ML032731177

Licensing Basis Document Change Request (LBDCR) 01-03-2, Fuel Storage Requirements, Technical Specification 4.2

ML032731177
Person / Time
Site:	Millstone
Issue date:	09/18/2003
From:	Price J Dominion Nuclear Connecticut
To:	Document Control Desk, NRC/FSME
References
B18972
Download: ML032731177 (39)

Text

Dominion Nuclear Connecticut, Inc.

Millstone Power Station 4 t Do *i*

Rope Ferry Road Waterford, CT 06385 SEP I 8 2003 Docket No. 50-245 B18972 RE: 10 CFR 50.90 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Power Station, Unit No. 1 Licensing Basis Document Change Request (LBDCR) 01-03-2 Fuel Storaae Requirements, Technical Soecification 4.2 Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC), hereby proposes to amend Operating License DPR-21 by incorporating the attached proposed changes into the Technical Specifications of Millstone Unit No. 1.

DNC is proposing to change Design Features Technical Specification 4.2, Fuel Storage." This Technical Specification Change implements the following proposed changes:

• Eliminates all credit for Boraflex as a neutron absorber.

• Reduces the number of fuel assemblies allowed to be stored in the spent fuel pool from 3229 to 2959. Fuel will be prohibited from being stored in 270 specific storage rack locations. This is necessary to support the elimination of all credit for Boraflex.

• Changes the required spent fuel pool keff to < 0.95. This is necessary to support the elimination of all credit for Boraflex.

• Eliminates the Design Features requirements on new fuel storage, since Millstone Unit No. 1 is a plant that has ceased power operation and will no longer receive new fuel.

There are no physical changes in the plant hardware necessary to implement these changes. provides a discussion of the proposed changes and the Safety Summary, including the analyses demonstrating the proposed changes do not involve a Significant Hazards Consideration. Attachments 1 and 2 provide marked-up and retyped versions of the current Millstone Unit No. 1 Technical Specifications respectively. Attachment 3 provides a summary of the criticality analysis. Attachment 4 provides the Criticality Benchmark and Determination of Upper Sub-critical Limit (USL) for the criticality calculations.

Site Operations Review Committee and Management Safety Review Committee The Site Operations Review Committee and Management Safety Review Committee have reviewed and concurred with the determinations.

U.S. Nuclear Regulatory Commission B18972/Page 2 Schedule DNC requests approval of the proposed amendment by June 30, 2004, to support elimination of Boraflex testing requirements. Once approved, the amendment shall be implemented within 60 days.

State Notification In accordance with 10 CFR 50.91(b), a copy of this License Amendment Request is being provided to the State of Connecticut.

There are no regulatory commitments contained within this letter.

If you have any questions or require additional information, please contact Mr. David W. Dodson at 860-447-1791, extension 2346.

Very truly yours, DOMINION NUCLEAR CONNECTICUT, INC.

J. a Price Si OeVice President - Millstone Sworn to and subscribed before me this J day of v 42 v , 2003 Notary Public My Commission expires IOTARYPlBLIC MY COMMSSION WUORES /IU2005 W0%. %JlGGI IV^L FMaya;;

U.S. Nuclear Regulatory Commission B18972/Page 3

Enclosures:

1. Evaluation of Proposed Changes and Safety Summary Attachments:

1. Proposed Technical Specification Changes (Mark-Up)

2. Proposed Technical Specification Pages (Retyped)

3. Criticality Analysis Summary

4. Criticality Benchmark and Determination of Upper Sub-critical Limit (USL) for Criticality Calculations cc: H. J. Miller, Region I Administrator D. G. Holland, NRC Project Manager, Millstone Unit No. 1 J. R. Wray, NRC Inspector, Region 1, Millstone Unit No. .1 Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127

ENCLOSURE 1 Millstone Power Station, Unit No. 1 Licensing Basis Document Change Request (LBDCR) 01-03-2 Fuel Storage Requirements Evaluation of Proposed Changes and Safety Summary

U.S. Nuclear Regulatory Commission B18972/Enclosure 1/Page 1 ENCLOSURE 1 Millstone Power Station, Unit No. 1 Licensing Basis Document Change Request (LB DCR) 01-03-2 Fuel Storage Requirements Evaluation of Proposed Changes and Safety Summary

Subject:

Application for Amendment to Technical Specification (TS) 4.2, "Fuel Storage," to

• Eliminate All Credit For Boraflex As A Neutron Absorber

• Reduce The Allowable Number Of Fuel Assemblies To Be Stored In The Spent Fuel Pool

• Change The Required Spent Fuel Pool kff to < 0.95

• Eliminate The Design Features Requirements On New Fuel Storage

1.0 DESCRIPTION

The proposed amendment would revise Millstone Power Station, Unit No. 1 (MP1) Design Features TS 4.2, Fuel Storage," to address the following objectives:

• Eliminates all credit for Boraflex as a neutron absorber.

• Reduces the number of fuel assemblies allowed to be stored in the spent fuel pool from 3229 to 2959. Fuel will be prohibited from being stored in 270 specific storage rack locations. This is necessary to support the elimination of all credit for Boraflex.

• Changes the required spent fuel pool kff to 0.95. This is necessary to support the elimination of all credit for Boraflex.

• Eliminates the requirements on new fuel storage, since MP1 is a plant that has ceased power operation and will no longer receive new fuel.

There are no physical changes in the plant hardware necessary to implement these changes.

2.0 PROPOSED CHANGE

Specifically, the proposed changes would revise the following:

Design Features Section 4.2, Fuel Storage," section 4.2.1

• Section 4.2.1 is proposed to be marked as deleted.' Since MP1 is a plant that has ceased power operation and will no longer receive new fuel, there is no need for any new fuel storage Design Features requirements.

U.S. Nuclear Regulatory Commission B18972/Enclosure 1/Page 2 Design Features Section 4.2, Fuel Storage," Section 4.2.2

• Section 4.2.2 is proposed to have three changes. The first is to change the required kff from 0.90 to 0.95. The second change is to add the phrase and with no fuel allowed in the storage locations shown in Figure 4.1." The third change is to add Figure 4.1.

All of these changes are being made to be consistent with the criticality analysis submitted with this proposed amendment.

Design Features Section 4.2, "Fuel Storage," Section 4.2.3 Section 4.2.3 is proposed to have one change, which changes the maximum number of fuel assemblies allowed to be stored in the spent fuel pool from 3229 to 2959.

This change is being made to be consistent with the criticality analysis submitted with this proposed amendment.

3.0 BACKGROUND

MP1 was shut down for a normal refueling outage on November 4, 1995, and has not operated since. On November 19, 1995, transfer of all fuel assemblies from the reactor vessel into the spent fuel pool (SFP) was completed. On July 17, 1998, it was decided to permanently cease further operation of the plant. The Certification to the NRC of permanent cessation of operation and permanent removal of fuel from the reactor vessel in accordance with 10 CFR 50.82(a)(1)(i)

& (ii) was filed on July 21, 1998, at which time the facility license no longer authorized operation of the reactor or placement of fuel in the reactor vessel.

The MP1 SFP holds fuel assemblies, control rods, and small vessel components. The spent fuel storage racks consist of two types of storage racks. The spent fuel storage racks for the spent fuel assemblies are designed to assure sub-criticality in the SFP. The storage racks are an interconnected honeycomb array of square stainless steel boxes forming individual cells for spent fuel storage. 1045 storage cells contain Boraflex sheets on four sides, and 2184 storage cells contain boron carbide ( 4C) plates for neutron adsorption. The MP1 Technical Specification Section 4.2 provides requirements for fuel storage (new and spent fuel).

The reasons for the proposed changes are:

• Boraflex is subject to long-term degradation, and elimination of credit for Boraflex as a neutron absorber is a conservative action to ensure that the pool is maintained in a safe sub-critical condition.

• Since MP1 is a plant that has ceased power operation and will no longer receive new fuel, there is no need for any new fuel storage Design Features requirements.

• The proposed changes will support elimination of Boraflex testing requirements.

4.0 TECHNICAL ANALYSIS

The MP1 spent fuel storage racks consist of two types of storage racks. About two thirds of the fuel storage cells use B4C plates for reactivity control. About one third of the fuel storage cells use Boraflex for reactivity control. Proposed TS Figure 4.1 shows a general layout of the pool.

In Figure 4.1, the section 1-6 racks use B4C plates for neutron poison material, and the sections 7-10 racks use Boraflex neutron poison material. MP1 is a boiling water reactor that has an SFP with no soluble boron in the water.

