and calculations of a concrete modular storage and transportation system, which is owned and has been held in ennMance by Transnuclear West Inc.

2)

The information is of a type customarily held in ennMaace by Transnuclear West Inc. and not customarily disclosed to the public. Transnuclear West Inc. has a rational basis for determuung the types ofinformation customarily held in ennMaace by it.

3)

The information is being transmitted to the Commission in confidence under the provisions of 10 CFR 2.790 with the understandmg that it is to be received in confidence by the Commission.

4) ne information, to the best of my knowledge and belief L not available in public sources, and any disclosure to third parties has been made pursuant to regulatory provisions or proprietary agreements which provide for maintenanca of the information in confidence 5)

Public disclosure of the information is likely to cause substantial harm to the competitive position of Transnuclear West Inc. because:

a)

A similar product is manufactured and sold by competitors of Transnuclear West Inc.

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Attachment F Page 2 of 2 b)

Development of this information by Transnuclear West Inc. required thousands of

' man-hours and hundreds of thousands of dollars. To the best of my knowledge and belief, a competitor would have to undergo similar expense in generating equivalentinformation.

4 i

c)

In order to acquire such information, a competitor would also require considerable time and inconvenience related to the development of a design and analysis of a dry spent fuel storage system.

d)

The information required significant effort and expense to obtain the licensing approvals necessary for application of the information. Avoidance of this expense would decrease a competitor's cost in applying the information and marketing the product to which the information is applicable.

e)

The information consists of description of the design and analysis of a dry spent fuel storage and transportation system, the application of which provides a competitive economic advantage.

The availability of such information to competitors would enable them to modify their product to better compete with Transnuclear West Inc., take marketing or other actions to improve their product's position or impair the position of Transnuclear West's product, and avoid developing similar data and analyses in support of their processes, methods or apparatus.

f)

In pricing Transnuclear West's products and services, significant research, development, engineering, analytical, licensing, quality assurance and other costs and expenses must be included. 'Ihe ability of Transnuclear West's competitors to utilize such information without similar expenditure of resources may enable them to sell at prices reflecting significantly lower costs.

Further the deponent sayeth not.

,//

10 Robert M. Grenier '

President and Chief Operating Officer Transnuclear West Inc.

Subscri and swom to me before this 9* day of October,1998 by Robert M. Grenier.

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% TON Kim Stone, NotaryMblic l

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0 ATTACHMENT A j

Summary of Amendment Channes INTRODUCTION This amendment changes the fuel qualification specification for both PWR and BWR fuel. This document describes the changes in summary form and gives background on why they are i

necessary.

REVISION HISTORY Revision 2: This revision incorporates the responses to the NRC " Request for Additional Information on Fuel Qualification Amendment Application (TAC No. L22252)" dated March 17, 1998. This revision completely replaces Revision 1 of the application which was submitted in November of 1996. A summary of changes included in this revision is provided below.

1. Revision history updated in Attachment A.

2. Attachment B revised to include responses to March 1997 RAI and add supplemental l

,O information supporting the RAI responses.

V i

3. Attachment C, CS AR Changed Pages, revised to: (A) Update design criteria to include revised design pressures and maximum burnup; (B) Update cladding temperature limits based on RAI responses; (C) Update fuel qualification tables and shielding results based on the revised cladding temperature limits; and (D) Revise DSC internal pressure based on response to RAI question 7-4.

4. Attachment D, C of C Changed Pages, revised to: (A) Clarify the description of the initial l

equivalent enrichment requirement: (B) Update the qualification tables as discussed in Attachment C; and (C) Update criticality acceptance curve to limit PWR burnup to 50 GWd/MTU.

l

5. Attachment E, Calculation Packages, revised to include latest revision of calculations NUH004.0409 (Revision 2) and NUH004.0510 (Revision 3). Additional calculations NUH004.0427 (Revision 0) and NUH004.0242 (Revision 0) have been added to support the i

1 response to RAI question 7-4. Calculations revised to incorporate RAI comments including:

l (A) Provide justification for use of PNL-6189 for high burnup fuels; (B) Provide justification

(

that the selected fuel types bound the dry storage cladding temperature limits; (C) Justify the j

. selected cladding oxide thickness for high burnup PWR fuels; (D) Evaluate the containment pressure analysis for high burnup fuels; (F) Reduce the maximum fuel bumups to l 50,000 MWD /MTU (PWR) and 45,000 MWD /MTU (BWR) to facilitate licensing; (G)

Update the cladding temperature calculation to incorporate the above items; and (H) Update l-j the fuel qualification calculation to incorporate the revised cladding temperature limits.

i November 1996 72-1004 Amendment Page A.1 Revision 1

m This revision also includes a report prepared by Dr. E. R. Gilben (the author of PNL-6189) which provides additional i- 'rmation supporting this application.

Revision 1: This revision incorporated the responses to the NRC " Request for Additionhl Information on Fuel Qualification Amendment Application (TAC No. L22252)" dated August 26,1996. This revision completely replaced the original application which was submitted in February of 1996. A summary of changes included in this Revision 1 is provided below.

1. Revision history added to Attachment A.

2. Attachment B revised to include responses to August 1996 RAI.

3. Attachment C, CSAR Changed Pages, revised to: (A) Provide tables listing the fuel assembly designs which can be stored in the standardized NUHOMS* system; (B) Update the cladding temperature limit calculation results to include the dependence on fuel burnup; (C) Update the shielding analysis and occupational exposure results to include the peak calculated doses for the fuel burnup and cooling time parameters qualified by this amendment; (D) Update the 1

DSC leakage accident dose calculations to include the fuel burnup and cooling time parameters qualified by this amendment; (E) Revise the radiological acceptance criteria to 1

include the cask radial surface dose rate; (F) Revise the PWR and BWR fuel specifications to refer to CSAR Chapter 3 for acceptable fuel designs (see item A above); (G) Revise the alternate nuclear parameters to limit the maximum burnup to 40,000 mwd /MTU for PWR p-fuel and 35,000 mwd /MTU for BWR fuel; (H) Clarify units shown on the fuel qualification tables; (I) Revise PWR qualification table to exclude assemblies excluded by the criticality L

acceptance curve; (J) Revise both qualification tables to incorporate revised cladding temperature limits and radiological acceptance criteria - maximum PWR burnup reduced to 55 GWd/MTU; and (K) Update criticality acceptance curve tv limit PWR burnup to 55 GWd/MTU.

