ML18128A049

License Amendment Request Regarding Proposed Technical Specifications Changes for Spent Fuel Storage and New Fuel Storage

ML18128A049
Person / Time
Site:	Millstone
Issue date:	05/03/2018
From:	Mark D. Sartain Dominion Energy Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML18128A049 (393)

Text

PROPRIETARY INFORMATION-WITHHOLD UNDER 10 CFR 2.390

~

Dominion Energy Nuclear Connecticut, Inc.

Dominion 5000 Dominion Boulevard, Glen Allen, VA 23060 Dominion Energy.com May 3, 2018

=--' Energy 10 CFR 50.90 U.S. Nuclear Regulatory Commission Serial No.: 18-039 Attention: Document Control Desk NRA/DEA RO Washington, DC 20555 Docket No.: 50-423 License No.: NPF-49 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 .

LICENSE AMENDMENT REQUEST REGARDING PROPOSED TECHNICAL SPECIFICATIONS CHANGES FOR SPENT FUEL STORAGE AND NEW FUEL STORAGE Pursuant to 10 CFR 50.90, Dominion Energy Nuclear Connecticut, Inc. (DENC) is submitting a license amendment request to. revise the following Technical Specifications (TS):

• TS 1.40 "Storage Pattern"

• TS 1.41 "3-0UT-OF-4 and 4-0UT-OF-4"

• TS 3/4.9.1.2 "Boron Concentration"

• TS 3/4.9: 13 "Spent Fuel Pool - Reactivity"

• TS. 3/4.9.14 "Spent Fuel Pool - Storage Pattern" .

• Figure 3.9-1 "Minimum Fuel Assembly Burnup Versus Nominal Initial Enrichment for Region 1 4-0UT-OF-4 Fuel Storage Configuration"

• Figure 3.9-2 "Region 1 3-0UT-OF-4 Storage Fuel Assembly Loading Schematic"

• Figure 3.9-3 "Minimum Fuel Assembly Burnup and Decay Time Versus Nominal Initial Enrichment for Region 2 Storage Configuration"

• Figure 3.9-4 "Minimum Fuel Assembly Burnup and Decay Time Versus* Nominal Initial Enrichment for Region 3 Storage Configuration for Assemblies from Pre-Uprate (3411 MWt) Cores" * .

• Figure 3.9-5 "Minimum Fuel Assembly Burnup and Decay Time Versus Nominal Initial Enrichment for Region 3 Storage Configuration for Assemblies from Post-Uprate (3650 MWt), Co_res"

• TS 5.6.1.1 "Criticality."

Conforming changes to the associated Technical Specifications Bases (TSB) are also included in this license amendment request for information only.

DENC performed a criticality safety evaluation for fuel assembly storage in the Millstone Power Station Unit 3 (MPS3) Spent Fuel Pool (SFP) storage racks and New Fuel Storage Racks (NFSR) to support the proposed TS change using a new methodology described in .

Attachment 5 contains information that is being withheld from public disclosure under 10 CFR 2.390. Upon separation from Attachment 5, this letter is decontrolled.

Serial No.18-039 Docket No.: 50-423 Page 2 of 5 This proposed amendment implements the following conditions associated with fuel assembly storage at MPS3:

• Increase the TS minimum SFP soluble boron concentration.

• Revise allowed storage patterns and initial enrichment/burnup/decay time for fuel assemblies in the SFP to meet keff requirements under normal and accident conditions as follows:

o Region 1 will have an updated configuration as follows:

• Each Region 1 fuel storage rack will contain storage locations designated as Region 1A and Region 1B.

• Region 1A - Region 1A can store fresh fuel with an enrichment

~ 4.75 wt% U-235 and no integral burnable poison without restriction.

• Region 1A - Region 1A can store fresh fuel with an enrichment

~ 5.0 wt% U-235 and with a minimum of twelve (12) Integral Fuel Burnable Absorber (IFBA) rods.

• Region 1A - Region 1A can store fuel with an enrichment~ 5.0 wt% U-235 and with a minimum of 2.0 GWd/MTU burnup.

• Region 1B - Region 1B can store fresh fuel with an enrichment

~ 5.0 wt% U-235 and no integral burnable absorber without restriction.

o Regions 2 and 3 will have new burnup curves. The Region 3 burnup curve will include credit for decay time. * *

• The proposed change will permit the storage of any fuel assembly with an enrichment~ 5.0 wt% U-235 that contains a Rod Cluster Control Assembly (RCCA) in Region 2 without restriction (e.g., fresh fuel with no integral burnable absorber)

• For consistency, the criticality analysis for the New Fuel Storage Racks is also being revised using the updated methods for the SFP criticality analysis. This reanalysis will not require a TS change.

Information provided in the attachments to this letter is summarized below:

- Attachment 1 provides an evaluation of the proposed TS changes.

Attachment 2 provides proposed marked-up TS pages.

Attachment 3 provides marked-up TS Bases pages (for information only).

Attachment 4 contains the Criticality Safety Evaluation Checklist.

Attachment 5 contains the Criticality Safety Evaluation Report (Proprietary).

Attachment 6 contains the Criticality Safety Evaluation Report (Non-proprietary).

Attachment 7 contains the MPS3 Boron Dilution Analysis Attachment 8 contains the Holtec International, Inc. (Holtec) affidavit for withholding proprietary information from public disclosure.

Serial No.18-039 Docket No.: 50-423 Page 3 of 5 Attachment 9 contains the Westinghouse Electric Company, LLC (Westinghouse) application and accompanying affidavit for withholding proprietary information from public disclosure.

Since Attachment 5 contains information proprietary to Holtec and Westinghouse, it is supported by affidavits signed by the owners of the information. The affidavits set forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390.

Accordingly, it is respectfully requested that the proprietary information be withheld from public disclosure in accordance with 10 CFR2.390.

DENG has evaluated the proposed amendment and determined that it does not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for this determination is included in Attachment 1. DENG has also determined that operation with the proposed change will not result in any significant increase in the amount of effluents that may be released offsite or any significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed change.

The proposed TS change has been reviewed and approved by the MPS3 Facility Safety Review Committee.

DENG requests approval of the proposed amendment by April 30, 2019. DENG plans to fully implement the revised TS within 90 days after NRC approval of this proposed license amendment.

Should you have any questions in regard to this submittal, please contact Wanda D. Craft at (804) 273-4687.

Sincerely,

~- VICKI L. HULL NOTARY PUBLIC Mark D. Sartain REG. # 140542 Vice President - Nuclear Engineering and Fleet Support COMMONWEALTH OF VIRGINIA MY COMMISSlON EXPIRES 5131/2022 COMMONWEAL TH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, who is Vice President - Nuclear Engineering and Fleet Support of Dominion Energy Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

• Acknowledged before me this3 ~ti day of /11,g y , 2018.

My CommlMloo Expl~, 5 - 3 I * ;;/:I . . ~- L~

Notary Public

Serial No.18-039 Docket No.: 50-423 Page 4 of 5 Commitments made in this letter: None Attachments:

1. Discussion of Change

2. Marked-up Technical Specifications Pages

3. Marked-up Technical Specifications Bases Pages (for information only)

4. Criticality Analysis Checklist

5. Criticality Safety Evaluation Report - (Proprietary)

6. Criticality Safety Evaluation Report - (Non-proprietary)

7. Spent Fuel Pool Boron Dilution Analysis

8. Holtec, International Affidavit

9. Westinghouse Electric Company, LLC Application and Accompanying Affidavit

Serial No.18-039 Docket No.: 50-423 Page 5 of 5 cc: U.S. Nuclear Regulatory Commission - Region I Regional Administrator 2100 Renaissance Blvd.

Suite 100 King of Prussia, PA 19406-2713 Richard V. Guzman Senior NRC Project Manager U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 C2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No.18-039 Docket No. 50-423 Attachment 8 HOLTEC, INTERNATIONAL AFFIDAVIT DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

U.S. Nuclear Regulatory Commission Affidavit Supporting Dominion Energy Nuclear C:onnecticut Millstone Power Station Unit 3 License Amendment Request AFFIDAVIT PURSUANT TO 10 CFR 2.390 I, Kimberly Manzione, being duly sworn, depose and state as follows:

( 1) I have reviewed the information described in paragraph (2) which is sought to be withheld, and am authorized to apply for its withholding.

(2) The information sought to be withheld is information provided in Letter Serial Number 18-039 from Dominion Energy Nuclear Connecticut, Inc. to the NRC "Dominion Energy Nuclear Connecticut Inc. Millstone Power Station Unit 3 License Amendment ~equest Regarding Proposed Technical Specifications Changes for Spent Fuel Storage and New Fuel Storage." This letter contains Holtec Proprietary information.

(3) In making this application for withholding of proprietary infonnation of which it is the owner, Holtec International relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec'. 552(b)(4) and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10CFR Part 9.17(a)(4), 2.390(a)(4), and 2.390(b)(l) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all "confidential commercial information",

and some portions also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory ,comtnission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v.FDA, 704F2dl280 (DC Cir. 1983).

1 of 5

U.S. Nuclear Regulatory Commission Affidavit Supporting Dominion Energy Nuclear Connecticut Millstone Power Station Unit 3 License Amendment Request AFFIDAVIT PURSUANT TO 10 CFR 2.390 (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Holtec's competitors without license from Holtec International constitutes a competitive economic* advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.

c. Information which reveals cost or price information, production, capacities, budget levels, or commercial strategies of Holtec International, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future Holtec International customer-funded development plans and programs of potential commercial value to Holtec International;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4.a, 4~b and 4.e above.

(5) The information sought to be withheld is being submitted to the NRC in confidence. The information (including that compiled from many sources) is of a sort customarily held in confidence by Holtec International, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Holtec International. No public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence .. Its initial designation as 2 ofS

U.S. Nuclear Regulatory Commission Affidavit Supporting Dominion Energy Nuclear Connecticut Millstone Power Station Unit 3 License Amendment Request AFFIDAVIT PURSUANT TO 10 CFR 2.390 proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within Holtec International is limited on a ,;need to know" basis.* *

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his designee ), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside Holtec International are limited

• to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and* others with a legitimate need for the information, and then only iri

• accordance with appropriate regulatory provisions or proprietary agreements.

(' (8) The information classified as proprietary was developed and compiled by Holtec International at a significant cost to Holtec International. This information is classified as proprietary because it contains detailed descriptions of analytical approaches and methodologies not available elsewhere. This information would provide other parties, including competitors, with information from Holtec lnternational's technical database and the results of evaluations performed by Holtec International. A substantial effort has been expended by Holtec International to develop this information. Release of this information would improve a competitor's position because it would enable Holtec's competitor to copy our technology and offer it for sale in competition with our company, causing us financial lllJUry.

3 of 5

U.S. Nuclear Regulatory Commission Affidavit Supporting Dominion Energy Nuclear Connecticut Millstone Power Station Unit 3 License Amendment Request AFFIDAVIT PURSUANT TO 10 CFR 2.390 (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to Holtec International's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Holtec International's comprehensive spent fuel storage technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology, and includes development of the expertise to determine and apply the appropriate evaluation process.

The research, development, engineering, and analytical costs comprise a substantial investment of time and money by Holtec International.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

Holtec International's competitive advantage will be lost if its competitors are able to use the results of the Holtec International experience to normalize

. or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to Holtec International would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Holtec International of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

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U.S. Nuclear Regulatory Commission Affidavit Supporting Dominion Energy Nuclear Connecticut Millstone Power Station Unit 3 License Amendment Request AFFIDAVIT PURSUANT TO 10 CFR 2.390 ST ATE OF NEW JERSEY )

) ss:

COUNTY OF CAMDEN )

Kimberly Manzione, being duly sworn, deposes and says:

That she has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.

Executed at Camden, New Jersey, this 30th day of January, 2018.

Kimberly Manzione Licensing Manager Holtec International Subscribed and sworn before me this 30th day of January, 2018.

Erika Grandrimo NOTARY PUBLIC STATE OF NEW JERSEY MY CO'.\i\-llSSION EXPIRES January 17.2022 5 of 5

Serial No.18-039 Docket No. 50-423 Attachment 9 WESTINGHOUSE ELECTRIC COMPANY, LLC APPLICATION AND ACCOMPANYING AFFIDAVIT DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Westinghouse Non-Proprietary Class 3

@Westinghouse Westinghouse Electric Company 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 940-8542 11555 Rockville Pike e-mail: greshaja@westinghouse.com Rockville, MD 20852 CAW-18-4731 April 4, 2018 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

Millstone Unit 3 LAR Proprietary Class 2 Marked Pages (Proprietary)

The Application for Withholding Proprietary Information from Public Disclosure is submitted by Westinghouse Electric Company LLC ("Westinghouse"), pursuant to the provisions of paragraph (b )( 1) of Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations. It contains commercial strategic information proprietary to Westinghouse and customarily held in confidence.

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-18-4731 signed by the owner of the proprietary information, Westinghouse. The Affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Dominion.

Correspondence with respect to the proprietary aspects of the Application for Withholding or the Westinghouse Affidavit should reference CAW-18-4 731, and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 2, Suite 259, Cranberry Township, Pennsylvania 16066.

© 2018 Westinghouse Electric Company LLC. All Rights Reserved.

CAW-18-4731 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

I, James A. Gresham, am authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC ("Westinghouse") and declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.

Executed on: _{4--'--(+-f/+--Jf__

James A. Gresham, Manager Regulatory Compliance I

l__

3 CAW-18-4731 (1) I am Manager, Regulatory Compliance, Westinghouse Electric Company LLC ("Westinghouse"),

and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) I have personal know ledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 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 should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

4 CAW-18-4731 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(iii) There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

1 5 CAW-18-4731 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(__

(iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, is to be received in confidence by the Commission.

(v) The information sought to be* protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in CE-18-139 Revision 1 -Attachment 1, "Millstone Unit 3 LAR Proprietary Class 2 Marked Pages" (Proprietary), for submittal to the Commission, being transmitted by Dominion letter. The proprietary information as submitted by Westinghouse is that associated spent fuel pool storage information for criticality analyses, and may be used only for that purpose.

(a) This information is part of that which will enable Westinghouse to perform spent fuel pool analyses.

(b) Further, this information has substantial commercial value as follows:

(i) Westinghouse plans to sell the use of similar information to its customers for the purpose of spent fuel pool analysis.

6 CAW-18-4731 (ii) Westinghouse can sell support and defense of industry guidelines and acceptance criteria for plant-specific applications.

(iii) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and non-proprietary versions of a document, furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CPR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CPR 2.390(b)(l).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CPR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

Dominion Letter for Transmittal to the NRC The following paragraphs should be included in your letter to the NRC Document Control Desk:

Enclosed are:

1. "Millstone Unit 3 Spent Fuel Pool Analysis Proprietary Marking Identification" (Proprietary)

2. "Millstone Unit 3 Spent Fuel Pool Analysis Proprietary Marking Identification" (Non-Proprietary)

Also enclosed are the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-18-4731, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice.

As Item 1 contains information proprietary to Westinghouse Electric Company LLC ("Westinghouse"), it is supported by an Affidavit signed by Westinghouse, the owner of the information. The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the Nuclear Regulatory Commission ("Commission") and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations.

Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CPR Section 2.390 of the Commission's regulations.

Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse Affidavit should reference CAW-18-4731 and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 2, Suite 259, Cranberry Township, Pennsylvania 16066.

Serial No.18-039 Docket No. 50-423 Attachment 1 DISCUSSION OF CHANGE DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 Attachment 5 contains information that is being withheld from public disclosure under 10 CFR 2.390. Upon separation from Attachment 5, this letter is decontrolled.

Serial No.18-039 Docket No. 50-423 Attachment 1, Page 1 of 20 Discussion of Change Table of Contents 1.0 Summary Description ...........................................................................................2 2.0 Detailed Description of Change ........................................................................... 3 2.1 TS 1.40 - Storage Pattern and TS 1.41 0UT-OF-4 and 4-0UT-OF-4 .......... 3 2.2 TS 3/4.9.1.2 - Boron Concentration .................................................................... 3 2.3 TS 3/4.9.13 - Spent Fuel Pool - Reactivity ......................................................... 3 3.0 Discussion ...........................................................................................................7 3.1 Current MPS3 Spent Fuel Pool Configuration .................................................... 8 4.0 Technical Evaluation Summary ............................................................................ 9 4.1 Introduction ......................................................................................................... 9 4.2 Fuel Storage Criticality Analysis - General ........................................................ 10 4.3 New Fuel Storage Racks Criticality Analysis - Normal Storage and Accident Conditions ......................................................................................................... 10 4.4 Spent Fuel Pool Criticality Safety Evaluation - Normal Storage Conditions ...... 10 4.5 Spent Fuel Pool Criticality Safety Evaluation - Accident Conditions ................. 11 4.6 Boron Dilution ................................................................................................... 11

4. 7 Storage of Non-fuel Components and Non-standard Fuel Assemblies ............. 12 4.8 Spent Fuel Pool Storage - Other Items ............................................................ 12 4.9 Implementation Considerations ........................................................................ 13 4.10 Conclusions ...................................................................................................... 13 5.0 Regulatory Evaluation .............. :........... :............................................................. 14 5.1 Applicable Regulatory Requirements and Criteria ............................................ 14 5.2 No Significant Hazards Consideration .............................................................. 14 5.3 Precedents ........................................................................................................ 18 5.4 Conclusion .......................................................................................... :............. 19 6.0 Environmental Considerations ........................................................................... 20 7.0 References ........................................................................................................20 Attachment 5 contains information that is being withheld from public disclosure under 10 CFR 2.390. Upon separation from Attachment 5, this letter is decontrolled.

Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 2 of 20 1.0 Summary Description Dominion Energy Nuclear Connecticut, Inc. (DENG) hereby proposes to amend Operating License NPF-49 by incorporating the attached proposed changes into the Technical Specifications (TS) of the Millstone Power Station Unit 3 (MPS3). DENG is proposing to change the following TS:

• TS 1.40 "Storage Pattern"

• TS 1.41 "3-0UT-OF 4 and 4-0UT-OF-4"

• TS 3/4.9.1.2 "Boron Concentration"

• TS 3/4.9.13 "Spent Fuel Pool - Reactivity"

• TS 3/4.9.14 "Spent Fuel Pool - Storage Pattern"

• Figure 3.9-1 "Minimum Fuel Assembly Burnup Versus Nominal Initial Enrichment for Region 1 4-0UT-OF-4 Fuel Storage Configuration"

• Figure 3.9-2 "Region 1 3-0UT-OF-4 Storage Fuel Assembly Loading Schematic"

• Figure 3.9-3 "Minimum Fuel Assembly Burnup and Decay Time Versus Nominal Initial Enrichment for Region 2 Storage Configuration"

• Figure 3.9-4 "Minimum Fuel Assembly Burnup and Decay Time Versus Nominal Initial Enrichment for Region 3 Storage Configuration for Assemblies from Pre-Uprate (3411 MWt) Cores"

• Figure 3.9-5 "Minimum Fuel Assembly Burnup and Decay Time Versus Nominal Initial Enrichment for Region 3 Storage Configuration for Assemblies from Post-Uprate (3650 MWt) Cores"

• TS 5.6.1.1 "Criticality."

The associated Bases for TS 3/4.9.1.2, 3/4.9.13, and 3/4.9.14 are also being modified to address the proposed changes and are provided for information only. Changes to the TS Bases are controlled in accordance with the MPS TS Bases Control Program.

The Region 3 Spent Fuel Pool (SFP) storage racks contain Boraflex which is currently not credited as a neutron absorber. This proposed change will not credit Boraflex. Regions 1 and 2 racks contain BORAL as a neutron absorber which will continue to be credited.

To meet the SFP criticality requirements, the following changes are being proposed:

• Revise allowed storage patterns for fuel assemblies in the SFP to meet kett requirements under normal and accident conditions:

o Region 1 will no longer require the use of cell blocking devices.

• Update the surveillance requirements for storing fuel assemblies in Region 1, Region 2, and Region 3.

• Include new burnup curves for Regions 2 and 3. The Region 3 burnup curve will include credit for decay time.

The proposed change has been reviewed and confirmed to accommodate all fuel currently in the SFP and New Fuel Storage Racks (NFSR), and fuel assembly designs anticipated in the future.

Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 3 of 20 2.0 Detailed Description of Change 2.1 TS 1.40- Storage Pattern and TS 1.41- 3-0UT-OF-4 and 4-0UT-OF-4 These definitions are being deleted. They apply to the required fuel storage configuration imposed by cell blocking devices in designated Region 1 storage locations. Since the proposed change will remove the requirement for cell blocking devices, these definitions will no longer apply.

2.2 TS 3/4.9.1.2- Boron Concentration The proposed change to TS 3/4.9.1.2 will increase the licensed minimum SFP boron concentration from 800 ppm to 2600 ppm.

LCO 3.9.1.2 and Surveillance Requirement 4.9.1.2 LCO 3.9.1.2 is updated to increase the licensed minimum SFP boron concentration from 800 ppm to 2600 ppm. The updated boron concentration will maintain kett ~ 0.95 under the bounding postulated accident conditions, including allowance for biases and uncertainties.

• 2.3 TS 3/4.9. 13 - Spent Fuel Pool - Reactivity The TS title is being changed to "Spent Fuel Pool - Storage" to address the new SFP storage configurations and requir~ments. The criticality safety evaluation demonstrates that the SFP will maintain kett < 1.0 in an unborated water environment assuming each SFP rack storage location contains a fuel assembly with the highest reactivity allowed for that location. Also, the criticality safety evaluation demonstrates that the SFP will maintain kett ~ 0.95 at all times if the SFP soluble boron concentration is ~ 2600 ppm; thus, existing TS 3.9.13.a, which regards berating the SFP, will be deleted because the SFP boron concentration requirements will now be solely governed by TS 3/4.9.1.2.

Existing TS 3.9.13.b refers to four (4) burnup curves. These curves are being replaced with two (2) burnup curves, one for Region 2 and one for Region 3. The Region 3 burnup curve includes credit for decay time.

For simplicity, the proposed new burnup curves will now be designated as Figures 3.9-2 and 3.9-3. Figure 3.9-1 will become the new Fuel Assembly Loading Schematic for Region 1 spent fuel racks (this figure designates which storage locations are Region 1A and Region 1B). TS 3/4.9: 14, which specifies installation and removal of cell blocking devices, is being deleted since the proposed change will no longer require usage of cell blocking devices.

Two items need to be defined regarding enrichment and burnup. Initial enrichment, when used to compare to fuel storage requirements, is the "maximum initial planar volume averaged as-built U-235 enrichment" in the assembly. If the assembly has axial blankets the lower enriched fuel is not credited in determining the enrichment.

Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 4 of 20 Also, fuel burnup when used to compare to fuel storage requirements is the volume averaged burnup of the assembly as determined using the measured reaction rates with no reduction for measurement uncertainty.

Each Region 1 rack will designate each storage location as a Region 1A or Region 1B storage location. Region 1B storage locations will be permitted to store fresh fuel with an initial enrichment~ 5.0 wt% U-235 and with no burnable absorber. Region 1B is defined in the proposed Fuel Assembly Loading Schematic in Figure 3.9-1. In Region 1, the two rows adjacent to the west SFP wall are designated Region 1B, and the remaining locations designated as Region 1A.

Fuel may be stored in a Region 1A storage location if it meets one of the following criteria:

• Initial enrichment~ 4.75 wt% U-235, or

• Initial enrichment ~ 5.0 wt% U-235 and fuel burnup (measured) > 2.0 GWD/MTU, or

• Initial enrichment ~ 5.0 wt% U-235 and contains a minimum of twelve (12)

Integral Fuel Burnable Absorber (IFBA) rods.

There will no longer be a burnup curve specifically for pre-uprate fuel assemblies.

The proposed change accommodates a rated thermal power greater than any historical MPS3 cycle. These changes will also be reflected in the proposed changes to the TS Surveillance Requirements.

Changes to the TS are as follows:

LCO 3.9.13 Action a. will now only refer to Surveillance Requirement 4.9.13.1.1, Action b. will refer to Figure 3.9.2 (Region 2 burnup curve), and Action c. will refer to Figure 3.9.3 (Region 3 burnup curve), and will require immediate action if an assembly does not meet the relevant requirements. In Region 2, this LCO only applies to fuel assemblies that do not contain a RCCA. Action a. in the existing TS will be deleted since the proposed change to TS 3/4.9.1.2 will provide the requirement to borate the SFP should boron concentration fall below 2600 ppm.

Region 1A requirements include:

• Fuel assemblies with enrichment~ 4.75 wt% U-235 may be stored in Region 1A.

• A fuel assembly with enrichment > 4.75 wt % and ~ 5.0 wt% U-235 may be stored in a Region 1A storage location if fuel burnup is~ 2.0 GWD/MTU, or if it contains at least 12 IFBA rods.

Region 1B requirements include:

• Fuel assemblies with enrichment ~ 5.0 wt% U-235 may be stored in Region 18.

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Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 5 of 20 Region 2 requirements include:

• Fuel assemblies with a combination of initial enrichment and burnup in the "Acceptable" domain of the new burnup curve (Figure 3.9-2) may be stored in Region 2.

• A fuel assembly with enrichment_:::. 5.0 wt% U-235 that contains a RCCA may be stored in Region 2.

Region 3 requirements include:

• Fuel assemblies with a combination of initial enrichment, burnup, and decay time in the "Acceptable" domain of the new burnup curve (Figure 3.9-3) may be stored in Region 3.

• Decay Time is defined as the time elapsed since a fuel assembly was last used at power in a reactor core. A burnup curve is applicable to a fuel assembly if the fuel assembly decay time is greater than or equal to the burnup curve decay time.

Each burnup curve in Figures 3.9-2 and 3.9-3 is defined by a set of polynomial fit coefficients with the following format:

BU [GWD/MTU] = a 4

• wt% 4 + a3

• wt% 3 + a2

• wt% 2 + a1

• wt% 1 + a0 wt%: maximum initial planar volume averaged, as-built U-235 enrichment.

SR 4.9.13 Surveillance Requirements (SR) 4.9.13.1.1, 4.9.13.1.2, 4.9.13.1.3, and 4.9.13.1.4 are updated to instruct personnel to confirm that (1) initial enrichment, burnup, IFBA loading (Region 1), decay time (Region 3), and location of assemblies are acceptable based on the requirements in new Figures 3.9.1, 3.9.2, 3.9.3, and the Region 1A restrictions in Surveillance Requirement 4.9.13.1.1 or (2) that the assembly contains a RCCA for storage in Region 2.

TS Figure 3.9-1 TS Figure 3.9-1 is replaced. The new figure shows the bounding Region 1 fuel storage loading schematic, including which storage locations are designated as Region 1A and 1B.

TS Figure 3.9-2 TS Figure 3.9-2 is replaced. This new figure shows the minimum required fuel assembly burnup as a function of initial enrichment to permit storage in Region 2 (for assemblies that do not contain a RCCA).

Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 6 of 20 TS Figure 3.9-3 TS Figure 3.9-3 is replaced. This new figure shows the minimum required fuel assembly burnup and decay time as a function of initial enrichment to permit storage in Region 3.

TS Figure 3.9.4 TS Figure 3.9-4 is deleted.

TS Figure 3.9-5 TS Figure 3.9-5 is deleted.

TS 3/4.9.14 TS 3/4.9.14 is deleted. Cell blocking devices will no longer be required.

TS 5.6.1.1 The proposed change to TS 5.6.1.1 will update the summary of the fuel storage configuration and burnup/decay time requirements for each region consistent with the proposed TS changes above. This update will also indicate that kett will remain less than 1.0 with no soluble boron in the SFP water and ~ 0.95 with credit for soluble boron. TS 5.6.1.1 currently indicates that kett will remain less than ~ 0.95 with no soluble boron in the SFP water. A description of the NFSR will be added to TS 5.6.1.1, including that the NFSR will maintain kett ~ 0.95 under flooded conditions and

~ 0.98 under optimum moderator conditions.

Seriai' No.18-039 Docket No.: 50-423 Attachment 1, Page 7 of 20 3.0 Discussion DENC proposes to amend Operating License NPF-49 by incorporating the attached proposed changes into the MPS3 TS. A summary of the proposed changes is provided in Section 1.0 above. The supporting criticality safety evaluation report is included as Attachment 5 (Proprietary) and Attachment 6 (Non-proprietary).

Three un\que regions have been analyzed for the SFP, each with a somewhat different rack design.

• Region 1 racks have a flux trap design and contain BORAL neutron absorber panels. Region 1 racks are sub-divided into Region 1A and Region 1B.

Region 1B occupies two rows of Region 1 rack storage cells adjacent to the west SFP wall. Region 1B credits neutron leakage at the interface with the Region 1 wall.

• Region 2 racks contain BORAL neutron absorber panels, but do not use a flux trap design. A fuel assembly must meet the requirements of the Region 2 burnup curve to be stored in this region. However, any fuel assembly with initial enrichments 5.0 wt% U-235 that contains a RCCA may also be stored in Region 2 without restriction.

• Region 3 racks have a flux trap design and contain Boraflex neutron absorber material. Boraflex is not credited in the criticality safety analysis and is modeled as SFP water. A fuel assembly must meet the requirements of the Region 3 burnup curve, which credits decay time, to be stored in Region 3.

The rack design and the modeling of the racks are discussed in the criticality safety evaluation.

There are 1779 storage locations currently in the MPS3 SFP as follows:

• Region 1 (contains BORAL as a poison) o Five 7 x 10 racks

• Region 2 (contains BORAL as a poison) o Three 7 x 9 racks o One 7 x 10 rack o One 9 x 10 rack o Four 9 x 9 racks o There is one 9 x 9 rack that is licensed per Reference 1 but has not been installed. The proposed change continues to also assume that this rack is installed.

• Region 3 (contains Boraflex, which is not credited) o Twenty-one 6 x 6 racks.

Serial No.18-039 Docket No.: 50-423 Attachment 1 , Page 8 of 20 The criticality safety evaluation was performed for each region independently. The results verified that no adverse boundary effects occur at region interfaces. The analysis includes soluble boron credit.

• The present TS SFP boron concentration requirement of the SFP is 800 ppm of soluble boron. The proposed change will increase this requirement to 2600 ppm.

3. 1 Current MPS3 Spent Fuel Pool Configuration The MPS3 SFP contains 350 Region 1 storage locations, 673 Region 2 storage locations, and 756 Region 3 storage locations, for a total of 1779 fuel storage locations. An additional Region 2 rack with 81 storage locations is licensed to be placed in the spent fuel pool, if needed. With this additional rack installed, the Region 2 storage capacity is 754 storage locations. The total SFP storage capacity is limited to no more than 1860 fuel assemblies. The proposed license amendment will not make any changes to the licensed number of spent fuel racks or storage locations (1860).

Region 1 storage currently includes a 3-out-of-4 configuration which employs a cell blocker on the required empty cells. This configuration and the requirement for the use of cell blockers is eliminated in the proposed configuration.

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Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 9 of 20 4.0 Technical Evaluation Summary

4. 1 Introduction The analysis and results in this section are summarized from the criticality safety evaluation report (Attachment 5 [Proprietary]/Attachment 6 [Non-proprietary]). The topics presented here have the most significant impact with regard to the proposed license amendment.

The proposed license amendment does not result in any equipment modifications to the plant or any changes regarding how equipment is operated and maintained.

There are no changes in how fuel assemblies are handled and moved, nor are there any changes in how they are inserted into or removed from a SFP or NFSR storage location. There are no changes to how RCCAs are handled and moved, nor are there any changes in how they are inserted into or removed from a fuel assembly.

There is no change to how personnel qualify and verify fuel assembly storage in either the SFP or the NFSR. The existing four burnup curves are being replaced with two burnup curves, one for Region 2 and one for Region 3. The Region 3 burnup curve will credit decay time. The burnup curves no longer distinguish between pre-uprate and post-uprate fuel and apply to all fuel currently in the SFP and NFSR, as well as anticipated future fuel designs. The proposed change is analyzed for a maximum power level of 3725 megawatts thermal.

Each Region 1 rack will have storage locations designated as Region 1A and Region 1B. A proposed new Fuel Assembly Storage Schematic (TS Figure 3.9-1) specifies which storage locations are Region 1A and which are Region 1B.

The administrative process for verifying proper fuel assembly loading using the new Region 1 storage configuration and new Region 2 and 3 burnup curves will be the same as the process used for the present burnup curves and storage configuration.

Also, the response to a fuel assembly misloading event remains the same. There are no changes regarding how fuel assemblies are handled and moved, or in the administrative means used to ensure that fuel assemblies are not dropped or misloaded.

The TS requirement for SFP soluble boron will increase from 800 ppm to 2600 ppm.

The process of controlling and measuring boron concentration, and responding should the concentration be found to be below the requirement, will remain the same.

Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 1O of 20 4.2 Fuel Storage Criticality Analysis - General For each normal operational and postulated accident scenario (both in the NFSR and the SFP) the criticality safety evaluation conservatively assumes that each fuel assembly is of a design to maximize fuel reactivity. The limiting design is a bounding composite of current and historical fuel designs that is expected to remain bounding for future fuel designs. The analyses assume bounding conservative depletion conditions for depleted fuel and account for biases and uncertainties as described in the criticality safety evaluation.

4.3 New Fuel Storage Racks Criticality Analysis - Normal Storage and Accident Conditions Results of the criticality safety evaluation show that the NFSR maintain kett .::. 0.95 under normal storage conditions, accounting for biases and uncertainties.

The following postulated accident conditions were analyzed:

• NFSR area fully flooded (with water)

• NFSR area with optimum moderation (fully flooded with foam).

The results also show that the NFSR maintain kett.::. 0.95 for the area fully flooded scenario, and maintain kett.::. 0.98 for the optimum moderation scenario.

4.4 Spent Fuel Pool Criticality Safety Evaluation - Normal Storage Conditions Results of the SFP criticality safety evaluation show that the proposed SFP storage configuration will maintain kett <1.0 with O ppm of soluble boron in the SFP water, and kett.::. 0.95 with the SFP filled with 600 ppm soluble boron for normal fuel assembly storage conditions (including biases and uncertainties). Each Region 1 rack has storage locations designated as Region 1A or Region 1B per the configuration specified in new Figure 3.9-1.

Fresh fuel with enrichments 5.0 wt% may be stored in Region 1B. In order to remain bounded by the criticality safety evaluation, fuel assemblies must meet one of the following requirements for Region 1A storage:

• Enrichment.::. 4. 75 wt% U-235, or

• Enrichment.::. 5.0 wt% U-235 with fuel burnup ~ 2.0 GWD/MTU, or

• Enrichment.::. 5.0 wt% U-235 with at least 12 IFBA rods Fuel assemblies may be stored in Region 2 if they are in the acceptable burnup domain of the initial enrichment/burnup curve in Figure 3.9-2, or contain a RCCA (no burnup restriction).

Fuel assemblies may be stored in Region 3 if they are in the acceptable burnup/decay time domain of the initial enrichment/burnup/decay time curve in Figure 3.9-3.

Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 11 of 20 4.5 Spent Fuel Pool Criticality Safety Evaluation - Accident Conditions The SFP criticality safety evaluation has analyzed the following postulated accident conditions:

• single fuel assembly mislead in the spent fuel racks

• multiple fuel assembly mislead in the spent fuel racks

• loss of SFP cooling including partial voiding

• dropped assembly into the racks with grid damage (optimum fuel pin pitch)

• mislead of an assembly between fuel racks

• fuel handling error (two fresh fuel assemblies out of rack in close proximity)

• seismic event.

The bounding accident is the multiple fuel assembly mislead in Region 2 spent fuel racks. This scenario assumes a 6 x 4 storage array of the limiting fresh fuel assembly (5.0 wt% U-235 enrichment with no burnable poison) surrounded by fresh fuel assemblies with 5.0 wt% U-235 enrichment and 32 IFBA rods in a 20x20 storage cell array.

The criticality safety evaluation shows that 2600 ppm of soluble boron maintains keff ~

0.95 for this bounding scenario, accounting for biases and uncertainties.

The proposed change will increase the minimum SFP boron concentration to 2600 ppm.

4. 6 Boron Dilution The proposed change will increase the SFP minimum soluble boron concentration requirement from 800 ppm to 2600 ppm. No equipment that could contribute to or mitigate a boron dilution event will be changed as part of this proposed change.

Thus, no new avenues for a boron dilution event will be created. There are no proposed changes regarding boron concentration maintenance or response to a boron dilution event.

DENG performed a SFP boron dilution analysis that assumes a dilution from 2300 ppm to 700 ppm soluble boron (summarized in Attachment 7). The systems that could dilute SFP boron, either by direct connection to the spent fuel pool or by a potential pipe crack/break, were analyzed via a bleed and feed methodology. The analysis demonstrates that sufficient time is available to detect and mitigate a boron dilution event prior to reaching a concentration of 700 ppm.

Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 12 of 20

4. 7 Storage of Non-fuel Components and Non-standard Fuel Assemblies The criticality safety evaluation concludes that non-fuel components may be placed in any spent fuel rack storage location since these components are less reactive than fuel. Non-fuel components can also be placed in the guide tubes of any fuel assembly because the fuel lattice is under moderated.

The criticality safety evaluation evaluated each existing non-standard fuel assembly currently stored in the MPS3 SFP using the same methods used for standard fuel.

The reactivity of non-standard fuel assemblies in the SFP is bounded by the limiting assembly design. The criticality safety evaluation provides a list of the non-standard fuel assemblies analyzed.

4.8 Spent Fuel Pool Storage- Other Items Seismic Response:

The criticality safety evaluation analyzed the reactivity impact of a postulated seismic event under the proposed storage requirements, and found that the spent fuel racks maintain kett ~ 0.95. The evaluation considered fuel assembly and spent fuel rack motion during the event. The reactivity impact of the seismic event is bounded by the multiple fuel misload.

The MPS3 SFP is currently licensed to store 1860 fuel assemblies per MPS3 TS 5.6.3 (this includes the capacity of one Region 2 rack that has not been placed into the SFP). The proposed change will not change the number of fuel storage locations or fuel assemblies, or physically change any of the spent fuel racks. Thus, the SFP seismic/structural loading requirements for the proposed change are bounded by the existing TS.

Radiological and Thermal Impact:

DENC has considered the impact of the proposed change on the MPS3 licensing basis fuel handling accident dose consequence assessment and the SFP heat load.

The proposed change does not alter the existing limits on enrichment, burnup limits, peaking factors, or gap fractions in the MPS3 core. Therefore, there is no impact on the radiological assessment or the SFP heat load.

Safety Analysis Limits:

Other than the proposed changes to the TS which specify SFP Region 1 storage configuration, the SFP soluble boron concentration, and the reactivity credit taken for assemblies containing RCCAs in Region 2, there are no changes to any TS LCO or operating or safety-related setpoints associated with the proposed change.

Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 13 of 20 4.9 Implementation Considerations DENG plans to fully implement the revised TS within 90 days after NRC approval of this proposed license amendment.

4. 10 Conclusions Implementation of the proposed license amendment is safe and will not negatively affect plant operation. The proposed change will make no modifications to plant equipment or how equipment is operated or maintained. In particular, there are no changes to how fuel is handled, including how fuel is moved, inserted into, and removed from SFP and NFSR storage locations. There are no changes to how RCCAs are handled, including how they are moved, inserted into, and removed from a fuel assembly. There are no changes to qualifying and verifying fuel storage in the SFP. There are no changes to the required response to a fuel misloading or drop event. Also, since the proposed license amendment does not modify plant equipment or its operation and maintenance, including equipment used to maintain SFP soluble boron levels, the proposed license amendment will not impact a boron dilution event or plant response to it.

SFP fuel storage requirements will continue to be maintained by administrative means to ensure compliance with the proposed Region 1 storage configuration and fresh fuel enrichment requirements, the proposed burnup curves for Regions 2 and 3, and the allowance to credit RCCAs contained in fuel assemblies for Region 2. The consequences of, or plant response to, a fuel misload event are not changed.