U.S. Nuclear Regulatory Commission B18972IEnclosure 1/Page 3 Boraflex is subject to long-term degradation, and elimination of credit for Boraflex as a neutron absorber is conservative to ensure that the pool is maintained in a safe sub-critical condition.

The modifications to Design Features sections 4.2.2 and 4.2.3 are necessary to end credit for Boraflex as a neutron absorber.

To accommodate the elimination of Boraflex credit, for the storage racks containing Boraflex, Transnuclear Incorporated performed a revised criticality analysis. A summary of this revised criticality analysis is provided in Attachment 3. Attachment 4 provides the Criticality Benchmark and Determination of Upper Sub-critical Limit (USL) for the criticality calculations, as documented by Transnuclear Incorporated.

The revised criticality analysis showed, for the fuel storage cells that currently use Boraflex for reactivity control, if fuel is limited to a 3-out-of-4 fuel storage configuration, kff of the racks will be maintained < 0.95, on a 95/95 basis, without credit for any Boraflex. This criticality analysis conservatively assumes that all fuel in the SFP is at the most reactive condition possible, using the most reactive fuel design and at the most reactive normal operating temperature. Abnormal and accident conditions are also considered in the analysis. The revised criticality analysis for the Boraflex storage racks has no effect on the other storage racks in the SFP.

This revised criticality analysis therefore utilizes requirements which are different from the current Technical Specification requirements, in that the current Technical Specifications allow fuel to be stored in any storage location, and the SFP keff limit is < 0.90.

Thus there are two changes that are being made in the proposed Design Features Technical Specifications to implement no Boraflex credit. First, the allowable kff of the SFP is being increased from 0.90 to 0.95. Second, as required by proposed TS Figure 4-1, 270 specific fuel storage locations will be designated as "non-fuel" storage locations.

The use of 0.95 as the SFP k limit is an NRC accepted standard, as documented in the Standard Review Plan (section 9.1.2) and Regulatory Guide 1.13. Also, it is an accepted industry standard as documented in ANSI/ANS-57.2, "American National Standard Design Requirements for LWR Spent Fuel Storage Facilities at Nuclear Power Plants."

The requirement to designate 270 specific storage locations as non-fuel" locations will be accomplished by administrative controls. Procedures will be revised to prohibit fuel storage in the 270 specific locations designated by proposed TS Figure 4-1. Fuel has already been removed from these 270 specific storage locations. Fuel movements in the MP1 SFP are performed in accordance with procedures that require any fuel movement to be tracked by Material Transfer Forms. These procedures and Material Transfer Forms require dual verification that the fuel assembly is being removed from, and placed into, the correct fuel storage location. Also, since MP1 is a plant that has ceased power operation and will no longer receive new fuel, the amount of fuel movement in the SFP is minimal, such that the likelihood of a fuel movement error is very low. The criticality analysis has considered the impact of a hypothetical single fuel assembly misloaded in the worst possible configuration. The criticality analysis shows that a single misloaded fuel assembly will not cause kff to exceed 0.95, on a 95/95 basis. The criticality analysis is especially conservative since it assumes that all fuel in the pool is at its maximum reactivity at any time In life. In fact, most of the fuel in the pool is fuel that has been discharged with very low reactivity.

The attached criticality analysis (Attachment 3), and proposed TS Figure 4.1, document that the 270 prohibited fuel storage locations are generally a repeating 3-out-of-4 pattem. There are two fuel locations that are an exception to this repeating 3-out-of-4 pattern, and the justification for

U.S. Nuclear Regulatory Commission B18972/Enclosure 1/Page 4 these two locations are explained in the attached criticality analysis. The attached criticality analysis also explains that while fuel is prohibited from these 270 non-fuel" storage locations, non-fuel items may be stored in selected "non-fuel" locations.

The proposed change to the maximum number of fuel assemblies allowed to be stored in the SFP from 3229 to 2959 is made to reflect the reduction of 270 storage locations. Lowering the number of allowed fuel storage locations is a conservative action that will not impact any other storage rack or heat load analyses.

The deletion of the Design Features Section 4.2.1 requirements for new fuel storage is acceptable since MP1 is a plant that has ceased power operation, does not currently hold any new fuel, and will no longer receive new fuel. Since irradiated fuel cannot be removed from the SFP (due to radiological considerations) and placed into the new fuel racks, and new fuel can no longer be received, there is no possibility of fuel being placed in the new fuel storage racks, and hence no need for new fuel storage Design Features requirements.

In conclusion, the proposed changes do not impact the ability of the spent fuel storage racks to maintain their design function, to keep the fuel in a sub-critical and cooled condition under all normal conditions and postulated accidents.

5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration Dominion Nuclear Connecticut, Inc. (DNC) has evaluated whether or not a Significant Hazards Consideration (SHC) is involved with the proposed changes by focusing on the three standards set forth in 10 CFR 50.92(c) as discussed below:

The proposed changes do not Involve an SHC because the changes do not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

Accidents previously evaluated are the fuel handling accidents described in the Decommissioned Safety Analysis Report (DSAR), and a seismic event, which is considered as part of the spent fuel rack design.

Since there are no changes to plant hardware, nor any changes in how fuel is moved, there are no changes to the probability of a fuel handling accident. The consequences of a fuel handling accident are not affected, since none of the inputs to the fuel handling accident is affected.

The proposed changes affect the criticality analysis of the spent fuel storage racks. The spent fuel racks will continue to be able to perform their design function, which is to maintain the stored fuel in a sub-critical and cooled condition under all normal and postulated accident conditions. There are no physical hardware changes to the plant from these proposed changes. The revised criticality analysis submitted with these proposed changes demonstrates that fuel will be maintained in a sub-critical condition during all normal and postulated accident conditions, including the seismic event. Since there is no change in the ability of the fuel storage racks to maintain a sub-critical condition due to a seismic event, there is no change in the probability or consequences of this accident.

U.S. Nuclear Regulatory Commission B18972/Enclosure 1/Page 5 Reducing the amount of fuel storage is a conservative action, and the spent fuel racks were designed and licensed to allow empty, partially filled, or completely full storage racks. Thus the fuel racks will continue to be able to perform their design function to maintain the fuel in a coolable condition.

The change to the new fuel storage racks is to delete the Technical Specification requirements for the new fuel storage kff limits. Since MP1 is a plant that has ceased power operation and will no longer receive new fuel, there is no need for these Technical Specification requirements. There are no new fuel related accidents previously analyzed, therefore this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

In summary, the proposed changes do not involve an increase in the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

Since there are no changes to the plant equipment, there is no possibility of a new or different kind of accident being initiated or affected by equipment issues.

Reducing the number of fuel assemblies to be stored in the pool, and discontinuing credit for Boraflex are conservative changes that do not introduce any new or different kind of failure modes.

The changes made primarily affect the nuclear criticality analysis and do not create a new or different kind of accident. Changes in eliminating Boraflex credit, restricting fuel in certain storage locations, and changing the allowable kff limit are all impacts to the nuclear criticality analysis for the SFP. The SFP criticality analysis is part of the basic design of the system and is not an accident. The ability to maintain the SFP kff less than or equal to 0.95, as well as within the 10 CFR 50 Appendix A, General Design Criteria for Nuclear Power Plants," Criterion 62 "Prevention Of Criticality In Fuel Storage And Handling," (reference 6) criteria of sub-critical, have been evaluated. Criticality impacts are more appropriately discussed under the margin of safety criterion.

The change to the new fuel storage racks is to delete the Technical Specification requirements for the new fuel storage kff limits. Since MP1 is a plant that has ceased power operation and will no longer receive new fuel, there is no need for these Technical Specification requirements. Since Millstone 1 currently has no new fuel and new fuel cannot be received, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

In summary, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

The margin of safety relevant to the SFP is defined as (1) SFP kff remains sub-critical by an acceptable margin, and (2) the spent fuel in the SFP remains adequately cooled so that the fission product barriers remain intact.

U.S. Nuclear Regulatory Commission B18972/Enclosure 1/Page 6 The industry and regulatory accepted value for required sub-criticality margin in the SFP is to ensure that the kff of the SFP remains < 0.95 under all normal and postulated accident conditions. This is documented in the Standard Review Plan, Regulatory Guide 1.13, and ANSI/ANS-57.2, "American National Standard Design Requirements for LWR Spent Fuel Storage Facilities at Nuclear Power Plants." The current MP1 Technical Specifications require a more conservative value of 0.90 for SFP k. The proposed Design Features Technical Specification changes the maximum SFP keff from 0.90 to 0.95. This is not a significant reduction in the margin to safety since the proposed value of 0.95 is consistent with the accepted regulatory guidance for sub-criticality margin. The proposed criticality analysis demonstrates that the SFP keff remains < 0.95 on a 95/95 basis under all normal and postulated accident conditions, thus the required margin of criticality safety has been maintained.