4. Attachment D, C of C Changed Pages, revised to: (A) Refers to CSAR Chapter 3 for a list of suitable fuel designs; (B) Revise the radiological acceptance criteria to include the cask radial surface dose rate; (C) Revise the alternate nuclear parameters to limit the maximum burnup to 40,000 mwd /MTU for PWR fuel and 35,000 mwd /MTU for BWR fuel; (D) Update the qualification tables as discussed in Attachment C; (E) Specify that licensees must account for enrichment and burnup uncenainties when using qualification tables; (F) Specify that fuel with initial enrichments less than 2 w/o U-235 must be qualified per the alternate nuclear parameters; and (G) Update criticality acceptance curve to limit PWR burnup to 55 1

GWd/MTU.

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l S. Attachment E, Calculation Packages, revised to include latest revision of calculations NUH004.0409 (Revision 1) and NUH004.0510 (Revision 2). Calculations revised to incorporate RAI comments includ ng: (A) Provide revised DSC leakage analysis; (B)

Calculate the impact on public and worker doses; (C) Include the transfer cask radial surface dose as an acceptance criterion for fuel storage; (D) Remove the assumed 5% factor from gamma dose calculations and instead calculate doses based on actual gamma spectra; (E)

_O Calculate neutron, gamma,'and total dose rates for the HSM door and cask axial surfaces; (F)

U November 1996 72-1004 Amendment Page A.2 i

Revision 1 J

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Determine which PWR and BWR fuel designs are acceptable for storage; (G) Revise the 4

cladding temperature limit calculation to incorporate fuel data for burnups up to the maximums allowed by this amendment; and (H) Provide supporting documentation for cladding stress limit vs. storage temperature limit curves.

- This revision also incr~ sed the NR heavy metal weight from 0.472 MTU to 0.475 MTU. This change was made to bout i existing PWR fuel designs. The PWR decay heat and dose rate calculations were updated to incorporate this change.

NEED FOR CHANGES Fuel Qualification Table Fuel qualification for storage in NUHOMS' is currently based on the parameters of fuel assemblies used as design basis for the NUHOMS' system. These design parameters have been translated into twelve separate fuel qualification limits (size, weight, enrichment, burnup, cooling time, initial heavy metal, heat generation, gamma source strength, gamma spectrum, neutron source strength, neutron spectrum, and burnup-enrichment curve). Although this approach ensures that important safety parameters (ker, temperatures, and radiation exposure) do not

. exceed design values, many fuel assemblies which fall outside the intersection of these twelve parameters are unnecessarily precluded from storage. In addition, a series of independent fuel

. qualifications must be performed by the utility user in order to qualify each fuel assembly to be D.

stored. Some of these calculations are lengthy, or non-standard (for instance the calculation of I

neutron spectra), and may not be approached in the same manner by all utility users. We propose an equivalent approach to fuel qualification which focuses on the safety parameters rather than the parameters of the design basis fuel assemblies. The approach also has the unique advantages that all utilities would use the same simple, straightforward method to qualify fuel, and that this method would be " pre-reviewed" by the NRC Staff.

DESCRIPTION OF CHANGES

. Fuel Qualification Table This change affects the fuel specifications in Section 10.3.1 of the CSAlt. The affected CSAR pages are shown revised in Attachment C, and our suggested revisions to 3.e Certificate of

, Compliance are shown in Attachment D.

The Fuel Qualificatien Table calculations (proprietary) are also available for review in Attachment E. These records are normally retained in Transnuclear West's files; we have

. submitted them in order to provide the most complete package for your review.

~/9

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- November 1996 72-1004 Amendment Page A.3 Revision 1 J

Fuel Oualification-A Review of Existine Dry Storage Systems Metal Casks 4

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Fuel qualification specifications for early storage systems were limited to the fuel parameters used in their safety analysis, i.e. heat generation rate, fuel enrichment, burnup, and cooling time.

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That practice was continued as recently as 1990 when the four metal cask storage systems (GNSI l

CASTOR V/21, Westinghouse MC-10 NAC Sff, and NAC C-28) were " grandfathered" into the j

new 10CFR72 Subpart K mle.

This strategy requires the utility to independently calculate the fuel heat generation rate, and verify the fuel's burnup (via the plant's fuel accountability system), initial enrichment, and cooling time.

Concrete Casks (Site specific Licenses)

Historically, site-specific licenses granted under the provisions of 10CFR72 have made a distinction between the design basis fuel and other combinations of fuel parameters. This was done to allow the utility the flexibility to store fuel that is not strictly bounded by the design basis fuel parameters, but which could still be stored without violating system safety parameters. The utility, at its option could either qualify fuel by verifying initial enrichment, burnup, and cooling

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time, and calculating the decay heat, or it could calculate heat generation rate, gamma and

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neutron source terms, and gamma and neutron energy spectra. This arrangement (see Table 1) was more attractive than the earliest specifications, since it allowed the utility more flexibihty.

In the event the utility qualified its fuel by calculation, the process (particularly when it gets to the detail of the gamma and neutron spectra) becomes somewhat complicated and ambiguous.

g g

e' g

g e

a a

... =

e...

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Initial Enrichment, w/o U235 s 3.3 s 3.3 s 4.0 s 4.0 Decay Power, kW not spec'd s 1.0 0.66 s 0.66 Max Bumup, mwd /MTU s 33.000 not spec'd s 40,000 not spec'd Cooling Time, yrs 25 not spec'd 2 10 not spec'd Gamma Source, g/sec-assy not spec'd 1.1e15 not spec'd 1.6e15 Gamma Energy Spectrum not spec'd per SAR not spec'd per TR Neutron Source, n/sec-at,sy not spec'd 2.0e07 not spec'd 1.5e08 Neutron Energy Spectrum not spec'd per SAR not spec'd per TR Table 1 Fuel Qualification Parameters for Two NUHOMS* Site Licenses O

November 1996 72-1004 Amendment Page A.4 j

Revision 1

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

Concrete Casks (Certificates)

The first Certificate of Compliance (after the new Subpart K rule) was issued to Sierra Nuclear's VSC-24 dry storage system. Tliis fuel specification differs from previous ones in several ways.