The criticality safety evaluation shows that the NFSR will maintain kett.:: 0.95 in the fully flooded condition and kett ,:: 0.98 in the optimum moderation condition.

Furthermore, the criticality safety evaluation shows that the SFP will maintain kett .::

0.95 under normal and postulated accident conditions with credit for soluble boron.

The SFP will also maintain kett < 1.0 with no soluble boron under normal conditions under the most reactive fuel storage configuration allowed by the proposed change.

Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 14 of 20 5.0 Regulatory Evaluation

5. 1 Applicable Regulatory Requirements and Criteria Appendix A to Title 10 of the Code of Federal Regulations, Part 50 (10 CFR 50),

General Design Criterion (GDC) 62, "Prevention of criticality in fuel storage and handling," states that "criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations." The NRC has established a 5% subcriticality margin (i.e., k-effective (keff) less than or equal to 0.95) for nuclear power plant licensees to comply with GDC 62.

10 CFR 50.68 subpart (b), regarding New Fuel Storage Racks, specifies that "(2) The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.

(3) If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need_ not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used."

• Also, Subpart (b)(4) of 10 CFR 50.68, "Criticality accident requirements," specifies, "if credit is taken for soluble boron, the k-effective of the SFP storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0. 95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water."

5.2 No Significant Hazards Consideration DENG has performed the significant hazards consideration for the proposed license amendment by addressing the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 15 of 20 The proposed change will not affect plant equipment or structure, including the SFP, NFSR, or fuel handling equipment, including how the equipment is operated and maintained. There are no changes to the equipment for fuel handling or how fuel assemblies are handled, including how fuel assemblies are inserted into and removed from SFP and NFSR storage locations. There is no change to administrative means to verify correct fuel assembly storage in the SFP, or the required response to a fuel assembly mislead or drop event. There are no changes to how RCCAs will be handled, including how RCCAs are inserted into or removed from a fuel assembly or other location such as a SFP storage location. Also, since the proposed change does not modify plant equipment or its operation and maintenance, including equipment used to maintain SFP soluble boron levels, the proposed change will not impact a boron dilution event or plant response to it.

The criticality safety evaluation -concluded that the NFSR limiting accident is the fully flooded condition with each storage location loaded with a maximum reactivity fuel assembly. The NFSR maintains kett ~ 0.95 for this postulated scenario including uncertainties and biases. The NFSR also maintains kett ~

0.98 for the optimum moderation scenario including uncertainties and biases.

Thus, the consequences of a previously evaluated NFSR related accident is not significantly increased. There is no change to the plant equipment or its operation and maintenance due to the proposed change. Thus, the probability of a flooding accident that could impact the NFSR is not significantly increased.

Regarding the SFP, the Region 1 storage configuration will change. The Region 2 and 3 burnup curves will be updated and reduced in number. The process of choosing fuel assembly storage locations will not change, except that the Region 1 storage configuration and Region 2 and 3 burnup requirements will be updated, and fuel assemblies containing RCCAs may be stored in Region 2 without consideration of the burnup curve. The physical handling, insertion, removal, and storage of fuel assemblies in SFP racks will not change. The MPS3 program for choosing fuel assembly storage locations, for fuel handling, and for assuring that the fuel assemblies are placed into correct locations will remain in place. Thus, the probability of a fuel assembly misleading or a fuel assembly drop in the SFP will not significantly increase due to the proposed change.

Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 16 of 20 Several postulated accidents for the SFP were reviewed for the proposed change which included postulated fuel assembly misload and drop scenarios.

The criticality safety evaluation for the SFP concluded that the limiting accident, which bounds the other scenarios, is a multiple misload of fuel assemblies into each Region 2 fuel storage location. The criticality safety evaluation concluded that a SFP soluble boron concentration of 2600 ppm will maintain keff.:::. 0.95, including uncertainties and biases, for this postulated scenario. The minimum TS soluble boron concentration will be increased from 800 ppm to 2600 ppm. MPS3 has maintained SFP soluble boron concentration greater than 2600 ppm for many years, so the proposed change will not affect the routine maintaining of the boron concentration.

There are no changes to plant equipment, including its operation and maintenance, as a result of the proposed change, including equipment associated with maintaining SFP soluble boron concentration or possible flow paths that could contribute to a boron dilution event. Thus, no new avenues for a boron dilution event will be created. There is no change -regarding how the plant maintains boron concentration or responds to a boron dilution event. The criticality safety evaluation for the postulated boron dilution event shows the SFP maintains keff.:::. 0.95 at 600 ppm soluble boron. Thus, there is no significant increase in the probability or consequences of a boron dilution accident.

The MPS3 SFP is currently licensed to store 1860 fuel assemblies which include a Region 2 rack that has not been placed in the SFP (TS 5.6.3).

Thus, the SFP seismic/structural loading requirements for the proposed change are bounded by the existing TS. The criticality safety evaluation shows that keff will be maintained .:::. 0.95 during a postulated seismic event.

Thus, there is no increase in the consequences of a seismic event.

In each of the above scenarios the proposed change does not significantly increase the probability of an accident previously evaluated, and maintains required keff margin. Therefore, it is concluded that the probability or consequences of a previously evaluated accident do not significantly increase.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

There is no change to any plant equipment, including how equipment is operated and maintained. There will be no changes to equipment used to handle fuel assemblies (or any heavy load) over the NFSR or the SFP.

There is no change regarding how the fuel assemblies are stored, inserted

Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 17 of 20 into, and removed from fuel storage locations. There is no change to how RCCAs will be inserted into or removed from a fuel assembly or other location, or otherwise how RCCAs are handled. Thus, there are no new accidents created over and above the existing postulated accidents of a fuel mislead or a fuel assembly drop in the SFP or a flooding event in the NFSR area.

Also, since there is no change to the plant equipment or how equipment is operated and maintained, the probability of a new type of accident that could impact the SFP or NFSR is not significantly increased.

The criticality safety evaluation for the first time at MPS3 specifically analyzes a boron dilution event. However, the overall accident analyzed is the potential for a SFP criticality, and the boron dilution event is another potential initiator of the postulated SFP criticality accident. Also, the possibility of a SFP boron dilution event has always existed at MPS3 and the proposed change does not newly create or change the possibility of such an event occurring.

The criticality safety evaluation for the first time at MPS3 specifically analyzes a multiple fuel mislead event. As with the postulated boron dilution event, the possibility of a multiple fuel assembly mislead has always existed at MPS3 and the proposed change does not newly create or change the possibility of such an event occurring. Also, this postulated event was analyzed for the MPS2 spent fuel pool criticality LAR which the NRC approved in June 2016 (Reference 3).

Since the proposed change will not change fuel/RCCA handling equipment or how fuel assemblies and RCCAs are handled and stored, nor will it change any other plant equipment, there is no mechanism for creating a new or different kind of accident not previously evaluated. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change does not involve a significant reduction in a margin of safety.

The licensing requirement for the SFP is that keff remain~ 0.95 under normal and postulated accident conditions with credit for soluble boron. The criticality safety evaluation concluded that this requirement is m,et for the bounding postulated accident of a multiple mislead of fuel assemblies into each Region 2 fuel storage location. The analyses apply to all of the fuel assemblies currently stored in the MPS3 SFP and to future anticipated fuel designs.

Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 18 of 20 In addition, the criticality safety evaluation concludes that the SFP will maintain keff < 1.0 with O ppm soluble boron in the SFP under normal conditions with the maximum allowed reactivity fuel assembly stored in each fuel storage location.

The criticality safety evalua'ti_on also allows the following storage configurations. In each case the storage configuration does not increase reactivity assuring that keff margin is maintained:

• Storing non-fuel components in any spent fuel rack storage location where fuel assemblies are allowed

• Storing non-fuel components in the guide tubes of any fuel assembly.

The criticality safety evaluation evaluated non-standard fuel stored in the MPS3 to determine where they can be stored in the SFP. This information is used to maintain keff margin when storing non-standard fuel assemblies.

The licensing requirement for the NFSR is that keff remain.::: 0.95 for the fully flooded scenario, and .::: 0.98 for the optimum moderation scenario. The criticality safety evaluation concludes that these requirements are met assuming each fuel storage location is loaded with a maximum reactivity fuel assembly (5.0 wt% U-235 enrichment with no burnable poisons).

Therefore, all the margins of safety- are maintained, and the proposed change does not involve a significant reduction in a margin of safety Based on the above information, DENG concludes that the proposed license amendment involves no significant hazards consideration under the criteria set forth in 10 CFR 50.92(c) and, accordingly, a finding of no significant hazards consideration is justified.

5.3 Precedents The proposed changes to the MPS3 technical specifications are similar in fundamental aspects to those referenced in this NRG Safety Evaluation Report:

1. NRG License Amendment and associated SER under cover letter "Millstone Power Station, Unit No. 2 - Issuance of Amendment Re: Technical Specification Changes for Spent Fuel Storage (TAC NO. MF0435)," from R. V. Guzman (NRG) to D. A. Heacock (Dominion Nuclear Connecticut, Inc.), June 23, 2016.

Serial No.18-039 Docket No.: 50-423 Attachment 1, Page 19 of 20 The current MPS3 application includes unique aspects that reflect DENC's understanding of NRC staff expectations for the content and supporting analyses of spent fuel criticality submittals. The enclosed submittal content was developed based on insights and discussion between DENG and NRC staff which occurred during a pre-submittal meeting:

1. Summary of April 26, 2016, Pre-Application Teleconference with Virginia Electric and Power Company for Increase in Maximum Fuel Enrichment for New Fuel Storage Racks and Spent Fuel Pool (CAC Nos. MF7432 and MF7433).

5.4 Conclusion Based on the considerations discussed above, there is reasonable assurance that (1) the health and safety of the public will not be endangered by the proposed changes, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the requested license amendments will not be inimical to the common defense and security or to the health and safety of the public.

Serial No.18-039 Docket No.: 50-423 Attachment 1 , Page 20 of 20 6.0 Environmental Considerations DENC has reviewed the proposed license amendment for environmental considerations.

The proposed license amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7 .0 References

1. NRC Amendment and associated SER under cover letter "Millstone Nuclear Power Station, Unit No. 3 - Issuance of Amendment Re: Increasing Spent Fuel Storage Capacity (TAC No. MA5137)," letter from V. Nerses (NRC) to R. G.

Lizotte (Northeast Nuclear Energy Co.), November 28, 2000.

2. NRC Amendment and associated SER under cover letter "Millstone Power Station, Unit No.3 - Issuance of Amendment Re: Spent Fuel Pool Criticality (TAC NO. MD8251)," letter from C. J. Sanders (NRC) to D. A. Heacock (Dominion Nuclear Connecticut, Inc.), March 26, 2010.

3. NRC Amendment and associated SER under cover letter "Millstone Power Station, Unit No. 2 - Issuance of Amendment Re: Technical Specification Changes for Spent Fuel Storage (TAC NO. MF0435)," from R. V. Guzman (NRC) to D. A.

Heacock (Dominion Nuclear Connecticut, Inc.), June 23, 2016.

Serial No.18-039 Docket No. 50-423 Attachment 2 MARKED-UP TECHNICAL SPECIFICATIONS PAGES DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

July 28, 2016 DEFINITIONS 1.40 Deleted VENTING 1.41 Deleted 1.39 VENTING shall be the controlled process of discharging air or gas from confinement to maintain temperature, pressure, humidity, concentration, or other operating co ition, in such a manner that replacement air or gas is not provided or required during VENTI G. Vent, used in system names, does not imply a VENTING process.

NT FUEL POOL STORAGE PATTERNS:

RN refers to the blocked location in a Region 1 storage rack and all adjacent and diagonal Reg1 1 (or Region 2) cell locations surroundi e blocked location. The blocked location is for criticali ntrol.

3-0UT-OF-4 AND 4-0UT-OF-4 1.41 (a) Areas of the Regio spent fuel racks with fuel wed in every storage location the 4-0UT-OF-4 Region 1 storage a (b) Are of the Region 1 spent fuel racks which contain a cell bloc device in very 4th location for criticality control, are referred to as the 3-0UT- -4 Region 1 storage area. A STORAGE PATTERN is a subset of the 3-0UT-OF-4 Re storage area.

CORE OPERATING LIMITS REPORT (COLR) 1.42 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.6. Unit Operation within these operating limits is addressed in individual specifications.

1.43 Deleted 1.44 Deleted MILLSTONE - UNIT 3 1-7 Amendment No. ;w, ~.@, n, +oo, 89, 2 Febrnary 25, 2014 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1.2 The soluble boron concentration of the Spent Fuel Pool shall be greater than or equal to~ppm.

- 2600 APPLICABILITY:

Whenever fuel assemblies are in the spent fuel pool.

ACTION:

a.

b. With the boron concentration less than ppm, suspend the movement of all assemblies within the spent fuel pool and loads over the spent fuel racks.

SURVEILLANCE REQUIREMENTS 4.9.1.2 Verify that the boron concentration in the fuel pool is greater than or equal to ee ppm I at the frequency specified in the Surveillance Frequency Control Program.  %

MILLSTONE - UNIT 3 3/4 9-la Amendment No. H, H&, +&-9, ~ '

~

March 26, 2010

/ STORAGE REFUELING OPERATIONS J/

3/4.9.13 SPENT FUEL POOL - ltEACTIVITY Insert# 1 on next page LIMITING CONDITION FOR OPERATION The Reactivity Condition of the Spent Fuel Pool shall be such that keff is less than eq 1 to 0.95 at all times.

Whenever fuel assemblies are in the spent fuel pool.

ACTION: ith keff greater than 0.95:

a. e Spent Fuel Pool until keff is less than or equal

b. Initiate imme

• te action to move any fuel assem which does not meet the requirements of

• ures 3.9-1, 3.9-3, 3.9-4, or . -5 to a location for which that fuel assembly is all ed.

SURVEILLANCE REQUIREMENTS e p ed in Region 1 "4-0UT-OF-4" fuel storage are within the enrichment an umup lim of Figure 3.9-1 by checking the fuel assembly's design and bu -up documentatio 4.9.13.1.2. Ensure that all fuel ssemblies to be placed in Reg 2 fuel storage are within the enrichment, dee time, and bumup limits of Figure 3. -3 by checking the fuel assembly's de gn, decay time, and bum-up documentatio .

at all fuel assemblies used exclusively in pre-uprate 11 Mwt) conditions are to be placed in Region 3 fuel storage are within the enric ent, decay time, bumup limits of Figure 3.9-4 by checking the fuel assembly's des n, decay time, and bum-up documentation. Ensure that all fuel assemblies used in post- rate (3650 Mwt) conditions which are to be placed in Region 3 fuel storage are "thin the enrichment, decay time, and bum-up limits of Figure 3.9-5 by checking the fue assembly's design, decay time, and burn-up documentation. t MILLSTONE - UNIT 3 3/49-16 Amendment No.~. -1:-§.8., -l-8-9, ziffl

Insert# 1 to TS 3.9.13 Spent Fuel Pool - Storage Limiting Condition for Operation 3.9.13 The spent fuel storage requirements necessary to maintain kettWithin limits shall be met.

APPLICABILITY: Whenever fuel assemblies are in the spent fuel pool.

ACTION:

a. For a fuel assembly stored in Region 1A - initiate immediate action to move any assembly which does not meet Surveillance Requirement 4.9.13.1.1. to Region 1B.

b. For a fuel assembly stored in Region 2 that does not contain a Rod Cluster Control Assembly - initiate immediate action to move any assembly which does not meet the requirements of Figure 3.9-2 to a location for which that fuel assembly is allowed.

c. For a fuel assembly stored in Region 3 - initiate immediate action to move any assembly which does not meet the requirements of Figure 3.9-3 to a location for which that fuel assembly is allowed.

SURVEILLANCE REQUIREMENTS

---

NOTE---------------------------------------------------------

The Region 1 Fuel Storage Loading Schematic (Figure 3.9-1) designates each storage location as either Region 1A or Region 1B.

Regarding fuel assemblies that contain a Rod Cluster Control Assembly for storage in Region 2 - if the enrichment and burnup of a given assembly is not in the "Acceptable" domain of Figure 3.9-2 (e.g., the assembly requires a Rod Cluster Control Assembly to be stored in Region 2), then the assembly must be located in an acceptable Region 1 storage location before its Rod Cluster Control Assembly can be inserted or removed.

Initial enrichment is the maximum initial planar volume averaged as-built U-235 enrichment in the assembly. If the assembly has axial blankets the lower enriched fuel is not credited in determining the enrichment. Also, fuel burnup is the volume averaged burnup of the assembly as determined using the measured reaction rates.

Insert# 1 to TS 3/4.9.13 Spent Fuel Pool - Storage Limiting Condition for Operation and Surveillance Requirements (continued) 4.9.13.1.1. Ensure that all fuel assemblies to be placed into a Region 1A storage location, with an initial enrichment greater than 4.75 w/o U-235, have achieved a fuel burnup greater than or equal to 2.0 GWD/MTU or contain a minimum of twelve (12) Integral Fuel Burnable Absorber (IFBA) Rods by checking the fuel assembly's location, design, and burnup documentation.

Fuel assemblies with an initial enrichment less than or equal to 4. 75 w/o U-235 may be stored in Region 1A without restriction.

4.9.13.1.2. Ensure that all fuel assemblies to be placed into Region 1Bare stored consistent with the Fuel Storage Loading Schematic specified in Figure 3.9-1 by checking the fuel assembly's storage location. All fuel assemblies with an initial enrichment less than or equal to 5.0 w/o U-235 may be stored in Region 1B without restriction.

4.9.13.1.3. Ensure that all fuel assemblies to be stored in Region 2 - that do not contain a Rod Cluster Control Assembly - are within the enrichment and burnup limits of Figure 3.9-2 by checking the fuel assembly's design and burnup documentation. A fuel assembly that contains a Rod Cluster Control Assembly may be stored in Region 2 without restriction.

4.9.13.1.4. Ensure that all fuel assemblies to be stored in Region 3 are within the enrichment, burnup, and decay time limits of Figure 3.9-3 by checking the fuel assembly's design, burnup, and decay time documentation.

~(6. embet 28, 2000 REFUELING OPERATIONS SPENT FUEL POOL - STORAGE PATTERN LIMITING CONDITION FOR OPERATION Each STORAGE PATTERN of the Region 1 spent fuel pool racks shall require that:

Prior to storing fuel assemblies in the STORAGE PATTERN per Figur locking device for the cell location must be installed.

b. Prior to remo fa cell blocking device from the ce ocation per Figure 3 .9-2, the STORAGE PA'. must be vacant of a ored fuel assemblies APPLICABILITY:

ACTION:

erify that 3.9.14 is satisfied with no fuel assemblies stored in the STORA RN prior to installing and removing a cell blocking device in the spent fuel racks.

ITS 3.9.14 is DELETED MILLSTONE - UNIT 3 3/4 9-17 Amendment No. ;9,4-89 I

November 28, 2000 Figure 3.9-1 Minimum Fuel Assembly Bumup Versus Nominal Initial Enrichment for Region 1 4-0UT-OF-4 Fuel Storage Configuration 8

7 6

5 4

3 2

1 4.00 4.25 4.50 4.75 Initial Fuel Enrichment (w/o U-235)

MILLSTONE - UNIT 3 3/49-18 Amendment No.~.~

Replace with new Region 1 Fuel Storage Loading Schematic (next page)

Figure 3.9-1 Region 1 Fuel Storage Loading Schematic 11111111111111 11111111111111 11111111111111 11111111111111 11111111111111 .-----------,

Region 1A 11111111111111 11111111111111 11111111111111 11111111111111 Region 18 11111111111111

.______ _ ____, ~ West SFP Wall I

November 28, 2000 Figure 3.9-2 Region 1 3-0UT-OF-4 Storage Fuel Assembly Loading Schematic Region 2 or Region 1 4-0UT-OF-4 may be placed along this face This face must be along Region 2 or Region 1 t--........---il'----.aoi,-----t'---~---,<f----'I the wall of the spent fuel 4-0UT-OF-4 may be pool, or other Region 1 placed along this face 3-0UT-OF-4 storage Cell Blocker location sembly Storage location MILLSTONE - UNIT 3 3/4 9-19 Amendment No . ~ ' 1-Br Replace with new Region 2 burnup curve (next page)

FIGURE 3.9-2 Min imum Fuel Assembly Burnup versus Initial Enrichment for Reg ion 2 Storage Configuration (Fuel Assemblies without Rod Cluster Control Assemblies) 45 I

40 I

[.....,,-"""' i I

I I

I _.,.- ~ I S- 35 ACCEPTABLE

~ I I V

~

........ L,../

~ 30

~ o Analyzed V C.

g... 25 - V NOT ACCEPTABLE I

I a:i

-Poly. (Bounding)

V QI

20

/ I

~

QI

~ 15 /

V Iv= -0.16936x 3

- 0.04949x 2 + 20.SSlx - 40.098 I

.$J E

V QI

~ 10 V 5

/

V V I 0

/ I 2.0 2.2 2.4 2.6 2.8 3.0 3.2 3.4 3.6 3.8 4.0 4.2 4.4 4.6 4.8 5.0 Max Planar Average Assembly Initial Enrichment (wt% U-235)

March 26, 2010 Figure 3.9-3 Minimum Fuel Assembly Burnup and Decay Time Versus Nominal Initial Enrichment for Region 2 Storage Configuration

~A .,_~~--+~~~-+'.,_._~~-+-~---,-~+-~~~it--.........,,._-+-1l!"-~~-1

'J NA !; . Il l' 1/,U I

.60 For assemblies from Post-Uprate (3650 Mwt) Cores, the nominal fuel enrichment of blankets must be .:S 2.6 w/o U-235 , and nominal blanket length must be at least 6 inches on both ends of the fuel. Fuel batches A, B, C, and D may not be stored in Region 2.

MILLSTONE - UNIT 3 3/4 9-20 Amendment No.+8-9, 24,g Replace with new Region 3 burnup curve (next page)

FIGURE 3.9-3 Minimum Fuel Assembly Burnup and Decay Time versus Initial Enrichment for Region 3 Storage Configuration

- o Years Decay 50

): 0 Years Analyzed

--- 3 Years Decay 45

-

• 3 Years Analyzed i 40 - * *9 Years Decay ACCEPTABLE

~

• 9 Years Analyzed

~ 35 -

• 18 Years Decay g- t;, 18 Year Analyzed E

30 -

• 25 Years Decay ai

• 25 Year Analyzed fili 25 ~ ===:::::i===~ - l - - - -~ .A-=-..,....,_-.,,,,.~ -+----+-----+-----------4

~

~

<C 20 - i - - - - - - - i - - - - -~ - + - - - - - +-t NOT ACCEPTABLE tt-----;

iS

~ 15 - + - - - - - + - - - -.h'e....--:.~- - - - + - - - - - + - - - - - t - - - - - - + - - - - - - 1

<(

5 0 -+---l- ---1-----f------+-----+----+------+------;

1.5 2.0 2 .5 3.0 3.5 4.0 4 .5 5.0 Max Planar Average Assembly Initial Enrichment (wt% U-235)

The burnup curve equations have the following polynomial format (bound ing):

BU [GWD/MTU] = a 4

• wt% 4 + a 3

• wt% 3 + a 2

• wt% 2 + a 1

• wt% 1 + a 0 Burnup Credit Curve Polynomial Coefficients Region Decay Time (Years) a4 a3 a2 a1 ao 3 No Credit -0 .2459 4.208 -26.80 88.70 -92.00 3 3 Years -0.2338 4.00 1 -25.48 84.34 -87 .47 3 9 Years -0 .2153 3.684 -23.46 77.66 -80.54 3 18Years -0 .2020 3.458 -22.02 72.88 -75.59 3 25 Years -0 .1964 3.361 -21.40 70.84 -73 .47

iviarcn Lo , LU 1 u Figure 3.9-4 Minimum Fuel Assembly Burnup and Decay Time Versus Nominal Initial Enrichment for Region 3 Storage Configuration for Assemblies from Pre-Uprate (3411 Mwt) Cores 40

)

t,..

Q

~

-=

c.,

Q,, 30

--+-- 0 year decay time

......tr-5 year decay time 10 year decay time C

J.

CQ = *"" 20 year decay time

-= ~

~

20 0+---+---+----+----1-------1----1 2.00 2.50 3.00 3.50 4.00 4.50 5.00 Initial Fuel Enrichment (w/o U-235)

MILLSTONE - UNIT 3 3/4 9-21 Amendment No ...J:-&.9 2-48

!FIGURE 3.9-4 is DELETED

March 26, 2010 Figure 3.9-5 Minimum Fuel Assembly Burnup and Decay Time Versus Nominal Initial Enrichment for Region 3 Storage Configuration for Assemblies from Post-Uprate (3650 Mwt) Cores (i 0.0 00

~

~  :*1:1,ot11 l-+~!'""--t-+-i---li'-o!--'i-~'!--t---¥--,lf-,..'-J,9~" ' 1 - - - - - t - - - -

~

a.

!,. 30.000 1--:-__;;..--+-;.....;..---+-+--+--...;_""'*'er--t'H~!-----!--+----+-.,.......;-_:;

I!

IJ I\IA,CCE ;,-.t;Bt.E

!... i~.ooo l-'--'--~-!-+--_.,_....__-1--,.-1-+;1C,CM-~-1-~1--....._-4--4--l. ...__-1---------'---i IL I~. 00 2 00

For assemblies from Post-Uprate (3650 Mwt) Cores, the nominal fuel enrichment of blankets must be s.,2.6 w/o U-235, and nominal blanket length must be at least 6 inches on both ends of the fuel.

MILLSTONE - UNIT 3 3/4 9-22 Amendment No . ~

!FIGURE 3.9-5 is DELETED .I

Mttfeh 26, 2010 DESIGN FEATURES 5.6 FUEL STORAGE

/ jinsert # 2 on next page CRITICALITY The spent fuel storage racks are made up of 3 Regions which are designed and shall be maintained to ensure a Keff less than orequal to 0.95 when flooded with unborated water. The storage rack Regions are:

a. egion 1, a nominal 10.0 inch (North/South) and a nominal 10.455 i (Ea West) center to center distance, credits a fixed neutron abso r (BORAL) within rack, and can store fuel in 2 storage configurations*

(1) With er it for fuel bumup as shown in Figure .9-1, fuel may be stored in a "4-0UT- -4" storage configuration.

(2)

b. Region 2, a nomi absorber (BO ) within the rack, and with credit fo el bumup and fuel decay time as sh n in Figure 3.9-3, fuel may be stored in all av

• ble Region 2 storage C. egion 3, a nominal 10.35 inch center to center distance, with credit bumup and fuel decay time as shown in Figure 3.9-4 for assemblies used exclusively in pre-uprate (3411 Mwt) cores or Figure 3.9-5 for assemblies use n post-uprate (3650 Mwt) cores, fuel may be stored in all available Region 3 storage locations. The Boraflex contained inside these storage racks is not credited.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 45 feet.

MILLSTONE ~ UNIT 3 5-6 Amendment No. 3-9, W, 8-9, ~

Insert# 2 to TS 5.6.1.1 Fuel Storage - Criticality 5.6.1.1 The New Fuel Storage Racks, a nominal 22.125 inch center to center distance, credit a fixed neutron absorber (BORAL) within the rack and are designed and shall be maintained with:

a. Kett less than or equal to 0.95 with the storage racks fully loaded with the highest reactivity fuel and flooded with potential moderators,

b. Kett less than or equal to 0.98 with the storage racks fully loaded with the highest reactivity fuel and optimum moderation of the racks.

The spent fuel storage racks are rnade up of 3 Regions which are designed and shall be maintained to ensure a Kett less than 1.0 when flooded with unborated water, and Kett less than or equal to 0.95 with 600 ppm soluble boron in the spent fuel pool water. The storage rack regions are as follows:

a. Region 1, a nominal 10.0 inch (North/South) and a nominal 10.455 inch (East/West) center to center distance, credits a fixed neutron absorber (BORAL) within the rack. Each Region 1 fuel storage rack contains two storage sub-regions - Region 1A and Region 1B:

(1) Region 1A- Fuel assemblies meeting one of the following three criteria may be stored in Region 1A storage locations:

i. initial enrichment less than or equal to 4.75 w/o U-235, or ii. initial enrichment less than or equal to 5.0 w/o U-235 with a fuel burnup greater than or equal to 2.0 GWD/MTU, or iii. initial enrichment less than or equal to 5.0 w/o U-235 that contain a minimum of 12 Integral Fuel Burnable Absorber (IFBA) rods.

(2) Region 1B - Fuel assemblies with an initial enrichment less than or equal to 5.0 w/o U-235 shall be stored per the Fuel Storage Loading Schematic shown in Figure 3.9-1 (the two rows against the spent fuel pool west wall are designated Region 18).

b. Region 2, a nominal 9.017 inch center to center distance, credits a fixed, neutron absorber (BORAL) within the rack and either fuel burnup as shown in Figure 3.9-2 or takes credit for containing a Rod Cluster Control Assembly.

c. Region 3, a nominal 10.35 inch center to center distance, credits fuel burnup and decay time as shown in Figure 3.9-3. These racks contain Boraflex which is not credited.

Serial No.18-039 Docket No.: 50-423 Attachment 3 MARKED-UP TECHNICAL SPECIFICATIONS BASES PAGES (For Information Only)

DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

LnLJttt l~U. lV 1911.) VUO FOR INFORMATION ONLY March 9, 2010 3/4.9 REFUELING OPERATIONS and the use of Rod Cluster Control Assemblies for certain assemblies in BASES Region 2.

3/4.9.1.2 BORON CONCENTRATION IN SPENT FUEL POOL

___ ,_Ll.0 During normal Spent F~ rool operation, the spent fuel racks are capable of maintaining Keff at less than or equal to 0. 95 in an unborated water environment. This is accomplished in Region 1, 2, and 3 storage racks by the combination of geometry of the rack spacing, the use of fixed neutron absorbers in some fuel storage regions, the limits on fuel burnup, fuel enrichme and minimum fuel decay time, ai:i.d the JJ:i,e of blocking devices in eertaifl fuel ~te,111ge location .

The boron requirement in the spent fuel pool specified in 3 .9 .1.2 ensures that in the event of a fuel assembly * * * * * *

• a~~embly, the Keff of the s en misload accident, which involves the misloading of 3/4.9.2 INSTRUMENTATION multiple fuel assemblies, The source range neutron flux monitors are used during refueling operations to monitor the core reactivity condition. The installed source range-neutron flux monitors are part of the Nuclear Instrumentation System (NIS). These detectors are located external to the reactor vessel and detect neutrons leaking from the core.

There are two sets of source range neutron flux monitors:

(1) Westinghouse source range neutron flux monitors, and (2) Gamma-Metrics source range neutron flux monitors.

The Westinghouse monitors are the normal source range monitors used during refueling activities. Gamma-Metrics source range neutron flux monitors are an acceptable equivalent control room indication for the Westinghouse source range neutron flux Monitors in MODE 6, including CORE Alterations, as follows:

with the core in place within the reactor vessel or, with the Gamma Metrics source range neutron flux monitor(s) coupled to the core.

Reactor Engineering shall detennine whether each monitor is coupled to the core.

This limiting condition for operation requires two source range neutron flux monitors be OPERABLE to ensure that redundant monitoring capability is available to detect changes in core reactivity. To be OPERABLE, each monitor must provide visual indication in the control room. In addition, at least one of the two monitors must provide an OPERABLE audible count rate function in the control room and containment.

MILLSTONE - UNIT 3 B 3/4 9-la Amendment No. H, @, 8-, ~ ' +/--1-9,

+/-W

FOR INFORMATION ONLY LBDCR No. 07 MP3 037 fol, 12, 2007 REFUELING OPERATIONS BASES

~ -STORAGE 3/4.9.13 SPENT FUEL POOL - REACTIVITY ~pnsert # Bl on next page I unng norma spent e poo operat10n, t e spent ue rac s are capa e o mamtammg at less than or equal to 0.95 in an unborated water environment.

aintaining Keff at less than or equal to 0.95 is accomplished in Region 1 3-0UT-O -4 storage rac by the combination of geometry of the rack spacing, the use of fixed neutr absorbers in t racks, a maximum nominal 5 weight percent fuel enrichment, and t use of blocking devices 1 certain fuel storage locations, as specified by the interface re irements shown in Figure 3.9- .

Maintaining Keff at s than or equal to 0.95 is accomplished i egion 1 4-0UT-OF-4 storage racks by the combinatio of geometry of the rack spacing, use of fixed neutron absorbers in the racks, and the lim1 on fuel enrichment/fuel b up specified in Figure 3.9-1.

Maintaining Keffat less than or equa o 0.95 is complished in Region 3 storage racks by the combination of geometry of the rac acing, and the its on fuel enrichment/fuel bumup and fuel decay time specified in Fig 3.9-4 for assemblies u exclusively in the pre-uprate (3411 Mwt) cores and Figure 3.9 for assemblies used in the pos date (3650 Mwt) cores.

Fixed neutron absorbers are credited in the Region 3 fuel storage r ks.

The limitations escribed by Figures 3.9-1, 3.9-2, 3.9-3, 3.9-4, and . -5 ensure that the reactivity of the fu ssemblies stored in the spent fuel pool are conservatively *thin the assumptions of e safety analysis.

inistrative controls have been developed and instituted to verify that the fuel enric ent, fuel bumup, fuel decay times, and fuel interface restrictions specified in Figures

, 3.9-2, 3.9-3, 3.9-4, and 3.9-5 as well as restrictions specified in the Note on Figures 3.9-3 nd 3.9-5 are complied with.

3/4.9.14 SPENT FUEL POOL- STORAGE PATTERJ..J' ~ DELETED The limitations of this specification ensure that the reactivity conditions of the Regio

- storage racks and spent fuel pool keff will remain less than or equal to ERN for the Region 1 storage racks will be es

• ed and expanded from the of the spent fuel pool per Figure 3.9-2 to ensure definition and con f the

• n 1 3-0UT-OF-4 Boundary to other Storage Regions and minimize the number of boundaries where a fuel mis lacement incident can occur.

MILLSTONE - UNIT 3 B 3/4 9-9

Insert# 81 to TSB 3/4.9.13 Spent Fuel Pool - Storage During normal spent fuel pool operation, the spent fuel racks are capable of maintaining Ken at less than 1.0 in an unborated water environment, and less than or equal to 0.95 with 600 ppm soluble boron in the spent fuel pool water.

Maintaining Ken less than or equal to 0.95 is accomplished in Region 1A storage rack locations by the combination of geometry of the rack spacing, the use of fixed neutron absorbers (BORAL) in the racks, and prohibiting storage of fuel assemblies with an enrichment greater than 4.75 w/o U-235 unless its fuel burnup is greater than or equal to 2.0 GWD/MTU or it contains twelve (12) or more IFBA rods.

Maintaining Kett less than or equal to 0.95 is accomplished in Region 1B storage rack locations by the combination of geometry of the rack spacing, the use of fixed neutron absorbers (BORAL) in the racks, a maximum 5.0 weight percent initial fuel enrichment, and specifying which storage locations are designated as Region 1B.

Maintaining Kett less than or equal to 0.95 is accomplished in Region 2 storage racks by the combination of geometry of the rack spacing, the use of fixed neutron absorbers (BORAL) in the racks, and the limits on initial fuel enrichment/fuel burnup specified in Figure 3.9-2. As an alternative, maintaining Ken less than or equal to 0.95 can also accomplished in Region 2 storage rack locations if the assembly has a maximum initial enrichment less than or equal to 5.0 weight percent and contains a Rod Cluster Control Assembly.

Maintaining Ken less than or equal to 0.95 is accomplished in Region 3 storage racks by the combination of geometry of the rack spacing and the limits on initial fuel enrichment/fuel burnup and fuel decay time specified in Figure 3.9-3. Fixed neutron absorbers are not credited in the Region 3 fuel storage racks The limitations described by Figures 3.9-1, 3.9-2, 3.9-3, the burnup/lFBA requirement in Region 1A, and the use of Rod Cluster Control Assemblies in Region 2 ensure that the reactivity of the fuel assemblies stored in the spent fuel pool is conservatively within the assumptions of the safety analysis.

Administrative controls have been developed and instituted to verify that the initial fuel enrichment, fuel burnup, and fuel decay times specified in Figures 3.9-1, 3.9-2, 3.9-3, the burnup/lFBA requirement in Region 1A, and the presence of a Rod Cluster Control Assembly (Region 2) are complied with.

Initial enrichment, when used to compare to fuel storage requirements, is the maximum initial planar volume averaged as-built U-235 enrichment in the assembly. If the

, assembly has axial blankets the lower enriched fuel is not credited in determining the enrichment. Fuel burnup when used to compare to fuel storage requirements is the volume averaged burnup of the assembly as determined using the measured reaction rates with no reduction for measurement uncertainty.

Serial No.18-039 Docket No. 50-423 ATTACHMENT 4 CRITICALITY ANAL VSIS CHECKLIST DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.18-039 Docket No. 50-423 Attachment 4, Page 2 of 9 Purpose of submittal YES Identify and add margin, MUR uprate, boron credit, control rod credit, SFP wall credit, update methods and codes, provide more core design flexibility.

License changes requested YES Burnup curves, storage pattern, minimum soluble boron, remove axial blanket restrictions, remove cell blocker requirement.

Summary of physical changes YES Remove cell blockers.

Summary of analytical scope YES SFP and new fuel storage criticality safety analysis including normal storage and fuel handling, abnormal conditions, and boron dilution analysis.

Summary of requirements and guidance YES Requirements documents referenced YES Multiple.

Guidance documents referenced YES DSS-ISG-2010-01, NEI 12-16 draft Acceptance criteria described YES Describe reactor operating parameters YES Bounds historical and anticipated.

Describe all fuel in pool YES Geometric dimensions (Nominal and YES Tolerances)

Schematic of guide tube patterns YES Only one pattern.

Material compositions YES Describe future fuel to be covered NO None proposed. Composite bounding design used.

Geometric dimensions (Nominal and NO None proposed.

Tolerances)

Schematic of guide tube patterns NO None proposed.

Material compositions NO None proposed.

Describe all fuel inserts YES Geometric dimensions (Nominal and YES BPRA, WABA, RCCA, source rods, in-core Tolerances) thimble (tolerances for in-SFP components).

Schematic (axial/cross-section) NO Commonly used inserts.

Material compositions YES Describe non-standard fuel YES

Serial No.18-039 Docket No. 50-423 Attachment 4, Page 3 of 9 Geometric dimensions YES Fuel rod storage canister and fuel reconstitution considered.

Describe non-fuel items in fuel cells NO Non-fuel items can only be stored in cells that are allowed to store fuel assemblies. Therefore, non-fuel items displace fuel.

Nominal and tolerance dimensions NO New fuel vault & Storage rack description YES Nominal and tolerance dimensions YES Schematic (axial/cross-section) YES Rack diagrams Material compositions YES Spent fuel pool, Storage rack description YES Region 1: BORAL, flux trap Region 2: BORAL, non-flux trap Region 3: Boraflex (not credited), flux trap Nominal and tolerance dimensions YES Schematic (axial/cross-section) YES Rack diagrams Material compositions YES Other Reactivity Control Devices {Inserts) NO No installed rack inserts (optional RCCA credit).

Nominal and tolerance dimensions NO Schematic (axial/cross-section) NO Material compositions NO New fuel rack analysis description YES BORAL installed in New Fuel Storage racks.

Storage geometries NO All cells qualified for new fuel storage.

Bounding assembly design(s) YES Integral absorber credit NO Accident analysis YES Spent fuel storage rack analysis description YES Storage geometries YES Region lA has all-cell storage. Region 18 restricted to two rows adjacent to SFP wall (credits leakage). Regions 2 and 3 have all-cell fuel storage.

Bounding assembly design(s) YES Composite bounding assembly.

Soluble boron credit YES Boron dilution analysis YES

Serial No.18-039 Docket No. 50-423 Attachment 4, Page 4 of 9 Burnup credit YES Decay time credit YES Region 3.