The proposed changes conservatively reduce the amount of fuel that can be stored, and therefore do not affect the SFP cooling analysis. Therefore, the spent fuel In the SFP remains adequately cooled so that the fission product barriers remain intact.

The removal of Technical Specification requirements for the new fuel storage kf limits does not affect the margin of safety since new fuel can no longer be received.

Therefore, based on the above, the proposed changes do not involve a significant reduction in a margin of safety.

In summary, in accordance with 10 CFR 50.92, DNC has reviewed the proposed changes and has concluded that they do not involve an SHC. The basis for this conclusion is that the three criteria of 10 CFR 50.92(c) are not compromised.

5.2 Applicable Regulatory Requirements/Criteria 5.2.1 Regulations The regulatory basis for DESIGN FEATURES Section 4.2 Fuel Storage, is to ensure that stored fuel assemblies are maintained in a cool-able and sub-critical condition.

10 CFR Part 50 Appendix A, "General Design Criteria for Nuclear Power Plants," Criterion 62 "Prevention Of Criticality In Fuel Storage And Handling," (reference 6) requires that criticality in the fuel storage and handling system be prevented.

5.2.2 Design Bases (UFSAR)

DSAR Section 5.2 The MP1 design basis Fuel Handling Accident involves the dropping of a spent fuel assembly or other component onto the SFP storage area. The analysis assumes the rupture of all fuel rods in four fuel assemblies. Section 5.2 of the DSAR demonstrates that resulting doses are within 10 CFR Part 100 limits.

DSAR section 3.2.1.1

U.S. Nuclear Regulatory Commission B18972/Enclosure 1/Page 7 DSAR section 3.2.1.1 specifies the new fuel storage design bases. The principal design basis is to ensure that new fuel stored in the new fuel storage vault maintains a kff of less than 0.90 dry and kcf, of less than 0.95 in the flooded position.

DSAR section 3.2.1.2 DSAR section 3.2.1.2 specifies the spent fuel storage design bases. This section states the design basis is to ensure that the fuel stored in the SFP maintains a kff of less than 0.90 at all times, including postulated criticality accidents.

5.2.3 Approved Methodologies NUREG-0800 (reference 1), U.S. NRC Standard Review Plan, Section 9.1.2, revision 3, provides guidance to the NRC staff on the acceptable spent fuel storage ken value.

Memorandum to T. Collins from L. Kopp, (reference 2), Guidance On The Regulatory Requirements For Criticality Analysis Of Fuel Storage At LWR Power Plants", dated August 12, 1998, provides guidance concerning the regulatory requirements for criticality analysis, used by the Reactor System Branch, of new and spent fuel storage at light water reactors.

Regulatory Guide 1.13, revision 2 (draft) (reference 3), provides guidance to the NRC staff on the acceptable spent fuel storage kff value.

NUREG/CR-6361, "Criticality Benchmark Guide for LWR Fuel in Transportation and Storage Packages" (reference 4), 1997, provides information needed for determining the Upper Sub-critical Limit (USL).

5.2.4 Analysis The criticality analysis performed to support these proposed changes use the SCALE-4.4, Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation for Workstations and Personal Computers, CCC-545, ORNL.

The maximum kff value of 0.95 for spent fuel storage is in accordance with NUREG-0800 (reference 1), Regulatory Guide 1.13, (reference 3), Memorandum to T. Collins from L Kopp, dated August 12, 1998, (reference 2), and ANSI/ANS-57.2-1983 (reference 5).

The criticality analysis provided is in accordance with GDC 62 (reference 6). This criticality analysis justifies discontinuation of Boraflex credit. As a result of the analysis, there is a reduction of 270 in the number of available fuel storage locations in the SFP.

The proposed change to the maximum number of fuel assemblies allowed to be stored in the SFP from 3229 to 2959 is made to reflect the reduction of 270 storage locations. Lowering the number of allowed fuel storage locations is a conservative action that will not impact any other storage rack or heat load analyses.

The deletion of the Design Features requirements for new fuel storage is acceptable since MP1 is a plant that has ceased power operation and will no longer receive new fuel. Since irradiated fuel cannot be removed from the SFP (due to radiological considerations) and placed into the new fuel racks, and new fuel can no longer be received, there is no possibility of fuel being

U.S. Nuclear Regulatory Commission B18972/Enclosure 1/Page 8 placed in the new fuel storage racks, and hence no need for new fuel storage Design Features requirements.

5.2.5 Conclusion The criticality analysis demonstrates the MP1 SFP Boraflex racks can be maintained with a kef

• 0.95 on a 95/95 basis, without credit for Boraflex, provided fuel is stored in a 3-out-of-4 storage pattern. This kff value complies with 10 CFR 50 Appendix A, GDC 62.

The proposed change to the maximum number of fuel assemblies allowed to be stored in the SFP from 3229 to 2959, is a conservative action that will not impact any other storage rack or heat load analyses.

The deletion of the Design Features requirements for new fuel storage is acceptable since MP1 is a plant that has ceased power operation and will no longer receive new fuel.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

DNC has evaluated the proposed change against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.22. DNC has determined that the proposed changes meet the criteria for categorical exclusion set forth in 10 CFR 51.22(c). The proposed amendment also does not involve irreversible consequences in accordance with 10 CFR 50.92(b).

This determination is based on the fact that the changes are being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to use of a facility component located within the restricted area, as defined by 10 CFR 20, or that changes an inspection or a surveillance requirement, and the amendment requests meets the following specific criteria:

(i) The proposed change involves no significant hazards consideration.

As demonstrated above, the proposed changes do not involve a significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released off site.

No changes are being made in the types or amounts of any radiological effluents that may be released offsite during normal operation and design basis accidents.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

U.S. Nuclear Regulatory Commission B18972/Enclosure 1/Page 9 The proposed changes will not result in changes in the hardware of the facility. There will be no change in the level of controls or methodology used for processing radioactive effluents or handling of solid radioactive waste. There will be no change to the normal radiation levels within the plant. Therefore, there will be no increase in individual or cumulative occupational exposure resulting from the proposed changes.

7.0 REFERENCES

1 NUREG-0800, U.S. NRC Standard Review Plan, Section 9.1.2, revision 3.

2 Memorandum to T. Collins from L. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at LWR Power Plants," August 12, 1998.

3 Regulatory Guide 1.13, revision 2 (draft), "Spent Fuel Storage Facility Design Basis."

4 NUREG/CR-6361, Criticality Benchmark Guide for LWR Fuel in Transportation and Storage Packages," 1997.

5 ANSI/ANS-57.2-1983, ANS Design Requirements for LWR Spent Fuel Storage Facilities at Nuclear Power Plants."

6 10 CFR Part 50 Appendix A Criteria 62, "Prevention of Criticality in Fuel Storage and Handling."

ATTACHMENT 1 Millstone Power Station, Unit No. I Licensing Basis Document Change Request (LBDCR) 01-03-2 Fuel Storage Requirements PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

U.S. Nuclear Regulatory Commission B18972lAttachment 1/Page 1 Licensing Basis Document Change Request (LBDCR) 01-03-2 Marked UP Pages Technical Specification Page and Section Number Title of Section Amendment Numbers 4.2 Fuel Storage 4.0-1 Amendment 111 Figure 4.1 Millstone Unit No. 1 Spent Fuel Pool 4.0-2

SEP-18-2G02 14:15 VCe~,Iy I I I. w7e're g&LL%. ... e 4.0 4.0 DESIGN FEATURES 4.1 Site Location The Unit 1 Reactor Building Islocated on the site at Millstone Point in Waterford. Connecticut. The nearest site boundary on land is 2063 feet northeast of the reactor building (1627 feet northeast of the elevated staocl). which is the minimum distance to the boundary of the exclusion area as described in 10 CFR 100.3. No part of the site that is closer to the reactor building than 2063 feet shall be sold or leased except to Dominion Nuclear Connecticut, Inc. or its corporate affiliates for use in conjunction with normal utility operations.

4.2 Fuel Storage r-epl ace W'Iw 4.2.1 e new fuel storage facility shall be-such that the K,,ff dry is less an His les~s than 0.95.

4.2.2 The Ke,,of the spent fuel storage pool shall be less than or equal to o~gs If. This Kffvalue is satisfied with fuel assemblies having a maximum k-infinity of 1.24 in the normal reactor configuration at cold conditions, and an average U-235 enrichment of 3,8 welght percent or less) a i &'

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4.2.3 The number of fuel assemblies stored in the spent fuel storage pool shall not exceed bundles.