First, the utility is required to calculated the fuel aasembly heat generation rate, gamma, and neutron source terms, regardless of the fuel parameters. This unnecessarily limits the number of assemblies which can be qualified for storage, The VSC fuel specification includes a permission to store fuel above the design basis 35,000 e

mwd /MTU burnup (up to 51,800 mwd /MTU) if the utility can demonstrate that the fuel cladding temperature, neutron source, and gamma source are within limits. This was a substantial departure from previous practices for two reasons. First is that the utility is required to calculate cladding temperatures, not heat generation rates. Second, the utility must perform complex heat transfer calculations to derive cladding temperatures ifit chooses to store fuel with the burnup exceeding 35,000 mwd /MTU.

The Certificate of Compliance for the standardized NUHOMS* dry storage system is the latest generation fuel specification. Some of the major differences between it and its predecessors are:

No allowance for alternate fuel specifications (i.e. beyond the design basis burnups). This is, e

l,(

in fact, specifically prohibited "Any deviation constitutes an Unanalyzed Condition and l

Violation of the Certificate of Compliance."

The utility is not permitted to qualify fuel by independent cladding temperature calculations j

(as in the Sierra VSC-24).

The utility is prohibited from qualifying fuel by independent radiological source term calculations (as in the Sierra VSC-24, and all other storage licenses known to us).

A comparison between the Certificate of Compliance for the NUHOMS*-24P and the VSC-24 systems is shown in Table 2.

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November 1996 72-1004 Amendment Page A.5 1

Revision 1 l

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Initial Enrichment, w/o U235 s 4.2 s 4.2 s 4.0 not permitted Decay Power, kW s1 cale clad temp s1 not permitted Max Burnup, mwd /MTU s 35,000 s 51,800' s 40,000 not permitted Min Burnup, mwd /MTU BU Curve BU Curve BU Curve not permitted Cooling Time, yrs 25 25 25 not permitted Gamma Source, g/sec-assy 6.8e15 6.8e15 7.5e15 not permitted Gamma Energy Spectrum per SAR per SAR per SAR not permitted Neutron Source, n/sec-assy 1.2e08 1.2e08 2.2e08 not permitted Neutron Energy Spectrum per SAR per SAR per SAR not permitted Table 2 Comparison of VSC-24 and NUHOMS*-24P Fuel Specifications What is the "Eneineered Fus! Specification"?

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The Engineered Fuel Specification is a new approach to qualifying fuel for storage that focuses on parameters which directly impact safety (ker, radiological exposure, etc.) rather than fuel parameters (burnup, cooling time, etc.).

We believe that the fuel specification should be simple, that the utility should not have to perform calculations in order to qualify fuel, and that the NRC safety reviewers should be able to review the calculations under which most, if not all, of the fuel assemblies are qualified. In A

addition, we feel that the fuel qualification parameters should not be " multi-specified". By this, Q

we mean that simultaneously specifying the fuel enrichment, burnup, enrichment vi urnup, b

cooling time, neutron source, gamma source, neutron spectrum (in several energy groups),

gamma spectrum (in several energy groups), and heat generation rate does not enhance the safety or the value of the dry storage system, but rather makes it less effective, and unnecessarily complicated and confusing for both the safety reviewer and the end user. For these reasons, we propose the " Engineered Fuel Specification" as a replacement for the fuel specification currently in the NUHOMS* Certificate of Compliance, SER, and SAR.

Where does the "Eneineered Fuel Soccification" come from?

The Fuel Qualification Tables were constructed by performing lookups in the OCRWM Characteristics Database (CDB)2 for hundreds of combinations of fuel burnups, enrichments, and cooling times. Since the Tables do not introduce any new safety analysis (they only define what fuel falls within the bounds of the important system safety parameters), there is neither design impact on the storage system, nor any change in the basis for the design. Instead, the calculations currently performed by utilities are performed "in advance" and there will be no "unanalyzed conditions" for anything on the table with a required cooling time.

' For burnup 2 35.000, utility must evaluate cladding temperature and radiological source terms.

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2. Characteristics of Potential Repository Wastes," DOE /RW-0184 R1, Office of Civilian Radioactive Waste Management, July,1992.

s November 1996 72-1004 Amendment Page A.6 Revision I l

! C Complete backup for the Fuel Source Term Tables is provided in the proprietary Attachment E.

l The methodology is simple and ample backup is provided so that safety reviewers can audit any l

combination of fuel enrichment, bumup, at:d cooling time.

How are Imoortant Safety Parameters Met?

The NUHOMS* dry fuel storage system was designed to safely store both PWR and BWR spent fuel assemblies for a wide range of fuel initial enrichments, burnups, and cooling times. For the purpose of demonstrating the safety of the design, the PWR configuration was determined to be limiting. A design basis PWR fuel assembly with the following properties was chosen for input to shielding, criticality, thermal, and structural calculations:

l Fuel De s ig n.............................................................. B &W 15 x 15 L

' Metric Tons Initial Heavy Metal.............................. 0.475 MTIHM

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Initial Enrichment..................................................... 4.0% 'U 23 5 B um up.................................................................... 40,000 mwd /MTU Coolin g Time........................................................ 5 years Other combinations offuelparameters may be stored without exceeding the system safety

. parameters. We seek to define those acceptable combinations of parameters using an' approach.

l we call the Fuel Qualification Table. The Fuel Qualification Table is used along with other L

factors needed to assure safe operation of the storage system. The factors that are important to 1

A system safety and the manner in which we propose that they be controlled are:

,h l

Criticality Safety Only intact, unconsolidated fuel j

l Only fuel designs qualified in SAR Limit maximum initial enrichment e

Minimum pool boron content (for PWR systems) e

' Burnup-enrichment criticality curve for the NUHOMS*-24P

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Fuel Cladding and Concrete Temperatures Fuel Qualification Table e

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Radiological Exposure Fuel Qualification Table How is the " Engineered Fuel Specification" Used?

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There are two Fuel Qualification Tables: one each for PWR and BWR fuel. They are shown in L

the proposed NUHOMS' fuel specification. The Fuel Qualification Tables provide a repeatable, l

' reliable system for qualifying fuel from the standpoint of heat and radiological source terms. The tables are based on preserving two parameters to be the same, or less, as thosefor design basis fuel: the fuel cladding temperature (and concrete temperatures), and the total dose rate on the

[p exterior of the horizontal storage module and the transfer cask. Note that this is more

,i conservative than satisfying design limits.