Integral absorber credit YES Minimum historical integral absorber loading used for some fuel in the multiple misload accident analysis (Region 2 limiting). Optional IFBA credit in Region 1A.

Other credit YES SFP wall credited for Region 18, control rod credit for Region 2, and decay time credit for Region 3.

Fixed neutron absorbers YES BORAL in Regions 1 and 2, Boraflex in Region 3 (not credited).

Aging management program YES Coupon monitoring (BORAL).

Accident analysis YES Temperature increase YES Assembly drop YES Into racks (broken grids pitch change) and between racks.

Single assembly misload YES Multiple misload YES Defines a maximum reactivity batch of 24 un-poisoned fresh 5.0 wt%

assemblies clustered together surrounded by minimally poisoned fuel.

Boron dilution YES Other YES Fuel Handling Accident Fuel out of rack analysis YES Handling YES Movement YES Inspection YES

~~o ComputE!f t<>cl~s, Cross Sections cind Validation .

Overview Code/Modules Used for Calculation of kett YES SCALE6.0/CSASS - KENO V.a Cross section library YES ENDF/B-VII 238 Group List all the isotopes used YES All TRITON isotopes.

Convergence checks YES 1000 generations skipped, specific convergence checks on key cases.

Code/Module Used for Depletion Calculation YES SCALE6.0/TS-depl - KENO V.a Cross section library YES ENDF/B-VII 238 Group List all the isotopes used YES All SCALE 6.0 (TS-DEPL addnux=3)

Serial No.18-039 Docket No. 50-423 Attachment 4, Page 5 of 9 Convergence checks YES Same input (step sizes, neutron histories) as was confirmed to achieve convergence in prior LAR with same fuel design. Additional convergence demonstration cases provided.

Validation of Code and Library YES Major Actinides and Structural Materials YES Minor Actinides and Fission Products YES 1.5% bias (NUREG/CR-7109)

Absorbers Credited YES B (BORAL, IFBA, soluble boron), Cd (RCCA).

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Rack model YES Full storage area with structure and concrete.

Boundary conditions YES Void.

Source distribution NO Uniform in fissile material.

Geometry restrictions NO Limiting fuel design YES Fuel density YES Bounding high.

Burnable Poisons NO No credit.

Fuel dimensions YES Composite bounding design.

Axial blankets NO No credit.

Limiting rack model Storage vault dimensions and materials YES Temperature YES Multiple regions/configurations NO Flooded YES Low density moderator YES Eccentric fuel placement YES Tolerances Fuel geometry YES Fuel pin pitch YES Fuel pellet OD YES Fuel clad OD YES Fuel content Enrichment YES Density YES Integral absorber NO No credit.

Rack geometry Rack pitch YES Cell wall thickness YES

Serial No.18-039 Docket No. 50-423 Attachment 4, Page 6 of 9 Storage vault dimensions/ materials YES Code uncertainty YES Biases Temperature YES Code bias YES Moderator Conditions Fully flooded and optimum density YES moderator

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Depletion Model Considerations Time step verification YES Verified (similar to prior LARs).

Convergence verification YES Verified (similar to prior LARs).

Simplifications YES Bounding grid volume homogenized in fuel lattice water, constant depletion conditions (except reduced power for final 40 days).

Non-uniform enrichments NO Post Depletion Nuclide Adjustment YES Reduced volatile fission product content (same as prior LARs). Remove very low concentration isotopes.

Cooling time YES Time of peak reactivity (5 days) except for decay time credit cases.

Limiting depletion parameters Burnable Absorbers YES Maximum WABA. IFBA and BPRA have been used at MP3. Maximum WABA bounds maximum IFBA and allows for possible WABA use. Unbounded BPRA from early cycles dispositioned.

Integral absorbers NO Maximum lfBA plus source rods bounded by maximum WABA.

Soluble Boron YES Bounding Fuel and Moderator Temperature YES Calculated using bounding high fuel assembly power history and bounding high core moderator exit temperature Power YES Bounding high power reduced near end of depletion to maximize depleted fuel reactivity.

Control rod insertion NO Bounded by WABA depletion, including maximum IFBA plus control rod insertion. Unbounded CR insertion from early cycles are dispositioned Atypical Cycle Operating History YES Bounded or justified.

Serial No.18-039 Docket No. 50-423 Attachment 4, Page 7 of 9 Rack model YES Three rack models (Regions 1-3), two configurations in Region 1 Boundary conditions YES Periodic X-Y.

Source distribution YES Uniform in fissile material Geometry restrictions YES Region 18 must be 2 rows adjacent to SFP wall.

Design Basis Fuel Description YES Composite bounding design.

Fuel density YES Bounds all fuel assemblies.

Burnable Poisons YES Optional credit in Region 1A. Minimum IFBA credit in some fuel in multiple misload analysis.

Fuel assembly inserts YES Optional control rod credit, evaluation of depleted BPRA storage.

Fuel dimensions YES Present and past MP3 designs bounded.

Axial blankets NO No credit taken.

Configurations considered YES Evaluated for each rack type, with and without boron.

Borated YES Un borated YES Multiple rack designs YES Three rack types.

Alternate storage geometry YES Region 18.

Reactivity Control Devices YES Fuel Assembly Inserts YES Optional RCCA credit in Region 2.

Storage Cell Inserts NO Storage Cell Blocking Devices NO Axial burnup shapes Uniform/Distributed YES NUREG/CR-6801 and uniform, justified with MP3 shapes.

Nodalization YES NUREG/CR-680118 nodes.

Blankets modeled NO No credit except in disposition of unbounded discharged fuel.

Tolerances/Uncertainties Fuel geometry Fuel rod pin pitch YES Fuel pellet OD YES Cladding OD YES Axial fuel position YES Fuel content Enrichment YES

Serial No.18-039 Docket No. 50-423 Attachment 4, Page 8 of 9 Density YES Assembly insert dimensions and materials) NO Conservative RCCA model used.

Rack geometry Flux-trap size (width) YES Rack cell pitch YES Rack wall thickness YES Neutron Absorber Dimensions YES Rack insert dimensions and materials NO None present.

Code validation uncertainty YES Criticality case uncertainty YES Depletion Uncertainty YES Use ISG 2010-01 (5% burnup worth).

Burnup Uncertainty YES 4% of measured burn up (developed from measurement uncertainty).

Biases Design Basis Fuel design YES Built into Base Model Minor actinides and fission product worth YES 1.5% of worth (NUREG/CR-7109).

Code bias YES Temperature YES Built into Base Model Eccentric fuel placement YES Built into Base Model lncore thimble depletion effect YES lncore thimble included in fuel depletion model.

NRC administrative margin YES Grid growth YES Fuel clad creep also included.

Modeling simplifications Identified and described YES Interface configurations analyzed YES Between dissimilar racks YES Interface region bias and uncertainty determined based on major components.

Between storage configurations within a rack YES Region lA, 18.

Interface restrictions NO No restrictions needed.

Fuel handling equipment YES Bounding analysis.

Administrative controls YES Described.

Fuel inspection equipment or processes YES Bounding analysis.

Fuel reconstitution YES Evaluated with limitations.

Serial No.18-039 Docket No. 50-423 Attachment 4, Page 9 of 9 Boron dilution YES 0 ppm kett< 1.0 including biases and uncertainties.

Normal conditions YES kett< 0.95 with minimum dilution analysis boron.

Accident conditions YES kett< 0.95 with TS minimum SFP boron.

Single assembly mislead YES Bounded by multiple misload.

Fuel assembly misplacement YES Bounded by multiple misload.

Neutron Absorber Insert Mislead NO Bounded by multiple misload.

Multiple fuel mislead YES Maximum anticipated reactivity full fresh fuel batch misload.

Dropped assembly YES Between and in racks, bounded by multiple misload.

Temperature YES Partial voiding due to boiling considered. Bounded by multiple misload.

Seismic event/ other natural phenomena NO No interface effect from shifting racks due to Region design.

Summary of results YES Burnup curve interpolation YES Bounding polynomial coefficients.

Intermediate Decay time treatment NO Pass/ fail.

New administrative controls NO Technical Specification markups YES Code validation methodology and bases YES NUREG 6698 Method.

New Fuel YES Depleted Fuel YES MOX YES HTC YES Convergence YES Trends YES Bias and uncertainty YES Range of applicability YES Analysis of Area of Applicability coverage YES

Serial No.18-039 Docket No. 50-423 ATTACHMENT 6 CRITICALITY SAFETY EVALUATION REPORT DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 2 of 300 Table of Contents 1 Introduction .................................................................................. 17 2 Acceptance Criteria and Regulatory Guidance ......................... 18 3 Reactor and Fuel Design Description ........................................ 24 3.1 Fuel Description ............. ~ ......................................................................................... 24 3.2 Fuel Inserts Description ........................................................................................... 27 3.2.1 Burnable Absorbers ................................................................................... 27 3.2.2 Control Rods ...............................................................................................29 3.2.3 Sources and Other lnserts ...................................... .-.................................. 30 3.3 Non-Standard Fuel in the Pool ................................................................................ 31 4 Storage Rack Description ........................................................... 33 4.1 New Fuel Storage Area ............................................................................................ 33 4.2 Spent Fuel Pool ........................................................................................................ 36 4.2.1 Region 1 ...................................................................................................... 37 4.2.2 Region 2 ...................................................................................................... 41 4.2.3 Region 3 ...................................................................................................... 45 5 Overview of the Method of Analysis .......................................... 50 5.1 New Fuel Storage Area ............................................................................................ 50 5.2 Spent Fuel Storage Racks ....................................................................................... 51 5.2.1 Storage Geometry ...................................................................................... 51 5.2.2 Bounding Fuel Design ................................................................................ 53 5.2.3 Soluble Boron Credit .................................................................................. 54 5.2.4 Burnup Credit .............................................................................................54 5.2.5 Other Credit ................................................................................................57 5.2.6 Neutron Absorbers ..................................................................................... 57 5.2.7 Accident Analysis .......................................................................~ ............... 57 5.2.8 Normal Operations ..................................................................................... 58 6 Cross Sections, Computer Codes, and Validation .................... 59 6.1 Cross Sections and Computer Codes .................................................................... 59 6.1.1 CSAS5 ......................................................................................................... 59 6.1.2 Isotopes Used ............................................................................................. 60

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 3 of 300 6.1.3 TRITON ........................................................................................................ 63 6.2 Uncertainty in Depleted Fuel Isotopic Content ...................................................... 64 6.3 Validation of Criticality Analysis ............................................................................. 65 6.3.1 Major Actinides and Structural Materials .................................................. 65 6.3.2 Minor Actinides and Fission Products ...................................................... 69 6.3.3 Temperature Dependence .......................................................................... 69 6.3.4 Absorbers Credited .................................................................................... 70 7 Criticality Safety Analysis of the New Fuel Storage Area ......... 71 7.1 New Fuel Storage Area KENO Model ...................................................................... 71 7.2 Limiting Fuel Design ................................................................-................................ 74 7.2.1 Fuel Dimensions and Materials ................................................................. 74 7.2.2 Fuel Density, Burnable Poisons, and Axial Blankets ............................... 75 7.3 Limiting Rack Model ................................................................................................ 77 7.3.1 NFSR Materials and Dimensions ............................................................... 77 7.3.2 Temperature and Flooding ........................................................................ 79 7.3.3 Low Density Moderator .............................................................................. 80 7.3.4 Asymmetric Fuel Placement ...................................................................... 82 7.3.5 Summary of the Base Case for the NFSA Analysis ................................. 83 7.4 Biases and Uncertainties for the New Fuel Storage Area Analysis ...................... 84 7.5 Accident Conditions ................................................................................................ 88 7.5.1 Optimum Moderation ................................................................................. 88 7.5.2 Dropped/Misplaced Assembly ................................................................... 88 7.5.3 Seismic Event ............................................................................................. 88 8 Depletion Modeling and Burnup Effects .................................... 89 8.1 Depletion Method Overview .................................................................................... 89 8.2 Bounding Fuel Assembly Depletion Power ............................................................ 90 8.3 Bounding Depletion Boron ...................................................................................... 93 8.4 Bounding RCS Temperature .................................................................................... 94 8.5 Bounding Fuel Temperature .................................................................................... 95 8.6 Bounding Axial Burn up Profiles ............................................................................. 97 8.7 Burnable Absorbers ................................................................................................. 98 8.8 Control Rod History ............................................................................................... 100 8.9 In-core Thimble ...................................................................................................... 102 8.1 O Reduced Power before Storage ............................................................................ 103

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 4 of 300 8.11 Grid Growth and Clad Creep Depletion Effects .................................................... 104 8.12 Grids ............................................................................................, ........................... 104 8.13 Instrument and Guide Tube Design ...................................................................... 105 8.14 Comparison of CASMO and TRITON Depletion Reactivity .................................. 106 8.15 TRITON Depletion Model Summary ...................................................................... 108 9 Spent Fuel Rack Analysis ......................................................... 11 O 9.1 Region 1 Analysis .................................................................................................. 11 O 9.1.1 Region 1 Fuel Storage .............................................................................. 11 O 9.1.2 Region 1 Modeling Assumptions ............................................................ 11 O 9.2 Region 1 Infinite Lattice KENO Model ................................................................... 112 9.2.1 BORAL Blisters ........................................................................................ 114 9.2.2 SFP Normal Operation Water Temperature and Density ....................... 114 9.2.3 Simplified TRITON Depletion Input for Burn up Credit ...... :.................... 115 9.2.4 IFBA Credit ................................................................................................ 115 9.2.5 Asymmetric Fuel Placement .................................................................... 115 9.2.6 Code Validation Bias and Uncertainty .................................................... 117 9.2.7 Measured Burnup Uncertainty ................................................................. 117 9.2.8 Summary of Bias and Uncertainty........................................................... 119 9.2.9 Region 1 Infinite Lattice k-eff and Margin Calculation ........................... 121 9.3 Region 1 KENO SFP Wall Credit Model ................................................................ 124 9.4 Region 1 Analysis Summary ................................................................................. 127 9.5 Region 2 Analysis .................................................................................................. 128 9.5.1 Region 2 Fuel Storage .............................................................................. 128 9.5.2 Region 2 Modeling Assumptions ............................................................ 128 9.6 Region 2 Infinite Lattice Model .............................................................................. 129 9.6.1 BORAL Blisters ........................................................................................ 132 9.6.2 SFP Normal Operation Water Temperature and Density ....................... 132 9.6.3 TRITON Depletion Input for Burnup Credit ............................................. 132 9.6.4 Asymmetric Fuel Placement .................................................................... 132 9.6.5 Fuel Geometry Changes with Burnup ..................................................... 132 9.6.6 Horizontal Burnup Tilt .............................................................................. 137 9.6.7 Code Validation Bias and Uncertainty .................................................... 141 9.6.8 Summary of Bias and Uncertainty........................................................... 142 9.6.9 Unbounded Historical Fuel ...................................................................... 144 9.6.10 Region 2 k-eff and Margin Calculation - Fresh fuel ............................... 145

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 5 of 300 9.6.11 Region 2 Control Rod Credit.................................................................... 147 9.6.12 Region 2 k-eff and Margin Calculation - Burnup credit ......................... 150 9.7 Region 2 Analysis Summary ................................................................................. 155 9.8 Region 3 Analysis .................................................................................................. 156 9.8.1 Region 3 Fuel Storage .............................................................................. 156 9.8.2 Region 3 Modeling Assumptions ............................................................ 156 9.9 Region 3 Infinite Lattice Model .............................................................................. 157 9.9.1 SFP Normal Operation Water Temperature and Density ....................... 160 9.9.2 TRITON Depletion Input for Burnup Credit ............................................. 160 9.9.3 Asymmetric Fuel Placement .................................................................... 160 9.9.4 Fuel Geometry Changes with Burnup ..................................................... 160 9.9.5 Horizontal Burnup Tilt .............................................................................. 161 9.9.6 Code Validation Bias and Uncertainty .................................................... 163 9.9.7 Summary of Bias and Uncertainty........................................................... 163 9.9.8 Unbounded Historical Fuel ...................................................................... 165 9.9.9 Region 3 k-eff and Margin Calculation Without Decay Time Credit ...... 16~

9.9.1 O Region 3 k-eff and Margin Calculation With Decay Time Credit ........... 173 9.1 O Region 3 Analysis Summary ................................................................................. 182 10 Interface Analysis ...................................................................... 183 10.1 Interfaces Between Storage Configurations Within a Rack ................................ 183 10.2 Interfaces Between Dissimilar Racks ................................................................... 183 10.2.1 MPS3 Region lnterfaces ........................................................................... 184 10.2.2 Interface Analysis Method ....................................................................... 184 10.3 Region 1-2 Interface ............................................................................................... 185 10.3.1 Region 1-2 Interface with Region 2 Fresh Fuel.. ..................................... 189 10.3.2 Region 1-2 Interface with Region 2 Depleted Fuel ................................. 189 10.4 Region 2-3 Interface ............................................................................................... 193 10.4.1 Region 2-3 Interface (Fresh / Fresh) ........................................................ 195 10.4.2 Region 2-3 Interface (Fresh / Depleted) .................................................. 195 10.4.3 Region 2-3 Interface (Depleted/ Fresh) .................................................. 198 10.4.4 Region 2-3 Interface (Depleted / Depleted) ............................................. 202 10.5 Interface Analysis Summary ................................................................................. 205 11 Normal Conditions .................................................................... 206 11.1 Fuel Handling ......................................................................................................... 206

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 6 of 300 11.2 Fuel Inspection ....................................................................................................... 208 11.3 Non-standard Fuel .................................................................................................. 208 11.3.1 Fuel Rod Storage Canister ....................................................................... 208 11.3.2 Reconstituted Fuel ................................................................................... 210 11.4 Normal Condition Boron Credit. ............................................................................ 215 12 Accident Analysis ...................................................................... 221 12.1 Accident Condition Soluble Boron Requirement ................................................. 221 12.1.1 Loss of Cooling ........................................................................................ 221 12.1.2 Single Mis-placement ............................................................................... 222 12.1.3 Multiple Mis-load ...... .-............................................................................... 226 12.1.4 Dropped Assembly ................................................................................... 232 12.1.5 Handling Error .......................................................................................... 233 12.1.6 Seismic Event ........................................................................................... 234 12.2 Accident Analysis Summary ................................................................................. 234 13 Summary and Conclusions ....................................................... 235 13.1 New Fuel Storage Area ..................... .- ... *................................................................. 235 13.2 Spent Fuel Pool .................................. ;..... ;............................................................. 236 13.3 Bounding Fuel Design Values ............................................................................... 236 13.4 Bounding Depletion Condition Input .................................................................... 236 13.5 Summary of Loading Constraints ......................................................................... 237 14 References ................................................................................. 243 Appendix A: Validation for Criticality Analysis Using Laboratory Critical Experiments ....................................................... 24 7 A.1. Overview..................................................................................... 247 A.2. Definition of the Range of Parameters to Be Validated .......... 248 A.3. Selection of the Critical Benchmark Experiments ................... 248 A.3.1 Selection of the Fresh U0 2 Critical Benchmark Experiments ............................. 248 A.3.2 Selection of MOX Critical Experiments ................................................................. 258 A.4. Modeling and Calculating k of the Critical Experiments ......... 260 A.5. Statistical Analysis of the Data ................................................. 271 A.5.1 Statistical Analysis of the U02 Critical Experiments ........................................... 273 A.5.2 Statistic;;tl Analysis of MOX Critical Experiments ................................................ 287 A.5.3 Subcritical Margin ..................................................................................... .- ............ 292

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 7 of 300 A.6. Area of Applicability (Benchmark Applicability) ..................... 293 A.7. Summary and Recommendations ............................................ 294 A.8. Temperature Bias ...................................................................... 296 A.9. Appendix References ................................................................ 300

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 8 of 300 List of Tables Table 2.1: Use of DSS-ISG-2010-01 ........................................................*.............................. 21 Table 3.1: Fuel Design Dimensions ...................................................................................... 24 Table 3.2: Fuel Design Tolerances ....................................................................................... 25 Table 3.3: Feed Fuel Design and Enrichment History for Millstone Unit 3 ........................ 26 Table 3.4: Description of Burnable Poisons ........................................................................ 28 Table 3.5 Description of Control Rods ................................................................................. 29 Table 3.6: Description of In-core Thimble ............................................................................ 30 Table 4.1: New Fuel Storage Rack Dimensions ................................................................... 34 Table 4.2: Axial Heights of the New Fuel Storage Area KENO Model (lnches) .................. 35 Table 4.3: Region 1 Rack Dimensions and Tolerances ........................................................ 40 Table 4.4: Region 1 Rack Axial Alignment ............................................................................ 41 Table 4.5: Region 2 Racks ...................... ~ ............................................................................... 42 Table 4.6: Region 2 Rack Dimensions and Tolerances ........................................................ 44 Table 4.7: Region 2 Rack Axial Alignment ............................................................................ 44 Table 4.8: Region 3 Rack Model X-Y Dimensions (cm) ....................................................... 47 Table 4.9: Region 3 Rack Model Axial Dimensions ............................................................. 49 Table 5.1 - Storage Geometries ............................................................................................. 51 Table 6.1: Isotopes Used in the Criticality Analysis ............................................................ 60 Table 6.2: Fission Product Gases and Volatiles .................................................................. 61 Table 6.3: Fission Gas Release Fractions ............................................................................ 62 Table 6.4 ....................................................................................................................NOT USED Table 6.5 ....................................................................................................................NOT USED Table 6.6: Summary of Validation Bias and Uncertainty From Major Actinides and Structural Materials ................................................................................................................68 Table 7 .1 : EPRI Dry Concrete Composition ......................................................................... 77 Table 7.2: k-eff of NFSR vs. Concrete Composition ............................................................ 78 Table 7.3: Axial Location of Active Fuel for Various Fuel Design Types ........................... 78 Table 7.4: k-eff of NFSR vs. Active Fuel Axial Location ...................................................... 79 Table 7.5: NFSR Reactivity Sensitivity to Water Density .................................................... 79 Table 7.6: NFSR Reactivity Sensitivity to Water Density .................................................... 80 Table 7.7: k-eff of NFSR vs. Assembly Asymmetric Position ............................................. 83

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 9 of 300 Table 7.8: MPS3 NFSR Bias and Uncertainty Cases ....................*...................................... 84 Table 7.9: MPS3 NFSR Maximum k-eff, Full Density Water ..........................*..................... 85 Table 7.10: Maximum k-eff for MPS3 NFSR, Optimum Moderation *................................... 87 Table 8.1: Bounding Burnup Averages Relative Assembly Power versus Burnup........... 92 Table 8.2: RCS Thermal Hydraulic History for MPS3 .......................................................... 95 Table 8.3: Example of Nodal Depletion Conditions (20 GWd/MTU) ..................*.................. 96 Table 8.4 ................................................................................*...................................NOT USED Table 8.5 .....................................................................................................................NOT USED Table 8.6: Depletion with Different Burnable Absorbers (5.0 wt%, 54 GWDIT) .................. 98 Table 8.7: Depletion with IFBA and WABA Simultaneously (Node 16) .............................. 99 Table 8.8: Control Rod Dimensions ........................................................................*........... 100 Table 8.9: Control Rod Use Sensitivity Cases ....................................................*.........*.... 102 Table 8.1 O: In-Core Thimble and Instrument Tube Dimensions .....*.....*.....*...**................ 102 Table 8.11: Low Power EOC Effect on SFP Reactivity (5.0 w/o, 54 GWDIT) ......*.............. 103 Table 8.12: SFP Grid Modeling Effect (Node 16, Fresh Fuel, O ppm soluble boron) ........ 105 Table 8.13: Effect of Depleting with Grids (Node 16, O ppm soluble boron) ..................... 105 Table 8.14: Instrument and Guide Tube Dimensions ..........................................*............. 105 Table 8.15: Comparison of CASMO and TRITON Depletion Worth ....................**............ 107 Table 8.16: Depletion Parameters for TRITON Depletion Model.. ..................................... 108 Table 9.1: SFP Normal Operation Water Temperature and Density .................................. 114 Table 9.2: Validation Bias and Uncertainty (EALF :S 0.35 eV, Appendix A) .........**............ 117 Table 9.3: Summary of Region 1 Biases, Uncertainties, and Conservatism .................... 120 Table 9.4: Selected Tolerance Results for Region 1 (Fresh 4.75 wt% fuel) ...................... 121 Table 9.5: Worth of Burnup and Minor Actinides and Fission Products .*........................ 122 Table 9.6: Region 1 Total Bias, Uncertainty, and Margin (0 ppm soluble boron) ............. 123 Table 9.7: Region 1 Wall Model Base and Sensitivity Cases ................*.................*.......... 126 Table 9.8 .........................*.........................................***.................................**....*......NOT USED Table 9.9 ..........................................*..*......................................................................NOT USED Table 9.10: Horizontal Burnup Tilt Bias ............................................*.................*....*.......... 141 Table 9.11: Region 2 Validation Bias and Uncertainty ..............................................*........ 142 Table 9.12: Summary of Region 2 Biases, Uncertainties, and Conservatism ................... 143 Table 9.13: Selected Tolerance Results for Region 2 (Fresh fuel) ...................*................ 145 Table 9.14: Region 2 Total Bias, Uncertainty, and Margin (0 ppm soluble boron) ........... 146

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 1O of 300 Table 9.15: MPS3 Control Rod Data .................................................................................... 148 Table 9.16: Region 2 KENO Control Rod Credit Cases (5.0 w/o fresh fuel) ...................... 149 Table 9.17: SFP Region 2 Limiting Axial Shapes ................................................................ 150 Table 9.18: Enrichment Tolerance Calculation Comparison ............................................. 152 Table 9.19: Region 2 Base Cases (0 ppm boron) ................................................................ 152 Table 9.20: Region 2 Total Bias, Uncertainty, and Margin ................................................. 153 Table 9.21 ..................................................................................................................NOT USED Table 9.22: Region 3 Validation Bias and Uncertainty ....................................................... 163 Table 9.23: Summary of Region 3 Biases, Uncertainties, and Conservatism .............. :.... 164 Table 9.24: Depletion Conditions Comparison for Assembly 824 .................................... 166 Table 9.25: Modeling Assembly 824 with Measured Design Inputs ................................. 167 Table 9.26: Temperature and Burnup Shape Test Cases .................................................. 168 Table 9.27: SFP Region 3 Limiting Axial Shapes ................................................................ 169 Table 9.28: Region 3 Total Bias, Uncertainty, and Margin (0 years decay) ....................... 171 Table 9.29: Uncertainty Cases for Region 3 Decay Time Burnup Curves ........................ 175 Table 9.30: Region 3 Total Bias, Uncertainty, and Margin (3 years decay) ....................... 177 Table 9.31: Region 3 Total Bias, Uncertainty, and Margin (9 years decay) ....................... 178 Table 9.32: Region 3 Total Bias, Uncertainty, and Margin (18 years decay) ..................... 179 Table 9.33: Region 3 Total Bias, Uncertainty, and Margin (25 years decay) ..................... 180 Table 9.34: Region 3 Burnup Credit Curve Coefficients .................................................... 181 Table 10.1: Region 1-2 Interface with Fresh Fuel ............................................................... 189 Table 10.2: Region 1-2 Interface (Region 1 Fresh Fuel and Region 2 Depleted Fuel) ...... 189 Table 10.3: Region 1-2 Interface Model Selected Tolerances (Fresh/Depleted) ............... 190 Table 10.4: Region 1-2 Interface Model Tolerance Comparison (Fresh / Depleted) ......... 191 Table 10.5: Region 1-2 Interface Model Margin (Fresh/ Depleted) .................................... 192 Table 10.6: Region 2-3 Interface (Fresh/ Fresh) ................................................................. 195 Table 10.7: Region 2-3 Interface (Fresh / Depleted) ........................................................... 196 Table 10.8: Region 2-3 Interface Model Selected Tolerances (Fresh/ Depleted) ............. 196 Table 10.9: Region 2-3 Interface Model Tolerance Comparison (Fresh / Depleted) ;........ 197 Table 10.10: Region 2-3 Interface Model Margin (Fresh/ Depleted) .................................. 198 Table 10.11: Region 2-3 Interface (Depleted / Fresh) ................. ~ ........................................ 199 Table 10.12: Region 2-3 Interface Model Selected Tolerances (Depleted/ Fresh) .......... 199

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 11 of 300 Table 10.13: Region 2-3 Interface Model Tolerance Comparison (Depleted/ Fresh) ....... 200 Table 10.14: Region 2-3 Interface Model Margin (Depleted/ Fresh) .................................. 201 Table 10.15: Region 2-3 Interface (Depleted/ Depleted) .................................................... 202 Table 10.16: Region 2-3 Interface Model Selected Tolerances (Depleted/ Depleted) ...... 203 Table 10.17: Region 2-3 Interface Model Tolerance Comparison (Depleted / Depleted) .. 203 Table 10.18: Region 2-3 Interface Model Margin (Depleted/ Depleted) ............................. 204 Table 11.1: Region 2 and 3 Fuel Rod Storage Canister Model Results ............................. 210 Table 11.2: Fuel Reconstitution Assembly Isolation .......................................................... 212 Table 11.3: Fuel Reconstitution Fuel Rod Replacement .................................................... 213 Table 11.4: Region 1-3 Normal Condition Boron Requirement Cases (600 ppm) ............. 217 Table 11.5: Region 1 Bias, Uncertainty, and Margin Comparison ..................................... 218 Table 11.6: Region 2 Bias, Uncertainty, and Margin Comparison (fresh fuel) .................. 219 Table 11.7: Region 2 Bias, Uncertainty, and Margin Comparison (depleted fuel) ............ 220 Table 12.1: Loss of Cooling Event ...................................................................................... 222 Table 12.2: Single Fuel Assembly Mis-placement .............................................................. 225 Table 12.3: Millstone Unit 3 Fuel Batch History .................................................................. 227 Table 12.4: Multiple Mis-load Base Case Results ............................................................... 229 Table 12.5: Multiple Mis-load Tolerance Results (2550 ppm boron) .................................. 230 Table 12.6: Multiple Mis-load Bias, Uncertainty and Margin .............................................. 231 Table 12.7: Dropped Assembly Results .............................................................................. 233 Table 12.8: Dropped Assembly Results .............................................................................. 233 Table 13.1: Bounding Fuel Design Values ......................................................................... 235 Table 13.2: Bounding Fuel Design Values ......................................................................... 236 Table 13.3: Bounding Depletion Conditions for Burnup Credit ........................................ 237 Table 13.4: MPS3 SFP Storage Constraint Summary ........................................................ 238 Table 13.5: Burnup Credit Curve Polynomial Coefficients ................................................ 242 Table A.3.1: Selection Review of OECD/NEA Criticality Benchmarks .............................. 250 Table A.4.1: U02 Critical Experiment Results with SCALE 6.0 and ENDF/B-Vll.. ............ 261 Table A.4.2: HTC Critical Experiment Results with SCALE 6.0 and ENDF/B-Vll. ............. 267 Table A.4.3: Results of Low Enriched MOX Critical Experiments Calculated with SCALE

...............................................................................................................................................270

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 12 of 300 Table A.5.1: Bias and Uncertainty Based on the EALF Trend from U02 Critical Experiments .......................................................................................................................... 279 Table A.5.2: U02 Critical Experiment With Soluble Boron Results with SCALE 6.0 and ENDF/B-Vll .............................................................................................................................284 Table A.5.3: Bias and Uncertainty Based on the EALF Trend from U02 Critical Experiments ..........................................................................................................................286 Table A.5.4: Bias and Uncertainty Based on the EALF Trend from MOX Critical Experiments .................................................................................................:************** .......... 292 Table A.6.1: Area of Applicability (Benchmark Applicability) ........................................... 293 Table A. 7 .1: Summary of the Trend Analysis ..................................................................... 295 Table A.7.2: Final Bias and Uncertainty .............................................................................. 296 Table A.8.1: LCT-046 with Full Thermal Expansion Calculated with SCALE 6.0 and ENDF/B-Vll .............................................................................................................................297

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 13 of 300 List of Figures Figure 3.1: 17x17 Fuel Design ............................................................................................... 25 Figure 3.2: Fuel Rod Storage Container Schematic ............................................................ 32 Figure 4.1: Millstone Unit 3 New Fuel Storage Area ............................................................ 33 Figure 4.2: Axial Drawing of a Millstone Unit 3 New Fuel Storage Rack Module .............. 35 Figure 4.3: MPS3 SFP Rack Orientation ................................................................................ 36 Figure 4.4: MPS3 Region 1 Rack Dimensions (X-Y) ............................................................. 38 Figure 4.5: MPS3 Region 1 Rack Dimensions (X-Z) ............................................................. 39 Figure 4.6: MPS3 Region 2 Rack Dimensions (X-Y) ............................................................. 42 Figure 4.7: MPS3 Region 2 Rack Dimensions (X-2) ............................................................. 43 Figure 4.8: Location of Region 3 Racks in the SFP ............................................................. 45 Figure 4.9: X-Y Drawing of a Region 3 Rack Cell.. ............................................................... 46 Figure 4.10: X-Y Drawing of a Region 3 Rack Wrapper ....................................................... 46 Figure 4.11: Axial Drawing of a Region 3 Rack Cell ............................................................ 48 Figure 5.1: KENO Region 1A/1 B Wall Credit Model .............................................................. 52 Figure 7.1: Cutaway View of the New Fuel Storage Area Model ......................................... 72 Figure 7.2: Top View of Part of the New Fuel Storage Area Model.. ................................... 73 Figure 7.3: Full New Fuel Storage Area Model ..................................................................... 73 Figure 7.4: MPS3 Fuel Assembly Stack Density .................................................................. 75 Figure 7.5: k-eff of NFSR vs. Water Density ......................................................................... 81 Figure 7.6: Asymmetric Skewing of Assemblies ................................................................. 82 Figure 8.1: Bounding Burnup Averaged Relative Assembly Power ................................... 92 Figure 8.2: MPS3 Bounding Cycle Average Soluble Boron ................................................ 93 Figure 8.3 ..................................................................................................................NOT USED Figure 8.4 ..................................................................................................................NOT USED Figure 8.5: Cycle Average Control Rod Insertion .............................................................. 101 Figure 9.1: X-Y Representation of Region 1 KENO Model ................................................. 112 Figure 9.2: 3D Representation of Region 1 KENO Model (full model and cutaway) ......... 113 Figure 9.3: Asymmetric Fuel Placement (4x4 in a 6x6 model) ........................................... 116 Figure 9.4: KENO Region 1 Wall Credit Model .................................................................... 125 Figure 9.5 ..................................................................................................................NOT USED Figure 9.6 ..................................................................................................................NOT USED

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 14 of 300 Figure 9.7: X-Y Representation of Region 2 2x2 KENO Model .......................................... 130 Figure 9.8: 3D Representation of Region 2 2x2 KENO Model ............................................ 131 Figure 9.9: Zircaloy-4 Grid Growth [38] ............................................................................... 135 Figure 9.10: ZIRLO Grid Growth Data [42] .......................................................................... 136 Figure 9.11: Horizontal Burnup Tilt KENO Model #1 .......................................................... 138 Figure 9.12: Horizontal Burn up Tilt KENO Model #2 .......................................................... 139 Figure 9.13: Quadrant Tilt vs. Burnup ................................................................................ 140 Figure 9.14: Region 2 Bounding Burnup Credit Curve ...................................................... 154 Figure 9.15: X-Y Representation of Region 3 KENO Model ............................................... 158 Figure 9.16: Axial Representation of Region 3 KENO Model ............................................. 159 Figure 9.17: Region 3 Horizontal Burnup Tilt KENO Model #1 .......................................... 162 Figure 9.18: Region 3 Horizontal Burnup Tilt KENO Model #2 .......................................... 162 Figure 9.19: Region 3 Bounding Burnup Credit Curve (0 Years Decay) ........................... 172 Figure 9.20: Region 3 Bounding Burnup Credit Curves .................................................... 181 Figure 10.1 : X-Y Representation of the Region 1-2 Interface Model ................................. 186 Figure 10.2: X-Y Representation of the Region 1-2 Interface Showing Flux Traps .......... 188 Figure 10.3: X-Y Representation of the Region 2-3 Interface Model ................................. 193 Figure 11.1: k-eff of Two Fresh 5 wt% Fuel Assemblies in Unborated Water .................. 207 Figure 11.2: Location of MPS3 Fuel Handling Equipment ................................................ 207 Figure 11.3: Region 2 Fuel Rod Storage Canister Model ................................................... 209 Figure 11.4: Region 3 Fuel Rod Storage Canister Model ................................................... 209 Figure 11.5: Region 2 Reconstitution Model.. ..................................................................... 211 Figure 11.6: Region 3 Reconstitution Model.. ..................................................................... 212 Figure 11.7: Region 3 Fuel Assembly MR71 Model ............................................................ 214 Figure 12.1: Region 1-2 Interface Model for Between-rack Mis-placement ..................... 223 Figure 12.2: Region 2-3 Interface Model for Between-rack Mis-placement ...................... 224 Figure 12.3.................................................................................................................NOT USED Figure 12.4: Region 2 KENO Multiple Mis-load Model (partial X-Y view) .......................... 228 Figure 12.5: Region 2 Dropped Assembly Model .............................................................. 232 Figure 13.1: Region 1A/1B Orientation ................................................................................ 239 Figure 13.2: Region 2 Burnup Credit Curve ........................................................................ 240 Figure 13.3: Region 3 Burnup Credit Curves ...................................................................... 241

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 15 of 300 Figure A.5.1: Distribution of the Calculated k's Around the Mean for All the U02 Benchmarks .......................................................................................................................... 275 Figure A.5.2: Q-Q Normality Plot for the U02 Benchmarks ............................................... 275 Figure A.5.3: Calculated k for the U02 Critical Benchmarks as a Function of EALF ...... 280 Figure A.5.4: Calculated k for the U02 Critical Benchmarks as a Function of Pin Diameter

...............................................................................................................................................281 Figure A.5.5: Calculated k for the U02 Critical Benchmarks as a Function of Fuel Pin Pitch ....................................................................................................................................... 282 Figure A.5.6: Calculated k for the U02 Critical Benchmarks as a Function of Enrichment

...............................................................................................................................................283 Figure A.5.7: Calculated k for the U02 Critical Benchmarks as a Function of Soluble Boron ..................................................................................................................................... 285 Figure A.5.8: Calculated k for the Critical Benchmarks as a Function of Plutonium Content .................................................................................................................................. 288 Figure A.5.9a: Distribution of Calculated ks for the MOX Critical Benchmarks .............. 289 Figure A.5.9b: Q-Q Normality Plot for the MOX Benchmarks ........................................... 290 Figure A.5.10: Calculated MOX Critical k as a Function of EALF ..................................... 291 Figure A.8.1: LCT-046 Corrected Calculated k per Case ................................................... 298 Figure A.8.2: LCT-046 k versus Temperature .................................................................... 299 Annotation of proprietary information herein corresponds to the specific reason(s) for claiming the information as proprietary as delineated in the respective Affidavit executed by the owners of the information. The annotations used in Attachment 5 are provided as follows:

1) Holtec proprietary information - denoted with "H" superscripts:

2) Westinghouse proprietary information - denoted with "a,c" superscripts:

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 16 of 300 L"ISt 0 f Acronyms an d Abb rev1a *rions Burnable Poison Rod Assembly is a group of boron containing rodlets held BPRA together by a plate that rests on the assembly top nozzle.

DOE Department of Energy (USA)

EALF Energy of the Average Lethargy of a neutron causing Fission EOL End of Life FA Fuel Assembly FRSC Fuel Rod Storage Container FP Fission Product GWd/MTU Gigawatt*day per Metric Ton (Tonne) of Uranium. A unit of burnup.

Haut Taux de Combustion. This is a set of critical experiments done in France HTC that uses fuel that represents the uranium and plutonium content of 4.5 wt% fuel burned to 37.5 GWd/MTU ID Inner Diameter Integral Fuel Burnable Absorber which is a ZrB 2 coating placed on the outside of IFBA the fuel pellet. A Westinghouse product.