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ATTACHMENT 2 Millstone Power Station, Unit No. I Licensing Basis Document Change Request (LBDCR) 01-03-2 Fuel Storage Requirements PROPOSED TECHNICAL SPECIFICATION PAGES (RETYPED)

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The Unit I Reactor Building Is located on the site at Millstone Point in Waterford, Connecticut. The nearest site boundary on land is 2063 feet northeast of the reactor building (1627 feet northeast of the elevated stack), which Isthe minimum distance to the boundary of the exclusion area as described in 10 CFR 100.3. No part of the site that is closer to the reactor building than 2063 feet shall be sold or leased except to Dominion Nuclear Connecticut, Inc. or its corporate affiliates for use In conjunction with normal utility operations.

4.2 Fuel Storage 4.2.1 DELETED 4.2.2 The Keff of the spent fuel storage pool shall be less than or equal to 0.95.

This Keff value is satisfied with fuel assemblies having a maximum k-infinity of 1.24 in the normal reactor configuration at cold conditions, and an average U-235 enrichment of 3.8 weight percent or less, and with no fuel allowed in the storage locations shown in Figure 4.1. I 4.2.3 The number of fuel assemblies stored in the spent fuel storage pool shall not exceed 2959 bundles. I Millstone - Unit I 4.0-1 Amendment No. 406, 40 444,

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ATTACHMENT 3 Millstone Power Station, Unit No. 1 Ucensing Basis Document Change Request (LBDCR) 01-03-2 Fuel Storage Requirements Criticality Analysis Summary

U.S. Nuclear Regulatory Commission B18972/Attachment 3/Page 1 ATTACHMENT 3 Millstone Power Station, Unit No. 1 Licensing Basis Document Change Request (LBDCR) 01-03-2 Fuel Storage Requirements Criticality Analysis Summary

Background

The Millstone Unit No. 1 (MP1) spent fuel storage racks consist of two types of storage racks.

About two thirds of the fuel storage cells use boron carbide ( 4C) plates for reactivity control.

About one third of the fuel storage cells use Boraflex for reactivity control. Figure 4.1 shows a general layout of the pool. Infigure 4.1, the section 1-6 racks use B4C plates for neutron poison material, and the sections 7-10 racks use Boraflex neutron poison material.

The fuel storage racks which contain Boraflex have a nominal center to center spacing of 6.30 inches, with a fuel storage cavity square dimension of 6.06 inches. The nominal rack thickness is 0.075 inches of Stainless Steel, with a 0.090 inch cavity between adjacent rack walls. The cavity between adjacent rack walls contains the Boraflex.

Purpose To perform criticality calculations to demonstrate that the MP1 spent fuel racks which contain Boraflex, can be maintained with a k < 0.95 on a 95/95 basis, without credit for Boraflex.

Credit for Boraflex will not be necessary, provided the fuel is placed in a 3-out-of-4 fuel storage configuration, with every 4" storage location not containing any fuel. Figure 4.1 shows the 3-out-of-4 pattem, which results in 270 fuel storage locations being designated as non-fuer locations. Calculations will be performed using KENOVa from the SCALE-4.4 code package.

References (1) SCALE-4.4, Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation for Workstations and Personal Computers, CCC-545, ORNL.

Assumptions

• An analysis of a fully reflected 11 x 11 fuel storage rack will bound all Boraflex fuel storage racks since it is the largest in the pool.

• The only credit taken for radial leakage will be a layer of water between the rack and the mirror boundary. The thickness of that layer will conservatively be half the minimum distance between adjacent racks in the pool.

• The volume normally containing Boraflex will be replaced and modeled as full density water. No credit will be taken for any Boron-1 0.

U.S. Nuclear Regulatory Commission B1 8972/Attachment 3/Page 2

• The top end fitting, plenum and bottom end fitting is modeled as full density water. The water is a less efficient absorber while having an insignificant effect on reflection relative to the hardware. No credit is taken for axial leakage.

• The most reactive fuel assembly type will be used to represent all assemblies.

• The fuel inventory will be modeled using 95% density uranium dioxide (U02) with an effective enrichment based on a maximum assembly K-infinity of 1.24, in the cold in reactor lattice configuration. This will include the combined effects of bumup and burnable absorbers (like gadolinium) in the manufactured fuel, but still be conservative since each assembly will be at the maximum lifetime reactivity. The K-infinity value of 1.24 is based on the corresponding Technical Specification (TS) limit in the Design Section of TS. All fuel manufactured for MP1over its entire operational history has been designed to a maximum K-infinity of 1.24, in the cold in reactor lattice configuration.

• Nominal dimensions will be used and positive reactivity effects will be statistically combined (based on being independent) and added to the nominal kff. The following will be analyzed for possible increases in reactivity: manufacturing tolerances, eccentric assembly positioning, and channel swelling.

• The spent fuel pool kff is not to exceed 0.95.

• General Electric manufactured all fuel present in the pool, with the latest fuel design being GE-10.

• To neutronically isolate the Boraflex racks from the B4C racks, each Boraflex rack face adjacent to a B4C rack will have every other compartment empty of fuel.

Methodolo y The criticality analysis is performed using the 44 group SCALE 4.4 code system for the MP1 Spent Fuel Pool (SFP). The analysis will be based on, and support the 3-out-of-4 fuel loading configuration shown in Figure 4-1.

The first step is to determine the most reactive assembly that exists. Using that assembly, an equivalent enrichment is determined that corresponds to a K-infinity of 1.24. The most reactive fuel assembly type with the effective uranium-235 (U-235) enrichment corresponding to the maximum K-infinity of 1.24 will be used to represent all fuel in the MP1 SFP.

The next step is to model a nominal 11 x 11 array in a 3-out-of-4 configuration and determine the value for the system kff. The 11 x 11 array is chosen because it represents the largest rack in the pool. The array will be set up such that at least one face will have every other compartment empty on the interface to the B4C racks.

The model is then modified to account for abnormal conditions. Abnormal conditions include manufacturing tolerances, temperature changes, eccentric location of the assembly in the compartment, and zircalloy (Zr) channel swelling. Conditions regarding Boraflex are not included since no credit is taken for Boraflex. Fuel density Is not included since the most reactive condition is used in the nominal analysis. Each condition is modeled independently so a delta-K for each abnormal condition is found. The positive reactivity changes are statistically combined by way of root mean square, since they are independent, and a net bias is obtained.

Several accident scenarios are then considered and the most limiting is determined. These conditions include abnormal moderator temperature, dropped assembly that becomes

U.S. Nuclear Regulatory Commission BI8972/Attachment 3/Page 3 positioned next to or on top of the rack, a seismic event, and single worst case mis-loaded fuel assembly.

The most limiting accident k is then used in calculating the final results. The results are compared to the upper sub-critical limit (USL). The acceptance criteria is kff + net bias +

3sigma < USL, where sigma is the statistical deviation associated with the Monte Carlo method in KENOVa.

Determination of Representative Fuel Assembly Of the General Electric fuel types resident in the MP1 SFP, the 8 x 8 fuel assembly with two small waterholes and a channel of 100 mills yields the highest K-infinity and is therefore the most reactive fuel type. An effective enrichment that yields an in-core K-infinity of 1.24 is then determined through an iterative process. A nominal in-core assembly pitch of 6 inches is used with a mirror reflected boundary condition at all faces. The equivalent enrichment determined from this process is 2.085 w/o U-235.

The most reactive fuel assembly has the following dimensions:

• Fuel Pellet OD (outside diameter) is 0.411 inches of 95% Theoretical Density U02

• Clad OD is 0.483" of Zr

• Clad ID (inside diameter) is 0.419 inches

• Fuel Rod Pitch is 0.64 inches

• Fuel Channel is 5.48 inches square (Outer Dimension) with 0.1 inch wall thickness of Zr

• Fuel array is 8x8 with 62 fuel rods and 2 water rods Pool Models The first model developed is the nominal case that will be used to calculate the baseline ken.

The largest rack in the pool, which is an 11 by 11 array, is modeled with a mirror boundary condition.

The nominal model differs from the actual storage rack in the following conservative ways:

• Boraflex is replaced with full density water.

• Each fuel assembly is replaced with the most reactive fuel assembly design at its most reactive state anytime in life.

• A reflective boundary condition is located at the top and bottom of the active fuel. The top end fittings, bottom fittings and plenum are not modeled.

• An operating temperature of 363 degrees K is used for the water and the water is at full density. This is greater than the maximum normal operating temperature (150 degrees F), and is conservative due to a positive temperature coefficient. The positive temperature coefficient is verified by a resultant increasing in reactivity in the boiling accident case.