.L November 1996 72-1004 Amendment Page A.7 i

l Revision 1

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To use the Fuci Qualification Tables, the utility user simply looks up the required cooling time in i

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the appropriate Fuel Qualification Table using the assembly's fuel enrichment and bumup.

. Assemblies which de not meet the minimum required cooling time cannot be qualified for storage unless they are qualified by independent utility calculations. We anticipate this to be the exception, rather than the rule (as is the current practice).

What are the Advantages of the " Engineered Fuel Specification"?

The Engineered Fuel Specification offers several distinct advantages to both the safety reviewer

. and utility user.

i-Advantages to the Safety Reviewer.

The sarety bases are clearly stated.

The safety bases are meaningful, The qualification criteria are not ambiguous.

e All utility users can be expected to come to the same qualification conclusion for a given e

fuel assembly.

No reliance on the utility to independently calculate heat generation rates, fuel cladding e

. temperatures, gamma source terms, gamma spectra, neutron source terms, neutron spectra, etc. in order to qualify fuel. NRC can review the Fuel Qualification Tables prior L

to certification.

O&

Advantages to the Utility.

L e

For most cases, computer calculations are not necessary to qualify fuel (i.e. heat generation, gamma or neutron source term calculations).

Fewer fuel assemblies are unnecessarily precluded from storage.

e Reduction in radiological exposure since fewer low-dose fuel assemblies would be e

unnecessarily precluded from storage, Qualification criteria are not ambiguous.

e Conclusion The. ways in which fuel is qualified for dry storage have changed drastically as the systems have evolved. The NUHOMS* system fuel specification is the most restrictive yet. It presents its users o

l with twelve independent specifications en fuel parameters-some requiring complicated calculations. Instead of enhancing system safety, this multi-specification actually unnecessarily disqualifies a substantial number fuel ass:mblies which would be more suitable for storage from i

. an ALARA standpoint. These restrictions are also inconsistent with permissions historically granted on both site specific and certified licenses which were designed to give the utility users L

more flexibility in fuel qualification.

.We feel that this engineered approach provides more value to the NRC safety reviewers, the utility users, and the public than the existing methods by which fuel is qualified for dry storage.

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November 1996 '

72-1004 Amendment Page A.8 Revision 1 l

a In summary, we have proposed an alternate approach to fuel qualification for storage. It does not f(

. introduce new safety calculations, nor does it change any of the bases for safety of the NUHOMS* system. Instead, this approach focuses on the real safety parameters and treats the fuel qualification parameters in a simple, complete manner that has not been done before to our knowledge.

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-INovember 1996' 72-1004 Amendment Page A.9 Revision 1 4

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ATTACHMENT B Responses to March 17.1997 RAI (TAC No. L22252) 7-1 Provide technicaljustificationfor the use of the methodology and datafrom PNL-6189,

" Recommended Temperature Limitsfor Dry Storage ofSpent Light Water Reactor Zircaloy-Clad Fuel Rods in inert Gas," to perform Calculation NUH004.0409.

The cask models developed in PNL-6189 were based on spentfuel burnup of 30 MWD /MTU and, although briefly discussed, do not take into accountfuels with higher burnups such as 55 MWD /MTU.

l Response: The attached (see Attachment E of this submittal) Transnuclear West j

calculation, NUH004.0409, has been revised to include a justification for use of the PNL-6189 methodology for higher burnup fuels. This justification is included in Section 3 of the calculation. The PNL-6189 methodology can be used if the long term failure mechanism for the cladding is still stress rupture due to creep induced by the high temperature and internal rod pressure and if the fuel rods still maintain resistance to gross degradation at higher burnups. These two criteria are discussed andjustification is provided in the revised calculation. Additional details and justification are given in

'I Section 7.1 of the following Reference. A copy of this Reference is included with this j

, Oj

response, b

Reference:

Gilbert E. R., " Supporting Information for VECTRA Calculation Package Revision 1 NUH004.0409," Questions 7.1,7.2,7.3, Dated May 19,1997.

7-2 Provide the technicaljustification that the selectedfuel types,for both boiling water reactors (BWR) andpressurized water reactors (PWR), would bound the calculated dry storage clad !ing temperature limits.

l The cladding temperature limit, as calculated using the PNL-6189 methodology, is a j

function of the cladding hoop stress. Hoop stress is calculated using thefollowing equation:

crsove = P x r/t Where.

l l

P rod internalpressure r

midwallcladding radius [(r + r)/2]

o s

t cladding thickness l

The cladding dimensions (diameter and thickness) vary among differentfuel types. These variations results in a different calculated hoop stressfor a given rod's internal pressure.

I(d VECTRA 's selection ofBabcock and Wilcox 15 x 15 cladding dimensions with an r/t ratio l-October 1998 72-1004 Amendment Page B.1 Revision 2

p 7.61 does not conservatively bound the r/t ofother PWRfuel types (i.e., Westinghouse 15 x 15, r/t ratio of 8.22; Westinghouse 17 x 17, r/t ratio of 7.81; and Westinghouse 14 x 14, r/t ratio of 8.88). The same is true with respect to the selection ofAdvanced Nuclear l

Fuels 9 x 9 fuel (r/t ratio of 6.57) as conservatively bounding the r/t ratio of all BWR fuels (i.e., GE 2 7x 7, r/t ratio of 8.30; and 8 x 8 r/t of 7.05). Assuming that the other

~

parameters (rod internalpressure, fission gas temperature, cladding oxide thickness) l provided by VECTRA are acceptable, the calculated dry storage cladding temperature l

limit would be a lower value than thatpresented in VECTRA's Calculation l

NUH004.0409, Revision 1. Changing the boundingfuel type could impact the decay time requirements delineated in thefuel quah'fication tables.

}

In addition, there are two minor corrections that need to be implemented in Calculation NUH004.0409, Revision 1:

l a:

Hoop stress calculations in thisfile are based on a rod internalpressure that is l

above the atmospheric pressure (i.e., psig), instead of using the absolute pressure

'(i.e., psia). This approach is inconsistent with the methodology prescribed in i

. PNL-6189.

b:

The calculationfile has analyses up to a burnup of 60,000 MWD /MTU.