ISG Interim Staff Guidance from the NRG MOX Mixed Oxide fuel. Contains both U0 2 and Pu0 2

• MPS3 Millstone Power Station Unit 3 MW/MTU Megawatt per Metric Ton (Tonne) of Uranium. A unit of specific power.

MWd/MTU Megawatt*day per Metric Ton (Tonne) of Uranium. A unit of burnup.

NCS Nuclear Criticality Safety New Fuel Storage Area. Referring to the concrete pit and steel racks used to NFSA hold the new fuel. Sometimes used interchangeably with NSFR.

New Fuel Storage Racks. Referring to the racks inside the New Fuel Storage NFSR Area. Sometimes used interchangeably with NSFA.

OD Outer Diameter pcm 0.00001 in k (acronym from percent mille) ppm Parts per million by weight PWR Pressurized Water Reactor RCCA Rod Cluster Control Assembly (control rod)

RCS Reactor Coolant System RFA Robust Fuel Assembly is a fuel design manufactured by Westinghouse.

RSS square Root of the Sum of the Squares RTP Rated Thermal Power SER *Safety Evaluation Report SFP Spent Fuel Pool SNF Spent Nuclear Fuel ss Stainless Steel VF Volume Fraction Wet Annular Burnable Absorbers. This is a Westinghouse removable burnable WABA absorber product.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 17 of 300 1 Introduction A SFP criticality analysis is performed for Millstone Unit 3 (MPS3) to allow for a measurement uncertainty recapture power uprate, to incorporate spent fuel pool (SFP) soluble boron credit, to update computer codes and analysis methods, to eliminate the need for empty storage cells and cell blockers, and to increase identified margin to k-effective (k-eff) limits. A new fuel storage area (NFSA) criticality analysis is performed for MPS3 to update computer codes and analysis methods and to increase identified margin to k-eff limits. The current enrichment limit for the MPS3 NFSA and SFP is 5.0 wt% U-235.

Criticality analysis scope includes normal fuel storage, normal fuel handling, abnormal fuel storage conditions, abnormal fuel handling events, and storage of non-standard fuel-bearing components. A boron dilution analy~is is provided that identifies potential SFP dilution events, dilution water source flow rates and volumes, means of detection, time required for detection and mitigation, and minimum dilution event soluble boron concentration.

Changes to the operation and administration of the SFP are summarized as follows:

• Modify storage requirements in Technical Specifications 1.40, 1.41, 3/4.9.13 and 3/4.9.14, including Figures detailing storage patterns and minimum fuel burnup versus fuel enrichment.

• Eliminate the requirement for cell blockers in Region 1.

• Modify SFP soluble boron requirements in Technical Specifications 3/4.9.1.2.

• Modify the description of SFP criticality control in Technical Specification 5.6.1.1.

No changes to the operation and administration of the NFSA are proposed.

Fuel burnup when used to compare to fuel storage requirements is the volume averaged burnup of the assembly as determined using the measured reaction rates with no reduction for measurement uncertainty. Enrichment when used to compare to fuel storage requirements is the maximum planar volume averaged as-built initial enrichment in the assembly. If the assembly has axial blankets, the lower enriched fuel is not credited in determining the enrichment.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 18 of 300 2 Acceptance Criteria and Regulatory Guidance The Code of Federal Regulations Title 1O Part 50 Section 68 (b).4 states:

"If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritica/), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water."

This analysis shows at a 95 percent probability and 95 percent confidence level that if the fuel loaded in the SFP meets the Technical Specification requirement for enrichment, burnup, and decay time, the SFP k-eff will be less than 0.95 crediting soluble boron and less than 1.0 with unborated water as specified in 10CFR50.68. [1]

Further, Title 1O Part 50 Section 68 (b) paragraphs 2 and 3 state:

"(2) The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.

(3) If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with /ow-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed O. 98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation o,r if fresh fuel storage racks are not used."

This analysis of the fresh fuel storage racks shows that the storage racks meet the requirements of paragraphs 2 and 3 when fully loaded with fresh 5.0 wt% U-235 fuel assemblies.

Meeting 10CFR50.68 also satisfies 10CFR50, Appendix A, General Design Criterion 62 [2]

which states:

"Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations."

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 19 of 300 Guidance for the regulatory review is found in the Standard Review Plan, NUREG-0800 Section 9.1.1, "Criticality Safety of Fresh and Spent Fuel Storage and Handling." [3] Below are the eleven specific areas of review from this guidance with where the information is found in this report. (Note that there are actually 13 areas of review but the last two do not apply to License Amendment Requests.)

1 . Fuel assembly design to verify that appropriate fuel assembly data were used.

Fuel assembly design data is found in Section 3.

2. Fuel storage rack design to verify that appropriate fuel storage rack data were used.

Storage rack design data is found in Section 4.

3. Evaluation of performance effectiveness of the neutron absorbing materials in the fresh and spent fuel racks.

BORAL is credited in the NFSA and in Region 1 and 2 SFP racks. A BORAL coupon monitoring program description has been provided to the NRC. [5, 6]

4. Computational methods and related data to verify that acceptable computational methods and data were used.

Computational methods are described in Sections 5 and 6.

5. Computational method validation to verify that the validation study is thorough and uses benchmark critical experiments that are similar to the normal-conditions and abnormal conditions models and to verify that the neutron distribution coefficient (K(eff))

bias and bias uncertainty values are conservatively determined.

Validation is summarized in Section 6 and the details are provided in Appendix A.

6. Identification of normal conditions to verify that the scope of specified normal conditions is comprehensive.

Range of normal conditions is identified in Section 11

7. Normal-conditions models to verify that normal conditions are modeled conservatively and that all modeling approximations and assumptions are appropriate.

Normal conditions models and the tolerances and uncertainties in these models are described in Sections 7 - 11.

8. Identification of abnormal conditions to verify that the scope of considered abnormal conditions is comprehensive.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 20 of 300 Abnormal conditions are described in Section 7 and 12.

9. Abnormal-conditions models to verify that abnormal conditions are modeled*

conservatively and that all modeling approximations and assumptions* are appropriate.

Abnormal condition models are described in Section 12.

1O. Analysis of normal and credible abnormal conditions to verify that the analysis is complete and logically sound and that assumptions, limits, and controls are clearly stated.

The analysis is contained in Sections 7 through 12. The limitations of the analysis are listed in Section 13.

11 . Analysis conclusions to verify the applicant's conclusions regarding maintaining subcriticality for all normal and credible abnormal conditions.

The conclusion of the analysis is in Section 13.

Guidance for spent fuel pool criticality analysis is given in DSS-ISG-2010-01. [4]

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 21 of 300 Table 2.1: Use of DSS-ISG-2010-01

1. Fuel Assembly Selection MPS3 fuel designs and variations are described. Sections 3, The NCS analysis must Variations are bounded by the analysis for both 7.2, 8.7, adequately bound all designs depletions and criticality calculations by use of a hybrid 8.9, 8.11, and variations within a design. assembly design and biases as needed. Zr was used 8.12, and to conservatively represent all Zr based grid and clad 8.13 allo s.

2. Depletion Analysis Used 5% depletion uncertainty only for isotopic Sections a.i. Depletion uncertainty concentration uncertainty. Code validation covers 6.2, 6.3, (5%) covers only isotopic major actinide cross section uncertainty. Bias and 9.2.8, concentration uncertainty. uncertainty in the worth of fission products and minor 9.6.8, and actinides is covered by bias of 1.5% of the fission 9.9.9 roduction and minor actinides worth.

2. Depletion Analysis Followed. Section 6.2, a.ii. Reactivity decrement 9.2.7, should not include the worth of 9.6.8, 9.9.9 the burnable absorbers.

2. Depletion Analysis Bounding values are identified and used Section 8 b.i. Bounding values should be used.

2. Depletion Analysis Fuel and moderator temperatures are maximized Section 8 b.ii. Use the more limiting based on high specific power. Depletion at reduced bounding parameter when a power is performed near EOL without changing conflict occurs. moderator and fuel tern eratures.

2. Depletion Analysis Bounding values were used for all parameters or Section 8.15 b.iii. Non-bounding values justification provided.

are outside sea e of ISG.

2. Depletion Analysis All removable burnable absorbers are identified and Section 8.7 c.i. All removable burnable the most limiting burnable absorbers are used.

absorbers must be considered. Justification rovided for earl c cle BPRA.

2. Depletion Analysis Reference depletions use maximum WABA, which Section 8.7 c.ii. Limiting integral bounds the effect of maximum IFBA. Maximum IFBA burnable absorbers should be plus some WABA is also considered.

used.

2. Depletion Analysis WABA are conservatively modeled in the depletion Section 8.7 c.iii. Model the burnable analysis as full length with no cutback region and absorbers appropriately. contain the maximum desi n B-10 loadin .

2. Depletion Analysis The depletion model correctly accounts for competing Section 8.15 c.iv. Consider competing effects.

effects

2. Depletion Analysis Rodded operation is bounded by the standard Sections 8.8 d.i. Spectrum hardening from depletion. Early cycle exceptions are justified. and 9.9.8 rodded operation should be considered.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 22 of 300

2. Depletion Analysis NUREG/CR-6801 axial burnup profiles were used, Section 8.6 d.ii. Effect of control rods on which include rodded operation effects.

the axial burnup profile should be considered

3. Criticality Analysis NUREG/CR-6801 axial burnup profiles were used and Section 8.6

a. Axial Burnup Profile justified by comparison to MPS3 profiles. Analysis was also done with uniform burnup and the most limitin of the two rofiles was used.

3. Criticality Analysis The rack dimensions and materials are taken from the Section 4

b. Rack Model manufacturer's drawings.

i. Model inputs should be traceable.

3. Criticality Analysis BORAL is modeled with minimum certified areal Sections

b. Rack Model density. BORAL coupon program is used to verify 4.2, 5.2.6, ii. Efficiency of the neutron conformity of the rack absorber and the analysis. and 9.6.11 absorber should be established. RCCA credit assumes maximum anticipated RCCA depletion over the portion of absorber inserted during lant operation.

3. Criticality Analysis RCCA credit assumes maximum anticipated RCCA Sections

b. Rack Model depletion over the portion of absorber inserted during 4.2, 5.2.6, iii. Conservative degradation plant operation. Conservative (low) BORAL B-10 and 9.6.11 should be used. content used in analysis is confirmed by coupon monitorin ro ram.

3. Criticality Analysis Either the maximum uncertainty from either side is Section 10

c. Interfaces - If not used or the biases and uncertainties for the interface determining interface model model are calculated.

biases and uncertainties, use the maximum uncertainties from either side.

3. Criticality Analysis Normal conditions are considered. Section 11

d. Normal Conditions - All normal conditions such as movement of fuel and inspections should be considered.

3. Criticality Analysis Normal initial conditions are considered as base Section 12

e. Accident Conditions conditions for the accident anal sis.

4. Criticality Code Validation NUREG/CR-6698 was followed for the validation. Appendix A NUREG/CR-6698 endorsed

4. Criticality Code Validation The HTC critical experiments are included in the Appendix A

a. Area of Applicability analysis.

i. Include the HTC criticals

4. Criticality Code Validation Appropriate critical experiments are used. Appendix A

a. Area of Applicability ii. Use appropriate criticals

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 23 of 300

'~9.i~.~~c~ from os 010;.01?******

A large sample of critical experiments is used. Appendix A

4. Criticality Code Validation Correlation avoided by use of experiments from Appendix A

a. Area of Applicability multiple critical facilities.

iv. Be sure the set is not hi hi correlated.

4. Criticality Code Validation Trend analysis is performed on the major parameters. Appendix A

b. Trend Analysis The trend analysis finds the best linear fit. No trends Adequate, appropriate, are rejected to be conservative. The most limiting bias not rejected. and uncertainty for the area of applicability is applied assuming both that all trends are real and there are no trends.

4. Criticality Code Validation The statistical approach recommended in NUREG/CR- Appendix A

c. Statistical Treatment 6698 is used. The variance of the population about

i. Use the variance of the the mean rather than the variance of the mean is used.

o ulation about the mean

4. Criticality Code Validation The statistical approach recommended in NU REG/CR- Appendix A

c. Statistical Treatment 6698 is used. The correct confidence factors were ii. Use correct confidence used.

factors.

4. Criticality Code Validation Normality testing was performed and the assumption Appendix A

c. Statistical Treatment of normality justified.

iii. Consider Normalit

4. Criticality Code Validation Lumped Fission Products are not used. Section 6.1

d. Lum ed Fission Products

4. Criticality Code Validation 5% of the delta k of depletion is used to cover isotopic Sections

e. Code-to-Code uncertainty. The analysis in the North Anna Power 6.2 & 8.14 Comparisons Station LAR dated May 2, 2017 confirmed this is appropriate for SCALE/TRITON by comparisons to CASM0-4 and 5.

5. Miscellaneous Precedence is not quoted as a licensing basis.

a. Precedence References used were carefully chosen to be

b. References applicable to the point being made.

c. Assum tions Assumptions are identified and justified.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 24 of 300 3 Reactor and Fuel Design Description 3.1 Fuel Description MPS3 uses a 17x17 lattice fuel with a center instrument tube and 24 guide tubes. Four fuel designs have been used but all the designs are similar for the criticality analysis. The initial fuel design, which is designated "Standard," used all lnconel grids. The currently used fuel design is the Westinghouse RFA-2 design. Table 3.1 provides fuel dimensions.

Table 3.1: Fuel Design Dimensions (Dimensions in inches)

• • *.** .** *.* Standard VSH .*. RFA/RFA-2. NGF.>>

Pellet Diameter 0.3225 0.3225 0.3225 0.3225 Clad Inner Diameter 0.329 0.329 0.329 0.329 Clad Outer Diameter 0.374 0.374 0.374 0.374 ZIRLO Zircaloy-4 Opt.

Clad Material Zircaloy-4 Opt.

ZIRLO ZIRLO ZIRLO Rod Pitch 0.496 0.496 0.496 0.496 Guide Tube and Instrument Tube Inner Diameter 0.450 0.442 0.442 0.442 Outer Diameter 0.482 0.474 0.482 0.482 Grid Volume (cubic inches) 1 [ t*e [ re [ re [ ]a,e 1

The gnd volume 1s the volume of the grids plus sleeves m the active fuel ignoring the bottom grid. The grids are a zirconium alloy except for the Standard fuel which used all lnconel grids.

The fuel pellet is dished and chamfered. The fuel batch stack density (density of the pellet reduced by the dishing and chamfering and takes no credit for annular pellet blankets) has ranged from [ ]a,e of the U02 theoretical density. Figure 7.4 shows that the maximum assembly stack density is [ ]a,e. Two standard deviation variation of fuel assembly stack density within a batch is 0.5%. Manufacturing tolerances for the fuel are found on Table 3.2. MPS3 fuel has had reduced enrichment axial blankets since Cycle 3 and annular pellet axial blankets since Cycle 5. Axial blanket length is nominally 6 inches on each end of the fuel pins. The annular pellets have the same outer diameter as standard pellets but have a void center with a diameter of 0.155 inches.

The active length of the fuel is 144 inches [ ]a,e. The distance from the bottom of the fuel assembly to the bottom of the active fuel has varied by design from [

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 25 of 300

]a,c_ Axial blanket enrichments of 0.74 wt% (Batches 6 and 7) and 2.6 wt% have been used. All fuel pins in an assembly have the same central zone enrichment and the same blanket enrichment.

The assembly pitch in the core is 8.466 inches (21.50 cm). The location of the guide tubes in the assembly is given on Figure 3.1. Table 3.3 lists the fuel type and enrichment of each fuel Batch.

T a bl e 3.2 . Fue ID es1gn T o Ierances Pellet Diameter [ re Clad Inner Diameter [ re Clad Outer Diameter [ re Rod Pitch [ re Guide Tube and Instrument Tube Inner Diameter [ t*e Guide Tube and Instrument Tube Outer Diameter [ t*e Stack Density [ re Figure 3.1: 17x17 Fuel Design F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F F (F = Fuel, G = Guide Tube, I = Instrumentation Tube)

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 26 of 300 Table 3.3: Feed Fuel Design and Enrichment History for MPS3 ii'.~~$'~~~~;~,~:!,~~ ,: ::1,,ij~1R 1 Standard 2.42, 2.9, 3.4 1

~~::c~i~t :

N/A 2 Standard 3.5, 3.8 N/A 3 Standard 4.1, 4.5 0.74 4 V5H 4.2, 4.5 0.74 5 V5H 4.4 0.74 6 V5H 4.6 0.74 7 RFA 4.4, 4.8 2.6 8 RFA 4.4, 4.8 2.6 9 RFA 4.2, 4.7 2.6 10 RFA-2, NGF 4.7, 4.95 2.6 11 RFA-2 4.0, 4.95 2.6 12 RFA-2 4.7, 4.95 2.6 13 RFA-2 4.1, 4.9 2.6 14 RFA-2 4.1, 4.95 2.6 15 RFA-2 4.1, 4.95 2.6 16 RFA-2 4.1, 4.95 2.6 17 RFA-2 4.1, 4.95 2.6

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 27 of 300 3.2 Fuel Inserts Description 3.2.1 Burnable Absorbers Two types of burnable poison have been used at MPS3. Pyrex BP was used in Cycles 1 and 2.

Integral Fuel Burnable Absorber (IFBA) has been used for the remaining cycles. WABA will also be considered in this analysis to allow possible future use of WABA.

Pyrex burnable absorbers consist of an annulus of borosilicate glass with an air filled central hole and an inner and outer SS clad. The WABA design is also annular, but has RCS water in the inner hole and Zircaloy-4 clad. Pyrex and WABA assemblies are composed of rods inserted in the fuel assembly guide tubes suspended from a baseplate that rests on the fuel assembly top nozzle. The rods are often referred to as "fingers." There are up to 24 fingers in a Pyrex or WABA assembly. Table 3.4 provides the information needed to model Pyrex and WABA burnable absorbers.

An IFBA rod is a fuel rod with a thin ZrB 2 coating on the surface of some of the fuel pellets. The pellets at the top and bottom 6 inches of an IFBA rod are annular. MPS3 fuel assemblies with IFBA have contained 32 to 128 IFBA rods. Table 3.4 provides IFBA modeling data. BP and IFBA have not been used in the same fuel assembly at MPS3.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 28 of 300 Table 3.4: Description of Burnable Poisons Dimension Pyrex BP IFBA WABA Cycles Used 1&2 3 -18 None Inner Clad ID (cm) 0.428 N/A 0.572 Inner Clad OD (cm) 0.460 N/A 0.6782 BP ID (cm) 0.483 0.8192 0.706 BP OD (cm) 0.853 0.8198 0.808 Outer Clad ID (cm) 0.874 N/A 0.8357 Outer Clad OD (cm) 0.9677 N/A 0.968 Clad Material SS-304 N/A Zirc-4

[ re a,e Poison Loading (mg B-10/cm) [ re* [ 1

[ re BP Material 8203 in Pyrex ZrB2 84C in A'203 glass BP Material Density (g/cc) [ ]8*e [ re [ ]a,e 108 (Batch 5)

Poison axial length (in) 144** 120 (Batch 6-16) 144**

122 (Batch 17+)

Poison lower bound axial 18 (Batch 5) location above bottom of fuel O** 12 (Batch 6-16) O**

stack (in.) 11 (Batch 17+)

• Calculated using a 8 20 3 loading of 12.5 w/o and natural B-10/8 isotope abundance,

• • Conservatively assumed to be full length for fuel depletion. WABA has not been used at MPS3. North Anna WABA has 128 inches of absorber (-8 inches of un-poisoned fuel at top and bottom)

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 29 of 300 3.2.2 Control Rods There have been three control rod designs used in MPS3. Table 3.5 summarizes the relevant design features. The first generation of MPS3 control rods were Westinghouse Hf-Zr type. A transition to the second generation (Westinghouse Ag-In-Cd) began in Cycle 3. A transition to the third generation of control rods (AREVA Ag-In-Cd) began in Cycle 13.

A control rod assembly has 24 fingers. When fully inserted, the control rod neutron absorber extends above the top of the active fuel, but can leave a maximum of 5.25 inches of the bottom of the active fuel un-poisoned. At the bottom of the control rod is a SS end plug.

Table 3.5 Description of Control Rods

• ** ~0111p911e11f * **:** .< .**

D~sign

.<. *. . Characteristic\

< . *. :.< ::.. *<: .. '/ >><>: .*. *:.

Control Rod Absorber Diameter 0.866 cm Clad Inner Diameter 0.874 cm Clad Outer Diameter 0.968 cm Absorber Material (Cycles 1-4) Hf (95.3 - 95.4 w/o)-Zr(4.5 w/o)

Ag(80 w/o)- ln(15 w/o)-

Absorber Material (Cycle 5 - present)

Cd(5 w/o)

Maximum Distance from the bottom end of the 5.25 inches absorber to the bottom end of the active fuel Material and d1mens1onal tolerances are not provided because of excess margin in control rod credit cases.

Lower 12 inches of absorber in' the AREVA design has a slightly smaller absorber diameter.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 30 of 300 3.2.3 Sources and Other Inserts Primary source rods (a single rod displacing a Pyrex BP rod in a 24 BP assembly) and secondary source rods (4 rod assembly) were used in Cycle 1. Because source rods do not contail') a strong absorber material the impact of a source rod is less than a BPRA rod.

Secondary source assemblies (SSA) have been used throughout MPS3's history, though not in every cycle. SSA with 4 pins were used in assemblies concurrently with 6 finger Pyrex BP assemblies during Cycle 2. SSA with 4 pins and with 6 pins have been used in combination with IFBA. Secondary source pins consist of SS clad containing Sb-Be source material and inert spacers and can be modeled during fuel depletion as solid SS rods with an outer diameter of 0.381 inches. SSA are weaker absorbers and displace less water than 24 finger WABA.

Therefore, depletion effects of 24 finger WABA bound those of SSA alone. Depletion effects of SSA and IFBA in the same fuel assembly are evaluated in the determination of the bounding fuel assembly depletion conditions.

Prior to operation, thimbles for the in-core flux monitoring are inserted into the instrument tubes of about one third of the assemblies. Since these thimbles displace water they have a small effect on fuel depletion. The dimensions of the thimbles are given in Table 3.6.

Table 3.6: Description of In-core Thimble In-Core Instrument Thimble Thimble Inner Diameter 0.51 cm Thimble Outer Diameter 0.76 cm Thimble Material Stainless Steel 316 Fraction of assemblies with in-core thimble 30% (58 of 193}

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 31 of 300 3.3 Non-Standard Fuel in the Pool There are a few items in the spent fuel pool that are not standard fuel assemblies. If the item does not have fuel in it, it can be placed in any cell that is allowed to have fuel. The following items in the MPS3 SFP contain fuel but are not standard fuel assemblies:

• Fuel rod storage container (FRSC)

• Reconstituted fuel assemblies o With SS replacement pins o Without SS replacement pins (MR71 has two open lattice locations)

The key design features of the FRSC are:

• Square lattice

• 0.937 inch storage tube pitch

• 52 SS fuel rod storage tubes

• Storage tube OD 0.625 inch (48 tubes) and 0.750 inch (4 tubes)

• Storage tube wall thickness 0.035 inch (48 tubes) and 0.049 inch (4 tubes)

The storage canister currently contains one rod from fuel assembly G81, two from L74, and three from M71. After reconstitution, assembly M71 was renamed MR71. Removed rods were replaced with stainless steel pins except in MR71, which has two remaining empty fuel rod locations. Missing rods in MR71 are locations H01 and J01 (peripheral locations near the middle of one face of the assembly). Figure 3.2 is a schematic view of the FRSC.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 32 of 300 Figure 3.2: Fuel Rod Storage Container Schematic J8.\1':>

Oi.15 0\ . .(.o2'5 O.D. 'Ill .O?,~ WA\..\..

O"t ..i50 (:),0 "".049 WALl.. {4 TUBES)

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 32 of 300 Figure 3.2: Fuel Rod Storage Container Schematic J8. rte;,

I I

Oi.15 0\ . .<:"o2'5 O.D . ... . 0'3':> WA\.\..

01. . .iSO o.O .-. ,049 WA'-\.. l-4 TUBES) 0\ . .<:"o2.'5 O.D. "'.0"3":> WA\..\..

01. ..iSO o.O -,. ,049 WA\..\.. l-4 TUBES)

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 33 of 300 4 Storage Rack Description 4.1 New Fuel Storage Area The New Fuel Storage Rack (NFSR) is designed with four 6x4 storage cell rack modules.

BORAL neutron absorber panels are aligned parallel in a south to north orientation and are in between both the 2nd and 3rd cells and 4th and 5th cells of each rack module. The BORAL has a minimum B-10 areal density of 0.005 g/cm 2c. The X-Y dimensions of the NFSR modules are shown below in Table 4.1 . Figure 4.1 gives an overview of the area. Rack cells are depicted in the SE corner of the SW module inside the circle.

Figure 4.1: MPS3 New Fuel Storage Area

~ . .,....,.......,... ,-,...., . .... , ** .,w,w,-,.w,*...,.,,...._----.w,w,**www- - m - ~; , _if!~~tl~ (.f, ~At - - * - w * * * * * * *

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,-* - - - - r-I 1,,1,!

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 34 of 300 Table 4.1: New Fuel Storage Rack Dimensions Measurement Description Dimension (inches) Uncertainty (inches)

Cell ID 8 15/16 [ ]a,e Cell Wall Thickness 0.090 [ re Cell Pitch , North -South 22 1/8 [ re Cell Pitch , East-West 24 1/16 r re Gap Between Cell Center and Rack [ ]a,e 8 1,4 Module EdQe Gap Between Rack Modules, North-South 18 [ t*e Gap Between Rack Modules, East-West 8 5/8 [

1 t*e BORAL Sheet Thickness 0.080 N/A BORAL Sheet Width 76 [ re SS Wall Liner Thickness 1/4 r re 1

BORAL thickness is unimportant because it does not affect B-10 areal density. Th ickness variation of the very thin BORAL panels would very slightly change the thickness of the moderator between the fuel cell and the BORAL. For full flooding , the moderator region thickness (- 7 inches on each side) is too large for a slight change to have an impact. For a low density moderator, the effect of a slight change in moderator thickness would be offset by a change to the optimum moderator density.

The axial dimensions of the New Fuel Storage Racks are shown below in Table 4.2. Figure 4.2 gives an overview of the area. Figure 4.2 does not show the BORAL sheets positions between the cells, but it does appear the sheets are centered between the 2nd and 3rd assembly cell and between the 4th and 5th assembly cell. A sensitivity case in Section 7.4 (Table 7.8 "BORAL Radial Position") with BORAL sheets off center by 1" decreased reactivity. Therefore, centering the BORAL sheets between storage cells is conservative .

There is no BORAL monitoring program for the NFSR because normal operation of these racks is dry with no fuel or with fresh fuel. The BORAL is not subjected to radiation , heat, or a corrosive environment. There is no physical interaction of the BORAL with any other objects during fuel storage activities or when cells are empty.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 35 of 300 Figure 4.2:

2-7/8 .,

NCi

- SEE OETAIL...

Io-:ve*

164-3'8

( 13' 3/!l" I

~-

168-3'8

, 1* * - i llEf I i j I B

~l~

4 I C

!::LcVATION Table 4.2: Axial Heights of the New Fuel Storage Area KENO Model (Inches)

Parameter Height 1 Uncertainty Bottom of SS Pool Floor -1/4 [ t *c Bottom of Cell Base Plate 0 [ re Bottom of Fuel 3/8 N/A Assembly Bottom of BORAL 2 7/8 N/A 2

Bottom of Active Fuel 3.278 N/A Top of Active Fuel 147.278 N/A Top of BORAL 148 7/8 N/A Top of Cell 168 1/8 N/A 1

AII heights are relative to the bottom of the cell base plate.

2 Varies from 3.28 to 3.97 inch depending on design.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 36 of 300 4.2 Spent Fuel Pool There are three rack designs in the MPS3 SFP grouped into three areas designated Region 1, Region 2, and Region 3. Figure 4.3 shows the location of each Region in the SFP. Region 2 rack module AS has not been installed but is included in this analysis.

Figure 4.3: MPS3 SFP Rack Orientation H

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 37 of 300 4.2.1 Region 1 Region 1 racks are BORAL flux trap design. Each storage cell consists of a square stainless steel tube with a BORAL neutron absorber panel enclosed inside a stainless steel wrapper on the outer side of each of the four faces of the cell. There are 5 Region 1 7x1 O racks with 350 total storage cells. Figure 4.4 shows the rack geometry in the X-Y plane. The storage cell pitch is larger in the Y dimension than the X dimension. Figure 4.5 shows the rack geometry in the X-Z plane.

The rack design in the X direction (the direction of adjacent Region 1 racks) provides more flux trap space between adjacent storage cells (1.5 inch) than the nominal interior cell spacing (1.0 inch nominal). This eliminates the need to consider the effect of thicker BORAL wrappers on the outer cells which could otherwise modestly reduce the flux trap water region between racks (0.1 inch). An infinite lattice model conservatively represents the actual installed Region 1 rack spacing. Further, minimum rack-to-rack spacing is limited by the protrusion of the rack baseplate beyond the storage cell envelope. Therefore, seismic activity cannot result in Region 1 storage cells being closer together than in the infinite lattice model. Region 1 rack dimensions and tolerances are given in Table 4.3. Table 4.4 provides the axial alignment of key parts of the Region 1 racks. Fuel stack alignment is based on the Westinghouse RFA/NGF fuel design, however, the 148 inch BORAL envelopes the fuel stack in all MPS3 fuel designs.

A BORAL coupon monitoring program is used to confirm that the material condition of the BORAL in service remains bounded by the criticality analysis. The monitoring program description has been provided to the NRG [5,6]. Because Regions 1 and 2 racks are of the same age, manufacturer, and BORAL specification, the MPS3 coupon program includes both Regions. Region 1 and 2 analyses use the minimum B-1 O areal density ([ t Ref. 5) . Minimum as-built BORAL coupon areal density (average of measurements taken at several areas of the coupon) is [ t

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 38 of 300 Figure 4.4: MPS3 Region 1 Rack Dimensions (X-Y)

H

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 39 of 300 Figure 4.5: MPS3 Region 1 Rack Dimensions (X-Z) H

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 40 of 300 Table 4.3: Region 1 Rack Dimensions and Tolerances Item Description Dimension Tolerance 1 BORAL length 148 in. min. [ t 2 BORAL width [ r [ r 3 BORAL thickness [ r [ r BORAL position above top of 4

rack baseplate

[ t [ t Minimum as-built BORAL B-1 O 5

AD (coupons)

[ t [ t 6 Modeled BORAL B-10 AD [ t N/A***

7 BORAL sheathing thickness [ t

• Storage can height above top of 8

baseplate

[ t [ t 9 Baseplate thickness 0.75 in. N/A 10 Foot height [ r [ r 11 Cell ID 8.80 in. [ t 12 Cell pitch "y" 10.455 in. [ r 13 Cell pitch "x" 10.0 in. [ r 14 Cell wall thickness [ r [ r 15 Inner wrapper thickness [ t [ t 16 Outer wrapper thickness [ t [ r 17 Wrapper space [ r [ r 18 Flux trap width "y"** r r [ r 19 Flux trap width "x"** r r r r 20 Rack gap [ t [ t Fuel Stack position above top of 21 rack baseplate

[ t [ t 22 Fuel stack height 144 in. nominal [ re 23 BORAL extent above fuel [ t [ t 24 BORAL extent below fuel [ t [ t

• No tolerance available. Minimum tolerance is bounded by replacing sheathing on one side with water. Maximum tolerance is bounded by replacing water in the wrapper space with aluminum.

• • The flux trap tolerance is equal to the sum of the pitch and cell ID tolerances and is captured in the modeling of the individual components.

• • • Conservative low value. See Generic Letter responses for more details (Ref. 5 and 6)

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 41 of 300 Table 4.4: Region 1 Rack Axial Alignment Item Axial Position (in.)

Top of rack baseplate [ ]H Bottom of BORAL [ t Bottom of fuel stack [ t Top of fuel stack 144.00 Top of BORAL [ t Top of rack cell [ t 4.2.2 Region 2 Region 2 racks are BORAL non-flux trap design. There are two types of storage cells in the Region 2 racks. Every other cell is an actual cell (a square stainless steel tube). Actual cells are welded together at the cell corners to form a checkerboard arrangement of actual and resultant cells. Resultant cells are formed by the checkerboard spaces between actual cells.

BORAL neutron absorber panels are enclosed on the outer faces of each actual storage cell by a thin stainless steel wrapper. Stainless steel plates with enclosed BORAL are used to enclose the open side of resultant cells on each rack face .

There are 1O Region 2 racks with 754 total storage cells (including the uninstalled rack module).

Table 4.5 shows the Region 2 rack sizes. Figure 4.6 shows the rack geometry in the X-Y plane. Figure 4.7 shows the rack geometry in the X-Z plane.

The rack design includes a baseplate that extends a minimum of [ t beyond the storage cell envelope. Storage cells at the edge of each rack that adjoins another Region 2 rack create a flux trap with the two exterior BORAL panels separated by the rack-to-rack spacing. Nominal installed rack-to-rack spacing is 1 inch. The flux trap water gap cannot be less than allowed by baseplate contact [ t . which also results in an edge assembly cell pitch at the rack-to-rack boundary greater than the interior rack cell pitch. This eliminates the need to consider the effect of thicker BORAL wrappers on the outer cells and ensures that an infinite lattice model conservatively represents the actual installed Region 2 racks. It also ensures a seismic event cannot result in Region 2 storage cells being closer together than in the infinite lattice model. Region 2 rack dimensions and tolerances are given in Table 4.6. Table 4.7 provides the axial alignment of key parts of the Region 2 racks. Fuel stack alignment is based on the

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 42 of 300 Westinghouse RFA/NGF fuel design, however, the 148 inch BORAL envelopes the fuel stack in all MPS3 fuel designs.

Table 4.5: Region 2 Racks Number of Rack size Rack ID racks 5 9x9 A1-A5 1 9x10 B3 1 7x10 C1 3 7x9 E1-E3 Figure 4.6: MPS3 Region 2 Rack Dimensions (X-Y)

H

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 43 of 300 Figure 4.7: MPS3 Region 2 Rack Dimensions (X-Z)

H

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 44 of 300 Table 4.6: Region 2 Rack Dimensions and Tolerances Item Description Value Tolerance 1 BORAL lenqth 148 in. r t 2 BORAL width [ t [ r 3 BORAL thickness [ r [ r BORAL position above ]H 4

top of baseplate

[ t [

Minimum as-built 5 BORAL B-10 AD [ t [ t (coupons)

Modeled BORAL B-10 6

AD

[ t N/A**

[

7 BORAL clad thickness [ t t*

Storage can height ]H 8

above top of baseplate

[ [ t 9 Baseplate 0.75 in. N/A 10 Foot [ t [ t 11 Cell ID 8.80 in. r t 12 Cell pitch 9.017 in . [ r 13 Cell wall thickness [ t [ r 14 Inner wrapper thickness r t r t 15 Outer wrapper thickness [ r r t Wrapper space (between inside of ]H 16 wrapper and outside of

[ [ t storage cell)

• Based on filling the wrapper space with thicker than nominal cladding

• • Conservative low value. See Generic Letter responses for more details (Ref. 5 and 6)

Table 4.7: Region 2 Rack Axial Alignment Item Axial Position (in.)

Top of rack baseplate [ r Bottom of BORAL [ r Bottom of fuel stack [ r Top of fuel stack 144.00 Top of BORAL [ r Top of rack cell [ r

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 45 of 300 4.2.3 Region 3 Region 3 racks are Boraflex flux trap design . Each Region 3 rack module is made up of a 6x6 array of cells surrounded by a quarter inch stainless steel panel (referred to as a rack wall or rack shear wall). Twenty one racks with a total of 756 fuel storage locations are installed in the SFP. Figure 4.8 shows the orientation of the Region 3 racks in the SFP.

Figure 4.8: Location of Region 3 Racks in the SFP

~

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Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 46 of 300 There are Boraflex panels with wrappers on all cell walls that face another cell. No credit is taken for the Boraflex, which is modeled as water. Figure 4.9 is an X-Y plane drawing of a Region 3 cell. Figure 4.10 shows the X-Y profile of the Boraflex wrapper. Rack dimensions are shown in Table 4.8. Figure 4.11 is an axial drawing of the Region 3 spent fuel racks. Table 4.9 gives the KENO dimensions for the axial model.

Figure 4.9: X-Y Drawing of a Region 3 Rack Cell I

r

_......,_ _ _ _ _ ,+ - - - - -

• U h a1so~

CN:i10E) R.EJ:

~EE NOTe S Figure 4.10: X-Y Drawing of a Region 3 Rack Wrapper 8 <o<,, M A.X. - - - - - -

rI - - , G,O M\N - - - - - - ~ i - - o + - - 4i0 MIN TYP

'$ 0 MIN "TYP

___ T\.11!> DIRE.CT ION OWLY

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 47 of 300 Table 4.8: Region 3 Rack Model X-Y Dimensions (cm)

Parameter Dimension Tolerance

[

Cell ID 22.2250 t *c Cell Wall Thickness 0.1524 [ 1 a.c Wrapper Thickness 0.0508 [ r*c Raised Wrapper Width 19.4056 N/A 1 Flush Wrapper Width 2 X 1.1430 N/A 1 Wrapper Channel Thickness 0.2286 [ r,c:,

1 Wrapper Channel Width 19.3040 N/A Cell Pitch 26.2890 [ r,c Rack Wall (aka Shear Panel) 0.635 [ r,c2 Thickness Cell Wall to Rack Wall Separation 0.7747 :5 N/N Rack to Rack Separation 2.54 [ 1a,c" 1

The tail portion of the wrapper is not modeled in KENO . Wrapper thickness cases show less wrapper material is bounding.

2 No tolerance available. Conservative upper bound value for quarter inch thick, hot rolled steel plate is used.

3 Separation between cells at the rack module edge is important not the precise placement of the rack wall in the area between cells.

4 Minimum installed rack separation modeled. Uncertainty evaluated to address seismic event.

5 The wrapper channel thickness tolerance slightly affects the amount of water in the flux trap region of the rack, but that is unimportant to this analysis since no Boraflex is credited.

Therefore, no tolerance case will be run for the wrapper channel thickness.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 48 of 300 Figure 4.11: Axial Drawing of a Region 3 Rack Cell

- -- - - - - - - - - - 1 0 J ~ l l t f ' ltp - - - - - -

• I

• I I

~

5H ' f(~'

(2900 I 5~ 3040) ~ - -<Jto"- ~~lf~~.ft~;I

{t9001 ~IHO'IO) r I

loo ur

<;f:ClfON B

• B

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 49 of 300 Table 4.9: Region 3 Rack Model Axial Dimensions Dimension" Parameter (inches)

Bottom of Fuel Assembly 0

and Rack Shear Wall Bottom of Cell Walls 0.75 Bottom of Active Fuel 3.18 1 Bottom of Wrapper 5.25 Top of Active Fuel 147.18~

Top of Wrapper 147.37 Top of Rack Shear Wall 149.07 1

There were several different fuel assembly designs used at MPS3. Each fuel assembly design has the fuel starting at slightly different elevations. The base case uses an unrealistically short elevation and the fuel position" cases tested the highest elevation. The delta is incorporated as the "Fuel Axial Position" bias.

2 There are no uncertainties for the axial model except for the stack length. This is because "Fuel Axial Position" bias makes the model bounding.

3 AII fuel assembly designs have the same nominal stack length. See Note 1.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 50 of 300 5 Overview of the Method of Analysis 5.1 New Fuel Storage Area The criticality analysis of the New Fuel Storage Area (NFSA) is performed assuming two accident conditions, fully flooded with water and fully flooded with low density water (optimum moderation). Each storage cell of the NFSA is modeled with the most reactive fuel design at the highest enrichment (5 wt% U-235). No credit is taken for burnable absorbers. The analysis considers both centered and asymmetric placement of the assemblies in the storage cells. The analysis models the full NFSA with boundaries extended sufficiently to justify use of zero flux boundary conditions. A range of moderator temperatures from 32°F to 150°F is covered. The concrete composition , which is important to the optimum moderation cases, is conservatively determined.