Once the baseline kff is determined, the baseline case is modified in order to develop the independent bias models. There are 4 cases that will make up the biases. The first deals with the possibility of zircalloy channel swelling. The second and third models consider the effects associated with manufacturing tolerances in the pitch of the compartments in the rack, and the thickness of the stainless steel in the rack. Two models are developed at each range of pitch dimensions and the positive case is used as the delta-K. Likewise, two models are also

U.S. Nuclear Regulatory Commission B18972/Attachment 3/Page 4 developed with a thinner and a thicker stainless steel dimension, and again the positive case is used as the delta-K. The final bias model is the most reactive effect of combined eccentric locations of the assemblies within their compartments. That is, the situation where the assemblies shift to a worst-case configuration. The determination of this basis requires several cases to ensure that the greatest delta-K is found. The 4 delta-K biases are combined by way of root mean square to obtain the net bias.

Three accident cases are modeled using nominal dimensions. For the first accident case, the baseline model is modified to show a positive moderator temperature coefficient exists by changing only the water temperature. Therefore, the maximum temperature effect on reactivity occurs with full density liquid solid (no voids) water at the saturation temperature that corresponds with the hydrostatic pressure at the bottom of the pool. The second accident case considers a dropped fuel assembly on top of, or adjacent to the rack. An assembly dropped on top of the rack need not be modeled because the combined distance of the plenum and top end fitting regions will neutronically isolate the assembly from the rack. The case where an assembly is dropped adjacent to the rack is bounded by an infinite nominal 3-out-of-4 lattice array. For this, a smaller array may be used with an even matrix of assemblies. The mirror boundary condition is replaced with a periodic boundary at all faces. This case also bounds the potential loss of spacing between adjacent racks due to a seismic event since no credit is taken for inter-rack spacing or radial leakage In this case. The final accident model is a worst-case single assembly misload. This is modeled by modifying the baseline model by removing a peripheral corner assembly and re-positioning it in the center of the array in a normally water filled compartment. The misload accident case yields the highest kef, and is the most limiting accident case that will be used to compare to, and meet the acceptance criteria.

Results Table 1 shows the results from the KENOVa cases. All KENOVa cases are performed using 16 million neutron histories.

Table 1 KENOVa Results Case Description kff Sigma 1 Baseline out-of-4, Mirror Boundary Condition 0.8799 0.0002 2 Bias - Swelling of Channel 0.8973 0.0002 3 Bias - Pitch Tolerance 0.8829 0.0002 4 < Bias - Stainless Steel Thickness Tolerance 0.8843 0.0002 5 Bias - Eccentric Positioning 0.8836 0.0002 6 Accident - Boiling Water 0.8857 0.0002 7 Accident - Dropped Fuel Assembly out-of-4 with 0.8805 0.0002 Periodic Boundary Conditions I_ I 8 Accident - Mis-load 0.9174 0.0002 From the above table, case 1 shows that the baseline 3-out-4 fuel storage condition, with no Boraflex credit, results in a nominal keff of 0.8799. Cases 2, 3, 4 and 5 show the results of the bias cases. The net bias from abnormal conditions, cases 2 through 5, is a statistical combination of the delta-k values for each case. The delta-K is simply the difference between the positive reactivity change and the nominal case. Table 2 shows the delta-K values.

U.S. Nuclear Regulatory Commission B18972lAttachment 3/Page 5 Table 2 Delta-K Values Used To Determine Net Bias Case Description Delta-K 2 Channel Bulge 0.0174 3 Pitch 0.0030 4 SS Thickness 0.0044 15 1Eccentric Position 0.0037 Net Bias (RMS) 0.0186 From Table 1, the most limiting accident case Is the single misload of fuel assembly. The boiling condition and fuel assembly drop accidents are not limiting. Also, since case 7 models a 3-out-of-4 condition in a periodic boundary condition, no credit is taken for any spacing between racks. Thus there is no concern for a seismic event reducing the spacing between racks.

Proposed 3-out-of-4 Fuel Storage Pattern Proposed Technical Specification Figure 4.1 shows the required 3-out-of-4 fuel storage pattern to ensure that the criticality analysis remains valid. As shown in Figure 4.1, spent fuel pool locations 9-1 -J and 9-1 -T allow fuel storage, and spent fuel pool locations 9-2-J and 9-2-T do not allow fuel storage. These locations do not have the repeating 3-out-of-4 pattern that is evident in the other Boraflex rack locations. This was necessary to avoid moving the 2 fuel assemblies from spent fuel pool locations 9-1-J and 9-1-T. It is difficult to move fuel from these 2 locations due to wall interferences. As a result, it was verified by additional KENOVa calculations that spent fuel pool locations 9-2-J and 9-2-T are adequate substitutes as "non-fuel" locations, instead of spent fuel pool locations 9-1-J and 9-1-T. KENOVa calculations performed by Dominion Nuclear Connecticut, Inc. (DNC) confirmed that there was no adverse effect due to this change.

Storaae of "Non-Fuel" Components in "Non-Fuel" Locations To minimize the effect of the loss of the 270 fuel storage locations, which are being designated as non-fuel locations by this proposed change, it is desirable to be able to store certain non-fuel" components in certain non-fuer storage locations. Examples of such non-fuel" components are dummy fuel assemblies. Additional KENOVa calculations were performed to demonstrate that storage of "non-fuel" components is acceptable in "non-fuer locations that are in peripheral storage locations which are either next to a wall or where there is no adjacent storage rack. Locations, which meet these requirements are the "non-fuel" storage locations shown in Figure 4.1, in row 1 of sections 7, 8, 9 and 10. KENOVa calculations performed by DNC confirmed that there was no adverse effect on kff to allowing "non-fuer components to be stored in these selected "non-fuel" locations.

In addition, there are 2 Boraflex coupon trees that were specifically evaluated with KENOVa to allow them to be stored in any "non-fuer location. These coupon trees have a small amount of Stainless Steel and the Boraflex coupons. Specific KENOVa calculations performed by DNC showed that there was no adverse effect on kff to allowing these Boraflex coupon trees to be stored in any "non-fuel" location.

Procedural controls will be used to limit the storage of non-fuel components to the locations described above.

U.S. Nuclear Regulatory Commission B18972/Attachment 3/Page 6 USL Determination The USL for the spent fuel pool was determined as shown in the benchmarking information provided in Attachment 4. Table 3 shows the 6 independent variables used, the equations calculated from USLSTATS, the value of each variable that corresponds to the MP1 analysis model, and the USL result for each independent variable. The lowest USL result is used in the final calculation for the acceptance criterion.

Table 3 Variable Equation MS value (x) USL Enrichment =0.9360 + (2.0656E-03)

• X 2.08 0.9403 Pin Pitch =0.9418 (X 2 1.505) 1.6256 0.9418 Assembly Spacing =0.9427 ( X 2 7.404) 16.002 0.9427 H/X =0.9421 (X 5255.997) 137 0.9421 Water:Fuel Volume =0.9417 (1.6000< X < 3.8830) 2.087 0.9417 Ratio AEF =0.9413 (X 0.130) 0.15707 0.9413 The USL used for this analysis is 0.9403.

Conclusion The acceptance criterion is the following:

Maximum kff = Most Limiting Accident kf + net bias + 3sigma < USL Substituting the results obtained in this analysis:

kff = (0.9174) + (0.0186) + 3 ( 0.0002) < 0.9403 kOff = 0.9366 < 0.9403 The MP1 spent fuel pool Boraflex storage racks can be maintained with a kff < 0.95, on a 95195 basis, without credit for Boraflex. To accomplish this, fuel storage is not allowed in the 270 storage locations designated in Figure 4-1.

ATTACHMENT 4 Millstone Power Station, Unit No. 1 Licensing Basis Document Change Request (LBDCR) 01-03-2 Fuel Storage Requirements Criticality Benchmark and Determination of Upper Sub-critical Limit (USL) for Criticality Calculations

TRANSNUCLEAR, INC.

01/04101 m Criticality Benchmark SHEET OF

,-WAED BY And Determination of USL for CALC. NO 10970-01 kA6;;~= DATE 23L. Millstone Unit I Spent Fuel Pool RE_ 0 1.

Purpose:

To benchmark and determine the Upper Subcritical Limit (USL) for the Boraflex racks in the Millstone Unit 1 Spent Fuel Pool.

2.