NUHOMS* dry storage casks can only acceptfuel with a maximum burnup of 55,000 MWD /MTU. The additienal analysis representing a burnup between I

55,000 to 60,000 MWD /MTU needs to be removed, or its use needs to bejustified.

l ' C' Response: Transnuclear West calculation NUH004.0409 has been revised to provide justification that the B&W 15x15 Mark B and ANF 9x9 fuel assemblies for the PWR and.

l BWR respectively, will give the bounding (lowest) fuel cladding temperature limits from the list of qualified assemblies for the NUHOMS system. This justification is included in Sections 4.1 and 5.1 of the calculation for the PWR and BWR assemblies, respectively.

The PNL-6189 methodology utilizes the following formula which incorporates pressure l

and temperature into the hoop stress calculation.

L w.r, = (PD,,a /2t)(a)(T /T,)

o 2

Where:

. Choop.T2 hoop stress during storage at temperature T2 P

rod internal pressure at Ti Dmid midwall diameter at End of Life (EOL) conditions t

cladding thickness at EOL a

0.95 for PWR,0.90 for BWR T2 temperature of storage condition Ti temperature at which pressure P is calculated or measured It is correct that the dry storage cladding temperature limit would be a lower value for a

' h

. lower r/t ratio if the other parameters (rod internal pressure, maximum peak linear heat G

L October 1998.

72-1004 Amendment Page B.2 Revision 2

l generation rate (LHGR), void volume, fission gas temperature, cladding oxide thickness)

'0 are the same for the different fuel designs. When all these parameters are compared for the PWR and BWR fuel assemblies, the B&W 15x15 Mark B and the ANF 9X9 fuel assembly are shown to be the limiting fuel assemblies for the PWR and BWR fuel designs respectively. The justification is included in calculation NUH004.0409, provided in Attachment E of this submittal.

- Comment 7.2(a) has been incorporated in the revised NUH004.0409 calculation.

Comment 7.2(b) has been incorporated in the revised rJ C904.0409 calculation. The calculation now includes only the burnups for which the NUHOMS* system is being qualified. These burnups are 650,000 MWD /MTU for PWR fuel assemblies and

$45,000 MWD /MTU for BWR fuel assemblies.

73 Provide technicaljustificationfor the conservatism of the selected cladding oxide thicknessfor high burnup PWRfuels.

The maximum cladding oxide thickness of 60 microns (0.0024 in) selected be VECTRA does not appear to be applicablefor a high burnup PWRfuel. According to specialists in thefuelperformance crea, cladding oxidation thicknessfor a PWRfuel rod with a 60,000 MWD /MTU burnup could be as high as 140 microns (0.0055 in). Since the analysis must be bounding and consistent with the premise setforth in PNL-6189 (Page 3.9), the maximum cladding oxide thickness must be used when calculating hoop stress.

e.

Otherwise, the selected cladding oxide thickness data needs to be technicallyjustified.

R' soonse: The cladding oxide thicknesses used in the calculation, NUH004.0409, have e

been revised to be conservative based on measured data currently available for PWR fuel rods at higher burnups. Section 7.3 of the Supporting Information (refer to the response to question 7-1) included with this response also contains more details and graphs from recently published studies which support the use of 90 microns (0.00354 in) and 110 microns (0.00433 in) as the limiting oxide thicknesses for burnups of 45 and 50 GWD/MTU respectively for PWR fuel assemblies.

7-4 Evaluate the containmentpressure analysis with 100percentfuel rod rupture, assuming l.

a 30 percent releasefractionfor all sig?ificantfission gases (stable and radioactive) and 100 percent of thefill gasfor accident conditions, orjustify why such an evaluation does t

.~ tot need to be performed.

It is the staff's position that a 30 percent releasefraction of the totalfission gas inventory should be assumedfor containment pressure calculations. The applicant should account for stable and radioactiveforms ofKr, Xe, H, He, etc. The stajfalso modsfied the release fractionfor Xenonfrom 10 percent (as implied by NRC Regulatory Guide 1.25,

" Assumptions Usedfor Evaluating the Potential Radiological Consequences of a Fuel l

Handling Accident in the Fuel Handling and Storage Facilityfor Boiling and Pressurized Water Reactors, " March 1972) to 30 percent to be consistent with past practicefor pressure calculations and to conservatively accountfor higher burnup levels.

lV October 1998 72-1004 Amendment Page B.3 Revision 2 l

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g' Response: The containment pressure evaluation is performed to account for the higher fission gas inventory due to increased burnups. The analysis is documented in calculation tbj NUH004.0427, Revision 0, to determine the maximum operational pressures at normal, off-normal and accident conditions. This analysis assumes a 30% release fraction for all significant fission gases (Kr, Xe, H, He, etc.) and a 100% of the fill gas from each ruptured rod. The analysis results show that the maximum DSC internal pressures during normal and off-normal conditions are still within the design basis values used in the structural evaluation of DSC internal pressure cases. However, the DSC internal pressure during accident conditions (blockage of inlet and outlet vents) exceeds the previous accident condition design basis value of 50.3 psig. Another calculation was performed (Calculation NUH004.0242, Revision 0) which analyzed the DSC for this higher accident pressure. The results of this analysis show that the stresses within the DSC and basket components are still within allowables. The calculations NUH004.0427 and NUH004.0242 are provided with Attachment E of this submittal.

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October 1998 72-1004 Amendment Page B.4 Revision 2

Responses to August 26,1996 RAI (TAC No. L22252) 1.

Provide the calculated values ofgaseous and halogenfission product inventoryforfuel assemblies encompassing the range ofinitial enrichments, burnups, and acceptable cooling times in Tables 10.3-3 and 10.3-4 of the proposed CSAR license amendment.

Compare these inventories to the Kr-85 inventory assumed in the 100% fuel breach DSC leakage accident in Section 8.2.8 of the CSAR. Determine sf any of the qualifiedfuel assemblies would cause the calculated public dose to exceed the values currently listed in the CSARfor this postulated accident.

The CSAR accident analysisfor DSC leakage assumes that 30% of the Kr-85 inventory (6.300 Curies per MTIHM) of the stored spentfuel is released to the environment. Some of thefuelin thefuel qualification tables may either contain a higher inventory of Kr-85 or have a radiologically significant quantity ofother radioisotopes (i.e., H-3 and 1-129).

The radioisotopes can be releasedfrom thefuel as easily as Kr-85 and have not decayed sufficiently in the shorter allowable cooling timesfor some lower burnupfuel.