A single composite fuel design is created from the range of historical fuel design features to bound new fuel storage reactivity for current and anticipated future fuel designs.

Reduced enrichment and annular pellet axial blankets are not modeled. Using higher than actual enrichment in the blanket region is conservative. The stack fuel density (combination of density, dish, and chamfer) in the model conservatively bounds the highest expected stack density for MPS3. New fuel is conservatively assumed to have no U-234 or U-236.

The analysis uses SCALE 6.0 and the ENDF/B-VII cross section library. [7] Criticality calculations are done with the CSAS5 module (KENO-V.a) . [8]

Determination of the 95/95 k-eff includes uncertainties for manufacturing tolerances of both the fuel and rack calculated at both the full and optimum moderation. In order to cover the variation in zirconium alloys, pure zirconium is used in the analysis because the alloying elements reduce reactivity.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 51 of 300 5.2 Spent Fuel Storage Racks 5.2.1 Storage Geometry Storage geometries for each Region of the SFP are listed in Table 5.1. Region 1 is composed of subregions 1A and 1B. Region 1B occupies the 2 rows of storage cells in each Region 1 rack module closest to the West SFP wall. Figure 5.1 shows the Region 1A/1 B configuration which credits neutron leakage at the SFP wall. Asymmetric positioning of fuel in the storage cells is calculated as a bias.

Table 5.1 - Storage Geometries Region Geometry Credits Description 1A 4 out of 4 None s 4. 75 wt% U-235 1A 4 out of 4 ~ 12 IFBA s 5.00 wt% U-235

~ 2 GWD/MTU ,

1A 4 out of 4 Burnup s 5.00 wt% U-235 s 5.00 wt% U-235, 1B 4 out of 4 SFP wall Region limit is s two rows of rack cells nearest the West SFP wall.

2 4 out of 4 Burnup Burnup curve , Table 13.5 No burnup credit requ ired for fuel 2 4 out of 4 Control rod enrichments 5.0 wt% U-235 One burnup curve for each Burnup, 3 4 out of 4 minimum decay time (0, 3, 9, 18, decay time and 25 years), Table 13.5

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 52 of 300 Figure 5.1: KENO Region 1 A/1 B Wall Credit Model 11111111111111 11111111111111 11111111111111 11111111111111 11111111111111 ~

Region 1A 11111111111111 11111111111111 11111111111111 11111111111111 Region 1B 11111111111111

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 53 of 300 5.2.2 Bounding Fuel Design A single composite fuel design is created from the range of historical fuel design features to bound SFP reactivity tor historical and anticipated future fuel designs. Selection of a bounding design feature such as grid volume is different tor depletion analysis and SFP k-eff calculations.

Maximum grid volume is used in the depletion analysis. SFP rack criticality calculations use minimum grid volume and a bias is added if sensitivity cases indicate maximum grid volume produced higher k-eff.

Four fuel designs have been used at MPS3. Other than small changes in axial fuel stack position, these are the features that vary among these designs:

Guide tube/ instrument tube diameter (all similar)

Clad alloy (all are Zr based)

Grid material (3 designs are Zr based, one has lnconel in the fuel region)

Grid volume (variation in grid size and number of grids)

Fuel loading (combination of density, dish, and chamfer)

Reduced enrichment axial blankets (some with annular pellets)

Burnable Poison Types (Pyrex or IFBA) lnconel grids are significant neutron absorbers and do not need to be considered tor the SFP rack model. Clad alloys and grid materials are represented in the KENO rack model as Zr, since the alloying materials are small quantities and are stronger neutron absorbers than Zr.

Grid volume is maximized in TRITON depletions (a harder neutron spectrum produces higher reactivity depleted fuel isotopic content) and minimized in the SFP rack models.

Higher fuel density increases k-eff (verified in rack model tolerance cases), so a bounding high value is used. The larger of two guide tube options (greater water displacement) is used in the TRITON model and the smaller is used in the KENO rack model. Reduced enrichment and annular pellet axial blankets are not modeled except as part of a justification argument. Using higher than actual enrichment in the blanket region is conservative.

TRITON depletions are performed using the maximum WABA loading (24 fingers), bounding fuel density, bounding high power and temperatures, and instrument thimble water displacement by an incore detector thimble. Depletion with 24 WABA bounds depletion with maximum IFBA. WABA has not been used in MPS3, but performing depletions with WABA inserted tor the entire burnup range will allow tor potential future WABA use. MPS3 burnable poison configurations not bounded by 24 WABA are identified and justified.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 54 of 300 Nominal fuel assembly dimensions are used in the TRITON depletions. Clad creep and grid expansion with depletion are neglected in the depletion because both increase the amount of water in the fuel lattice, soften the neutron spectrum during depletion, reduce Pu production, and reduce fuel reactivity in the SFP. Clad creep and grid expansion effects are included as a bias in the SFP k-eff calculations. Grid expansion used in the analysis is based on modern clad material (ZIRLO). Older fuel with Zircaloy-4 grids that potentially have more grid expansion are identified and justified.

5.2.3 Soluble Boron Credit The spent fuel storage rack analysis for Regions 1, 2, and 3 credits soluble boron. To support use of boron credit, a boron dilution analysis is provided in Attachment 7 that initiates below the Technical Specifications minimum allowable normal soluble boron concentration (2600 ppm in proposed TS). The boron dilution analysis confirms that the time needed for a dilution event to reduce the soluble boron concentration to the minimum acceptable level (600 ppm, a level at which the criticality analysis normal operation k-eff remains less than 0.95) is greater than the time needed for actions to be taken to prevent further dilution. The boron dilution accident analysis confirms that operators have sufficient time to identify, diagnose and correct the cause of the inadvertent dilution, thereby preventing SFP k-eff from exceeding the regulatory limit.

5.2.4 Burnup Credit Fuel burnup is credited in Regions 1, 2, and 3. To conservatively calculate the isotopic content of depleted fuel, these depletion variables are considered:

• Fuel and moderator temperature

• Power history

• Axial burnup profile

• Soluble boron history

• Burnable absorber

• Control rod insertion

• Fuel inserts Bounding axial node depletion temperatures are determined in part by selecting a bounding high fuel assembly average power. The assembly average power selected bounds the power history (average lifetime depletion power) experienced by MPS3 fuel at each analyzed assembly average burnup. High assembly power results in high fuel and moderator

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 55 of 300 temperatures, which hardens the neutron spectrum, increases Pu production, and increases fuel reactivity in the SFP.

Axial burnup profiles from DOE database reviewed in NUREG/CR-6801 [9] are used within their respective burnup ranges. With fuel assembly power and the axial burnup profile known, the nodal moderator temperature is determined for each of the 18 axial nodes starting from the core inlet and integrating the enthalpy added in each successive node. Data used for the axial heat balance calculation includes bounding high core power, minimum historical measured RCS coolant flow, conservatively high core bypass flow, and conservatively high HFP RCS inlet temperature. These heat balance input values result in core average RCS outlet temperature 2.5 °F higher than any MPS3 historical cycle.

The bounding high assembly power produces a maximum assembly RCS exit temperature only 6.1 °F below saturation temperature. Due to the margin to prior cycle RCS exit temperatures and the proximity to saturation temperature, it is unlikely that a future cycle would operate with moderator temperatures in an actual fuel assembly higher than the analyzed bounding assembly.

Fuel temperatures are determined for each node by adding to the nodal moderator temperature a value for fuel temperature increase above local moderator temperature. The fuel temperature increase value is a function of fuel burnup and nodal power. Fuel temperature data is from the SIMULATE model used by Dominion for its licensed fuel management analysis [10] . Fuel temperatures are pellet average, which bound the resonance effective fuel temperatures.

Bounding high power history is selected to maximize fuel and moderator temperature during depletion. However, depleting at low power produces less Pm-149, which has the effect of reducing Sm-149 peaking after fuel discharge, increasing fuel reactivity in the SFP. High power history combined with reduced power near end of life (EOL) is used to capture both effects.

Depletion is performed at constant high power conditions for all but the last 20 days of depletion. Depletion for 40 days at half the bounding power is performed at EOL instead of 20 days of depletion at the bounding power. Bounding high power level fuel and moderator temperatures are retained throughout the depletion for simplicity and conservatism.

High depletion soluble boron hardens the neutron spectrum, increases Pu production, and increases fuel reactivity in the SFP. A constant soluble boron that bounds the cycle average boron for all historical cycles (one exception) and expected future cycle average boron is used

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 56 of 300 in the TRITON depletions. All fuel depleted in the unbounded cycle were depleted in another cycle. The average boron concentration of any two cycles is lower than the boron concentration used in the TRITON depletions.

Depleted fuel content is determined for each of the 18 axial fuel nodes of a fuel assembly with independent TRITON depletions. All isotopes produced during depletion are used in the SFP rack k-eff calculations except for those of very low concentration (less than 1E-1 O at/b-cm).

With no credit for fuel decay time, isotopic content decayed for 5 days after the at-power fuel depletion maximizes fuel reactivity. Isotopic changes due to decay time are calculated within the TRITON depletions. There is a post-TRITON reduction applied to some fission gases and volatile elements to account for the possibility of migration of these isotopes away from where they were produced.

Determination of bounding assembly depletion conditions includes consideration of neutron absorbers and fuel assembly inserts that displace water. Neutron spectrum hardening by absorbers and inserts increases Pu production during depletion and fuel reactivity in the SFP.

Some inserts (WABA, control rods, source rods, etc) are mutually exclusive because they occupy guide tubes. The following absorbers and water displacers are evaluated:

• Pyrex BP (used in early cycles)

• IFBA (currently used)

• WABA (potential future use)

• Primary and secondary sources

• In-core detector thimble

• Control rod insertion (S 10 steps cycle average)

Depletion analysis assumes all fuel assemblies are depleted with 24 WABA rods and an in-core thimble inserted. WABA is depleted with the fuel and is retained throughout the entire depletion.

With the exception of some early cycle assemblies in which Pyrex BP was used, this approach conservatively bounds the depletion history of all MPS3 fuel assemblies to date. In particular, BP are removed after one cycle , only about a third of the core has in-core thimbles , and only five out of 193 fuel assemblies reside in lead control bank locations.

The analysis also evaluates some IFBA / WABA combinations and confirms which combinations are bounded by depletion with 24 WABA. Depletion with maximum IFBA / maximum source rods is demonstrated to be bounded by depletion with 24 WABA. Fuel depleted with a potentially more limiting than assumed history (two cycles had > 1O steps average lead bank control rod insertion and some Cycle 1 fuel had 24 Pyrex BP) are identified and justified.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 57 of 300 Maximum reactivity occurs after approximately 5 days decay time. The burnup credit analysis with "no" decay time are performed with a TRITON decay time of 5 days. The analyses which credit the reactivity reduction resulting from additional decay time are performed by increasing the decay time used at the end of the TRITON fuel depletions.

A bias is calculated to account for maximum horizontal burnup tilts for Regions 2 and 3.

5.2.5 Other Credit A portion of the Region 1 analysis takes credit for neutron leakage at the boundary of Region 1 and the West SFP wall. A portion of the Region 2 analysis includes control rod credit (fresh 5.0 wt% U-235 fuel with a new or discharged control rod assembly inserted).

Credit for a minimal number of IFBA rods {12) is used in the Region 1 analysis for fresh 5.0 wt%

fuel. A small number of IFBA rods (32) are also credited in the multiple mislead analysis to represent the lower reactivity fuel assemblies in a maximum reactivity fuel batch. IFBA is not credited in the determination of normal storage minimum burnup requirements (loading curves).

5.2.6 Neutron Absorbers Regions 1 and 2 take credit for BORAL neutron absorbers. A BORAL coupon monitoring program is used to confirm that the material condition of the BORAL in service remains bounded by the criticality analysis. The monitoring program description has been provided to the NRC

[5,6] Because Region 1 and 2 racks are of the same age, manufacturer, and BORAL specification, the MPS3 coupon program covers both Regions. Although the MPS3 BORAL coupons have not experienced blisters, a blister bias is calculated for Region 1 and 2 racks. No B-1 O areal density uncertainty is calculated because a bounding low value is used in the SFP k-eff calculations.

5.2.7 Accident Analysis The accident analysis is dominated by the multiple mislead. In this analysis it is assumed that 24 co-located fresh 5.0 wt% fuel assemblies are simultaneously misloaded in the most reactive Region. These assemblies are surrounded by a large number of minimally poisoned fresh 5.0 wt% fuel assemblies. With credit for the TS minimum soluble boron and including biases and uncertainties, SFP 95/95 k-eff is less than 0.95.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 58 of 300 An assembly misplacement and drop are analyzed and are considerably less limiting than the multiple mislead. The assembly drop is modeled to occur between racks and into a rack cell .

The drop is assumed to damage the grids resulting in a fuel pin pitch increase that maximizes k-eff.

For a postulated boron dilution event, it is shown that the 95/95 k-eff is much less than 0.95.

Normal operation SFP temperatures (including the associated water density change) are included as part of the normal operation rack k-eff calculations. Analysis of loss of cooling moderator temperatures (up to and including boiling) shows that in an over-temperature accident the 95/95 k-eff is much less than 0.95.

5.2.8 Normal Operations Analysis is performed for all normal operating conditions including movement, inspection, and reconstitution. All fuel bearing containers in the SFP are identified and analyzed to determine the limitations on location in the pool.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 59 of 300 6 Cross Sections, Computer Codes, and Validation 6.1 Cross Sections and Computer Codes This analysis uses the CSAS5 [8] module of SCALE 6.0 [7] to calculate SFP k-eff. Depleted fuel isotopic content is calculated using the SCALE 6.0 t5-depl TRITON [11] module which uses KENO-V.a for the flux calculations needed to collapse neutron cross sections spatially and in energy for the depletion. All analyses are performed using the 238 group ENDF/8-VII cross section library (v7-238).

Code validation is consistent with the DSS-ISG-2010 [4] and NUREG/CRs 6698 [12] and 7109

[13].

6.1.1 CSAS5 The CSAS5 module uses BONAMI to provide resonance corrected cross sections in the unresolved resonance range and WORKER, CENTRM, and PMC to provide resonance-corrected cross sections in the resolved resonance range. This is followed by KENO V.a which uses the processed cross sections to calculate the k-eff of three dimensional system models.

Most of the SFP CSAS5 computer runs use a Monte Carlo sampling of 8000 generations, 16000 neutrons per generation, and 1000 generations skipped to achieve a converged stati?tical uncertainty in k-eff of less than 0.0001.

Unless otherwise specified, all of the k-eff values reported in this document are raw calculated k-eff values with no adjustment for bias and uncertainty. The final values to be compared to the criticality criteria are the calculated values plus the total bias and uncertainty (notated as "k 95195" or 95/95 k-eff).

A uniform initial source distribution in fissile material was used for the k-eff calculations (the default option in CSAS5).

K-eff convergence of KENO cases is verified by comparing the k-eff at the midpoint of retained generations (3500 retained generations) to the final k-eff (7000 retained generations) with the difference in k-eff expressed as the number of final sigma. Nearly all cases had a final k-eff within 2 sigma of the midpoint value (approximately+/- 15 pcm). Cases which failed the screening were either rerun with additional generations and generations skipped, or were shown to be non-limiting or otherwise non applicable cases.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 60 of 300 6.1.2 Isotopes Used New fuel is conservatively assumed to have no U-234 or U-236. Table 6.1 shows the 251 isotopes in the depleted fuel content used in the criticality analysis. Depending on the burnup, some of these isotopes had insignificant atom densities and were not placed in the output file by SCALE. Isotopes with a concentration less than 1 x 10*10 at/b-cm were removed to reduce code run time. Test cases showed that reducing low concentration isotopes had an insignificant effect on the SFP k-eff calculations.

Table 6.1: Isotopes Used in the Criticality Analysis loo 0t-1 Ga n Ga 7oGe 72 Ge 73Ge 74Ge 76Ge 14 As 1';)As 14 Se rose rr se /tsse 1

\!Se tsuse ts2se 1

\!Br ts1Br tsuKr ts2Kr ts3Kr ts4Kr ts';JKr ts°Kr ts';JRb ts 0Rb tir Rb s4Sr s6Sr tsrsr tltlsr tlt-/Sr 90Sr B9y 90y 91 y 9ozr 91 zr 92zr 93Zr 94Zr 9 ';) Zr i:J6zr 93Nb 94Nb \j';)Nb \j4Mo \j';)Mo \j 0Mo 1

Mo \JtsMo \j\JMo iuuMo \!\!Tc \JtsRu \j\JRu ,vvRu 1u1 Ru 1u2Ru

,u"Ru 1u4Ru 100 Ru lU3Rh iu:,Rh lUZPd lU4Pd rnsPd 106Pd ,ur Pd lUtlPd ,10Pd 107Ag IU::/Aq ,,omAq ,,, Aq IUtlCd 11UCd 111 Cd ,, 2Cd 113Cd 114Cd 11';Jmcd 11ocd 113ln , ,';J in ll4Sn n:,sn llosn 11 1 Sn ll tlsn 11 t-1 Sn 12usn i a sn 1z3Sn 1z4Sn "'"Sn 126Sn 12, Sb 123Sb 124Sb ,2';JSb 126Sb 122Te 123Te 124Te ,2';JTe 120Te 12rmTe 12tsTe 12\JmTe 13uTe 132Te 12/1 12t-1I 13UI 131 1 1351 1zoxe 12sxe 129Xe ,;:iuxe 131 Xe 132Xe 133Xe 134Xe 13';,Xe 13oxe 133Cs 134Cs 13';,Cs 13ocs 13/Cs 132Ba 133Ba 134Ba 13:,Ba 13oBa 1<l 1 Ba 13sBa 14uBa 13tlla 13\Jla 140La 13sce 139Ce 140Ce 141 Ce 142Ce 143Ce 144Ce 141 Pr 142Pr 143Pr 142Nd 143Nd 144Nd 14:,Nd '"

0 Nd 14/Nd 14tsNd i:,uNd 141 Pm '"

0 Pm 14tim pm '""Pm i:,1 Pm 147Sm 14ssm 14ssm

,sosm ,s, Sm 1s2sm 1s3Sm 1s4Sm 1s1 Eu 1s2Eu 1';J3Eu 1s4Eu 1ssEu 1 0

';) Eu 1 1

';) Eu i'cJ2Gd 1';)3Gd 1';)4Gd l';J';)Gd 1';)0Gd 1:, 1 Gd i:,tsGd loUGd l ';J\JTb IOUTb 1';JtsDy l oUDy 1o1 Dy 102Dy 103Dy 164Dy io:,Ho 166mHo 1t,4Er ,ooEr iorEr iot1Er 170Er 2UtlPb atiTh at-1Th Z<JUTh z;:izTh

"""Th "" 'Pa """Pa """U """U Z<l4u z;:i:,u £'.JOU Z<l/ u Z<ltlu z;:i:,Np z;:ioNp Z<l/ Np Z<ltlNp Z<l::/Np z;:i 0 Pu z;:i' Pu "j"Pu ""::iPu ""uPu 241Pu 242Pu 244Pu 241Am zqzAm 242mAm 243Am """Am 241 cm """Cm 243cm 244cm 245cm 240cm 247cm 24tscm """Bk 24\JCf z:,ucf 2s1 Cf

"""Cf

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 61 of 300 The isotopic atom densities used come directly from the initial fuel content or the depletion analysis, except for an adjustment for gaseous or volatile fission products (hereafter called fission gases) . Table 6.2 lists the fission gases used in the criticality analysis. Two of the isotopes, Cs-133 and Xe-131 , are about 80% of the reactivity worth of the fission gases. The treatment of the fission gases for this criticality analysis is the same as was done fo r a recently approved Millstone Unit 2 criticality analysis [14] and the key elements of the position are repeated here.

Table 6.2: Fission Product Gases and Volatiles OUKr, 0 ~Kr, 0 "Kr, 0 "Kr, 0 °Kr, 0 °Kr Noble Gases 12sXe, 12sXe , 129Xe , 130Xe , 131Xe, 132Xe , 133Xe , 134Xe , 13sXe , 13sXe OORb , OD Rb, RbOf Alkali Metals 133Cs, 134Cs, 13scs , 13scs, 137Cs 0

"'Br, ' Br Halogens 1271, 1291, 1301, 13\ 1351 Most of the fission gases remain in the active fuel near where they were created . The mobility of fission gases is important in assessing the consequences of reactor accidents. Regulatory Guide RG 1.183 provides conservative release fractions for fission gases. [15]

It is expected that RG 1.183 will be updated to reflect higher linear powers than were used in developing the current limits. PNNL-18212 Rev. 1, "Update of Gap Release Fractions for Non-LOCA Events Utilizing the Revised ANS 5.4 Standard," which was completed in June 2011 ,

provides a new analysis. [16] Table 2.9 of PNNL-18212 provides new limiting release rates.

PNNL-18212 Appendix C also provides an example of calculated release rates when the linear power is known ("PWR Gap Inventories Based on Realistic Power Histories"). The limiting stable release linear heat rate in that example is - 9 kw/ft from BOC to 32 GWd/MTU and declines thereafter. The product of the MPS3 assembly average linear heat rate and the bounding assembly relative power (Section 8.2) is 8.8 kw/ft through 30 GWd/MTU and declines thereafter. Because the peak linear heat rate magnitude and function of burnup is similar to the bounding MPS3 power history, the Appendix C example release rates may be considered representative for MPS3. Table 6.3 shows the fission gas release fractions from the current RG

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 62 of 300 1.183, the PNNL-18212 recommended change to the RG, the PNNL-18212 plant dependent release rate from Appendix C, and the release rate assumed for the MPS3 criticality analysis.

As shown on Table 6.3, the fission gas release fraction selected is generous compared to Appendix C of PNNL-18212 and is more bounding than the current RG (note that 1-131 is low worth).

Table 6.3: Fission Gas Release Fractions Isotopes Current Bounding Appendix C Implemented in RG Proposed in of PNNL- MPS3 Criticality 1.183 PNNL-18212 18212 Analysis K-85 0.10 0.38 0.13 0.20 1-131 0.08 0.08 0.02 0.05 Other Nobles Gases 0.05 0.08 0.02 0.05 Other Halogens 0.05 0.05 0.01 0.05 Alkali Metals 0.12 0.50 0.16 0.20 The release rates for PNNL-18212 assume cladding breach. For the criticality analysis the concern is for the fission products moving into the plenum and therefore no longer acting as effective absorbers. The boiling point of Cesium (951 K) is much higher than the clad temperature during normal fuel use (boiling point of water at 2250 psia is 618 K). It is unlikely that a significant amount of Cesium can migrate away from the fuel due to the clad acting as a condensing surface. This contention is supported by two observations. It is common to use Cs-137 as a measure of burnup. The agreement between Cs-137 and Nd-148 as a burnup measurement for chemical assays has been generally good [17-21]. Also, the BNFL burnup measurement device which is based on Cs-137 has agreed well with the reactor record burnup.

[22] For this analysis it is assumed that 20% of the Cs-133 is lost, and Cs-133 is about a third of the worth of the fission gases. Therefore, the fission gas loss assumptions are considered conservative.

L__ _

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 63 of 300 6.1.3 TRITON The t5-depl sequence of SCALE's TRITON [11] enables depletion calculations to be performed by coordinating iterative calls between cross-section processing codes (CENTRM and BONAM!) , KENO-V.a, and the ORIGEN-S point-depletion code. A 20 KENO model of the fuel assembly in the core provides the flux distribution needed to collapse the cross sections spatially and in energy for the ORIGEN-S depletion calculations. All the fuel pins are treated as a single depletion material. TRITON parameter addnux=3 is selected to obtain the maximum number of depletion isotopes (Table 6.1). Other than removing light elements (less than oxygen), removing isotopes with concentration too low to affect k-eff, and reducing the concentration of volatile nuclides, the depleted fuel content is taken directly from the TRITON output.

The flux solution for each TRITON depletion step uses a KENO calculation with input requesting 3000 generations and 3000 neutrons per generation (NPG ). By default, SCALE increases the number of generations to 3599 and skips 602 generations for the final tally. Past analyses have shown that this many neutron histories or fewer provide adequate convergence [14,23). The convergence verification in Reference 23 used essentially the same fuel design used in this analysis.

It is important that the depletion time steps are short enough to assure convergence. For this analysis time steps of 10, 40, 50, 50, 50 days are used for the low burn ups. The time steps for the rest of the calculations are uniform to meet the desired burnup. The "nlib" is set so that all the remaining time steps are approximately 50 days.

To further confirm adequate convergence, a high burnup credit depletion (54 GWd/MTU) was run four ways:

• Standard convergence (3599 generations, 3000 NPG, 602 skipped)

• More histories (7199 generations, 6000 NPG , 1202 skipped)

• More skipped (3480 generations, 3000 NPG , 1082 skipped)

• Twice as many depletion steps (3599 generations, 3000 NPG , 602 skipped)

A Reg ion 3 rack k-eff KENO case was run for each of the four sets of TRITON depleted fuel content. The largest k-eff difference between the standard convergence case and the other three cases was -24 pcm (slightly lower rack k-eff) .

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 64 of 300 Fuel reactivity after discharge changes fairly rapidly in the first few days due to short half life isotopes such as Xe-135 decaying. Previous analyses performed for Millstone Unit 2 [14] and North Anna [23] confirmed that peak reactivity occurs at 5 days (k-eff is essentially plateaued from 4-7 days). MPS3 and North Anna have the same fuel design. For MPS3, 5 day cooling is selected for peak reactivity with no decay time credit.

6.2 Uncertainty in Depleted Fuel Isotopic Content The MPS3 burnup credit analysis includes an uncertainty term equal to 5% of the depletion reactivity to account for uncertainty in fuel isotopic content given a known fuel assembly burnup.

An additional uncertainty term accounts for the uncertainty in the measured fuel burnup (Section 8.15). At the time of this submittal the NRC's interim staff guidance, DSS-ISG-2010-01 [4],

permits use of a historical estimate of 5% of the depletion reactivity for the uncertainty and a zero bias. The zero bias is supported by ORNL chemical assay work [19].

Depletion reactivity uncertainty allowance of 5% has been supported based on a conservative estimate of the uncertainty of state of the art fuel management analysis computer codes. Since SCALE has not been used for fuel management, a comparison of in-rack depleted fuel reactivity was performed using fuel isotopic content from the TRITON t5-depl sequence of SCALE, CASM0-4, and CASM0-5. CASM0-4 and CASM0-5 are widely used NRC licensed state of the art fuel management depletion codes. CASM0-4 and CASM0-5 are licensed for MPS3 fuel management use [10 , 24].

Results of the TRITON / CASM0-4 I CASM0-5 comparison study were previously reported for Millstone Unit 2 [14] and North Anna. [23] Both studies show that using TRITON atom densities results in somewhat higher SFP rack k-eff than the same calculation using CASMO atom densities. The analysis performed for North Anna is particularly applicable to MPS3 because both plants use essentially the same fuel design (17x17 lattice, discrete and integral boron-based burnable absorbers).

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 65 of 300 6.3 Validation of Criticality Analysis Criticality computer codes and cross sections are validated for their ability to predict k-eff. The MPS3 criticality validation attempts to reasonably match the MPS3 racks for isotopic content, neutron energy spectrum and geometry.

Due to isotopic limitations in the critical experiments the validation is done in two steps. The first step is to use laboratory critical experiments to validate the structural materials and major actinides in a variety of geometries which produce a range of neutron spectra. The second step is to validate the minor actinides and fission products. Since there is little to no use of these latter isotopes in critical experiments, this validation is based on the uncertainty in the cross section measurement.

6.3.1 Major Actinides and Structural Materials The validation for the major actinides and structural materials follows NUREG/CR-6698 [12].

Three hundred sixty eight (368) critical experiments were selected from the OECD/NEA handbook [26] and the HTC critical experiments [27] that match the conditions of the MPS3 new fuel storage area and spent fuel pool. These experiments were analyzed with SCALE 6.0 using the 238-group ENDF/B-VII cross-section library. The resulting predicted k's were then fit for trends on the key parameters influencing k. Using these trends, the most limiting bias and uncertainty in the area of applicability was determined. Although some of the trends may not be statistically significant, it is conservative to use all of the trends in determining the limiting bias and uncertainty. Table A.6.1 is the area of applicability for the validation. The MPS3 spent fuel pool is covered by the area of applicability of the validation. Specifically,

1. Enrichment: The benchmarks selected range from 2.35 to 4.74 wt% U-235. The fuel in the spent fuel pool ranges from 2.4 to 5 wt% U-235. The bias decreases with enrichment and the slope is small allowing for a small extrapolation for higher enrichments. Although no assemblies in the pool are below 2.35 wt% U-235, analyses will be performed for enrichments as low as 1.7 wt% U-235. For an enrichment of 1.7 wt% U-235 the extrapolated bias is 0.0028 and the limiting bias is at least 0.0033. This margin makes the extrapolation acceptable.

2. Spectrum: The benchmarks cover a wide range of spectrum by varying the pin pitch.

The Energy of the Average Lethargy causing Fission (EALF) of the benchmarks ranges from 0.0605 to 0.8432 eV. This covers the range of spectrum in borated and

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 66 of 300 non-borated conditions in the spent fuel pool and the full moderated condition in the new fuel storage area. Some extrapolation is required for the optimum moderator condition in the new fuel storage area.

3. Fuel Pin Pitch: The fuel pin pitch of the benchmarks ranges from 1.075 to 2.54 cm.

The MPS3 fuel pin pitch is 1.26 cm .

4. Assembly Spacing: The benchmarks include spacing between assemblies of O to 15.4 cm of water. The maximum separation between assemblies in the flux trap racks is less than [ t .The NFSA has a separation of about 40 cm. The critical experiments cover the entire relevant range since neutron transport through > 15.4 cm of full density water has small effect on k. If the water has decreased density, then the separation effectively decreases. The optimum density of water in the NFSA is less than 20%.

5. Boron Areal Density: The critical benchmarks selected cover a range of areal density up to 0.067 gm B-1 O/cm 2 . The maximum areal density credited for the MPS3 analysis is [ t . Cd containing experiments were included to cover credited control rods.

6. Soluble Boron: The benchmarks have soluble boron concentrations up to 5030 ppm.

The soluble boron credited to meet the k less than 0.95 and credited for the accident analysis is well within the range of experiments.

Details on the area of applicabil ity can be found in Appendix A.

For the U0 2 and MOX critical experiments, the change in the k predicted for the critical experiments as a function of a parameter is performed for five different parameters; enrichment, pin diameter, pin pitch, soluble boron content, and EALF. For the MOX set a trend on the Pu enrichment was also performed. Rather than eliminate any trend, the largest bias from any of the trends is selected. The largest bias was generally from the trend as a function of EALF.

Table 6.6 provides the biases and uncertainties as a function of EALF. The EALF tabulated is the maximum allowed EALF for a given bias and uncertainty.

For the spent fuel pool, the bias and uncertainty depends on the burnup since at low burnup the dominant fissile material is U-235 and at high burnup the dominant fissile material is Pu-239. In order to avoid trying to properly weight the critical experiments for the amount of U-235 and Pu-L

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 67 of 300 239, two sets of bias and uncertainty are employed; one from the fresh U0 2 critical experiments and one from the MOX critical experiments. The final bias and uncertainty employed is that which produces the highest 95/95 k. The U0 2 critical experiments have a higher bias but lower uncertainty than the MOX experiments. Since the uncertainty is statistically combined with other uncertainties it is not possible to determine which set is more limiting until the other uncertainties due to factors such as manufacturing tolerances are determined. For the new fuel storage area or unburned fuel in the spent fuel pool only the U02 set from Table 6.6 is used.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 68 of 300 Table 6.6: Summary of Validation Bias and Uncertainty From Major Actinides and Structural Materials (Calculate with both U02 and MOX bias and uncertainty and use the set that provides the highest 95/95 k)

U02 MOX Maximum Maximum EALF Bias Uncertainty EALF Bias Uncertainty (ev) (ev) 0.35 0.0034 0.0048 0.35 0.0017 0.0083 0.40 0.0034 0.0048 0.40 0.0018 0.0088 0.45 0.0037 0.0048 0.45 0.0019 0.0094 0.50 0.0039 0.0048 0.50 0.0021 0.0099 0.55 0.0042 0.0049 0.55 0.0022 0.0104 0.60 0.0044 0.0050 0.60 0.0023 0.0110 0.65 0.0047 0.0051 0.65 0.0025 0.0115 0.70 0.0049 0.0052 0.70 0.0026 0.0121 0.75 0.0052 0.0053 0.75 0.0027 0.0126 0.80 0.0054 0.0054 0.80 0.0029 0.0132 0.85 0.0056 0.0055 0.85 0.0030 0.0137 0.90 0.0059 0.0057 0.90 0.0031 0.0143 0.95 0.0061 0.0058 0.95 0.0033 0.0148 1.00 0.0064 0.0059 1.00 0.0034 0.0154 1.05 0.0066 0.0060 1.05 0.0035 0.0159 1.10 0.0069 0.0061 1.10 0.0037 0.0165 1.15 0.0071 0.0062 1.15 0.0038 0.0170 1.20 0.0074 0.0063 1.20 0.0039 0.0176

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 69 of 300 6.3.2 Minor Actinides and Fission Products Since there are few to no critical experiments that contain some of the isotopes used in this criticality evaluation, validation is done by estimating the maximum error in k-eff due to cross section measurement uncertainty. NUREG-7109 has shown that applying a bias of 1.5% of the worth of the minor actinides and fission products conservatively accounts for both the bias and uncertainty due to the minor actinides and fission products. [13) NUREG-7109 mainly addresses the 28 highest worth isotopes but on the last sentence of page 106 it states, "An upper value of 1.5% of the worth is also applicable for SNF isotopic compositions consisting of all nuclides in the SFP configuration." NUREG-7109 limits the applicability to certain cross section sets but ENDF/B-VII used here is one of those sets. The use of the 1.5% bias is part of the NRC's transport division in ISG-8 Rev.3. [28)

The minor actinides are defined as actinides not contained in the criticality validation benchmarks. Table 6.1 lists all the isotopes used in the analysis. The major actinides are U-234, U-235, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 . U-236 is not a major actinide although it has a significant worth in spent fuel. Am-241 is treated as a major actinide because it decays from Pu-241 and is included in the MOX critical experiments.

The fission products used are listed on Table 6.1. Pb-208 is neither a fission product nor an actinide but is included in the analysis of burned fuel. Its atom density is extremely small with no real impact on the criticality analysis. It is treated as a fission product.

6.3.3 Temperature Dependence All of the critical experiments utilized in Section 6.3.1 were done at room temperature. There is one set of critical experiments which were run as a function of temperature in the range of interest for spent fuel pools. There are a couple of sets of experiments with temperatures greater than 200 C [29, 30) but LEU-COMP-THERM-046 [30) is ideal for determining a bias as a function of temperature in the range of interest. LEU-COMP-THERM-046 is not used in the set of experiments from Section 6.3.1 since in general they are at elevated temperatures and as such represent a unique set. The analysis of this temperature dependent set is detailed in Appendix A Section 8.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 70 of 300 The analysis of the only set of thermal critical experiments in the International Handbook that uses elevated temperatures has shown a small increase in the bias with temperature. This increase can be conservatively handled by a bias from room temperature (293K) of 1.7E-05 t.k/t.oc.

6.3.4 Absorbers Credited BORAL is directly credited in the NFSA and SFP analyses. Control rod credit is also used in portions of the Region 2 analyses. Two different control rod absorber materials are evaluated (Ag-In-Cd and Hf/Zr). To support credit for BORAL in the validation area of applicability, 28 International handbook experiments and 19 HTC experiments with borated plates were included in the fresh fuel and burned fuel sets, respectively. To support control rod credit, 17 International Handbook experiments and 8 HTC experiments utilizing cadmium absorber plates were included in the fresh fuel and burned fuel sets, respectively. Although not all control rod nuclides are explicitly represented in the validation, the control rod credit cases include a large amount of margin to the limit.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 71 of 300 7 Criticality Safety Analysis of the New Fuel Storage Area 7 .1 New Fuel Storage Area KENO Model The description of the New Fuel Storage Area (NFSA) is given in Section 4.1. The SCALE CSAS5 (KENO) model for the NFSA is a three dimensional model of the entire rack including the concrete walls and floor. Table 4.1 and 4.2 provides the dimensions and tolerances.

Figure 7.1 shows a cutaway view of the NFSA KENO model. This may be compared to the actual rack shown on Figures 4.1 and 4.2. Structural components are omitted, and the area above the active fuel is simplified. Cell walls are assumed to be straight rather than flared out at the top. Except for the cell wall , the area above the active fuel is assumed to be water at the same density as throughout the rack. The top and bottom nozzles are also modeled as water.

Figure 7.2 shows a portion of the NFSA KENO model from the top with water removed. Figure 7.3 is a less detailed view of the whole NFSA model.

The New Fuel Storage Vault has concrete walls and floor that is lined with a % inch stainless steel liner. Behind that liner is a concrete wal l. It is unclear how thick the walls and floor is in some places, so the KENO model conservatively extended the wall and floor thickness to 100 cm (>3 ft). Sensitivity cases confirm that thicker walls are conservative for the k-eff calculations.

Unless otherwise noted , each of the KENO cases run use 8000 generations, 16000 neutrons per generation , and skips 1000 generations. The initial source distribution of neutrons is uniform in the fuel. Void boundary conditions are used on the six sides of the model.

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Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 72 of 300 Figure 7.1: Cutaway View of the New Fuel Storage Area Model

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 73 of 300 Figure 7.2: Top View of Part of the New Fuel Storage Area Model Figure 7.3: Full New Fuel Storage Area Model

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 7 4 of 300 7 .2 Limiting Fuel Design The NFSR analysis employs a composite fuel design that bounds the range of variation in MPS3 fuel.

7.2.1 Fuel Dimensions and Materials Variation in fuel dimensions and clad materials over the history of MPS3 is summarized in Table 3.1. To bound the neutronic effect of various clad materials, clad is modeled as zirconium.

Omitting alloying elements conservatively increases k-eff.

The difference in guide tube designs in Table 3.1 is small. Tolerance cases (Section 7.4) indicate that less water displacement in the lattice increases k-eff. Therefore , the guide tube design with the least water displacement (V5H) is used in the NFSR model. For the same reason, the minimum zirconium-based grid volume design (V5H) is used.

Tolerance cases (Section 7.4) confirm that adding more stainless steel rack structure (thicker storage cell walls) to the model reduces k-eff. Therefore, NFSR cell support structure is omitted and fuel assembly hardware above and below the active fuel region is modeled as moderator.

There are normally small amounts of U-234 and U-236 in the fuel that decrease k-eff. U in U02 is conservatively assumed to be composed of only U-235 and U-238.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 75 of 300 7.2.2 Fuel Density, Burnable Poisons, and Axial Blankets Fuel density is calculated as a fraction of theoretical U02 density (10.96 g/cm 3 maximum). The calculation credits the nominal dish and chamfer volume as a reduction in net stack density in order to preserve the correct amount of fuel in the central (non-annular pellet) zone. For example, if a pellet is produced that has a material density 97 percent of theoretical density (PTO) but has 2% volume reduction due to geometric dish and chamfer, the density used in the KENO model is 97% x 98% or 95.06 PTO. Note that the missing fuel in annular blanket pellets is not credited when calculating the stack density.

Figure 7.4 shows the PTO net stack density for each fuel assembly used at MPS3 through Cycle 17. Based on the as-built fuel data, a bounding net stack density of 95.5 PTO is used in the NFSR KENO models. A fuel density uncertainty tolerance of [ t *cwill be included to provide an estimate of the reactivity variation about the average within a fuel batch due to fuel loading variations. By inspection of Figure 7.4, the historical variation in stack density of fuel in most batches is less than 1% (maximum - minimum) .