References:

I) NUREG/CR-6361, Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages, 1997

2) SCALE-4.4, Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation for Workstations and Personal Computers, CCC-545, ORNL

3) Transnuclear document E-1 6687, Test Report for Qualification of the SCALE 4.4 Program on the Transuclear PC With Windows NT 4.0

4) Transnuclear document E-16154, Test Report for Qualification of the USLSTATS Computer Program Version 1.3.4 on Transnuclear's Pentium-based Desktop Computers

5) NU Letter to USNRC, 5/5/88, B12844.

6) CSAS25 input/output files listed in Table I of this calculation.

7) USLSTATS input/output files listed in Table 2 of this calculation.

3. Method and Assumptions The results for this benchmark will be used in the analysis of the Millstone Unit I Spent Fuel Pool. The input files for all benchmarks are taken from References 1. The files are run in the CSAS25 sequence of the SCALE4.4 computer code using the 44 group cross section library generated from ENDF/B-IV data (Reference 2) on PC with the Windows NT 4.0 operating system (Reference 3).

An upper subcritical limit (USL) will be determined as a function of several experimental variables. The values will then be applied using Method 1 "confidence band with administrative margin", described in Section 4.1.1 of Reference 1. The administrative margin will be 0.05, and the confidence level -y, will be 0.95. Excel spreadsheet fictions and the ORNL program USLSTATS (Reference 4) versionl.3.7 are used to do the statistical analysis of the data.

4. Critical experiment characteristics Critical experiments were chosen to represent a well-moderated U0 2 fuel rod lattice without boron. A total of 95 experiments were found to be sufficient for this benchmark, and are listed in Table 1.

Six independent variables were used in the analysis: enrichment, pin pitch, water to fuel volume ratio, assembly spacing, Hydrogen to fissile material ratio (H/X), and the average energy of the fission-causing neutron (AEF).

TRANSNUCLEAR, INC.

4;O^r4 01104/01 T Criticality Benchmark 2 O I 1fP9PFDBY And Determination of USL for CALC N 10970-01 xrneCs-UWCKM BY DATE A L/,3)o Millstone Unit 1Spent Fuel Pool M_ 0

5. Analysis of the Results The data are analyzed to determine if there is a trend in the bias as a function of the independent variables. A least mean squares linear regression is performed to fit the data of Kfr as a function of the independent variable. This is done with both the ULSTATS program and with the Excel LINEST function. The results agree. The USLSTATS results can be found in Table 2. Table 2 also shows the USL value for each variable based on the parameter value corresponding to the Boraflex racks in the Millstone Unit 1 Spent Fuel Pool (Reference 5). Figures 1 through 6 show the Excel Plots.

Table 1: Critical Experiment Models and Results Filename Enrichmen Pitch Water to H/X AEF Spacing KENOVa t (cm) Fuel (cm) k-eff Ratio Ans33all 4.74 1.350 2.302 138.4 0.233406 5.00 1.0063 0.0017 Ans33a2 4.74 1.350 2.302 138.4 0.196468 2.50 1.0102 0.0018 Ans33a3 4.74 1.350 2.302 138.4 0.173790 10.00 1.0046 0.0018 Ans33ebl 4.74 1.350 2.302- 138.4 0.242133 2.50 0.9986 0.0018 Ans33eb2 4.74 1.350 2.302 138.4 0.206996 5.00 1.0078 0.0017 Ans33epl 4.74 1.350 2.302 . 138.4 0.251432 2.50 0.9971 0.0018 Ans33e2 4.74 1.350 2.302 138.4 0.223883 5.00 0.9972 0.0017 Ans33sg 4.74 1.350 2.302 138.4 0.195249 5.00 1.0000 0.0017 Ans33sty 2.46 1.410 2.302 138.4 0.264089 2.50 0.9940 0.0019 Bw1484c1 2.46 1.636 1.841 216.1 0.187696 1.64 0.9923 0.0016 Bw1484c2 2.46 1.636 1.841 216.1 0.147501 1.64 0.9928 0.0017 Bw1484sl 2.46 1.636 1.841 216.1 0.137700 6.54 0.9923 0.0015 Dsn399-1 4.74 1.350 2.302 138.2 0.228781 1.80 1.0081 0.0018 Dsn399-2 4.74 1.350 2.302 138.2 0.189871 5.80 0.9993 0.0018 P2438al 2.35 2.032 2.918 398.7 0.094854 8.67 0.9961 0.0015 P2438ba 2.35 2.032 2.918 398.7 0.114344 5.05 0.9979 0.0015 P2438cu 2.35 2.032 2.918 398.7 0.113884 6.62 0.9997 0.0016 P2438sig 2.35 2.032 2.918 398.7 0.094511 8.39 0.9973 0.0016 P2438ss 2.35 2.032 2.918 398.7 0.094766 6.88 0.9980 0.0015 P2438zr 2.35 2.032 2.918 398.7 0.094473 8.79 0.9994 0.0018 P2615al 4.31 2.540 3.883 256.1 0.113619 10.72 0.9995 0.0015 P2615ba 4.31 2.540 3.883 256.1 0.114344 6.72 0.9979 0.0015 P2615cd1 4.31 2.540 3.883 256.1 0.114824 7.82 0.9976 0.0017 P2615cd2 4.31 2.540 3.883 256.1 0.114877 5.68 0.9975 0.0018 P2615cu 4.31 2.540 3.883 256.1 0.113884 8.15 0.9997 0.0016 P2615ss 4.31 2.540 3.883 256.1 0.113472 8.58 0.9990 0.0016 P2615zr 4.31 2.540 3.883 256.1 0.112644 10.92 0.9975 0.0016 P282711 2.35 2.032 2.918 398.7 0.096719 13.72 1.0021 0.0015 P282712 2.35 2.032 2.918 398.7 0.094017 11.25 0.9991 0.0015 P282713.in 4.31 2.540 3.883 256.1 0.115864 20.78 1.0115 0.0017 P282714 4.31 2.540 3.883 256.1 0.114385 19.04 1.0065 0.0017 P2827slg 2.35 2.032 2.918 398.7 0.094010 8.31 0.9975 0.0014

RANSNUCLEAR INC.

^ 01/04/01 mL CriticalityBenchmark HET 3 OF 1 DBY And Detennination of USL for C NO 10970-01