Response: VECTRA calculation NUH004.0510 has been revised (to Revision 2, provided in Attachment E) to include the quantity of Kr-85, H-3, and I-129 as a function of burnup and corresponding cooling time for the fuel included in the qualification table. Public doses at the site boundary have been calculated using this information and shown to

[

remain below the 10CFR72.106 allowables. CSAR Table 8.2-10 has been updated, as

(

shown in Attachment C of this document, to include the revised information. The bounding case remains the 40,000 mwd /MTU,5-year cooled case previously reported, but the calculated dose has increased from 53 mrem to 224 mrem due to the inclusion of H-3 in the calculation. Because the 40,000 mwd /MTU,5-year cooled case remaim bounding, the Kr-85 inventory is essentially unchanged. The maximum calculated skin dose is less than that currently reported in the CS AR and, therefore, has not been changed.

2.

Calculate the impact on public and worker occupational dosesfrom the higher neutron dose rates that would resultfrom the inclusion ofsome of thefuelin the quahfication table as indicated in the worksheets in Appendix A to Calculation NUH004.0510, Revision 1.

i These worksheets and this calculation are based on demonstrating that the HSM roof l

total (i.e., neutron and gamma) calculated dose rate as reported in the CSAR is not exceeded, but this dose rate location consists ofless than a 1% contributionfrom neutrons and more than 99% contributionfrom gamma radiation. However, the dose

[

rate at many other locations on the HSM and the transfer cask has a much larger l

contributionfrom neutron radiation. For example, Table 7.3 2 of the NUHOMS CSAR Revision 3A indicates that the HSM door total dose rate neutron component is 46% to 56% depending on whether the door is removed er in place. The transfer cask radial surface dose rate has a 28% neutron component. For the transfer cask top axial dose m

i OEtober 1998 72-1004 Amendment Page B.5 Revision 2

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rate, 79% of the total dose rate is due to neutrons. Some of thefuel quahfication I

worksheets show that the neutron source can be as high as twice the CSAR design basis neutron source. Thus,for HSM and TC dose rate locations where the neutron dose is a significantfraction of the total dose rate, a doubling of the neutron source term will j

significantly increase the total dose rate above the CSAR Section 7 calculated values, increases in HSM surface dose rate should be incorporated into a revised analysis of public dose (i.e., direct and air scatter dose rate at a distancefrom an array ofHSMs in an ISFSI). ' increases in TC dose rate directly afect the estimates of occupational dose to workersfor the process ofloadingfuel into the DSC and TC and placing the loaded DSC into an HSM. Although the revised calculated dose rates may still be less than the technical specification limit, the increased actual dose rates would result in a higher I

public and occupational dose.

Response: VECTRA calculation NUH004.0510 Revision 2 (provided as Attachment E) l includes the following changes to incorporate this comment:

The fuel qualification worksheets in Appendices A and B of NUH004.0510 have been revised to include both the HSM roof dose rate and the transfer cask radial surface dose rate as acceptance criteria. The HSM roof dose rate was selected because it is indicative of the surface dose rate over the bulk of the HSM. The cask radial surface dose rate was selected because it is indicative of the source from which most of the occupa:ional exposure is due.

f i

Dose rates for locations where the neutron dose is a significant fraction of the total e

N dose rates, including the HSM door and cask top and bottom ends, have been calculated in Appendix C of NUH004.0510. See the response to Comment 4 for more information.

Appendix E of NUH004.0510 has been added to estimate the impact of the fuel qualification on occupational exposures and offsite dose rates. Occupational exposures are conservatively shown to increase by less than 11%. Chapter 7 of the CSAR has been revised accordingly, as provided in Attachment C of this document.

Offsite dose rates must be calculated by the licensee for the actual conditions and fuel to be loaded at their site as required by CS AR Chapter 7 and 10CFR72.212.

Appendix E of NUH004.0510 includes a numerical estimate of the impact on the dose rate at a distance of 100 feet from the front of a 2x10 array of NUMOMS* HSMs.

3.

Explain how a conclusion of a 5% safety margin on photon source term is adequate to accountfor spectral efectsfor differentfuel source terms. Appendix C of VECTRA Calculation NUH004.0510 Revision 1 presents calculations of the effects ofphoton energy spectrum on calculated HSM roofdose rates that show spectrum effects offrom

+32.5% to -67.8% as compared to the relative source.

The inventory oflow burnupfuel at utilities and anecdotal evidencefrom the Oconee plant are not acceptable technicaljustification. The technical basisfor selecting a 5%

.f-safety margin as being adequate has not been demonstrated.

1 October 1998 72-1004 Amendment Page B.6 Revision 2 -

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Response: Calculation NUH004.0510 Revision 2 (provided in Attachment E) includes O'

gamma dose rate calculations using the actual photon spectrum for each burnup and enrichment case on both the PWR and BWR qualification tables. The use and justification of the 5% safety factor have, therefore, been deleted.

4.

Using previously reviewed and approved VECTRA shielding calculation computer codes and the same photon spectrum sources in Table 5 of VECTRA calculation NUH004.0510 Revision 1 Appendix C, analyze the photon energy source spectrum effects on HSMfront door and transfer cask radial and axial gamma dose rate.

The type and thickness ofshielding at these locations is sigmficantly differentfrom the l

HSM roof. Therefore, the change in surface dose ratefrom diferent photon source spectra may also be diferent thanfor the analyzed HSM roofcase.

Response: VECTRA calculation NUH004.0510 has been revised (to Revision 2) to include the following: (1) The transfer cask radial surface dose rate, including spectral effects, is now explicitly calculated for every case on the fuel qualification tables; (2) l Because the gamma spectrum is now explicitly included in the HSM roof and cask radial dose rate calculations, the spectral discussion in Appendix C of the calculation has been deleted; and (3) Appendix C of the revised calculation includes calculations of the l

gamma and neutron dose rates for the HSM door, cask top axial surface, and cask bottom axial surface as a function of burnup and cooling time as allowed by the qualification l f7 tables. The gamma dose rate calculations were performed using the same shielding codes

(-}

and models as were previously used and approved. Photon sources and spectra for each case were taken directly from the worksheets in Appendix A of the revised calculation.

CSAR Table 7.3-2, provided in Attachment C, has been revised to incorporate the dose rate results.

5.

Clanfy thefuel qualification Tables 10.3 3 and 10.3-4 of the proposed CSAR amended pages (C.8 and C.9), by appropriate labels, so that the user knows that the tabulated values between burnup and initial enrichment are minimum required years of cooling time after reactor core discharge.