Figure 7.4: MPS3 Fuel Assembly Stack Density a,c

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 76 of 300 Burnable poisons are not credited in the NFSR analysis.

All MPS3 fuel pins have dished and chamfered pellets, some may have reduced enrichment axial blankets, and some may use some annular pellets. In the NFSR KENO models, fuel is modeled as a right circular cylinder. Ignoring annular pellets overstates the amount of fuel in the assembly. Tolerance cases in Section 7.4 (increased fuel density and increased pellet OD) confirm that this modeling simplification is conservative. Fuel in the axial blanket region is modeled as 5.0 wt% fuel. This modeling simplification is conservative because modeling higher than actual fuel enrichment increases k-eff.

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Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 77 of 300 7 .3 Limiting Rack Model 7.3.1 NFSR Materials and Dimensions The new fuel storage area has concrete walls and a concrete floor and is lined with a ~ inch stainless steel liner. The KENO model encompasses the entire area and uses nominal dimensions described in Tables 4.1 and 4.2. Storage cells are stainless steel. BORAL is modeled with the minimum certified B-10 areal density (0.005 g/cm 2). A fresh 5.0 wt% U-235 fuel assembly is modeled in each storage cell.

Due to some uncertainty about the thickness of the concrete, the KENO model conservatively assumes a thickness of 100 cm (>3 ft) thick. Thickness sensitivity cases verified that k-eff is insensitive to concrete thicknesses greater than about 70 cm. The storage cell and concrete wall height result in over 50 cm of moderator above the top of the active fuel. Moderator displacement above the fuel by the assembly top nozzle and rack structural materials is not modeled.

Previous new fuel storage criticality calculations [23] performed to determine the effect of different concrete compositions showed that "EPRI Dry Concrete" composition [31] results in the highest NFSA k-eff of the options considered. The composition of EPRI Dry Concrete is shown in Table 7.1.

Table 7.1: EPRI Dry Concrete Composition Element Concentration (wt% )

Fe 0.45 H 0.0 C 14.00 0 42.52 Na 2.32 Mg 7.53 Al 2.72 Si 26.94 Ca 3.52 Density (glee) 2.90 To verify that EPRI Dry Concrete is a bounding composition for the MPS3 NFSR, KENO k-eff calculations were run for five different concrete compositions. The four additional concrete compositions come from the SCALE manual [7]. Table 7.2 shows the results. EPRI Dry Concrete will be used in the KENO NFSR models because it produces the highest k-eff.

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Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 78 of 300 Table 7.2: k-eff of NFSR vs. Concrete Composition Case Description k-eff CJ Llk EPRI Dry Concrete 0.90678 0 .00008 N/A SCALE Maonuson 0 .90643 0 .00008 -0.00035 SCALE Oak Ridge 0 .90630 0.00008 -0.00049 SCALE Requlatorv 0 .9061 1 0.00008 -0.00067 SCALE Rocky Flats 0 .90629 0.00008 -0.00050 MPS3 fuel assembly designs have an active fuel length of 144 inches. However, different fuel assembly designs have had different bottom nozzle heights, end plug heights and gap size between the bottom nozzle and end plug. The axial location of the active fuel relative to the New Fuel Storage Rack structure will change depending on the future fuel design type . Table 7 .3 shows the distance from the bottom of the fuel assembly to the bottom of the active fuel for current and past MPS3 fuel designs.

Table 7.4 shows that the reactivity of the NFSR is insensitive to small changes to the axial position of the active fuel. The base model uses the RFA-2 fuel axial position.

Table 7.3: Axial Location of Active Fuel for Various Fuel Design Types Design Characteristics (cm)

Component VSH Batch VSH Batch Standard RFA-2 NGF 6-7 8 End Plug Height [ Ja,c [ r,c [ r *c [ r,c [ r,c Gap (rods to bottom Ja,c Ja,c r *c r *c r *c

[ [ [ [ [

nozzle) a,c r,c ]a,c r,c r *c Bottom Nozzle HeiQht r 1 r r r r Bottom of Bottom Nozzle to [ r ,c [ ]a,c [ ]a,c [ ]a,c [ r *c Bottom of Active Fuel

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 79 of 300 Table 7.4: k-eff of NFSR vs. Active Fuel Axial Location Case Description k-eff a L'lk Bottom of Active Fuel at

[ ]a,c 0.90678 0.00008 NIA Bottom of Active Fuel at

[ re 0.90671 0.00008 -0.00008 7.3.2 Temperature and Flooding Under normal storage conditions , fuel in the NFSR is stored dry and at room temperature. The maximum normal temperature in the fuel building is less than 100°F. The maximum normal operating temperature of the adjacent SFP is 150°F. Analysis of the NFSR includes flooding with un-borated water (the SFP contains borated water) and with optimum density hydrogenous moderator.

Flooding of the NFSR is not part of normal fuel storage, therefore the temperature of the moderator is not known. The freezing point of water (32°F) is a practical lower bound for NFSR temperature. Because the primary effect of water temperature change on NFSR k-eff is related to the associated water density change, water density at 32°F is set to 1.0 glcm 3 to cover the maximum density of water at atmospheric pressure. The upper NFSR temperatu re is 3

assumed to be 150°F (0.98 glcm ).

For temperatures above 68°F a k-eff bias (1.7E-05 L1k/°C) is added based on KENO code validation results (Section A.8) . At 150°F the bias is 0.00078. Base case NFSR KENO results in Table 7.5 show that higher temperatures are limiting for low density moderator, but low temperatures are limiting for NFSR flooding with water. The validation temperature bias has not been added to the Table 7.5 k-eff values.

Table 7.5: NFSR Reactivity Sensitivity to Water Density Case Description k-eff a L'lk 32°F, 1.00 Qlcc 0.90509 0.00009 NIA 150°F, 0.98 glee 0.90136 0.00008 -0.00373 32°F, 0.065 glee 0.87071 0.00007 -0.01043 150°F, 0.065 glee 0.88114 0.00007 NIA

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 80 of 300 7.3.3 Low Density Moderator Table 7.6 and Figure 7.5 show the k-eff of the NFSA over a range of moderator densities.

These cases were run at 150°F or 32°F, with the assembl ies centered in their cells , and with EPRI dry concrete. NFSR k-eff is maximized when flooded with full density water.

The full density water case bounds low density moderator cases for both temperature ranges.

The calculated k-eff with full flood ing is at least 0.02 dk higher than the highest k-eff of the low density moderator cases. In addition, the regulatory limit of the full density case is 0.03 dk lower than the low density moderator limit. Therefore , the limiting condition is the fully flooded case.

Key bias and uncertainty cases in Section 7.4 confirm that when bias and uncertainty are included, fully flooded is the limiting condition.

Table 7.6: NFSR Reactivity Sensitivity to Water Density Water Density k-eff a Ak 0.000 qlee , 150°F 0.42128 0.00004 -0.48008 0.025 glee, 150°F 0.76546 0.00007 -0.13590 0.045 qlee , 150°F 0.87124 0.00007 -0.03012 0.050 glee , 150°F 0.87764 0.00007 -0.02372 0.055 glee, 150°F 0.88061 0.00007 -0.02075 0.060 glee , 150°F 0.88114 0.00007 -0.02022 0.065 glee , 150°F 0.87949 0.00007 -0.02188 0.070 qlee, 150°F 0.87608 0.00008 -0.02528 0.075 qlee, 150°F 0.87119 0.00007 -0.03017 0.250 glee, 150°F 0.62515 0.00006 -0.27622 0.500 glee, 150°F 0.66980 0.00008 -0.23157 0.750 qlee, 150°F 0.79893 0.00008 -0.10243 0.900 glee , 150°F 0.86851 0.00008 -0.03285 0.950 qlee, 150°F 0.88978 0.00008 -0.01158 0.980 glee , 150°F 0.90136 0.00008 NIA 0.055 qlee, 32°F 0.86951 0.00007 -0.03558 0.060 glee, 32°F 0.87127 0.00007 -0.03383 0.065 qlee , 32°F 0.87071 0.00007 -0.03438 1.000 glee, 32°F 0.90509 0.00009 NIA L

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 81 of 300 Figure 7.5: k-eff of NFSR vs. Water Density 1.00 0.90 0.80

~

a, I

<t 0.70 V'l LI.

z 0.60 0.50 0.40 0.00 0.20 0.40 0.60 0.80 1.00 Water Density (g/cc)

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 82 of 300 7.3.4 Asymmetric Fuel Placement Asymmetric placement of the fuel within the storage cells is evaluated using four different eccentric configurations. Figure 7.6 shows how the assemblies were oriented for each case.

Table 7.7 shows that the highest k-eff is achieved when the assemblies are shifted towards the B or D locations (both cases have essentially the same k-eff). The NFSR KENO model has assemblies asymmetrically skewed towards the B locations shown in Figure 7.6.

Figure 7.6: Asymmetric Skewing of Assemblies X=A O =B + =C { =D

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 83 of 300 Table 7.7: k-eff of NFSR vs. Assembly Asymmetric Position Case Description k-eff 0 ~k Fuel Assembly Centered in Cell 0.90509 0.00009 -0.00169 Fuel Assemblies Shifted Towards A 0.90661 0.00008 -0.00017 Fuel Assemblies Shifted Towards B 0.90678 0.00008 N/A Fuel Assemblies Shifted Towards C 0.90671 0.00008 -0.00007 Fuel Assemblies Shifted Towards D 0.90680 0.00008 0.00001 7.3.5 Summary of the Base Case for the NFSA Analysis Sections 7.1-7.3 describe the development of the limiting MPS3 NFSR model. The base case model includes these features:

1. All storage cells contain a 5 wt% U-235 fresh fuel assembly with no burnable absorbers.

2. The fuel contains no U-234 or U-236.

3. The fuel stack density is 95.5% of the U02 theoretical density.

4. The fuel model includes minimum water displacement guide tubes and grids

5. The fuel clad material is modeled as zirconium .

6. Rack structure except the storage cells, liner, and concrete is ignored.

7. EPRI dry concrete composition is used to represent the concrete.

8. Concrete thickness is conservatively assumed to be 100 cm .

9. The minimum certified BORAL B-10 areal density is used.

10. Fuel is asymmetrically positioned in each rack cell to maximize k-eff.

11. The fuel and moderator temperature for the fully flooded condition is 273 K with water density set to 1.0 gm/cm 3 .

12. KENO cases are run with 8000 generations, 16,000 neutrons per generation, and 1000 generations skipped.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 84 of 300 7 .4 Biases and Uncertainties for the New Fuel Storage Area Analysis Table 7.8 shows the results of manufacturing uncertainty and bias cases. A full set of cases were run at the limiting water temperature and density (32°F and 1.0 g/cm3). Cases that were found to be most significant at low density moderation in a previous analysis [23] are also included in the Table for comparison (150°F and 0.06 g/cm 3) to support the conclusion that flooding with water is the most limiting condition when biases and uncertainty are included.

Table 7.8: MPS3 NFSR Bias and Uncertainty Cases Case Description k-eff a t.k Nominal, 32 F (Base case)* 0.90678 0.00008 N/A Fuel Stack PTO, Increase 0.90726 0.00008 0.00047 Pellet OD, Increase 0.90700 0.00008 0.00022 Active Fuel Lenqth, Increase 0.90654 0.00008 -0.00024 Active Fuel Length, Decrease 0.90638 0.00008 -0.00040 Clad ID, Increase 0.90625 0.00009 -0.00053 Clad ID, Decrease 0.90697 0.00008 0.00019 Clad OD, Decrease 0.90838 0.00008 0.00160 Guide Tube ID, Increase 0.90691 0.00008 0.00013 Guide Tube ID, Decrease 0.90646 0.00008 -0.00032 Guide Tube OD, Increase 0.90624 0.00008 -0.00054 Guide Tube OD, Decrease 0.90691 0.00008 0.00013 Pin Pitch, Increase 0.90727 0.00008 0.00049 Cell Wall Thickness, Increase 0.90556 0.00008 -0.00122 Cell Wall Thickness, Decrease 0.90793 0.00008 0.00115 Cell Pitch, Increase 0.90665 0.00008 -0.00013 Cell Pitch, Decrease 0.90637 0.00008 -0.00041 Rack Pitch , Decrease 0.90687 0.00008 0.00009 BORAL Radial Position 0.90655 0.00008 -0.00023 Stainless Steel Liner, Decrease 0.90668 0.00009 -0.00010 Concrete Thickness, Increase (+100 cm) 0.90684 0.00009 0.00006 BORAL Cutouts** 0.90673 0.00008 -0.00005 Nominal, 150 F, 0.060 g/ee (Base case)

• 0.88152 0.00007 N/A Cell Pitch, Decrease, 0.060 glee 0.88285 0.00007 0.00132 Cell Wall Thickness, Decrease , 0.060 g/ee 0.88700 0.00008 0.00547 Concrete Thickness, Increase, 0.060 glee 0.88259 0.00008 0.00107 Pin Pitch , Increase, 0.060 g/ee 0.88186 0.00007 0.00033 Foam Flooded Fuel Buildinq 0.060 g/ee 0.88481 0.00007 0.00329

• Base cases differ from Table 7.6 due to asymmetric fuel placement.

• • The "BORAL cutouts" case models holes in the BORAL through which rack structure passes.

The uncertainty case results for the fully flooded condition are combined in Table 7.9 using Equation 7.1 :

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 85 of 300 Total Unc. = [ ~ ( k+/-un,. - k Naminal +2' (7.1)

The "Code Benchmarking Unc." and "Code Benchmarking Bias" are described in Section A.7.

The base case has an EALF of 0.24 eV. There is no "Code Temperature Bias" for the flooded cases because the cases are run at below room temperatu re. Table 7.9 shows that the MPS3 NFSR has 0.0242 .b.k Dominion margin.

Table 7.9: MPS3 NFSR Maximum k-eff, Full Density Water Case Description k-eff a ~k* Max ak*

Base Case Nominal, 32F (Base case) 0.9068 0.0001 N/A N/A Uncertainties Fuel Stack PTO, + 0.9073 0.0001 0.0005 0.0007 Pellet OD, + 0.9070 0.0001 0.0002 0.0004 Active Fuel Length, + 0.9065 0.0001 -0.0002 0.0000 Clad ID, - 0.9070 0.0001 0.0002 0.0004 Clad OD, - 0.9084 0.0001 0.0016 0.0018 GTID, + 0.9069 0.0001 0.0001 0.0003 GT OD, - 0.9069 0.0001 0.0001 0.0003 Pin Pitch, + 0.9073 0.0001 0.0005 0.0007 Cell Wall Thickness, - 0.9079 0.0001 0.0011 0.0014 Cell Pitch , + 0.9066 0.0001 -0.0001 0.0001 Rack Pitch , - 0.9069 0.0001 0.0001 0.0003 Wall Thickness, + 0.9068 0.0001 0.0001 0.0003 Stainless Steel Liner Thickness, - 0.9067 0.0001 -0 .0001 0.0001 KENO Case Uncertainty NIA NIA NIA 0.0002 Code Benchmarking Unc. NIA NIA NIA 0.0048 RSS of Uncertainties 0.0055 Biases BORAL Cutout Bias 0.9067 0.0001 0.0000 0.0002 Code Temperature Bias NIA NIA NIA 0.0000 Code Benchmarkinq Bias NIA NIA NIA 0.0034 Sum of Biases 0.0036 Summary Base Case k-eff 0.9068 Total Bias and Uncertainty 0.0091 NRC Administrative Margin 0.0100 Maximum k-eff 0.9258 10CFR50.68 Limit 0.9500 DOMINION Margin 0.0242

• .b.k is kr k 1 . Max L1k is .b.k + 2*RSS(o1 , o2)

L

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 86 of 300 To support the conclusion that the flooded case is more limiting than the low density moderator case, an estimate of the low density moderator total bias and uncertainty is calculated using six uncertainty and three biases values. Four uncertainty cases and one bias case are shown in Table 7.8. The two remaining uncertainties are the Monte Carlo uncertainty and code benchmarking uncertainty, and the two remaining biases are both code benchmarking biases.

The six uncertainty values accounted for most of the total uncertainty of the North Anna New Fuel Storage Area [23] .

Computer code validation bias and uncertainty requires extrapolation due to the high EALF values of the low density moderator cases. The water density 0.06 g/cm 3 water density case has an EALF of 1.34 eV. The upper end of the EALF range in the benchmark cases is 0.85 eV (Appendix A) . The required extrapolation is 0.5 eV. Based on a linear trend , the code validation bias increases 0.0024 dk (to 0.0081) and the uncertainty increases 0.0012 dk (to 0.0067). Due to the large extrapolation, the code validation bias and uncertainty will be determined by doubling the amount of extrapolation. Table 7.1 O shows the margin calculation for the low density moderator condition. There is over 0.06 dk margin to the limit based on this estimate.

The low density moderator condition is not limiting.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 87 of 300 Table 7.10: Maximum k-eff for MPS3 NFSR, Optimum Moderation Case Description k-eff a ilk Max Llk Base Case Nominal, 32F (0.065 g.cc) 0.8824 0.0001 N/A N/A Base Case for Biases & 0.8815 0.0001 N/A N/A Uncertainties (0.060 q/cc)

Uncertainties Pin Pitch,+ 0.8819 0.0001 0.0003 0.0005 Cell Wall Thickness, - 0.8870 0.0001 0.0055 0.0057 Cell Pitch, - 0.8828 0.0001 0.0013 0.0015 Concrete Thickness, + 0.8826 0.0001 0.0011 0.0013 KENO Case Unc N/A N/A N/A 0.0001 Code Benchmarking Unc. N/A N/A N/A 0.0078 RSS of Uncertainties 0.0099 Biases Foam Flooded Building Bias 0.8848 0.0001 0.0033 0.0035 Code Temperature Bias N/A N/A N/A 0.0008 Code Benchmarkinq Bias N/A N/A N/A 0.0105 Sum of Biases 0.0148 Summary Base Case k-eff 0.8824 Total Bias and Uncertainty 0.0246 NRC Administrative Marqin 0.0100 Maximum k-eff 0.9170 10CFR50.68 Limit 0.9800 DOMINION Marqin 0.0630

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 88 of 300 7 .5 Accident Conditions 7.5.1 Optimum Moderation It should be noted that it would be difficult for the Millstone NFSR to be flooded with low density moderator. This is because most of the NFSR is usually covered. Regardless, the accident scenarios of flooding the NFSR with water and optimum moderator were analyzed in Section 7.4. Based on calculated margin to the k-eff limits, flooding the NFSR with water bounds flooding with foam by about 3%..1k.

7.5.2 Dropped/Misplaced Assembly A dropped/misplaced assembly would be analyzed in an air moderated NFSR. This is because of the double contingency principle which specifies that it shall require at least two unlikely, independent and concurrent events to produce a criticality accident. This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions. In addition, the Millstone NFSR has stainless steel tube covers over the whole vault which prohibits misplacing an assembly or crushing another assembly. The rack structure does not provide enough open space between storage cells for a fuel assembly to be mis-placed. The maximum opening (assuming all the NFSR covers were off) is 8 1/16 inches wide. A fuel assembly is about 8.4 inches wide. Therefore, neither the misplaced assembly nor crushed assembly accidents need to be analyzed.

7.5.3 Seismic Event During a seismic event, free standing equipment shifts around which can move fuel closer or farther apart. However, the MPS3 NFSA has no free standing equipment except for the assemblies inside the stainless steel tubes. Fuel assemblies can shift inside the stainless steel tubes, but that scenario is covered by the base case model asymmetric fuel positioning.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 89 of 300 8 Depletion Modeling and Burnup Effects This section describes the MPS3 SCALE 6.0 (TRITON) depletion models and conservative depletion conditions suitable for use in the MPS3 SFP criticality calculations.

8.1 Depletion Method Overview TRITON depletions are used to determine the isotopic content of depleted fuel for Spent Fuel Pool criticality analysis, specifically to develop burnup curves (required minimum fuel burnup as a function of initial enrichment). Performing TRITON depletions requires a SCALE model of an axial segment of a fuel assembly that includes geometry, material content, and depletion conditions (fuel temperature, moderator temperature and density, soluble boron, presence of burnable absorbers and or control rods, and depletion power).

Conservatism (maximizing SFP fuel reactivity) is incorporated via use of a bounding fuel assembly design and by choice of input depletion conditions that bound anticipated actual fuel depletion conditions. The methodology associated with determination of conservative models and conditions is consistent with that used for the recently accepted License Amendment Requests. [14, 23]

Section 6 provides information on the computer code used (t5-depl sequence of SCALE 6.0 TRITON), the cross section library (238 group ENDF/B-VII), how it was run (number of neutrons followed, time step size, isotopes followed, and cooling time), and the validation.

Each fuel assembly is modeled using 18 equal size axial nodes. For each node the burnup is the product of the assembly average burnup and the relative axial burnup distribution (Section 8.6). The axial burnup distribution also provides the relative burnup averaged axial power distribution which is used to determine depletion conditions for each node. The following parameters are included in the considerations for each depletion node:

• Assembly power (Section 8.2)

• Soluble boron (Section 8.3)

• Moderator temperature and density (Section 8.4)

• Fuel temperature (Section 8.5)

• Axial burnup shape (Section 8.6)

• Control rod insertion history (Section 8.8)

• Horizontal burnup shape (Sections 9.6.6 and 9.9.5)

• Spacer grids (Section 8.12)

• Fuel density (Section 7.2.2)

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 90 of 300

• Variation in fuel dimensions (Section 8.11 )

• In-core thimble for flux map detectors (Section 8.9)

• Burnable absorber and history (Section 8.7)

• Decay time (Section 5.2.4)

• Reduced Power Prior to Discharge (Section 8.10)

TRITON models are conservative in the sense that fuel features and depletion conditions will be selected to accommodate past, present, and expected future fuel designs and depletion history in a way that maximizes Spent Fuel Pool (SFP) k. TRITON models are best estimate in the sense that uncertainties in fuel features (such as clad OD design tolerance) are not considered.

The methodology associated with determination of conservative models and conditions is consistent with previous analyses. [14, 23]

Each of these features or conditions will be evaluated using fuel design information , core design history, and operating history. Justification for the conservatism of each feature or condition will be provided using first principles, prior evaluations, or TRITON depletion sensitivity cases.

8.2 Bounding Fuel Assembly Depletion Power Fuel and moderator temperatures depend in part on the fuel assembly power. Depletion with higher fuel and moderator temperature increases fuel reactivity in the SFP via increased Pu production resulting from a harder depletion neutron spectrum. Therefore, a bounding high power is desired for depletion.

The assembly average power for depletion to a particular fuel burnup is chosen to bound the average power an actual assembly could sustain from initial use through the fuel burnup being analyzed (the burnup averaged assembly power). Burnup averaged nodal fuel and moderator temperatures are calculated for the depletion analysis using the highest .!;!urnup ~veraged relative ~ssembly .e,ower (BARAP).

The BARAP at the end of each cycle is the accumulated assembly burnup divided by the sum of the cycle burnups for all cycles the assembly has resided in the core. For each assembly in MPS3 Cycles 1-17, the burnup at the end of each cycle divided by the accumulated cycle burnup was calculated and plotted in Figure 8.1 versus the assembly burnup. Figure 8.1 also shows a bounding a line for the BARAP as a function of burnup. Table 8.1 shows the

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 91 of 300 breakpoints and values for the bounding (high) BARAP function. This BARAP function will be used as input to calculate depletion power, fuel temperature, and moderator temperature.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 92 of 300 Figure 8.1: Bounding Burnup Averaged Relative Assembly Power 1.60 1.40 a, 1.20 - - ~ - - -

~

0 Q.

JS E 1.00 +--,--+--.;--

a, c:i:

a,

!!a, 0.80 +--+---,--

a:

"O g:,

~

§! 0.60 *1-******************t-******************~*********+*********-********

c:i:

C.

C

i

'° 0.40

• 0.20 0 .00 *

• 0 10000 20000 30000 40000 50000 60000 Assembly Burnup (MWO/MTU)

Table 8.1: Bounding Burnup Averages Relative Assembly Power versus Burnup (Linearly interpolate for intermediate points)

Burnup (MWD/MTU) RPO 0 1.44 30000 1.44 52000 1.3 60000 1

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 93 of 300 8.3 Bounding Depletion Boron Spent fuel reactivity is increased by depletion at higher soluble boron. Cycle average soluble boron concentration is calculated by trapezoidal integration of boron versus cycle burnup for cycles 1 through 17.

Figure 8.2 shows that a boron concentration of 1050 ppm bounds all two cycle averages and all but one single cycle average. Cycle 6 was a 21 month cycle intended to be a transition to 24 month cycle and is not expected to be repeated. Fuel depleted in Cycle 6 was also depleted in at least one other cycle , so depletion with 1050 ppm boron bounds all assemblies depleted at MPS3 as well as expected future cycle average boron.

Figure 8.2: MPS3 Bounding Cycle Average Soluble Boron 1200 1100 900

liJ Cydc A*1e. Pred L\omn 200 100 0

l 2 3 4 5 6 7 8 9 10 H 12 13 14 .1:. 16 11 (ydt

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 94 of 300 8.4 Bounding RCS Temperature Higher depletion moderator temperature produces more plutonium and increases burned fuel k-eff in the SFP. Calculation of the bounding high Reactor Coolant System (RCS) temperature is performed using a simple heat balance with the input variables: 1) minimum RCS flow (bounds all historical cycles), 2) burnup averaged relative assembly power (maximum from Figure 8.1 ), 3) core power (maximum of all historical cycles increased by 2% to cover potential future power uprates) , 4) a high inlet temperature, and 5) nominal RCS pressure. The RCS flow is further reduced by a high bypass flow.

Rated core power, inlet temperature, RCS flow, and bypass flow data is given on Table 8.2.

Average fuel assembly moderator exit temperature is calculated via simple heat balance to determine most limiting historical T/H conditions. Table 8.2 shows the results of the exit temperature calculation. The most limiting cycle is Cycle 16. Exit temperature is taken as the key metric for bounding RCS temperature because the reactivity of depleted PWR fuel is dominated by the top region of the fuel. The following input is used for the bounding depletion moderator temperature and density calculation :

• Core power (3725 MWth; 2% higher than current license)

• Assembly power (Section 8.2)

• RCS pressure (2250 psia)

• Minimum vessel flow of historical cycles (391358 gpm)

• Maximum core bypass flow (5.60% with Cycle 4 as the exception)

• HFP core inlet temperature (556.6 °F)

Table 8.2 shows that the RCS temperature calculation inputs described above produce a bounding core average moderator outlet temperature (2.4 °F higher than any historical cycle).

These inputs are expected to remain bounding for future cycles.

The depletion moderator temperature of each of the 18 nodes of a fuel assembly is calculated as follows. The fuel assembly average power for this calculation is BARAP times the core average assembly power. The axial power profile is the appropriate NUREG normalized burnup profile or uniform profile (Section 8.6) . With the fuel assembly power and the axial power profile known, the nodal average moderator enthalpy (average of the enthalpy at the upper and lower boundaries of the node)is determined for each of the 18 axial nodes starting from the core inlet and integrating the enthalpy added in each successive node. Moderator temperature is determined from moderator enthalpy using a pressure of 2250 psia.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 95 of 300 Table 8.2: RCS Thermal Hydraulic History for MPS3 HFP RCS Inlet Vessel Bypass Core Inlet Outlet Outlet Power Pressure Inlet Mass Cycle Density Flow Flow Flow Enthalpy Enthalpy Temp.

(MW) (psia) Temp Flow (lbm/113) (gpm) Fraction (gpm) (Btu/lbm) (Btu/lbm) (F)

(F) (lbm/s) 1 3411 2250 558.4 46.295 424937 0.056 401141 41376 557.62 635.75 615.46 2 3411 2250 558.4 46.295 413333 0.056 390186 40246 557.62 637.94 616.89 3 3411 2250 558.4 46.295 404620 0.056 381961 39398 557.62 639.67 618.02 4 3411 2250 558.4 46.295 403168 0.076 372527 38425 557.62 641 .75 619.37 5 3411 2250 558.4 46.295 391522 0.056 369597 38122 557.62 642.41 619.80 6 3411 2250 558.4 46.295 391358 0.056 369442 38106 557.62 642.45 619.82 7 3411 2250 558.4 46.295 403018 0.056 380449 39242 557.62 639.99 618.23 8 3411 2250 558.4 46.295 400841 0.056 378394 39030 557.62 640.44 618.52 9 3411 2250 558.4 46.295 403263 0.056 380680 39266 557.62 639.94 618.20 10 3411 2250 558 .7 46.275 399290 0.056 376930 38862 558.00 641.18 619.00 11 3411 2250 558 .7 46.275 398401 0.056 376091 38775 558.00 641 .36 619.12 12 3411 2250 558.7 46.275 400661 0.056 378224 38995 558.00 640.89 618.82 13 3650 2250 556.6 46.414 402113 0.056 379595 39255 555.35 643.47 620.47 14 3650 2250 556.6 46.414 400202 0.056 377791 39068 555.35 643.89 620.74 15 3650 2250 556.6 46.414 400167 0.056 377758 39065 555.35 643.90 620.75 16 3650 2250 556.6 46.414 399976 0.056 377577 39046 555.35 643.94 620.77 17 3650 2250 556.6 46.414 400553 0.052 379724 39268 555.35 643.44 620.45 18 3650 2250 556.6 46.414 402787 0.052 381842 39487 555.35 642.95 620 .1 4 Boundinq 3725 2250 556.6 46.414 391358 0.056 369442 38205 555.35 647.75 623.19 8.5 Bounding Fuel Temperature Higher depletion fuel temperature increases depleted fuel reactivity in the SFP by increasing U-238 neutron capture and production of Pu. Calculation of fuel temperature is performed on a nodal basis and is dependent on local specific power, local moderator temperature, and fuel burnup.

Depletion fuel temperatures are determined for each fuel node using the nodal moderator temperature and fuel temperature difference from the moderator temperature, which is dependent on the nodal fuel power and burnup. The fuel temperature data is taken from the SI MULATE-3 model used by Dominion for its licensed fuel management analysis. [1 O] Fuel temperatures are pellet average, which bound the resonance effective fuel temperatures. Fuel temperature data (a function of power and burnup) used for this calculation is also integrated to the burnup of interest to obtain an appropriate depletion average temperature rather than a point value at a particular burnup.

Table 8.3 shows the results of the nodal depletion condition calculation for a 20 GWd/MTU fuel assembly modeled using NUREG/CR-6801 [9] axial burnup shape 8 (Section 8.6) .

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 96 of 300 Table 8.3: Example of Nodal Depletion Conditions (20 GWd/MTU)

Specific Moderator Burnup Fuel Moderator Node Power Density (GWd/MTU) Temp. (K) Temp. (K)

(MW/MTU) (glee) 1 (Bottom) 13.4 40.7 905 566 0.741 2 20.7 63.0 1031 568 0.736 3 23.0 70.0 1082 572 0.729 4 21.9 66.6 1062 575 0.722 5 21.1 64.1 1048 579 0.715 6 21.0 63.8 1050 582 0.708 7 21.3 64.8 1059 585 0.700 8 21 .9 66.7 1075 588 0.693 9 22.4 68.3 1089 591 0.685 10 22.7 69.1 1098 595 0.676 11 22.8 69.4 1104 598 0.668 12 22.8 69.3 1106 601 0.659 13 22.6 68.8 1105 604 0.650 14 22.1 67.3 1098 606 0.641 15 21.0 63.9 1077 609 0.632 16 18.7 56.8 1037 611 0.623 17 13.4 40.7 952 613 0.616 18 (Top) 7.5 22.7 817 614 0.612

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 97 of 300 8.6 Bounding Axial Burnup Profiles DSS-ISG-01-2010 [4] provides guidance on use of axial burnup profiles. Consistent with this guidance, MPS3 SFP burnup credit analysis uses NUREG/CR-6801 [9] profiles within their respective burnup ranges as well as uniform profiles at low burnups.

Reduced enrichment axial blankets were used in new fuel batches after Cycle 2. These axial blankets are conservatively ignored. NUREG/CR-6801 concludes, "Because the end effect for assemblies with low enrichment axial blankets is typically very small or negative, this approach

[using the non-blanketed shapes for axial blanketed fuel] will bound those assemblies."

DSS-ISG-01-201 O also directs that applications for plant designs that set the limiting profiles in NUREG/CR-6801 should provide a site specific justification for the axial burnup distributions.

The NUREG/CR-6801 bounding profile for burnup group 1 (> 46 GWd/MTU) was set by MPS3 profiles. The authors of NUREG/CR-6801 note that the MPS3 profiles were from un-blanketed fuel assemblies in cycles that underwent a transition to fuel with axial blankets. The un-blanketed fuel burnup profiles were influenced by neighboring blanketed fuel. Use of the group 1 shape for MPS3 burnup credit analysis is justified for these reasons:

1) The transition shapes are represented in the NUREG/CR-6801 database.

2) The transition shapes are from cycles with natural enrichment axial blankets.

Natural enrichment blankets affect the burnup shape more than low enriched blankets, so the transition effects represented in the data can be considered bounding.

3) There has only been one transition to axial blankets at MPS3, which is represented in the data.

Specific NUREG/CR-6801 burnup profiles used for the MPS3 analysis are listed in Table 9.17 (Region 2) and Table 9.27 (Region 3).

L_

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 98 of 300

8. 7 Burnable Absorbers Two types of burnable absorber have been used at MPS3. Pyrex BP was only used for Cycles 1 (24 fingers) and 2 (8 fingers) while IFBA has been used for the remaining cycles. WABA is considered in this analysis to allow for potential future use. Although MPS3 Pyrex BP had no absorber cutback (a region near the top and bottom of the fuel assembly with no burnable absorber), cutback regions are typical for modern core designs. MPS3 IFBA has had at least 11 inches of cutback at each end.

Pyrex BP, WABA, and IFBA absorb low energy neutrons and harden the depletion neutron spectrum , increasing spent fuel reactivity. Pyrex BP and WABA also displace moderator in the guide tubes , which further hardens the depletion neutron spectrum . Region 3 KENO rack model single fuel node (node 16) test cases (Table 8.6) confirm that depletion with maximum WABA (24 rods) bounds depletion with maximum IFBA (200 rods at [ reloading) ,

even if the IFBA assembly is assumed to contain a six finger secondary source assembly for the entire depletion. MPS3 has used 4 and 6 fingered source assemblies. Source assemblies cannot be used in combination with WABA or Pyrex BP.

Table 8.6 KENO rack model cases take no credit for residual poisons or water displacement of the WABA in the SFP. Table 8.6 cases are run with 1050 ppm soluble boron during depletion and with no soluble boron in the SFP.

Table 8.6: Depletion with Different Burnable Absorbers (5.0 wt%, 54 GWD/T)

Case Description SFP k-eff a .ilk 24 WABA w/Zr grids 0.93663 0.00006 N/A 8 Pyrex w/Zr grids 0.92927 0.00006 -0.00736 24 Pyrex w/lnconel grids 0.93807 0.00006 0.00144 IFBA w/secondary sources and Zr grids 0.93342 0.00006 -0.00322 Depletion with 24 Pyrex BP results in somewhat higher SFP reactivity than depletion with 24 WABA. However, only Cycle 1 fuel (Batches 2 and 3) are affected by this issue. Batch 2 and 3 fuel is dispositioned in the Region 2 and Region 3 SFP analyses. Depletion analysis for MPS3 is performed using 24 WABA which are left in the fuel assembly for all burnups. The WABA is conservatively assumed to have no absorber cutback.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 99 of 300 MPS3 has not had an assembly with IFBA and another burnable absorber at the same time.

However, other plants have used IFBA and WABA simultaneously. Depletion with maximum WABA, wh ich has been shown to bound depletion with maximum I FBA, is the standard configuration for MPS3 burnup credit analysis. To assess the effect of IFBA and WABA in combination , test cases are provided comparing the depletion effect of maximum WABA with that of maximum IFBA with varying numbers of WABA rods.

The burnups chosen for the test cases are 20 and 54 GWD/T which represent an assembly after one cycle of operation and after three cycles of operation. The fuel enrichments (2.6 and 5.0 wt% respectively) are consistent with the Region 3 burnup credit curve (Section 9.9). A Region 3 KENO rack model is used for the comparison. For these cases , fuel is represented by Node 16 isotopic content. Node 16 is the third node from the top of the fuel and is in a neutronically important axial location for depleted fuel.

Table 8.7 shows the results of the KENO comparison cases. For both burnups evaluated, the depletion with 24 WABA results in higher SFP k-eff than depletion with maximum IFBA and 4, 8, or 12 WABA. The 20 GWd/MTU cases credit 50% of the TRITON calculated residual IFBA content (less than 1% of the initial B-10 loading). The Node 16 results are not only representative of the whole fuel assembly, but are also conservative in that the Node 16 model does not credit the effect of the IFBA or W ABA cutback.

Table 8.7: Depletion with IFBA and WABA Simultaneously (Node 16)

Case Description SFP k-eff C1 llk 24 WABA Rods, 0 IFBA Pins 0.96715 0.00006 N/A 2.60 wt%, 20 GWD/T 4 WABA Rods, 200 IFBA Pins 0.96204 0.00006 -0 .00511 2.60 wt%, 20 GWD 8 WABA Rods, 200 IFBA Pins 0.96437 0.00006 -0.00278 2.60 wt%, 20 GWD 12 WABA Rods, 200 IFBA Pins 0.96686 0.00006 -0.00028 2.60 wt%, 20 GWD 24 WABA Rods, 0 IFBA Pins 0.93663 0.00006 N/A 5.00 wt%, 54 GWD/T 0 WABA Rods, 200 IFBA Pins 0.92780 0.00006 -0.00883 5.00 wt%, 54 GWD/T 4 WABA Rods, 200 IFBA Pins 0.93027 0.00006 -0 .00636 5.00 wt%, 54 GWD/T 8 WABA Rods, 200 IFBA Pins 0.93258 0.00006 -0.00405 5.00 wt%, 54 GWD/T 12 WABA Rods, 200 IFBA Pins 0.93484 0.00005 -0.00179 5.00 wt% , 54 GWD/T

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 100 of 300 8.8 Control Rod History Depleting fuel assemblies with control rods inserted hardens the neutron spectrum and increases depleted fuel reactivity. Only fuel assemblies in lead control bank locations (5 of 193 core locations) accumulate significant rodded exposure. Control rod insertion is mutually exclusive with any other insert component (Pyrex BP, sources, WABA). Control rod dimensions are listed in Table 8.8.

Table 8.8: Control Rod Model Dimension Value Absorber Pellet 0 .0 . (cm) 0.8661 Clad 1.0. (cm) 0.874 Clad 0.0. (cm) 0.9677 End Plug Length (cm) 2.8575 Absorber Pellet Material 1 95.5 w/o Hf, 4.5 w/o Zr*

(Cycles 1 - 4)

Absorber Pellet Material 2 80 w/o Ag, 15 w/o In, 5 w/o Cd (Cycles 4 - Present)

Clad Material SS-304 Step Size (cm/step) 1.5875 Assessment of the effect of rodded depletion history is performed to determine whether the standard maximum WABA depletion bounds depletion scenarios involving control rod insertion.

In particular, fuel assemblies can have IFBA and control rod insertion.

Figure 8.5 shows the cycle average control rod insertion of lead bank (D bank) fuel assemblies and the average insertion of the remaining control rods. Except for five fuel assemblies in Cycle 1 and five fuel assemblies in Cycle 4, 10 steps average insertion (6.25 inches) bounds the most rodded fuel assemblies in each cycle. Cycle 1 and 4 lead bank assemblies are justified in the Region 3 analysis (Section 9.9.8). Average lead bank insertion has been well below 1O steps for 10 cycles and is expected to remain so in future cycles.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 101 of 300 Figure 8.5: Cycle Average Control Rod Insertion 30 _ _ _ _,i_ j _____

20 Ill! D-Ban k li!l AII Banks 5

4 5 6 7 8 9 10 11 12 13 14 15 16 Cycle Table 8.9 contains results of Region 3 KENO rack model cases combining control rod and IFBA (200 rods) depletion history. MPS3 IFBA has had an un-poisoned cutback region of at least 11 inches on the top and bottom of the active fuel. For Table 8.9 cases, the depletion is performed with IFBA in nodes 2-17 but not in Nodes 1 and 18 (1 node= 8 inches). Sensitivity cases are provided at 20 and 54 GWd/MTU. WABA and control rods are assumed to remain inserted throughout the depletion.