) DAIE 25g4120 Millstone Unit I Spent Fuel Pool KIEV _ _ _ _ _

kECKE BY P2827u1 2.35 2.032 2.918 398.7 0.211617 11.83 0.9962 0.0017 P2827u2 2.35 2.032 2.918 398.7 0.175625 14.11 0.9983 0.0014 P2827u3. 4.31 2.540 3.883 256.1 0.384476 15.38 1.0009 0.0017 P2827u4 4.31 2.540 3.883 256.1 0.273686 15.32 1.0070 .0.0016 P3314bs1 2.35 1.684 1.600 218.6 0.172238 3.86. 0.9951 0.0016 P3314bs2 2.35 1.684 1.600 218.6 0.175347 3.46 0.9941 0.0015 P3314bs3 4.31 1.892 1.600 105.4 0.289730 7.23 0.9966 0.0016 P3314bs4 4.31 1.892 1.600 105.4 0.296121 6.63 1.0023 0.0017 P3314cd2 2.35 1.684 1.600 218.6 0.174323 3.04 0.9998 0.0020 P3314cu3 4.31 1.892 1.600 105.4 0.280592 10.36 0.9965 0.0016 P3314cu4 4.31 1.892 1.600 105.4 0.289876 7.61 1.0019 0.0016 P3314cu5 2.35 1.684 1.600 218.6 0.165325 5.24 0.9935 0.0017 P3314cu6 2.35 1.684 1.600 218.6 0.171740 2.60 0.9980 0.0014 P3314ss5 2.35 1.684 1.600 218.6 0.168590 7.80 0;9930 0.0018 P3602bb 4.31 1.892 1.600 105.4 0.299416 8.30 1.0053 0.0016 P3602bs1 2.35 1.684 1.600 218.6 0.176393 4.80 1.0013 0.0017 P3602bs2 4.31 1.892 1.600 105.4 0.295310 9.83 1.0053 0.0019 P3602cdi 2.35 1.684 1.600 218.6 0.179154 : 3.86 1.0030 0.0014 P3602cd2 4.31 1.892 1.600 105A 0.300122 8.94 1.0039 0.0017 P3602cu1 2.35 1.684 1.600 218.6 0.171736 7.79 1.0017 0.0015 P3602cu2 2.35 1.684 1.600 218.6 0.175462 5.43 0.9996 0.0017 P3602cu3 4.31 1.892 1.600 105.4 0.290218 13.47 1.0056 0.0018 P3602cu4 4.31 1.892 1.600 105.4 0.298483 10.57 1.0045 0.0017 P3602n11 2.35 1.684 1;600 218.6 . 0.180150 8.98 1.0037 0.0016 P3602n12 2.35 1.684 1.600 218.6 0.173530 9.58 1.0039 0.0017 P3602n13 2.35 1.684 1.600 218.6 0.165750 9.66 1.0036 0.0016 P3602n14 2.35 1.684 1.600 218.6 0.160689 8.54 0.9973 0.0015 P3602n21 2.35 2.032 2.918 398.7 0.094630 10.36 0.9984 0.0016 P3602n22 2.35 2.032 2.918 398.7 0.098028 11.20 1.0009 0.0013 P3602n31 4.31 1.892 1.600 105.4 0.314546 14.87 1.0088 0.0019 P3602n32 4.31 1.892 1.600 105.4 0.304293 15.74 1.0072 0.0017 P3602n33 4.31 1.892 1.600 105.4 0.294242 15.87 1.0077 0.0017 P3602n34 4.31 1.892 1.600 105.4 0.286765 15.84 1.0070 0.0018 P3602n35 4.31 1.892 1.600 105.4 0.281898 15.45 1.0020 0.0018 P3602n36 4.31 1.892 1.600 105A 0.273386 13.82 1.0004 0.0019 P3602n41 4.31 2.540 3.883 256.1 0.123201 12.89 1.0097 0.0016 P3602n42 4.31 2.540 3.883 256.1 0.116951 14.12 1.0072 0.0021 P3602n43 4.31 2.540 3.883 256.1 0.113131 12.44 1.0040 0.0017 P3602ss1 2.35 1.684 1.600 218.6 0.169699 8.28 1.0016 0.0016 P3602ss2 4.31 1.892 1.600 105.4 0.288965 13.75 1.0029 0.0016 P392611 2.35 1.684 1.600 218.6 0.173007 10.06 0.9993 0.0019 P392612 2.35 1.684 1.600 218.6 0.166382 10.11 1.0030 0.0016 P392613 2.35 1.684 1.600 218.6 0.158648 8.50 0.9984 0.0015 P392614 4.31 1.892 1.600 105.4 0.304508 17.74 1.0075 0.0017 P392615 4.31 1.892 1.600 105.4 0.294827 18.18 1.0077 0.0018 P392616 4.31 1.892 1.600 105.4 0.279892 17A3 1.0069 0.0016 P3926s1 2.35 1.684 1.600 218.6 0.159064 6.59 0.9932 0.0015 P3926s12 4.31 1.892 1.600 105.4 0.275653 12.97 1.0001 0.0017 P3926u1 2.35 1.684 1.600 218.6 0.405591 8.06 9.5 0.9948 0.96 0.0015 0.001

_32u _1__ 2.3 .6_84 .60 21.6 0.40_

P3926u2 2.35 1.684 1.600 218.6 1 0.340984 1 9.50 0.9962 0.0017

TRANSNUCLEAR, INC.

D^TE 01/04/01 _ Criticality Benchmark 4 it AMD BY And Determination of USL for C NO 10970-01 Mnntlt - DTE BM Millstone Unit I Spent Fuel Pool REV0 k 'KEG P3926u3 2.35 1.684 1.600 218.6 0.258156 9.19 0.9991 0.0018 P3926u4 4.31 1.892 1.600 105.4 0.641460 15.33 1.0020 0.0017 P3926u5 4.31 1.892 1.600 105.4 0.510424 19.24 1.0038 0.0017 P3926u6 4.31 1.892 1.600 105.4 0.451978 18.78 1.0006 0.0020 P62ft231 4.31 1.891 1.600 105.0 0.358545 5.67 1.0013 0.0019 P71M4f3 4.31 1.891 1.600 105.0 0.373236 5.19 1.0023 0.0017 P71M4v3 4.31 1.891 1.600 105.0 0.367731 5.19 0.9987 0.0020 P71M4v5 4.31 1 i.891 1.600 105.0 0.365618 5.19 1.0011 0.0016 P71f214r 4.31 1.891 1.600 105.0 0.368163 5.19 0.9959 0.0020 Pat8011 4.74 1.600 3.807 228.6 0.148334 2.00 1.0001 0.0017 Pat8012 4.74 1.600 3.807 228.6 0.143004 2.00 0.9992 0.0020 Pat8Ossl 4.74 1.600 3.807 228.6 0.148335 2.00 1.0020 0.0015 Pat80ss2 4.74 1.600 3.807 228.6 0.143878 2.00 0.9931 0.0022 Table 2

Results From USLSTATS USLSTATS Variable Equation MS Value USL File name (I)

MSENR Enrichment 0.9360 + (2.0656E-03)*X 2.08 0.9403 MSPITCH Pin Pitch - 0.9418 (X>s 1.505) 1.6256 0.9418 MSSPACE Assembly - 0.9427 (X >= 7.404) 16.002 0.9427 Spacing MSHX H/X = 0.9421 (X <=255.997) 137 0.9421 MSWF Water.Fuel = 0.9417 (1.6000<X<3.8830) 2.087 0.9417 Volume Ratio _0_9413 ______0_130___

MSAEF AEF =09413 (X>= 0.130) 0.15707 0.9413

TRANSNUCLEAR, INC.

. 01/04/01 __ Criticality Benchmark _ _ O F H

,"P6111RED8Y And Determination of USL for 10970-01 bae- DATE 91231 - Millstone Unit I Spent Fuel Pool REV 0 WECKM BY KENO k-eff as a Function of Enrichment toVO

- A h.

tO1OSD 0~ I I y 2.066E-03x+9.934E-01 I ODSO 1000 II 0.9950.

0.9900.

I 0.9850 0 0.5 1 5 2 2.5 3 3.5 4 45 5 Enrichment Figure 1: k-eff as a Function of Enrichment

TRANSNUCLEAR, INC.

DATE 01_4_01_

01/ 4/ 1 __

Criticality Benchmark

_SH___ __ __ __ __ ____ __

6 __ OF __ _

And Determination of USL for CA. 10970-01 W4.. DSTE 3/M Millstone Unit 1 Spent Fuel Pool 0 0EV

(~+IEa FoOoBY KENO kff as a Funceon o Pitch

• 160 1O000

y. 1999E-03x+9.970&01 0.9950 0.9900 0.9850 . . . .

0 0.5 1 *5 2 2.6 3 Pitch Figure 2: k-eff as a Function of Pin Pitch KEIJO k-eff as a Function of Assembly Spacing 10150 t1000

.tOWI0 0.9950 0Q9900 0.9850 0 6 v I 20 25 Assembly Spacing Figure 3: k-eff as a Function of Assembly Spacing

TRANSNUCLEAR, INC.

D^TE 01/04/01 Criticality Benchmark 7 7

a DBY And Determination of USL for CALC NO 10970-01

. . DATE 3 b . Millstone Unit Spent Fuel Pool 0 UaECKED 8Y KENO k-eff as a Function of H to X 1.0150 t.0100 t.0050 IL00 0.950 Q90900 tI 0 50 100 60, *200 250 300 350 400 450

- H:X Figure 4: k-eff as a Function of H to X Ratio KENO k-offas a Function of Water to Fuel Volume Ratio 10s0 1010 Y616254.t+ tOW

'10050 I 10000 -

II o  :

II R

E~ I 41E I

01900 4

11 0.9850

---C I 05 t Is 2 25 3 3.5 4 45 Wate t Fuel Voluna Ratlo Figure 5: k-eff as a Function of Water to Fuel Volume Ratio

I TRANSNUCLEAR, INC.

I ?

DATE 01/04/01 Criticality Benchmark SHEET __ OF And Determination of USL for CAM NO 10970-01 ASg

<- .. o 31231oD Millstone Unit I Spent Fuel Pool RE0 0 C#KCKED fYt KENO k-eff as a Function of AEF 1.050.

1.0100.

s0tom Y8 38n3x +9.89EO1 O 1.0000

• O S50 I'

0.9900 oAM 0.000000 0.100000 0.200000 0.300000 0.400000 0a.00000 0OOOO 0.700000 I AF Figure 6: k-eff as a Function of AEF

6. Conclusions The variable that yielded the lowest USL result was enrichment. The USL used for the Boraflex rack analysis in the Millstone Unit 1 Spent Fuel Pool will be 0.9403.

Appendix - Sample USLSTATS Output usistats: a utility to calculate upper subcriticat limits for criticality safety applications Version 1.3.6, December 15, 1998 Oak Ridge National Laboratory Input to statistical treatment from file:MSAEF.ln

Title:

PCBENCH AEF Proportion of the population = .995 Confidence of fit = .950 Confidence on proportion = 950

TRANSNUCLEAR, INC.