There is a possibility ofhuman error in looking up these tables and interpreting the values as some other time units.

Response: The first note for both Table 10.3-3 and 10.3-4 states,"Use burnup and enrichment to lookup minimum cooling time in years.". This note has been augmented by adding the statement " Minimum required years of cooling time after reactor core discharge" to the caption of both tables. Refer to Attachment C of this submittal.

6.

Revise the text that explains the proposed CSARfuel guahfication tables (Tables 10.3-3 and 10.3-4) to accountfor uncertainties in initial U enrichment andfuel burnup.

Justify the technical basisfor the numerical values recommendedfor this revision.

l O October 1998 72-1004 Amendment Page B.7 Revision 2

,1 Commercial nuclearfuel is maintained with a tolerance or uncertainty on the value of (O)

U*" enrichment. The utility's determination ofindividualfuel assembly integrated burnup is also subject to a level ofuncertainty based on suchfactors as: fuel management computational tool accuracy, core thermal output accuracy, incorefuel power distribution hardware and software accuracy, and tolerances on the initial uranium content offuel assemblies.

Response: The first note referring to both Table 10.3-3 and 10.3-4 has been revised to include the following statement," Licensee is responsible for ensuring that uncertainties in fuel enrichment and burnup are correctly accounted for during fuel qualification."

Refer to Attachment C of this submittal.

7.

Revise CSAR Tables 10.31 and 10.3-2 to include a list ofspecific PWR and BWRfuel designs that can be stored in the standardized NUHOMS design. The delineatedfuel designs should be identified by cladding type (i.e., zircaloy), manufacturer, and geometry (i.e., Westinghouse 15x15). All thefuel designs must be bounded by the standardized NUHOMS structural, criticality, thermal, mechanical, shielding, and accident analyses.

Thefuel specification only identifies standard PWR and BWRfuel designs without specifying what is meant by " standard". Fuel manufacturers periodically develop new or modifiedfuel designs which may have not been analyzedfor storage in the NUHOMS design without technical evaluation and analysis. Foreign nuclearfuel vendors have been involved in the U.S. market, but their designs have not been analyzed in the CSAR.

9 Response: CSAR Tables 10.3-1 and 10.3-2 have been revised, as shown in Attachment C, to refer to CSAR Chapter 3 for lists of acceptable fuel designs. Additional tables have been added to CSAR Chapter 3 which define these acceptable fuel designs. Verification that these designs are bounded by the structural, criticality, thermal, mechanical, shielding, and accident analyses is provided in Appendix D of VECTRA Calculation NUH004.0510, Revision 2 (provided in Attachment E of this submittal).

8.

Provide the supporting documentation and technical basisfor the curves ofgeneric cladding stress limit vs. dry storage cladding temperature limit in Figures 1 and 2 of Calculation NUH004.0409 Revision O.

Neither PNL-6189 nor the referenced calculationfiles (DUK003.0203 Rev. 0, and NUH004.0410 Rev.1) provide any generic limit curvesfor cladding stress vs. cladding temperature beyond a 15-year cooling period. Calculation NUH004.0409 gives limit curvesforfuel cooling times of up to 30 years without technical basis. The information provided in the above referenced calculationfiles are only appropriatefor the design basis PWR and BWR spentfuel with the standard burnup (40 GWd/MTUfor PWR and 35 GWd/MTUfor BWR) as analyzed in the CSAR.

Response: The generic curves of cladding stress vs. dry storage cladding temperature limits in Figures 1 and 2 of the subject calculation are obtained using the computer code C}/

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DATING [ Reference 13 in VECTRA calculation NUH004.0409, Revision 1]. A copy of October 1998 72-1004 Amendment Page B.8 Revision 2

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the documentation on DATING is included as Attachment A to VECTRA calculation

(")

NUH004.0409, Revision 1. The DATING code was developed for the United States Department Of Energy by Pacific Northwest Laboratory (PNL) to provide a method for incorporating the CSFM Model described in PNL-6189 on IBM Compatible Personal Computers. VECTRA calculation NUH004.0409 has been revised (to Revision 1 provided in Attachment E of this submittal) to include the above.

9.

Provide technicaljustification that the selected values ofhot condition PWRfuel rod pressure and temperature are applicable to PWR spentfuel with a maximum burnup of 60 GWdoiTU The values used by the applicant in calculating cladding stress vs. cladding temperature for PWR spentfuel are appropriatefor afuel with a burnup of45 GWdaiTU. The assumed corrected value usedfor hot condition gas temperature should be 459*C (not the 459*F as indicated on page 5 of Calculation NUH004.0409). This temperature corresponds to a hot condition value of afuel with a burnup of a 40 GWdaiTU (as presented in Calculation DUK003.0203 Rev. 0). The applicant used a maximum PWR internalfuel rod pressure of 2416 psia which corresponds to a 45 GWdafTU burnup

[from PNL-6189] to representfuel with a 60 GWdafTU burnup. This is a non-conservative value. As burnup increases, the production offission gas andfission gas releasefraction would also increase. The combined effect would result in a higher internal rod pressure which would lead to a lower dry storage (cladding temperature) t limit.

Response: VECTRA calculation NUH004.0409 has been revised to include the hot condition PWR fuel rod pressure, temperature, and cladding oxide thickness for various bumups. These values are considered representative of the hot conditions expected for PWR fuel at these burnups.

10.

Explain the discrepancy in long-term storage cladding temperature limitfor 5-year cooled PWR spentfuel.

Using the methodpresented in NUH004.0409, the cladding temperature limitfor 5-year cooledl'WRfuel was calculated to be approximately 703*F. However, Table 2 of Calculat... NUH004.0409 lists a value of 721*F, without anyjustification.

Response: VECTRA calculation NUH004.0409 has been rev' ed to incorporate this comment.

j 11.

Provide a technicaljustificationfor the BWRfuel rod internalpressure calculations' 1

assumption of an initial (cold unirradiated) free (gas) volume of 2.93 in' in Calculation NUH004.0409 Revision O. Justify why this value would apply to irradiatedfuels.

During irradiation, the nuclearfuelpellet to-cladding gap closes, pellets crack, and the pellet stack swells both radially and axially. As a result, the available volumefor thefill O}

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andfission gas to occupy inside the rod is smaller than the cold unirradiated value.