Table 8.9 results illustrate several points:

• Depleting with a Ag-In-Cd control rod is slightly more limiting than a Hf control rod.

• Depleting with 2 nodes of control rod insertion is more limiting than depleting with full length WABA.

• Depleting with full length W ABA bounds depleting with 200 I FBA pins plus a control rod insertion of 1O steps at both 20 and 54 GWD/MTU.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 102 of 300 Table 8.9: Control Rod Use Sensitivity Cases Enrich/ Reg 3 Case Description a ~k BU k-eff 2.60 wt%

Maximum WABA 0.96300 0.00006 N/A 20 GWDff Ag-In-Cd Control Rod, 2.60 wt%

0.96150 0.00006 -0.00150 1O Step* Inserted & 200 IFBA 20 GWDff 5.00 wt%

Maximum WABA 0.95081 0.00006 N/A 54 GWDff Ag-In-Cd Control Rod, 5.00 wt%

0.95090 0.00007 0.00009 1 Node Inserted & O IFBA 54 GWDff Hafnium Control Rod, 5.00 wt%

0.95011 0.00007 -0.00070 1 Node Insertion & O IFBA 54 GWDff Hafnium Control Rod, 5.00 wt%

0.97087 0.00007 0.02006 2 Node Inserted & O IFBA 54 GWDff Ag-In-Cd Control Rod, 5.00 wt%

0.95070 0.00006 -0.00011 10 Step* Inserted & 200 IFBA 54 GWDff

• 10 step cases have 19 axial nodes. The bottom 17 nodes are the modeled with the standard methodology. The top node is split up into a rodded section (6.25 inches) and an un-rodded section (1.75 inches) .

These cases confirm that the standard full length maximum W ASA depletion is acceptable for modeling fuel assemblies, with the exception of lead bank assemblies in Cycles 1 and 4.

8.9 In-core Thimble MPS3 has in-core instrument thimbles inserted in the fuel assembly center instrument tube in 58 of 193 assembly locations. The in-core thimble is a hollow, thin wall tube that displaces water and hardens the neutron spectrum during depletion and increase k-eff in the SFP. Table 8.1 O shows the relevant dimension.s of the in-core thimble and instrument tube. The in-core thimble is included in the TRITON models as part of the standard depletion. This is conservative because the instrument thimble is present in only 58 of 193 fuel assemblies in each core.

Table 8.10: In-Core Thimble and Instrument Tube Dimensions Dimension Value I.T. Inner Diameter (cm) 1.123 In-Core Thimble Housing 1.0. (cm) 0.5105 In-Core Thimble Housing 0.0. (cm) 0.7595 In-Core Thimble Housing Material ASME Type 316L SS

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 103 of 300 8.1 O Reduced Power before Storage Bounding high power history is used to maximize spent fuel reactivity. Higher power increases fuel and moderator temperature during depletion. However, depleting at low power produces less Pm-149, which has the effect of reducing Sm-149 peaking after fuel discharge. Reducing Sm-149 in SFP fuel increases k-eff. Therefore , high power history combined with reduced power near end of life (EOL) can result in higher k-eff in the SFP. Pm-149 has a 2.2 day half life, so reduced power for at least several days at EOL is required for substantial reduction in SM-149 peaking.

There are two plausible ways to achieve low power near EOL for sufficient duration: (1) Fuel placement in a low power location during the last cycle, and (2) power coastdown. A final cycle at low power would reduce fuel assembly burnup average moderator and fuel temperature substantially, which would tend to reduce fuel reactivity. Power coastdown would have far less impact on burnup average moderator and fuel temperature because of the relatively short duration. Power coastdown below 50% power is not likely due to economic considerations.

Simulation of power coastdown is performed by depleting at high power conditions for all but the last 20 days of depletion. Twenty days of operation is more than sufficient for Pm-149 to reach a reduced equilibrium level. For the last 20 days, depletion proceeds at 50% of the prior power (40 days at reduced power) retaining high power fuel and moderator temperatures for simplicity and conservatism. Table 8.11 shows the typical effect that reduced power near EOG has on depleted fuel reactivity. The cases shown use Node 16 isotopic content to represent the fuel in a Region 3 KENO rack model.

Table 8.11: Low Power EOC Effect on SFP Reactivity (5.0 w/o, 54 GWD/T)

Case Description SFP k-eff a ~k Low Power EOL 0.93663 0.00006 N/A Nominal Power EOL 0.93360 0.00006 -0.00303

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 104 of 300 8.11 Grid Growth and Clad Creep Depletion Effects Zircaloy based grids can experience irradiation growth with increasing fuel burnup. Grid growth can increase fuel pin pitch. Because fuel pins are constrained within the grid lattice, fuel pin pitch growth is assumed to be the same as grid growth. Fuel clad diameter tends to decline due to creep during fuel depletion prior to increasing again at high fuel burnup.

Expanding the pin pitch and reducing clad diameter during fuel depletion would soften the depletion neutron spectrum. To simplify the depletion model, grid expansion and clad creep down are conservatively ignored in the TRITON depletions but are analyzed in the spent fuel pool KENO models.

8.12 Grids Grids displace water and harden the neutron spectrum during depletion, increasing k-eff in the SFP. In the SFP, grids displace water and may decrease k-eff, depending on the boron concentration. These effects are confirmed using a Region 3 SFP single fuel node KENO rack model. In the TRITON depletion model and in the KENO rack model, Zr based grids are represented by homogenizing the grid and the water in the fuel lattice over the length of the fuel.

Water inside the guide tubes are modeled as pure water. The total grid volume and water displacement is conserved.

Table 8.12 shows that with no soluble boron it is conservative to use the minimum Zr grid volume in the spent fuel pool KENO model. Table 8.13 shows that it is conservative to use maximum Zr grid volume in the TRITON depletion model.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 105 of 300 Table 8.12: SFP Grid Modeling Effect (Node 16, Fresh Fuel, 0 ppm soluble boron)

Case Description SFP k-eff C1 .6k 5.00 w/o, O GWD/T, 1.25414 0.00007 N/A Minimum Zr Grid 5.00 w/o, 0 GWD/T, 1.25328 0.00007 -0.00085 Maximum Zr Grid Table 8.13: Effect of Depleting with Grids (Node 16, 0 ppm soluble boron)

Case Description* SFP k-eff C1 .6k 5.00 w/o, 54 GWD/T, 0.93561 0.00006 -0.00102 Min. Zr Grid Depletion 5.00 w/o, 54 GWD/T, 0.93663 0.00006 N/A Max. Zr Grid Depletion

• Both cases have minimum Zr grids in the KENO rack model.

8.13 Instrument and Guide Tube Design The difference in guide tube and instrument tube dimensions between MPS3 fuel designs is small. Table 8.14 shows the instrument and guide tube dimensions for each fuel type. The RFA-2 guide thimble displaces the most water and will tend to increase spent fuel reactivity during depletion due to spectrum hardening. Therefore, the RFA-2 instrument and guide tube dimensions are used in the TRITON depletion model.

The VSH guide thimble displaces the least amount of water. For the pool model, guide tubes that displace the least amount of water will increase the reactivity because the fuel lattice is under moderated. Therefore, the Region 1, 2, and 3 pool models use the V5H guide tube dimensions.

Table 8.14: Instrument and Guide Tube Dimensions Cross Section Area Fuel Type Inner Diameter (cm) Outer Diameter (cm)

(cm 2)

Standard 1.143 1.224 0.1506 V5H 1.123 1.204 0.1480 RFA-2 1.123 1.224 0.1862

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 106 of 300 8.14 Comparison of CASMO and TRITON Depletion Reactivity The information in this section was previously submitted for NRC review in a North Anna License Amendment Request [23] and is copied below. The analysis was not rerun for MPS3 because the North Anna and MPS3 analyses use the same fuel design and computer codes.

A depletion reactivity uncertainty of 5% of depletion worth (k of fresh fuel - k of depleted fuel; no credit for burnable poison worth) has been accepted based on a conservative estimate of the state of the art of fuel management analysis computer codes. Since SCALE has not been used for fuel management, a study has been performed to compare the delta k depletion predicted using the TRITON t5-depl sequence of SCALE to CASM0-4 [1 O], as well as CASM0-5. [24]

The study shows that using TRITON atom densities is more conservative than using CASMO atom densities. Similar results were found when CASMO was compared to TRITON for Millstone 2. [14] Section 6.2 describes the analysis.

To confirm that TRITON is acceptable for depletion analysis for North Anna, representative comparisons are provided for North Anna (17x17 fuel) Region 2 (all cells loaded) fuel rack k using TRITON, CASM0-4 and CASM0-5 to generate depleted fuel isotopic content. The same conservative depletion conditions were used in all 3 codes, including 24 BPRA, 1100 ppm soluble boron, and high moderator and fuel temperature. A single node (Node 15; nodes are numbered from 1 at the bottom of fuel to 18 at the top) was used as a reasonable representation of the fuel for this comparison. No grids were used in the depletion models for simplicity. Two enrichment and burnup combinations representing a low and high burnup credit requirement for Region 2 are modeled.

For CASM0-4, isotopes common to both CASM0-4 and TRITON (49 nuclides) were used in the KENO rack models. CASM0-5 has no lumped fission products, so all nuclides available in the SCALE standard composition library were retained. All depletions end with 5 days decay after shutdown.

Results of the KENO Region 2 rack k cases are provided in Table 8.15. For 2.45 w/o fuel depleted to 1O GWd/MTU and 49 nuclides, the TRITON depletion produces a SFP rack k approximately 0.008 D-k higher than CASM0-4, and 0.0018 D-k higher than the CASM0-5 depletion. With all available nuclides included, the TRITON case rack k is higher than the CASM0-5 case by 0.00035 D-k.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 107 of 300 For 5.0 w/o fuel depleted to 44 GWd/MTU and 49 nuclides, the TRITON depletion produces rack k approximately 0.015 ~k higher than the equivalent CASM0-4 depletion. With all available nuclides included, the TRITON case rack k is higher than the CASM0-5 case by 0.0014 ~k.

Table 8.15 results show that depletion with TRITON produces the highest rack k as compared to depletion with CASM0-4 and CASM0-5. TRITON results are much closer to CASM0-5 results, probably because CASM0-5 and TRITON use ENDF/B-VII cross sections and CASM0-4 uses an earlier cross section set. Use of the older CASM0-4 cross section data to produce isotopic content that is then used in a KENO rack model with newer (ENDF/B-VII) cross section data creates a potential mismatch that may explain some of the large difference in rack k results between TRITON and CASM0-4. There is no cross section difference present in the TRITON and KENO models used for the MPS3 analysis.

Table 8.15: Comparison of CASMO and TRITON Depletion Worth Number Monte Enrich. of Burnup Depletion Calculated Carlo Burnup (U235 w/o) Nuclides (GWd/MTU) Code Rack k Sigma Worth (~k) 2.45 N/A 0 N/A 1.025768 0.000059 N/A 2.45 49 10 TRITON 0.957684 0.000056 0.0681 2.45 49 10 CASM04 0.949563 0.000053 0.0762 2.45 49 10 CASM05 0.955930 0.000054 0.0698 2.45 iii... ALL 10 TRITON 0.951871 0.000053 . . . +, 0.0139 2.45 !!!';* ALL* 1Q <R; CASMOS d.951523 0.000056 11: o:*0742 5.0 N/A 0 N/A 1.190491 0.000064 N/A 5.0 49 44 TRITON 0.930972 0.000055 0.2595 5.0 49 44 CASM04 0.916284 0.000053 0.2742 5.0 ALL 44 TRITON 0.914146 0.000052 0.2763 5.0 ALL* 44 CASM05 0.912796 0.000054 0.2777

• Some minor nuclides not in SCALE 6.0 library

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 108 of 300 8.15 TRITON Depletion Model Summary Table 8.16 summarizes the key elements of the MPS3 TRITON depletion model.

Table 8.16: Depletion Parameters for TRITON Depletion Model Parameter Value Reference/Basis Assembly pitch 21 .50 cm Section 3.1 Pellet diameter 0.8192 cm Section 3.1 Clad inner diameter 0.8356 cm Section 3.1 Clad outer diameter 0.95 cm Section 3.1 Clad material Zirconium Section 3.1 Fuel rod pitch 1.2598 cm Section 3.1 Guide tube inner diameter 1.123 cm Section 3.1, 8.13 Guide tube outer diameter 1.224 cm Section 3.1, 8.13 Grid volume fraction (lattice water displacement outside fuel rods and guide 1.7% Section 3.1, 8.12 tubes)

Fuel pellet net density 95.5% Section 3.1 Burnable absorber 24 WABA Section 8.7 Inner clad I.D. 0.572 cm Section 3.2.1 Inner clad O.D. 0.6782 cm Section 3.2.1 BP I.D. 0.706 cm Section 3.2.1 BPO.D. 0.808 cm Section 3.2.1 Outer clad I. D. 0.8357 cm Section 3.2.1 Outer clad O.D. 0.968 cm Section 3.2.1 Clad material Zirconium Section 3.2.1 Absorber loading [ re Section 3.2.1 Absorber material B4C Section 3.2.1 Absorber material density [ t*c Section 3.2.1 Bounding assembly average depletion power Burnup Section 8.2 BARAP (MWD/MTU) 0 1.44 30000 1.44 52000 1.3 60000 1 RCS boron concentration 1050 ppm Section 8.3 Core average thermal power 3725 MW Section 8.4 Core inlet temperature 556.6°F Section 8.4 RCS pressure 2250 psia Section 8.4 Vessel flow 391358 gpm Section 8.4 Bypass flow fraction 0.056 Section 8.4 Bounding moderator temperature Varies Section 8.4 Bounding fuel temperature Varies Section 8.5

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 109 of 300 Axial burnup shapes Uniform and NUREG/CR-6801 shapes Section 8.6 In-core thimble material 316L Stainless Steel Section 8.9 In-core thimble I.D. 0.5105 cm Section 8.9 In-core thimble O.D. 0.7595 cm Section 8.9 Decay time 5 days or decay time credit Section 5.2.4,5.2.5,6.1 Reduced power at EOL 50% reduction for last 40 days Section 8.10 Volatile fission products Select nuclides reduced Section 6.1.2

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 11 O of 300 9 Spent Fuel Rack Analysis 9.1 Region 1 Analysis 9.1.1 Region 1 Fuel Storage A fuel assembly may be placed in Region 1 if it meets one of the following criteria:

1) The enrichment is :5 4.75 wt% U-235.

2) The enrichment is :5 5.0 wt% U-235 and the fuel assembly contains 12 or more IFBA rods.

3) The enrichment is :5 5.0 wt% U-235 and the assembly average burnup is~ 2 GWd/MTU.

4) The enrichment is :5 5.0 wt% U-235 and the assembly is stored in one of the two rows of rack cells adjacent to the West SFP wall.

Criteria 1-3 are confirmed by analysis using an infinite lattice 6x6 rack cell model. Criterion 4 analysis uses a finite model described in Section 9.3.

9.1.2 Region 1 Modeling Assumptions These modeling simplifications and assumptions are used in the Region 1 analysis:

a) The plenum, guide tubes above and below the active fuel , and top and bottom nozzles are modeled as water. This practice was investigated as part of the Millstone Unit 2 criticality analysis in Ref. 14 Section 2.6 and found to be acceptable.

b) The storage cell is modeled at its nominal length. The rack structure below the bottom of the storage cell tube is modeled as water. A water reflector extends above the storage tube at the top of the model.

c) Neither reduction in fuel enrichment nor annular pellets are modeled for axial blankets.

This is conservative because it increases SFP k-eff.

d) BORAL is conservatively modeled as the same length as fuel. Table 4.4 shows that using nominal dimensions, the BORAL (using minimum BORAL length) extends well below the fuel in current and prior fuel designs and extends slightly above the fuel in current and prior fuel designs. This obviates the need for fuel axial position tolerance and BORAL length tolerance .

e) The BORAL wrapper length is modeled equal to the active fuel stack length. BORAL wrappers actually extend beyond the upper and lower boundaries of the fuel stack.

Since the region above and below the fuel is insensitive (item b above) and the un-modeled BORAL is a strong neutron absorber, there is no need for a wrapper length tolerance.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 111 of 300 f) Zr-based clad materials are modeled as Zr.

g) BORAL clad thickness is approximated. BORAL clad thickness for thickness tolerance cases is conservatively assumed to be O (no clad) and thick enough to fill the design wrapper space (maximum).

h) Water density is determined assuming the pressure is 15 psia.

i) Grids are represented by homogenizing the grid and the water in the fuel lattice over the length of the fuel. Water inside the guide tubes are modeled as pure water.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 112 of 300 9.2 Region 1 Infinite Lattice KENO Model Drawings and dimensions of Region 1 fuel racks are provided in Section 4.2.1. The primary KENO Region 1 model is a 6x6 model with reflective axial and periodic X-Y boundary conditions. The actual rack baseplate protrudes past the outer cell walls such that storage cell pitch across the rack to rack interface is larger than the normal rack cell pitch. Figure 9.1 shows a 2-D representation of the Region 1 model. Figure 9.2 shows a 3-D cutaway representation of the Region 1 model. Fuel is stored in all cells of the infinite lattice model.

A second Region 1 model used to confirm that neutron leakage at the SFP wall is sufficient to support storage of un-poisoned 5.0 wt% fuel is detailed in Section 9.3.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 113 of 300 Figure 9.2: 30 Representation of Region 1 KENO Model (full model and cutaway)

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 114 of 300 9.2.1 BORAL Blisters BORAL blisters have been observed in surveillance coupons and in rack cells [34 , 35]. The size of blisters (depth and diameter) appears to be a partial function of manufacturer (particularly manufacture date). No blisters have been observed in the MPS3 coupons .

To approximate the effect on rack k-eff of possible future blistering of BORAL in the Region 1 racks, full panel blisters are modeled in the KENO rack model by replacing the water surrounding the BORAL in the wrapper space (nominally [ t ,approximately the thickness of BORAL clad) with void. This approach is conservative because full panel blisters have not been observed and because no blisters have been observed on the MP3 coupons.

The modeled blister thickness is only about 20% of that found in some coupons [36] , however, the blister model covers the entire surface of all BORAL panels and the BORAL sheathing tends to limit blister thickness.

9.2.2 SFP Normal Operation Water Temperature and Density Rack k-eff calculations are performed at four water temperatures (Table 9.1). Water density is calculated assuming atmospheric pressure.

Table 9.1: SFP Normal Operation Water Temperature and Density Temp. K Temp. F Density 273.15 32 1.000 293.15 68 0.998 316.48 110 0.991 338.71 150 0.980

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 115 of 300 9.2.3 Simplified TRITON Depletion Input for Burnup Credit Because the maximum Region 1 burnup credit is very small (<2 GWd/MTU}, a uniform axial burnup shape is appropriate. The small burnup requirement also makes nodal variations in depleted fuel isotopic content unimportant. To simplify the Region 1 model, a single fuel composition for the entire fuel stack is used. To ensure conservatism, the TRITON depletion uses the highest nodal moderator temperature (lowest moderator density) and highest nodal fuel temperature of the 18 nodes. Depletion soluble boron is arbitrarily increased to 2000 ppm to bound the at-power boron concentration near beginning of cycle.

The following non-standard TRITON input is used:

Moderator temp.: 613.5 K Fuel temperature: 1104 K Moderator density: 0.6145 glee Specific power: 60.9 MW/MTU Depletion days: 32.8 Soluble Boron: 2000 ppm Depletion steps for this very short depletion are reduced to a maximum of 9.4 days to more accurately capture relatively rapid isotopic changes in fuel content. A 5 day decay step is included.

9.2.4 IFBA Credit IFBA credit cases include 12 fuel rods with IFBA coating. A 12 inch IFBA cutback region is modeled at the top and bottom of the fuel. IFBA B-1 O content is reduced 10% below the lower of the two standard IFBA product loadings (Table 3.4) for conservatism. The minimum number of IFBA per assembly for fuel containing any IFBA at MPS3 is 32.. Although no standard lattice pattern exists for 12 IFBA, IFBA pins were modeled in a symmetric pattern adjacent to guide tubes as is typical of other IFBA patterns.

9.2.5 Asymmetric Fuel Placement Asymmetric placement of fuel in the storage cell can increase k-eff. The Region 1 analysis determines the effect of asymmetric placement by co-locating 16 fuel assemblies toward the center of the model into the corner of each storage cell in the 4x4 central region of the 6x6 Region 1 model. Figure 9.3 shows the orientation. Test cases showed that fuel assemblies centered in their storage cell maximizes k-eff.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 116 of 300 Figure 9.3: Asymmetric Fuel Placement (4x4 in a 6x6 model)

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< * * ' ~

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Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 117 of 300 9.2.6 Code Validation Bias and Uncertainty Validation of SCALE/KENO using ENDF/B-VII 238 group cross sections identified EALF as the primary bias and uncertainty trend. Region 1 KENO cases have EALF ::. 0.35 eV. The applicable bias and uncertainty for fresh fuel and depleted fuel (MOX validation cases) are shown in Table 9.2.

Region 1 rack k-eff cases at normal operating temperatures show that temperatures below room temperature are clearly limiting, therefore no code temperature bias is necessary.

To cover gaps in the validation for minor actinides and fission products, a bias of 1.5% of the minor actinide and fission product worth is added. [13]

Table 9.2: Validation Bias and Uncertainty (EALF ~ 0.35 eV, Appendix A)

Validation cases Bias (dk) Uncertainty (dk)

U02 0.0034 0.0048 MOX* 0.0018 0.0088

• The MOX values are for EALF ::. 0.4 eV and are conservative for Region 1 analysis.

9.2.7 Measured Burnup Uncertainty Measured fuel burnup and as-built fuel enrichment are used to qualify fuel for storage when burnup credit is required. Measured burnup uncertainty is included in the development of the burnup curves. Fuel burnup is the time integral of specific fuel power. Measured burnup uncertainty includes these effects:

1) Assembly relative power measurement uncertainty (flux maps)

2) Uncertainty for unmonitored assemblies (flux maps)

3) Calorimetric power uncertainty (core power)

4) Fuel assembly loading uncertainty (MTU per assembly)

5) Uncertainty from time integration of measured power (adequacy of the number of power vs time points).

A conservative estimate of assembly relative power measurement uncertainty is the peak pin power nuclear uncertainty factor. For MPS3 a value of 2.8% was determined using the CASM0-4 and SIMULATE-3 core design models. This value is conservative as an estimate for measured fuel assembly power uncertainty because it includes the effect of assembly power prediction uncertainty, which is not a component of measured burnup uncertainty. It is also conservative in that it includes peak pin predictive uncertainty, which is not a component of

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 118 of 300 measured burnup uncertainty. However, this value is only applicable to monitored fuel assemblies, roughly 30% of fuel assemblies in any given cycle.

The additional uncertainty for unmonitored assemblies is calculated using MPS3 flux maps to determine the change in measured assembly relative power caused by changing each unique monitored location from monitored to non-monitored one assembly at a time . Based on 681 measured test cases, the difference between non-monitored and monitored relative power is 0.15% +/- 1.22% with a -1.99% lower 95/95 tolerance limit. Therefore, the uncertainty of the flux map's unmonitored locations will be 2.0%.

MPS3 calorimetric uncertainty is +/-1 .7% RTP at present. In the event of a measurement uncertainty recapture uprate, it would be substantially lower. Fuel assembly loading variation within each batch (MTU per assembly) is too small to affect the total uncertainty if included by RSS. For example, for MP3 cycle 18 the maximum U02 weight for any fuel assembly is 1069 lbs and the minimum is 1063 lbs. This represents total variation of+/- 0.3% about the average.

Regarding time integration uncertainty, power distribution measurement is performed at least monthly (nominally 18 times per cycle). Fuel assembly relative power changes gradually through the cycle. Integration of the assembly power over burnup using an 18 point approximation assuming constant power between points introduces relatively little uncertainty.

To evaluate the integration uncertainty, 18 point integrated burnups were calculated for 193 fuel assemblies in the Cycle 15 core using SIMULATE-3 predicted assembly power. Integrated burnups were then compared to actual SIMULATE-3 burnups. The average burnup increase error over a cycle is 0.03%. The standard deviation of burnup increase differences is 0.24%.

Based on this calculation, a burnup integration uncertainty allowance of 0.5% is used for integration uncertainty.

These uncertainty contributors are assumed to be independent and are combined by root sum square to obtain a burnup measurement uncertainty estimate. The RSS of 2.8%, 2.0%, 1.7%,

0.3%, and 0.5% is 3.9%. A measured burnup uncertainty of 4% is used based on this assessment of uncertainty contributors.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 119 of 300 9.2.8 Summary of Bias and Uncertainty Table 9.3 summarizes the biases and uncertainties incorporated in the Region 1 analysis.

Consideration of early cycle Pyrex BP is not needed because all early cycle fuel has much lower enrichment and much higher burnup than the small amount of burnup credit required for Region

1. Consideration of early cycle depletion with more than 1O steps of D-bank insertion is not needed because all early cycle fuel has lower enrichment and much higher burnup than the small amount of burnup credit required for Region 1. A flux trap tolerance is not directly modeled but is a result of the combined tolerances for cell ID and cell pitch, which are modeled individually. No burnup tilt, clad creep or grid growth cases are run due to the low burnup credit burnup.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 120 of 300 Table 9.3: Summary of Region 1 Biases, Uncertainties, and Conservatism I An--"-a-ly-s-is--+r

• B=bl~~~~~~~ri~, ~ 00. Ex~~ana..ti_o_:::

Depletion Burnable absorbers C Boundinq BPRA (24 WABA bounds IFBA)

Depletion Soluble boron C 2000 ppm is arbitrarily hiQh Depletion Fuel and Water C Single node with bounding high power, fuel Temperature temp., and moderator temperature Depletion Specific power C Includes all core power uprates, bounding high power assembly history Depletion Axial burnup shapes C Uniform (single node due to low burnup)

Depletion Grids C Maximum Zr grids Depletion Volatile fission products C Reduced based on qap release fractions method Depletion Fuel density C 95.5% net bounds all fuel batches Depletion Power history C Bounding high constant specific power. No coastdown due to very low burnup credit Depletion lncore thimble C Included in base depletion Rack k-eff Fuel density U 95.5% f 1a,c net bounds all fuel batches Rack k-eff Grids (base case) C Minimum volume Zr grids Rack k-eff Boral B-10 C Minimum [

Rack k-eff Boral width Rack k-eff Boral thickness u r Rack k-eff Boral wrapper thickness u r Rack k-eff Boral wrapper width u [ r (welded end tab removed)

Rack k-eff Fuel pin pitch U 0.496 [

Rack k-eff Fuel pellet OD U 0.3225 [

Rack k-eff Fuel clad ID U 0.329 [

Rack k-eff Fuel clad OD U 0.374 [

Rack k-eff Guide tube ID u 0.442 r Rack k-eff Guide tube OD U 0.474 [

Rack k-eff Fuel stack heiqht Rack k-eff Burnup worth U 5% of burnup worth Rack k-eff Measured burnup U 4% of burnup worth (Section 9.2.7)

Rack k-eff Enrichment U [ ]a,c Rack k-eff Cell wall thickness u [ r Rack k-eff Rack cell pitch u [ r Rack k-eff Rack cell ID u r r Rack k-eff Code uncertainty u Section 9.2.6 Rack k-eff KENO case uncertainty u 2 standard deviations Rack k-eff Minor actinides + FP B 1.5% of worth to cover validation gaps (9.2.6)

Rack k-eff Code bias B Section 9.2.6 Rack k-eff Temperature C Most reactive temp. (32 F) used Rack k-eff Eccentric fuel placement C Most reactive position (centered) used Rack k-eff Boral blisters B Full panel blister 0.009 in. thick Rack k-eff NRC admin. margin B 1% L\k

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 121 of 300 9.2.9 Region 1 Infinite Lattice k-eff and Margin Calculation Region 1 KENO infinite lattice model k-eff calculations are performed for fresh 4.75 wt% fuel ,

5.0 wt% fuel with 12 IFBA, and 5.0 wt% fuel depleted to 2 GWd/MTU. Uncertainty and bias calculations are only performed fo r fresh fue l because of the small amount of burnup credit (only a small change in fuel characteristics over 2 GWd/MTU burnup) and the large amount of margin (IFBA credit and burnup credit).

All tolerance cases except the temperature change are performed at 32 °F (1.00 glee) with no soluble boron. The 32 °F base case has the highest k-eff of the four temperature cases (0.0085 dk higher than the 150 °F case). Most tolerance cases (cell wall thickness, cell ID, cell pitch, fuel clad OD and ID, pellet OD, BORAL wrapper thickness, guide tube OD and ID, fuel pin pitch, and BORAL thickness) were run bidirectional rather than assume monotonic behavior.

Eccentric positioning cases including fuel skewed toward the model center (Figure 9.3) and skewed into the same cell corner confirmed that the highest k-eff results from fuel centered in each rack cell. As examples, a few of the key KENO results are shown in Table 9.4.

Table 9.4: Selected Tolerance Results for Region 1 (Fresh 4.75 wt% fuel)

Enrichment Sensitivity (w/o U235) Tolerance K-eff Uncert. (dK) EALF Note 4.75 Base case 0.96745 0.00008 N/A 0.32 Base case full model 32F No wrapper tabs. Used as Wrapper base case for other 4.75 width 0.96766 0.00008 0.0004 0.32 tolerances.

4.75 Cell ID+ 0.97353 0.00008 0.0061 0.31 Increase cell ID 4.75 Cell Pitch- 0.97287 0.00008 0.0054 0.32 Decrease cell pitch 4.75 150F 0.95896 0.00008 -0.0085 0.35 Increase temperature 4.75 Asymmetry 0.96721 0.00008 -0.0002 0.35 Eccentric position Calculation of the burnup worth for the calculation of depletion uncertainty (5% of burnup worth) and measured burnup uncertainty (4% of burnup worth) is demonstrated in Table 9.5. An additional case crediting only the major actinides is also shown to demonstrate the calculation of minor actinide and fission product worth needed for the validation bias (1.5% of worth) calculation. Nuclides included in the major actinides case depleted fuel composition are U-234, U-235, U-238 , Pu-239, Pu-240, Pu-241 , Pu-242, Am -241 , and 0-16 .

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 122 of 300 Table 9.5: Worth of Burnup and Minor Actinides and Fission Products Enrichment Sensitivity (w/o U235) Tolerance K-eff Uncert. (dK) EALF Note Fresh fuel base case 5.0 N/A 0.97646 0.00008 N/A 0.33 for burnup worth 2 Burnup worth 2 5.0 GWd/MTU 0.95476 0.00008 0.0219 0.35 GWd/T 2 Minor actinide+FP 5.0 GWd/MTU 0.97292 0.00008 0.0184 0.34 worth Bias and uncertainty values resulting from KENO cases include the difference in k-eff between the worth case and the base case plus two times the root sum square of the KENO case uncertainties.

Table 9.6 shows the calculation of 95/95 k-eff margin to the limit. Uncertainty values are combined by root sum square. Bias values are added. NRC administrative margin of 0.01 dk is added to the total bias and uncertainty for the Dominion margin calculation. Dominion margin is 1.0 minus base case k-eff minus total bias and uncertainty. The IFBA and burnup credit cases used fresh fuel uncertainty and bias values because the scenarios are similar and because of the excess amount of margin to the limit for these cases.

For depleted fuel, margin is calculated using fresh fuel and depleted fuel code validation bias and uncertainty. The option with the least margin is shown.

The storage configuration with fresh 4.75 wt% fuel in each storage cell results in the least Dominion margin (0.0079 dk). The IFBA credit and burnup credit cases are non-limiting with approximately 0.02 dk margin. Satisfaction of the storage requirements shall use fuel assembly characteristics as follows:

1) The fuel enrichment is the maximum planar volume averaged as-built initial enrichment in the assembly (blanket enrichment is not credited by averaging with the higher enrichment central zone).

2) The fuel burnup is the volume averaged burnup of the assembly as determined using the measured reaction rates with no reduction for measurement uncertainty.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 123 of 300 Tabl e 9.6: Region 1 Total Bias, Uncertainty, and Margin (0 ppm soluble bo ron)

Burnup (GWd/MTU) 0 0 2 Enrichment (wt%) 4.75 5.00 5.00 IFBA Rods 0 12 0 Worths (dkl Minor Actinides and FP 0.0000 0.0000 0.0184 Burnup 0.0000 0.0000 0.0219 Uncertainties (dk)

Fuel Stack PTO 0.0007 0.0007 0.0007 Enrichment, +0.05 0.0022 0.0000 0.0000 Pellet OD 0.0004 0.0004 0.0004 Active Fuel LenQth 0.0002 0.0002 0.0002 Clad ID 0.0002 0.0002 0.0002 Clad OD 0.0021 0.0021 0.0021 GTID 0.0006 0.0006 " 0.0006 &

GTOD 0.0005 0.0005 " 0.0005 Pin Pitch 0.0009 0.0009 0.0009 Cell Wall Thickness 0.0023 0.0023 0.0023 Cell l.D. 0.0061 0.0061 0.0061 Cell Pitch 0.0054 0.0054 0.0054 Wraooer Thickness 0.0005 0.0005 < 0.0005 ,;j@

Wraooer Width 0.0004 0.0004 0.0004 BORAL width 0.0013 0.0013 0.0013 BORAL thickness 0.0010 0.0010 0.0010 Depletion Worth Unc. 0.0000 0.0000 0.0011 Burnup Measurement Unc. 0.0000 0.0000 0.0009 Code BenchmarkinQ Unc. 0.0048 0.0048 0.0088 KENO Case Uncertainty 0.0002 0.0002 0.0002 ASS OF UNCERTAINTIES O.Q104 0.0102 0.0127 Biases (dk)

Minor Actinides and Fission Products N/A N/A 0.0003 Clad Creep and Grid Growth** N/A N/A 0.0000 Radial Burnup Tilt** N/A N/A 0.0000 BORAL Blisters* 0.0009 0.0009 0.0009 Code Benchmarkinq Bias 0.0034 0.0034 0.0018 SUM OF BIASES 0.0043 0.0043 0.0029 Summarv Base Case k-eff 0.9675 0.9534 0.9548 Total Bias and Uncertainty 0.0147 0.0145 0.0156 NRC Administrative Marqin 0.0100 0.0100 0.0100 Maximum k-eff 0.9921 0.9779 0.9804 10CFR50.68 Limit 1.0000 1.0000 1.0000 Dominion Marqin (dk) 0.0079 0.0221 0.0196

• Due to small amount of depletion, fresh fuel values used at 2 GWd/MTU

• • Due to small amount of depletion, these values are negligible

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 124 of 300 9.3 Region 1 KENO SFP Wall Credit Model The Region 1 KENO wall model is derived from the Region 1 infinite lattice model. The number of storage cells is increased to match a Region 1 rack (7 x 10) and one side of the model includes the rack-wall gap, the SFP liner (0.25 inch stainless steel), and the SFP wall.

Reflective boundary conditions are used on all sides of the model.

Figure 9.4 shows a 2-D representation of the Region 1 wall model with 1O cm of water between the racks and the SFP liner. The 8x7 group of cells at the top of Figure 9.4 (Region 1A) have un-poisoned fresh 4.75 wt% fuel. The remaining two rows of cells adjacent to the SFP wall (Region 1B) have un-poisoned fresh 5.0 wt% fuel. This storage configuration allows storage of 5.0 wt% fuel with no credit for burnup or integral absorber.

Tolerance cases evaluate the significance of concrete content, concrete thickness, and rack-wall gap size. The tolerance cases demonstrate that the size of the gap between the rack and the SFP is unimportant and the concrete composition and thickness is unimportant.

KENO cases in Table 9.7 establish the sensitivity to rack-wall water gap size, concrete thickness, and concrete content (0.5. 1.0. and 1.5 volume fraction SCALE regulatory concrete, concrete replaced by water, and EPRI "dry" concrete). The results are insensitive to any of these perturbations. The Region 1 wall model highest k-eff (0.96707) is obtained with 1O cm rack-wall gap and 30 cm of concrete. All of the wall model k-eff values are lower than the Region 1 infinite lattice model. The infinite lattice analysis bounds the Region 1N1 B configuration.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 125 of 300 Figure 9.4: KENO Region 1 Wall Credit Model

• • • • • • • ~ Region 1A

• • • • • • • Region1B

[ ... *. .*. . .. I}j West SFP Wall I

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 126 of 300 Table 9.7: Region 1 Wall Model Base and Sensitivity Cases Enrichment Tolerance K-eff Uncert. EALF Note (wt% U235)*

4.75 /4.75 32F Base 0.96748 0.00008 0.316 No SFP wall (infinite lattice) 4.75 /5.0 Wall gap 0.96689 0.00008 0.317 20 cm gap, SS liner, 20 cm concrete 4.75 /5.0 Wall gap 0.96682 0.00008 0.317 16 cm gap, SS liner, 20 cm concrete 4.75 /5.0 Wall gap 0.96701 0.00008 0.317 10 cm gap, SS liner, 20 cm concrete 4.75 /5.0 Wall gap 0.96698 0.00008 0.317 5 cm gap, SS liner, 20 cm concrete 4.75 /5.0 Wall gap 0.96696 0.00008 0.317 2 cm gap, SS liner, 20 cm concrete 4.75 /5.0 Wall gap 0.96694 0.00009 0.318 Ocm gap, SS liner, 20 cm concrete 4.75 /5.0 Concrete 0.96707 0.00008 0.317 1Ocm gap, SS liner, 30 cm concrete 4.75 /5.0 Concrete 0.96686 0.00008 0.317 10 cm gap, SS liner, 40 cm concrete 4.75 /5.0 Concrete 0.96670 0.00008 0.317 10 cm gap, SS liner, 20 cm 0.5 VF cone 4.75 /5.0 Concrete 0.96700 0.00008 0.317 1Ocm gap, SS liner, 20 cm 1.5 VF cone 4.75 /5.0 Concrete 0.96687 0.00008 0.317 1Ocm gap, SS liner, 20 cm dry cone 4.75 /5.0 Concrete 0.96703 0.00008 0.317 1Ocm gap, SS liner, 20 cm water

• Region 1A/ Region 1B

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 127 of 300 9.4 Region 1 Analysis Summary Region 1 has been demonstrated to satisfy the SFP k-eff requirement for spent fuel pool analyses as follows:

• k-eff < 1.0 if flooded with unborated water (95% probability, 95% confidence) if credit for soluble boron is taken A fuel assembly may be placed in Region 1 if it meets one of the following criteria:

1) The enrichment is s; 4.75 wt% U-235.

2) The enrichment is s; 5.0 wt% U-235 and the fuel assembly contains 12 or more IFBA rods.

3) The enrichment is s; 5.0 wt% U-235 and the assembly average burnup is 2:: 2 GWd/MTU.

4) The enrichment is s; 5.0 wt% U-235 and the assembly is stored in one of the two rows of rack cells adjacent to the West SFP wall.

Region 1 analysis includes calculation and application of bias and uncertainty as well as identification of NRC administrative margin and Dominion margin to the k-eff limit. Satisfaction of the storage requirements shall use fuel assembly characteristics as follows:

1) The fuel enrichment is the maximum planar volume averaged as-built initial enrichment in the assembly (blanket enrichment is not credited by averaging with the higher enrichment central zone).