B DA(^TEX 01104/01 TI Criticality Benchmark SHEET _ OF __ _

And Determination of USL for CALC. NO 10970-01 ko& - DATE W Millstone Unit I Spent Fuel Pool REV. 0 WECKE BY Number of observations = 95 Minimum value of closed band = 0.00 Maximum value of closed band = 0.00 Administrative margin .= 0.05 Independent dependent deviation independent dependent deviation variable - x variable - y in y variable - x variable - y in y 2.33406E-01 1.00630E+00 1.70000E-03 2.95310E-01 1.00530E+00 1.90000E-03 1.96468E-01 1.01020E+00 1.80000E-03 1.79154E-01 1.00300E+00 1.40000E-03 1.73790E-01 1.00460E+00 1.80000E-03 3.00122E-01 1.00390E+00 1.70000E-03 2.42133E-01 9.98600E-01 1.80000E-03 1.71736E-01 1.00170E+00 1.500002-03 2.06996E-01 1.00780E+00 1.70000E-03 1.75462E-01 9.996002-01 1.70000E-03 2.51432E-01 9.97100E-01 1.80000E-03 2.90218E-01 1.00560E+00 1.80000E-03 2.23883E-01 9.97200E-01 1.70000E03 2.98483E-01 1.00450E+00 1.70000E-03 1.95249E-01 1.00000E+00 1.70000E-03 1.80150E-01 1.00370E+00 1.60000E-03 2.64089E-01 9.94000E-01 1.90000E43 1.73530E-01 1.0039QE+00 1.70000E-03 1.87696E-01 9.92300E-01 1.60000E-03 1.65750E-1 1.00360E+00 1.60000E-03 1.47501 E-01 9.92800E-01 1.70000E-03 1.60689E-01 9.97300E-01 1.50000E-03 1.377OOE-01 9.92300E-01 1.50000E-03 9.46300E-02 9.98400E-01 1.60000E-03 2.28781E-01 1.00810E+00 1.80000E-03. 9.80280E-02 1.00090E+00 1.30000E-03 1.89871 E-01 9.99300E-01 1.80000E-03 3.14546E-01 1.00880E+00 1.90000E-03 9.48540E-02 9.96100E-01 1.50000E-03 3.04293E-01 1.00720E+00 1.70000E-03 1.14344E-01 9.97900E-01 1.50000E-03 2.94242E-01 1.00770E+00 1.70000E-03 1.13884E-01 9.99700E-01 1.60000E-03 2.86765E-01 1.00700E+00 1.80000E-03 9.4511 OE-02 9.97300E-01 1.60000E-03 2.81898E-01 1.00200E+00 1.80000E-03 9.47660E-02 9.98000E-01 1.50000E-03 2.73386E-01 1.00040E+00 1.90000E-03 9.44730E-02 9.99400E-01 1.80000E-03 1.23201E-01 1.00970E+00 1.60000E-03 1.13619E-01 9.9950OE-01 1.50000E-03 1.16951E-01 1.00720E+00 2.10000E-03 1.14344E-01 9.97900E-01 1.50000E-03 1.13131E-01 1.00400E+00 1.70000E-03 1.14824E4-1 9.97600E-01 1.70000E-03 1.69699E-01 1.00160E+00 1.60000E03 1.14877E-01 9.97500E-01 1.80000E-03 2.88965E-01 1.00290E+00 1.6000E-03 1.13884E-01 9.99700E-01 1.60000E-03 1.73007E-01 9.99300E-01 1.90000E-03 1.13472E-01 9.99000E-01 1.60000E-03 1.66382E-01 1.00300E+00 1.60000E-03 1.12644E-01 9.97500E-01 1.60000E-03 1.58648E-01 9.98400E-01 1.50000E-03 9.67190E-02 1.00210E+00 1.500002-03 3.04508E-01 1.00750E+00 1.70000E-03 9.40170E-02 9.99100E-01 1.50000E-03 2.94827E-01 1.00770E+00 1.80000E-03 1.15864E-01 1.01150E+00 1.70000E03 2.79892E-01 1.00690E+00 1.600OOE-03 1.14385E-01 1.00650E+00 1.70000E-03 1.59064E-01 9.93200E-01 1.50000E-03 9.40100E-02 9.97500E-01 1.40000E-03 2.75653E-01. 1.00010E+00 1.70000E-03 2.11617E-01 9.96200E-01 1.70000E-03 4.05591E-01 9.94800E-01 1.50000E-03 1.75625E-01 9.98300E-01 1.40000E-03 3.40984E-01 9.96200E-01 1.70000E-03 3.84476E-01 1.00090E+00 1.70000E-03 2.58156E-01 9.99100E-01 1.80000E-03 2.73686E-01 1.00700E+00 1.60000E-03 6.41460E-01 1.00200E+00 1.70000E-03 i.72238E-01 9.951OOE-01 1.60000E-03 5.10424E-01 1.00380E+00 1.70000E-03 1.75347E-01 9.94100E-01 1.50000E-03 4.51978E-01 1.00060E+00 2.00000E-03 2.89730E-01 9.96600E-01 1.60000E-03 3.58545E-01 1.00130E+00 1.90000E-03 2.96121E-01 1.00230E+00 1.70000E03 3.73236E-01 1.00230E+00 1.70000E-03 1.74323E-01 9.99800E-01 2.00000E-03 3.67731 E-01 9.98700E-01 2.00000E-03 2.80592E-01 9.96500E-01 1.60000E-03 3.65618E-01 1.00110E+00 1.60000E-03 2.89876E-01 1.00190E+00 1.60000E-03 3.68163E-01 9.95900E-01 2.OOOOOE-03

TRANSNUCLEAR, INC.

w I DATE 01/04/01 TILE Criticality Benchmark _ 10

,*- 5 'BY And Determination of USL for c O 10970-01 jaWAk - DATE 323101 Millstone Unit I Spent Fuel Pool REV 0 1.65325E-01 9.93500E-01 1.70000-E03 1.48334£-01 1.00010E+00 1.70000E-03 1.71740E-01 9.98000E-01 I.40000E-03 1.43004E-01 9.99200E-01 2.00000E-03 1.68590E-01 9.93000E-01 1.80000E-03 1.48335E-01 1.00200E+00 1.50000E-03 2.99416E-01 1.00530E+00 1.60000E-03 I.43878E-01 9.93100E-01 2.20000E-03

• 1.76393E-01 1.00130E+00 1.700002-03 chi = 3.8947 (upper bound = 9.49). The data tests normal.

Output from statistical treatment PCBENCH AEF Number of data points (n) 95 Unear regression, k(X) 0.9989 + ( 8.6356E-03)*X Confidence on fit (1-gamma) input) 95.0%

Confidence on proportion (alpha) [inputj 95.0%

Proportion of population falling above lower tolerance interval (rho) inputl 99.5%

Minimum value of X 0,0940 Maxdmum value of X 0.6415 Average value of X 0.21570 Average value of k 1.00074.

Minimum value of k 0.99230 Variance of fit, s(kX)A2 1.9900E-05 Within variance, sw)r2 2.8562E-06 Pooled variance, s(p)A2 2.2757E-05 Pooled std. deviation, s(p) 4.7704E-03 C(alpha,rho)*s(p) 1.8501E-02 student-t @ (n-2,1-gamma) 1.66385E+00 Confidence band width, W 8.6681E-03 Minimum margin of subcriticality, C*s(p)-W 9.8329E-03 Upper subcritical limits: ( 9.40100E-02 <= X <= 0.64146 )

• • • • • *********** ****h***

USL Method I (Confidence Band with Administrative Margin) USL1 = 0.9402 + ( 8.6356E-03)*X (X < 0.12988 )

= 0.9413 Q(>= 0.130)

USL Method 2 (Single-Sided Uniform Width Closed Interval Approach) USL2 = 0.9804 + ( 8.6356E-03)*X (X < 0.12988 )

= 0.9815 (X>= 0.130)

USLs Evaluated Over Range of Parameter X X: 9.40E-2 0.17 0.25 0.33 0.41 0.49 0.56 0.64

TRANSNUCLEAR, INC.

01/04/01 ___ CriticalityBenchmark ______ II_11 And Detemination of USL for CAM L40 10970-01 Dy MiE 3 d-l Millstone Unit I Spent Fuel Pool REV 0 USL-1: 0.9410 0.9413 0.94i3 0.9413 0.9413 0.9413 0.9413 0.9413 USL-2: 0.9812 0.9815 0.9816 0.9815 0.9815 0.9815 0.9815 09815 Thus spake USLSTATS Finis.