October 1998 72-1004 Amendment Page B.9 Revision 2 I

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

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Response: VECTRA calculation NUH004.0409 has been revised to incorporate this

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comment. The methodology for calculating cladding temperature limit for BWR fuel l

assemblies is now identical to PWR fuel assemblies.

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October 1998 72-1004 Amendment Page B.10

Revision 2

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Ouestions and Answers from November 2.1994 Meeting U

(note that many of these responses have been superseded by Revision 1 of this submittal) 1.

Show that the efects ofchanging neutron and gamma spectra have been properly accountedfor.

We have included a 5% margin to account for differences in photon and neutron spectra which would drive calculated dose rates slightly higher or lower than that calculated for fuel with the design basis spectra.

In order to confirm the adequacy of this approach, we have performed ANISN calculations using representative burnup/ enrichment combinations from the Fuel Qualification Table. These calculations show that spectral effects (alone) drive dose rates downward for fuel with more than the design basis burnup (40,000 mwd /MTU) and upward for fuel with less than the design basis burnup. Appendix C of Calculation NUH004.0510, Revision 1 (Attached) concludes that the spectral effects for lower burnup fuel are outweighed by the overall lower source term.

2.

Demonstrate that the cladding temperature limits are appropriatefor the range of burnups and cooling times in the Fuel Quahfication Table.

i The original CSAR was written using one cladding temperature limit for each fuel type b

(PWR and BWR). Now, in order to address burnups and cooling times other than design basis fuel, we calculate the cladding temperature limit as a function of the fuel assembly burnup and cooling time. This is done using the same methodology (PNL-6189) as was used for the orig: il CSAR calculations.

l' The calculations are described in detail in Calculation NUH004.0409 (cladding temperature limits) and NUH004.0510 (Fuel Qualification Table). The '409 calculation l

develops the required cooling times given initial decay heat and burnup for PWR and BWR fuel. These results are combined with the shielding algorithms to define the Fuel Qualification Table in the '510 calculation.

Summarize the conservatisms in the Fuel Quahfication Table.

3.

A 5% safety factor in radiological source is applied to all Fuel Qualification Table e'

cases.

All required cooling times are rounded up to integer years.

The limiting parameter for either cladding temperature or shielding was used.

None of the conservatisms in the original safety analyses have been altered.

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l October 1998 72-1004 Amendment Page B.11 l

Revision 2

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Clarify that the same dose limits apply asfor thefuel specifications currently in the

. 4.

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CSAR.

The dose rate specifications in Chapter 10 of the CSAR are not changed by incorporating the Fuel Qualification Table.

5.

Clarify the instructionsfor use.

The instructions for use have been clarified and are noted in the draft SAR Section 10.3.1.

6.

What are the safety impacts of using the Fuel Quahfication Table approach.

Structural-No adverse effect because this change does not affect the weight of the fuel to be stored. Also, since the heat load of the fuel to be stored will not be allowed to exceed the design basis fuel assembly source term (1.0 kW/PWR assy,0.37kW/BWR assy), system temperatures will not rise above design values. This means that the temperatures used in the structural analysis will still be bounding with the use of the Fuel Qualification Table.

Criticality-No adverse effect because the maximum allowable enrichment is still limited to the original design values (4.0% PWR and BWR). Use of the Fuel Qualification Table does not allow storage of enrichments higher than 4.0E Thermal-No adverse effect because system temperatures will not exceed design values.

j This is because the fuel to be stored will'not be allowed to exceed the design basis fuel assembly source term (1.0 kW/PWR assy,0.37kW/BWR assy). Also, fuel cladding.

temperatures will still be below the limits as defined by PNL-6189. Fuel cladding temperatures were an important part in building the Fuel Qualification Table. Although

. the cladding temperature limit now depends on fuel cooling time and burnup, the basis for any given fuel assembly is still PNL-6189 as it was in the original analysis.

Mechanical-No adverse effect because the nuclear parameters of the fuel have no significant bearing on the fuel v;igist, size, etc.

Shieldiag-No adverse effect because the Fuel Qualification Table is designed to result in no condition which would result in the HSM surface dose rate exceeding the values calculated for the design basis fuel assembly. The dose rates calculated for the design basis fuel assembly were composed of a neutron and gamma component. Since radiological damage to tissue is defined in terms of REM, which is not sensitive to which

type of radiation or what mix make up the field, the important parameter to preserve is the total dose rate-not that of each component. For this reason, much more fuel can be stored using the Fuel Qualification Table that would otherwise be unnecessarily disqualified (usually due to neutron source). In this way, overall exposure reductions should be i

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October 1998 72-1004 Amendment Page B.12 Revision 2.

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possible because fuel planners have a wider selection of qualified longer-cooled fuel assemblies.

The HSM was selected as the system dose rate to preserve when constructing the Fuel j

Qualification Table because of the relative time span of storage vs. operations, and

]

because it is the HSM that results in the bulk of the exposure to the public and general j

plant population. Preserving the HSM surface dose rate does not, however, guarantee that j

other system dose rates (i.e. transfer cask, welding operations) will not change due to changes in the neutron / gamma dose rate ratio. These other dose rates will not, however, exceed specified dose rates even with large shifts in dose rate components. For instance, Section 1.2.6 of the NUHOMS* CoC limits DSC top end dose rates to 200 mrem /hr (flooded) and 400 mrem /hr (evacuated). Based on NUHOMS* operational experience, even a factor of two increase in neutron dose rate would not cause the observed DSC top end dose rates to exceed the specification. Likewise, the specifications for the transfer cask (Section 1.2.11 of the NUHOMS' CoC) are 200 mrem /hr (flooded) and 500 j

mrem /hr (dry). Even with a neutron dose increase of two times, the overall dose rate is j

still expected to be within specification.

f 7.

Ensure that the Fuel Qualification Table is not specific to only one reactor design.

There are two Fuel Qualification Tables: one for PWR fuel and one for BWR fuel designs.

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October 1998 72-1004 Amendment Page B.13 Revision 2

SUPPORTING INFORMATION FOR ATTACHMENT B This section contains the following report (s):

Gilbert, E. R.," Supporting Information for VECTRA Calculation Package Revision 1 NUH004.0409, Questions 7.1,7.2,7.3," Mohr and Associates, Richland Washington, May 1997.

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OO October 1998 72-1004 Amendment Page B.14 Revision 2

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