2) The fuel burnup is the volume averaged burnup of the assembly as determined using the measured reaction rates with no reduction for measurement uncertainty.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 128 of 300 9.5 Region 2 Analysis 9.5.1 Region 2 Fuel Storage A fuel assembly may be placed in Region 2 if it meets one of the following criteria:

1) The fuel assembly burnup and enrichment satisfy the burn up credit curve (Figure 9.14)

2) The fuel assembly enrichment is ~ 5.0 wt% and the fuel assembly contains a 24 finger control rod assembly.

Options 1 and 2 are confirmed by analysis using an infinite lattice rack cell model.

9.5.2 Region 2 Modeling Assumptions These modeling simplifications and assumptions are used in the Region 2 analysis:

a) The plenum , guide tubes above and below the active fuel, and top and bottom nozzles are modeled as water.

b) The storage cell is modeled at its nominal length. The rack structure below the bottom of the storage cell tube is modeled as water. A water reflector extends above the storage tube at the top of the model.

c) Neither reduction in fuel enrichment nor annular pellets are modeled for axial blankets.

This is conservative because it increases SFP k-eff.

d) BORAL is conservatively modeled as the same length as fuel. Table 4.7 shows that using nominal dimensions, the BORAL (using minimum BORAL length) extends well below the fuel in current and prior fuel designs and extends slightly above the fuel in current and prior fuel designs. This obviates the need for fuel axial position tolerance and BORAL length tolerance.

e) The BORAL wrapper length is modeled equal to the active fuel stack length. BORAL wrappers actually extend beyond the upper and lower boundaries of the fuel stack.

Since the region above and below the fuel is insensitive (item b above) and the un-modeled BORAL is a strong neutron absorber, there is no need for a wrapper length tolerance.

f) Fuel pin pitch growth is assumed to be the same relative magnitude as grid growth. This is reasonable because fuel pins are constrained within the grid lattice.

g) Clad OD is conservatively assumed to decrease linearly from Oto 20 GWd/MTU and remain at the minimum value at higher burnup. This assumption is conservative based on the data in Section 9.6.5.

h) Zr-based clad materials are modeled as Zr.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 129 of 300 i) Certain tolerances are only calculated for fresh fuel and not for depleted fuel for efficiency. This practice is justified because results for other tolerances calculated for fresh and depleted fuel that change only modestly and because the tolerances assumed to be constant with burnup are small compared to the total uncertainty and any changes would not be seen in the RSS total.

j) Boral clad thickness is approximated. Boral clad thickness for thickness tolerance cases is conservatively assumed to be O (no clad) and thick enough to fill the design wrapper space (maximum).

k) Water density is determined assuming the pressure is 15 psia.

I) Grids are represented by homogenizing the grid and the water in the fuel lattice over the length of the fuel. Water inside the guide tubes are modeled as pure water.

9.6 Region 2 Infinite Lattice Model Drawings and dimensions of Region 2 fuel racks are provided in Section 4.2.2. The KENO Region 2 model is a 2x2 model with reflective axial and periodic X-Y boundary conditions. The rack module to module spacing is ignored since the rack module baseplate protrudes past the outer cell walls such that storage cell pitch across the rack to rack interface is larger than the normal rack storage cell pitch. Figure 9.7 shows a 2-D representation of the Region 2 model.

Figure 9.8 shows a 3-D cutaway representation of the Region 2 model. Fuel is stored in all cells of the infinite lattice model. A 6x6 expanded version of the Region 2 model is used to evaluate the horizontal burnup tilt bias.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 130 of 300 Figure 9.7: X-Y Representation of Region 2 2x2 KENO Model

@@0@000000000000 00000@0000@00000@

@0800000800000000 00000 000000000000 00000 00 00 00000 00000 00 00 00000 000 000000000 000 000 000000000 00@

00000000000000000 00000000000000000 00 00 00 00 00 00 00 00 00 00 00 00 00000000000000000 00000000000000000 00000000000000000 00000000000000000 00 00 00 00 00 00 00 00 00 88 80 80 800000000008080 00800000000000000

• 00000880008000 00000800@08@90088 w 00 00 00 00 80 00 00 O 00 00 00

%@00000880000000 00000000000000000

• 000080000 000 000 800000000 000 08000 00 00 00000 000@8 00 00 0000 00000000000008@0 00000000000000000 00000000000000000 00000000000000000 00000000000000000 00000000000000000 0000000080000000 00008000000000000 00000 00 00 00800 0000 0 00 00 00000 ooo 000000000 * *

• ooa 8aaaoo*** ooa oaoooooaooaoooooa oooooooooaaooooao oo oo oo oo oe oo oo oa oo oo oo oo ooeoooooooooooooo 00000000000000000 00000000000000000 00000000000000000 oo oo oo oo oo oo oo ao oo oo oo oo oaooooo~ooooooooo ooooaoooooaoooooo 000000000900000000 oooaoaaooaoooooe*

oo oo ~o oo oo oo oo oo oo oo oo oo 00000000000000000 00000000000000000 000 000000000 000 000 000000000 00@

aoeoo co oo 00000 oooaa oo oo ooooe 000000000000000 00000000000000000 00000000000000000 ooaaoooooooaooaoo

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 131 of 300 Figure 9.8: 30 Representation of Region 2 2x2 KENO Model

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 132 of 300 9.6.1 BORAL Blisters BORAL blisters are modeled in the same way as the Region 1 analysis described in Section 9.2.1.

9.6.2 SFP Normal Operation Water Temperature and Density Rack k-eff calculations are performed at the same four water temperatures as Region 1 (Table 9.1 ).

9.6.3 TRITON Depletion Input for Burnup Credit Depleted fuel content is calculated as described in Section 8. Uniform and NUREG/CR-6801 burnup profiles are used to determine the most reactive profile.

9.6.4 Asymmetric Fuel Placement Asymmetric placement of fuel in the storage cell is evaluated by co-locating fuel in the 2x2 model toward the center of the model and by co-locating fuel in the same direction into the corner of each storage cell. As in the Region 1 calculation, centered placement in each cell maximized k-eff.

9.6.5 Fuel Geometry Changes with Burnup Zircaloy based grids can experience expansion during fuel assembly depletion, which can increase fuel pin pitch. Pin pitch tolerance cases confirm that with no soluble boron, increased pin pitch increases fuel rack k-eff. MPS3 fuel has used Zircaloy-4 and ZIRLO grids. For Region 2 burnup credit analysis, grid growth is included in a bias term along with clad creep.

Figure 9.9 shows grid growth data for Zircaloy-4 grids (red data points). The Zirc-4 grid growth model used in this analysis was added to the figure and its equation is displayed below. This model was developed using non-public data but fits reasonably well with the public data in Reference 38.

Similarly, Figure 9.10 shows the degree of expansion for ZIRLO grids. A grid growth model fit line is added over the burnup range of interest (maximum burnup credit for Region 2 is -40 GWd/MTU). The ZIRLO grid growth model used in this analysis was developed using non-public data. The growth model bounds or matches most of the public data in Reference 42.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 133 of 300 Grid expansion used in the KENO Region 2 model to determine fuel pin pitch expansion at all axial elevations (the fit line on Figures 9.9 and 9.10) is described by the following equations (X =

burnup in GWd/MTU , burnup > 20 GWd/MTU):

Growth (Zircaloy-4, %) = 0.0186X - 0.3726 Growth (ZIRLO, %) = 0.0147X- 0.2941 The standard Region 2 fuel model is based on ZIRLO grids, which exhibit less growth than Zircaloy-4. Fuel batches 1 through 5 used lnconel grids. One batch of fuel (Batch 6, introduced in Cycle 4 and discharged after Cycle 5) was built with Zircaloy-4 grids. Batch 7 (introduced in Cycle 5 and discharged after Cycle 9) was re-caged into assembly skeletons with Zircaloy-4 grids. For these 4.2 - 4.5 w/o fuel assemblies, the Region 2 burnup requirement is -36 GWd/MTU or less. At 36 GWd/MTU, the additional grid growth of Zircaloy-4 vs. ZIRLO is re. Table 9.12 shows that the pin pitch tolerance is [ re of the nominal dimension, and Table 9.20 shows that tolerance is conservatively worth 0.0012 dk at 35.65 GWD/MTU. Therefore, based on fuel pitch tolerance cases, the effect of additional Zircaloy-4 grid growth at 36 GWd/MTU is about 0.0007 dk. Burnup in excess of the Region 2 requirement (-0.0035 dk) and decay time of at least 12 years (-0.02 dk) is available to more than offset the grid effect for these Zircaloy-4 grid assemblies. Fuel batches 1 through 5 used lnconel grids.

SFP k increases with reduced clad outer diameter (OD) with no soluble boron. Clad behavior with burnup is a complex function of many variables. A simple and conservative approach is used for MPS3 SFP criticality calculations.

The zirconium alloys Zircaloy-4 and ZIRLO have been used for cladding at MPS3. Reference 25 shows maximum fuel clad diameter reduction for Zircaloy-4 of about 70 microns (approximately 0.7% of the MPS3 fuel clad OD).

The NRC approved Westinghouse PAD 4.0 model (37] predicts ZIRLO clad OD reduction of reoccurring from Oto about 20 GWd/MTU, followed by clad OD increase to

]a,e greater than the initial value. Growth of the outer oxide layer partially offsets the reduced clad OD. The "typical" oxide thickness at 20 GWd/MTU is about [

]a,e (37]. The net fuel clad maximum OD reduction is therefore [ r e.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 134 of 300 Fuel clad creep down is evaluated as a bias. Although models and measurements confirm that clad OD rebounds after reaching a maximum reduction , a conservative two segment linear function will be used. The clad OD is assumed to decrease linearly by [ reat 20 GWd/MTU and then remain at this minimum OD through the rest of the burnup range.

Criticality Safety Evaluation Report Se rial No.18-039 Docket No. 50-423 Attachment 6, Page 135 of 300 Figure 9.9: Zircaloy-4 Grid Growth [38]

0.9 0 .8 RXA Z ircaloy-4 0 .7 Q

0 .6 $

~

e... 0 ¢ Ia "t,

0 .5 0.4

¢

~

~

<I),

6 0 .3 i

~ j 0 .2 A

0 .1 MSTM 0 .0 0 10 20 30 40 50 60 FA b urnup (GWd/tU)

NOTE: Zirc-4 growth model line added.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 136 of 300 Figure 9.10: ZIRLO Grid Growth Data [42]

ZirloGrid Growth Da!a Base 0.80 1 " " - - - - - - - - - - - - - - - - - - - - - - - -

+Plant A

- Plant B +

0.60 +------,

• Plante xNGF O~im~ed ZIRLO Data r.

~

eQ.40 + ------I t Plant EOptimized ZIRLO Data 1 - - - - - ~ - _ , _ __ _

• +

C)

"O

• c A

a

.- I l X X

(.!) * + I I + -

• i  ! X t

t + X

+

q a=

t *~

• ' r A

t ~

X X

0.00 -+ii---------......--------i-------~----.. . .

20000 25000 30000 35000 40000 45000 50000 55000 60000 65000 70000 Bumup (MWDIMTU)

NOTE: ZIRLO growth model line added.

Pellet densification and/or swelling can be ignored because they do not change the fuel to water ratio. Verification that the effect of slightly changing the pellet diameter without changing the fuel mass is neutronically insignificant is provided by comparing the pellet OD and pellet density tolerance case results. One case changes fuel mass and geometry and the other changes only the fuel mass. Adjusted for the amount of fuel mass change represented in each case, the effect on k-eff is the same to within one half of the uncertainty of one KENO case.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 137 of 300 9.6.6 Horizontal Burnup Tilt The effect of horizontal burn up tilt is determined at 10, 20, 30, 36, and 40 GWd/MTU using a Region 2 6x6 rack cell model. Analogous to the calculation of asymmetric placement bias in Region 1, a 4x4 region of optimally oriented, maximally "tilted" fuel is surrounded by a buffer row of fuel with no tilt. A 4x4 region is large enough to dominate the SFP k-eff calculation. Periodic X-Y boundary conditions replicate this grouping of fuel across the entire fuel pool.

Each "tilted" fuel assembly is modeled as two burnup zones on diagonally opposite halves of the fuel assembly. All fuel pins in each diagonal half have the same burnup, and the average burnup of the two halves is equal to the assembly average burnup of a reference case un-tilted assembly. A uniform axial burnup is used for simplicity and because the intent of this calculation is to determine the effect of a horizontal burnup tilt.

Two different 4x4 region burnup tilt orientations are modeled (Figures 9.11 and 9.12). The lower burnup portion of each assembly is oriented towards the center of the 4x4 block (Figure 9.11) or the center of each 2x2 cluster (Figure 9.12). A tilt bias is calculated using the KENO rack model k-effs from tilted and uniform burnup cases.

The magnitude of assembly horizontal burnup tilt was determined using SIMULATE-5 model predictions for fuel at the end of each cycle. Figure 9.13 shows the fuel assembly tilt results for 16 cycles of fuel and the bounding value used for the TRITON depletion models.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 138 of 300 Figure 9.11: Horizontal Burnup Tilt KENO Model #1

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 139 of 300 Figure 9.12: Horizontal Burnup Tilt KENO Model #2

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 140 of 300 Figure 9.13: Quadrant Tilt vs. Burnup 30%

5' cc QI bl) 20%

~

QI ci::

a 10%

E QI

"'"' o SIMS Data,

~

, Cycles 1-6 cc C:

0%

~ o SIMS Data, "C

Cycles 7-16

"':::s

'! -10%

~

.!1.

i=

§" -20%

E

s cc

-30%

0 5 10 15 20 25 30 35 40 45 50 55 60 65 Burnup (GWD/MTU)

Table 9. 1O shows the results of the KENO burnup tilt cases. These tilt bias values are considered conservative , because only a small portion of the fuel population has burnup tilts near the analyzed values and because the Region 2 model represents multiple repeating clusters of optimally oriented fuel assemblies.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 141 of 300 Table 9.10: Horizontal Burnup Tilt Bias Assembly High Tilt Enrich. Burnup Low Burnup Burnup Bias (wt%) (GWd/MTU) Tilt (GWd/MTU) (GWd/MTU) Case k-eff Sigma EALF (dk)*

2.60 10 0 10 10 Base 0.96816 0.00006 0.28 N/A 3.25 20 0 20 20 Base 0.95659 0.00006 0.333 N/A 4.00 30 0 30 30 Base 0.94899 0.00006 0.379 N/A 4.50 36 0 36 36 Base 0.94532 0.00006 0.404 N/A 5.00 40 0 40 40 Base 0.95065 0.00006 0.426 N/A 2.60 10 32% 6.8 13.2 Inward 0.96854 0.00006 0.279 0.0005 3.25 20 21 % 15.8 24.2 Inward 0.95695 0.00006 0.332 0.0005 4.00 30 10% 27.0 33.0 Inward 0.94918 0.00006 0.379 0.0004 4.50 36 10% 32.4 39.6 Inward 0.94542 0.00006 0.405 0.0003 5.00 40 10% 36.0 44.0 Inward 0.95096 0.00006 0.426 0.0005 2.60 10 32% 6.8 13.2 4 Clusters 0.96861 0.00006 0.279 0.0006 3.25 20 21 % 15.8 24.2 4 Clusters 0.95706 0.00006 0.332 0.0006 4.00 30 10% 27.0 33.0 4 Clusters 0.94924 0.00006 0.378 0.0004 4.50 36 10% 32.4 39.6 4 Clusters 0.94543 0.00006 0.405 0.0003 5.00 40 10% 36.0 44.0 4 Clusters 0.95090 0.00006 0.426 0.0004

• Includes statistical uncertainty 9.6.7 Code Validation Bias and Uncertainty Validation of SCALE/KENO using ENDF/B-VII 238 group cross sections identified EALF as the primary bias and uncertainty trend . Region 2 KENO cases have EALF::; 0.35 eV with fuel burnup < 30 GWd/MTU EALF ::; 0.4 eV with fuel burnup < 40 GWd/MTU and EALF < 0.45 with burnup ~ 40 GWd.MTU. The applicable bias and uncertainty for fresh fuel and depleted fuel (MOX validation cases) are shown in Table 9.11.

As in the Region 1 analysis, the limiting temperature is 32 °F, therefore, the validation temperature bias is not applicable.

To cover gaps in the validation for minor actinides and fission products, a bias of 1.5% of the minor actinide and fission product worth is added. [13]

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 142 of 300 Table 9.11: Region 2 Validation Bias and Uncertainty Validation cases Bias (dk) Uncertainty (dk)

U02 (EALF < 0.35) 0.0034 0.0048 MOX (EALF < 0.35) 0.0017 0.0083 U02 (EALF < 0.4) 0.0034 0.0048 MOX (EALF < 0.4) 0.0018 0.0088 U02 (EALF < 0.45) 0.0037 0.0048 MOX (EALF < 0.45) 0.0019 0.0094 9.6.8 Summary of Bias and Uncertainty Table 9.12 summarizes the biases and uncertainties incorporated in the Region 2 analysis.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 143 of 300

. S ummary ofR eg1on 2 Ta bl e 912 e*1ases, unee rtam . f 1es, an dC onservaf ism Analysis Item/Type Explanation (B=bias, C=conservatism, wt . Xff U=uncertainty)  : *i ....../\......

Depletion Burnable absorbers C Boundinq BPRA (24 WABA bounds IFBA).

Depletion Soluble boron C 1050 ppm bounds cycle average boron.

Depletion Fuel and Water C Bounding high node-specific. Section 8.2, 8.4, Temperature 8.5.

Depletion Specific power C Bounding high power assembly history.

Depletion Axial burnup shapes C Uniform and NUREG/CR-6801.

Depletion Grids C Maximum Zr qrids Depletion Volatile fission products C Reduced based on qap release fractions Depletion Fuel density (base case) C 95.5% net bounds all fuel batches Depletion Power history C Reduced 50% for the last 40 days of depletion Depletion lncore thimble C Included in base depletion Rack k-eff Fuel density u 95.5% [ ]a.e net bounds all fuel batches Rack k-eff Grids (base case) C Minimum volume Zr grids Rack k-eff Boral B-10 C Minimum (r n Rack k-eff Boral width u[ r Rack k-eff Boral thickness u[ r Rack k-eff Boral wrapper thickness u[ r Rack k-eff Boral wrapper width ur r (welded]a,e end tab removed)

Rack k-eff Fuel pin pitch u 0.496 [

Rack k-eff Fuel pellet OD u 0.3225 [ t *e Rack k-eff Fuel clad ID u o.329 r la,e Rack k-eff Fuel clad OD u 0.374 [ t *e Rack k-eff Guide tube ID u 0.442 r ia,e Rack k-eff Guide tube OD u 0.474 [ r e Rack k-eff Fuel stack height u[ re Rack k-eff Burnup worth u 5% of burnup worth Rack k-eff Measured burnup u 4% of burnup worth Rack k-eff Enrichment u[ t *e Rack k-eff Cell wall thickness u[ r Rack k-eff Rack cell pitch u[ r Rack k-eff Rack cell ID u[ r Rack k-eff Code uncertainty u Section 9.6.7 Rack k-eff KENO case uncertainty u 2 standard deviations Rack k-eff Minor actinides + FP B 1.5% of worth.

Rack k-eff Code bias B Section 9.6.7 Rack k-eff Temperature C Most reactive temp. (32 F) used Rack k-eff Creep and qrid qrowth B Modified pin pitch and clad diameter Rack k-eff Horizontal burnup tilt B Section 9.6.6 Rack k-eff Eccentric placement C Most reactive position (centered in cell) used Rack k-eff Boral blisters B Full panel blister 0.009 in. thick Rack k-eff NRC admin. margin B 1% t.k

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 144 of 300 9.6.9 Unbounded Historical Fuel Use of 24 W ASA in the TRITON depletions leaves two batches of fuel potentially unbounded due to the burnable absorber used. In Cycle 1, some batch B and C fuel assemblies contained 24 Pyrex BPRA. Pyrex BP is somewhat more limiting than WABA because Pyrex rods displace slightly more water than WABA and have stainless steel clad instead of zirconium based clad .

The effect of Pyrex BPRA depletion vs WABA depletion is estimated to be <0.0015 dk in Table 8.6 (three cycles of depletion with BP). This small effect is more than offset by these considerations:

1) Batch 2 and 3 assemblies have at least 5 GWd/MTU more burn up than required for Region 2 storage (--0.025 dk) .

2) Batch 2 and 3 assemblies were discharged before 1992. Decay time of 25 years provides substantial additional margin for these assemblies since decay time is not credited in Region 2. Decay time of 25 years is worth >0.02 dk at 20 GWD/MTU (based on Region 3 burnup curves and fuel enrichment worth tolerances) .

Ten fuel assemblies from Cycles 1 and 4 were identified in Section 8.8 as having control rod insertion history that exceeds 10 steps cycle average insertion that is bounded by the standard 24 WABA depletion. Sensitivity cases in Section 8.8 estimated the reactivity effect of depletion with a control rod inserted 10 steps and with a control rod inserted 2 nodes (bounds all MPS3 cycles) to be -0.02 dk for 3 cycles of depletion. The assemblies are dispositioned for storage in Region 2 because they have over 5 GWd/MTU more burnup than required (--0.025 dk) and 25 years of uncredited decay time.

Additional discussion of unbounded fuel assemblies is provided in Section 9.9.8 for the Region 3 analysis.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 145 of 300 9.6.10 Region 2 k-eff and Margin Calculation - Fresh fuel With the exception of temperature changes, tolerance cases for Region 2 (fresh fuel with no control rod credit) are performed at 32 °F (1.00 g/cc) with no soluble boron. The 32 °F base case has the highest k-eff of the four temperature cases (0.011 dk higher than the 150 °F case).

Most tolerance cases (cell wall thickness, cell ID, fuel clad OD and ID, pellet OD , BORAL wrapper thickness, guide tube OD and ID, fuel pin pitch, and BORAL thickness) were run bidirectional rather than assume monotonic behavior. Statistically significant tolerances are monotonic and only one-sided analysis is needed. Eccentric positioning cases including fuel skewed toward the model center and skewed into the same cell corner confirmed that the highest base case k-eff results from fuel centered in each rack cell. As examples, a few of the key KENO results are shown in Table 9.13.

Table 9.13: Selected Tolerance Results for Region 2 (Fresh fuel)

Enrichment Sensitiviy (w/o U235) Tolerance K-eff Uncert. (dK) EALF Note 2.0 32F Base 0.96421 0.00006 N/A 0.189 Symmetric 2x2 32F 2.05 Enrichment 0.97159 0.00006 0.0076 0.190 Enrichment +0.05 2.0 Clad OD- 0.96544 0.00006 0.0014 0.188 Decrease clad OD 2.0 narrowboral 0.96650 0.00006 0.0025 0.188 Decrease BORAL width 2.0 Pin pitch+ 0.96476 0.00006 0.0007 0.188 Increase pin pitch 2.0 150F 0.95227 0.00006 -0.0110 0.214 Temperature 2.0 Asymmetric 0.96054 0.00006 -0.00349 0.191 Eccentric 2x2 Positioning Bias and uncertainty values resulting from KENO cases include the difference in k-eff between the worth case and the base case plus two times the root sum square of the KENO case uncertainties.

Table 9.14 shows the calculation of 95/95 k-eff margin to the limit. Uncertainty values are combined by root sum square. Bias values are added. NRC administrative margin of 0.01 dk is added to the total bias and uncertainty for the Dominion margin calculation. Dominion margin is 1.0 minus base case k-eff minus total bias and uncertainty.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 146 of 300 Table 9.14: Region 2 Total Bias, Uncertainty, and Margin (0 ppm soluble boron)

Burnup (GWd/MTU) 0 Enrichment (wt%) 2.03 Uncertainties (dk Fuel Stack PTO 0.0009 Enrichment, +0.05 0.0076 Pellet OD 0.0006 Active Fuel Length 0.0003 Clad ID 0.0003 Clad OD 0.0014 GTID 0.0005 GTOD 0.0006 Pin Pitch 0.0007 Cell Wall Thickness 0.0005 Cell l.D. 0.0003 Cell Pitch 0.0006 Wrapper Thickness 0.0005 Wrapper Width 0.0003 BORAL width 0.0025 BORAL thickness 0.0003 Code Benchmarking Unc. 0.0048 KENO Case Uncertainty 0.0001 RSS OF UNCERTAINTIES 0.0096 Biases (dk)

BORAL Blisters 0.0004 Code Benchmarking Bias 0.0034 SUM OF BIASES 0.0038 Summary Base Case k-eff 0.9687 Total Bias and Uncertainty 0.0133 NRC Administrative Margin 0.0100 Maximum k-eff 0.9920 10CFR50.68 Limit 1.0000 Dominion Margin (dk) 0.0080

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 147 of 300 9.6.11 Region 2 Control Rod Credit Control rods in the SFP may have reached their in-core design lifetime and been discharged.

Discharged control rods may have experienced some depletion of the absorber material near the bottom of the rod. The amount of depletion is dependent on the at-power residence time in the reactor. Conservative depleted absorber content is required to extend control rod credit to discharged control rods.

There have been three types of control rod used in MP3. Table 9.15 summarizes the relevant design features. The first generation of MPS3 control rods were Hf/Zr. A transition to the second generation (AIC) began when Cycles 3 and 4 had 6 AIC control rods and 55 Hf/Zr control rods. Cycle 5 completed the transition with 61 AIC control rods. A transition to the third generation of control rods (AIC) began in Cycle 13.

RCCAs are depleted for three cycles or fewer in lead bank locations. In all but two cycles, lead bank cycle average insertion is 1O steps or less. Therefore, control rod depletion for three cycles involving the bottom 15.9 cm of absorber will bound all but Cycle 1 and Cycle 4 control rods .

Only one control rod was used in a lead bank location in three cycles and was also present in Cycle 1 or Cycle 4. No RCCA was in a lead bank location in both Cycles 1 and 4. The bounding RCCA history is one cycle of 22 step depletion and 2 additional cycles of 1O step depletion. A bounding depletion history is therefore depletion of the bottom 1O steps (15.9 cm) of absorber for three cycles and depletion of the next 12 steps (total length 34.9 cm) above that region for one cycle.

To bound control rod depletion conditions, the following TRITON depletion model input is used:

• Nominal absorber density and dimensions (TRITON depletion)

• Node 18 depletion conditions (top of core)

• Deplete 3 cycle control rod absorber (bottom 10 steps) using 40 GWd/MTU uniform axial burnup shape at core average RPO.

• Deplete 1 cycle rod absorber (next 12 steps) using 20 GWd/MTU uniform axial burnup shape at core average RPO.

• Fuel enrichment of 4.0 w/o is representative of historical MP3 fuel that experienced the highest average at power D-Bank insertion. Recent cycles have operated with less D-Bank insertion so depleting the control rods with parameters from Cycles 1-4 is bounding.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 148 of 300 Justification for the 3 cycle absorber depletion strategy is as follows. MPS3 cycle length is approximately 20 GWd/MTU. Fuel assembly average power in lead bank locations tends to be above average (- 1.25 x cycle average or less). However, in the top 16 cm of the fuel (midpoint of depletion region) , local power even with no control rod inserted relative power tends to be well below average due to leakage (- 0.4 x axial average). Control rod insertion in the lead bank location assembly further reduces the local power substantially. Depletion for 40 GWd/MTU is approximately 33% more than the best estimate burnup (20 GWd/MTU x 1.25 x 3 cycles x .4 = 30 GWd/MTU) even before accounting for power depression by the control rod.

Justification for the 1 cycle absorber depletion strategy is similar, except that the basis is the Cycle 1 and Cycle 4 cycles in particular because those are the cycles with greater than 1O step cycle average control rod insertion. The maximum burnup accumulation in Cycle1 and Cycle 4 lead bank locations is 21 .2 GWd/MTU. The maximum cycle average relative un-rodded axial power of the two cycles considered at the 1 cycle absorber elevation is - 0.8. Control rod insertion in the lead bank location assembly further reduces the local power substantially.

Depletion for 20 GWd/MTU is approximately 18% more than best estimate burnup (21.2 GWd/MTU x 0.8 = 16.96 GWd/MTU) even before accounting for powe r depression by the control rod.

Table 9.15: MPS3 Control Rod Data CR Type Item Value Hf-Zr Absorber 95.3 - 95.4 w/o Hf, 4.5 w/o Zr 80 w/o Ag , 15 w/o In, 5 w/o Ag -In-Cd Absorber Cd (10.16 glee)

All Absorber OD 0.341 in.

All Clad ID 0.344 in.

All Clad OD 0.381 in.

All Clad Material SS-304 All Step size 0.625 in.

Fuel below All 5.25 in.

absorber All Lead bank 5 core locations All All banks 61 core locations D-bank insertion All Figure 8.5

@ HFP Other bank All Figure 8.5 insertion @ HFP Modeled D-bank 22 steps (1 cycle)

All insertion @ HFP 1O steps (3 cycles)

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 149 of 300 Table 9.16 presents the results of KENO control rod credit cases using fresh 5.0 w/o fuel.

Control rods extend from 5.25 (maximum) inches above the bottom of the active fuel to the top of the active fuel. Control rod structure below the bottom of the control rod absorber is modeled as water. Nominal fuel and control rod dimensions are used. The bottom 15.88 cm of control rod absorber is depleted material from the 40 GWd/MTU TRITON node 18 depletion conditions.

Control rod absorber from 15.88 cm to 34.93 cm is depleted material from the 20 GWd/MTU TRITON node 18 depletion conditions. The balance of the control rod absorber is fresh material.

Table 9.16: Region 2 KENO Control Rod Credit Cases (5.0 w/o fresh fuel)

Control K-eff Uncert EALF Note Rod None 0.96430 0.00006 0.189 Base case 2.0 wt% fresh AIC 0.93264 0.00008 0.479 5.0 wt% fresh , 32 F AIC 0.92268 0.00007 0.531 5.0 wt% fresh, 150 F Hf/Zr 0.93240 0.00007 0.484 5.0 wt% fresh , 32 F The highest k-eff control rod case is more than 0.03 dk less reactive than the Region 2 fresh fuel base case . A full accounting of bias and uncertainty for this configuration would include tolerances that would increase total bias and uncertainty (absorber OD, control rod clad OD, absorber composition) but no enrichment tolerance . Review of similar SFP control rod credit cases for North Anna [23] indicates that total bias and uncertainty is smaller for the control rod configuration than for no control rod credit.

The North Anna calculations and the large amount of margin indicated by the control rod credit base cases is sufficient to conclude that for Region 2, the control rod credit configuration is significantly non-limiting and there is no need to perform a full set of tolerance and bias cases.

Fuel assemblies of enrichment 5.0 wt% U-235 or less, each containing a control rod may be stored in Region 2 with no burnup required. Removal of the control rod must be performed with the assembly in a Region in which it qualifies for storage without the control rod.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 150 of 300 9.6.12 Region 2 k-eff and Margin Calculation - Burnup credit With the exception of temperature changes, tolerance cases for Region 2 (depleted fuel) are performed at 32 °F (1.00 glee) with no soluble boron. The 32 °F base case has the highest k-eff. Table 9.17 identifies the NUREG/CR-6801 Table 5 axial burnup shapes used to determine the TRITON depletion input.

Table 9.17: SFP Region 2 Limiting Axial Shapes Assembly Average Limiting Shape (Uniform Burnup or NUREG/CR-6801 (GWd/MTU) Table 5 Group) 10 Uniform

• 20 8*

30 5 35, 36 4 40,41 3

• Uniform and NUREG shapes checked . Limiting shape listed.

Not all uncertainty terms are calculated for each burnup/enrichment combination :

• Fuel density uncertainty is calculated for fresh fuel and used for all burnups.

• Clad OD uncertainty is used for fresh fuel and replaced by a bias for clad creep for depleted fuel.

• Some minor uncertainty items are calculated for fresh fuel and used for all burnups (shaded values in Table 9.20)

• Some bias and uncertainty items are calculated at two burnups to confirm no significant trend exists and the maximum value used at other burnups.

Minor uncertainties are not calculated at all burnups because they require significant resources and because the magnitude is small enough that modest variation would not significantly affect the RSS total uncertainty, which is dominated by enrichment uncertainty at low burnup and by burnup uncertainties at high burnup. Bias and uncertainty values resulting from KENO cases include the difference in k-eff between the worth case and the base case plus two times the root sum square of the KENO case uncertainties.

Due to the iterative nature of the burnup credit curve determination, some bias and uncertainty calculations are performed at an enrichment or burnup close to but not identical to the final

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 151 of 300 enrichment/ burnup point. This is acceptable because only the enrichment uncertainty term is a strong function of enrichment. The base case k-eff at 35.65 and 40.25 GWd/MTU is determined by linear interpolation versus burnup using cases 1 GWd/MTU apart. Enrichment sensitivity is calculated with depleted fuel (2 sets of TRITON depletions required) except at 20 GWd/MTU, which uses fresh fuel. For example, two sets of TRITON depletion cases were run at 1O GWD/MTU with either 2.55 wt% or 2.60 wt% while only two fresh fuel cases were run at 3.25 wt% or 3.30 wt% to calculate the 20 GWD/MTU enrichment sensitivity. Table 9.18 provides a comparison of enrichment tolerance calculated using fresh and depleted fuel at two different enrichment/ burnup combinations that supports using either method for depleted fuel.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 152 of 300 Table 9.18: Enrichment Tolerance Calculation Comparison Enrichment Assembly Average Enrichment Tolerance (wt% U-235) Burnup (GWd/MTU) (dk w/0.05 wt% increase) 2.55 0 0.0057 2.55 10U* 0.0046 4.0 0 0.0031 4.0 30 0.0032

• U indicates uniform burnup profile.

Table 9.19 contains the Region 2 KENO model k-eff base cases performed at 273 K (1.00 glee water density) with no soluble boron.

Table 9.19: Region 2 Base Cases (0 ppm boron)

Enrichment Burnup K-eff Uncert. EALF (w/o U-235) (GWd/MTU) 2.6 10U* 0.96570 0.00006 0.277 3.25 20 0.96323 0.00007 0.314 4.05 30 0.96135 0.00006 0.361 4.47 35 0.96332 0.00007 0.379 4.47 36 0.95940 0.00006 0.379 5.0 40 0.95791 0.00006 0.408 5.0 41 0.95366 0.00007 0.409

• U indicates uniform burnup profile.

Table 9.20 shows the calculation of 95/95 k-eff margin to the limit for depleted fuel in Region 2.

Shaded cells are values that were not calculated for that burnup and enrichment combination .

The reactivity effect of 4% burnup uncertainty (the k-eff increase resulting from 4% burnup reduction) is determined by linear interpolation of adjacent burnup worth vs. burnup data points that bound the reduced burnup rather than calculating 4% of the total burnup worth.

Uncertainty values are combined by root sum square. Bias values are added. NRC administrative margin of 0.01 dk is added to the total bias and uncertainty for the margin calculation. Margin is 1.0 minus base case k-eff minus total bias and uncertainty. For depleted fuel, margin is calculated using fresh fuel and depleted fuel code validation bias and uncertainty.

The option with the least margin is shown. The Region 2 burnup credit curve, which matches or bounds all analyzed points, is shown in Figure 9.14.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 153 of 300 Table 9.20: Region 2 Total Bias, Uncertainty, and Margin (Depleted fuel, 0 ppm boron)

Burnup (GWd/MTU) 10 20 30 35.65 40.25 Enrichment (wt%) 2.60 3.25 4.05 4.47* 5.00*

Worths (dk)

Minor Actinides and FP 0.0486 0.0663 0.0871 0.0942 0.1079 Burnup 0.0737 0.1367 0.1927 0.2159 0.2441 Uncertainties (dk)

Fuel Stack PTD 0.0009 0.0009 0.0009 0.0009 0.0009 Enrichment, +0.05 0.0046 0.0043 0.0032 0.0027 0.0000 Pellet OD 0.0002 0.0002 0.0002 0.0002 0.0002 Active Fuel Length 0.0003 0.0003 0.0003 0.0003 0.0003 Clad ID 0.0003 0.0003 0.0003 0.0003 0.0003 Clad OD (replaced by creep bias) 0.0000 0.0000 0.0000 0.0000 0.0000 GT ID 0.0004 0.0004. 0.0004 0.0004 0.0004 GTOD 0.0007 0.0007 0.0007 0.0007 0.0007 Pin Pitch 0.0009 0.0009 0.0012 0.0012 0.0012 Cell Wall Thickness 0.0004 0.0004 0.0004 0.0004 0.0004 Cell l.D. 0.0002 0.0002 0.0002 0.0002 0.0002 Cell Pitch 0.0013 0.0013 0.0015 0.0015 0.0015 Wrapper Thickness 0.0002 0.0002 0.0004 0.0004 0.0004 Wrapper Width 0.0003 0.0003 0.0003 0.0003 0.0003 BORAL Width 0.0024 0.0024 0.0024 0.0024 0.0024 BORAL Thickness 0.0003 0.0003 , 0.0003 0.0003 0.0003 Depletion Worth Unc. 0.0037 0.0068 0.0096 0.0108 0.0122 Burnup Measurement Unc. 0.0029 0.0050 0.0067 0.0058 0.0099 Code Benchmarking Unc. 0.0083 0.0083 0.0088 0.0088 0.0094 KENO Case Uncertainty 0.0001 0.0001 0.0001 0.0001 0.0001 ASS OF UNCERTAINTIES 0.0111 0.0130 0.0154 0.0157 0.0186 Biases {dk)

Minor Actinides and Fission Products 0.0007 0.0010 0.0013 0.0014 0.0016 Clad Creep and Grid Growth 0.0011 0.0023 0.0039 0.0044 0.0050 Radial Burnup Tilt 0.0006 0.0006 0.0004 0.0003 0.0005 Blisters 0.0002, 0.0002 0.0004 ' )f0.0004 0.0004 Code Benchmarking Bias 0.0017 0.0017 0.0018 0.0018 0.0019 SUM OF BIASES 0.0043 0.0058 0.0078 0.0082 0.0093 Summary Base Case k-eff 0.9657 0.9632 0.9614 0.9608 0.9568 Total Bias and Uncertainty 0.0154 0.0189 0.0232 0.0239 0.0280 NRC Administrative Margin 0.0100 0.0100 0.0100 0.0100 0.0100 Maximum k-eff 0.9911 0.9921 0.9946 0.9947 0.9948 10CFR50.68 Limit 1.0000 1.0000 1.0000 1.0000 1.0000 Dominion Margin (dk) 0.0089 0.0079 0.0054 0.0053 0.0052

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 154 of 300 Figure 9.14: Region 2 Bounding Burnup Credit Curve 45 . ,.....................,....................,.....................,......................,.....................,......................,....................., ...........,

40 5' 35 I-

~

~ 30 . ,.....r;;:::-;;:::-;;:::-~ - ~ -, ., .... ..................,....................1......

~ o "A~~lyzed Q.

...~

I 25 - Poly. (Bounding) NOT ACCEPTABLE CD CIJ

20 ..,...................+ ................... j ...... ...1............ ********************t* ....................;.................... ,

CIJ y = -0.16936x 3- 0.04949x2 + 2Q.SS1x - 40.098 f

~

15 + ............................... -+ i 1 r t

<10

~

-+----+---+---E-- - + - - - + - - - l - - t - - - + - - - -1+ - - - - - - + - - - + - - - + - - -

j .................... .;.. ...................................... ..;................... + ................... J 0 +----+*-*--+-- -,--*---+-*----+**-*--*-*-**

I

---

+---.*-*--**-*--+--->-

2.0 2.2 2.4 2.6 2.8 3.0 3.2 3.4 3.6 3.8 4.0 4.2 4.4 4.6 4.8 5.0 Max Planar Average Assembly Initial Enrichment (wt% U-235)

Satisfaction of the burnup curve requirement shall use fuel assembly characteristics as follows:

1) The fuel enrichment is the maximum planar volume averaged as-built initial enrichment in the assembly (blanket enrichment is not credited by averaging with the higher enrichment central zone) .

2) The fuel burnup is the volume averaged burnup of the assembly as determined using the measured reaction rates with no reduction for measurement uncertainty.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 155 of 300

9. 7 Region 2 Analysis Summary Region 2 has been demonstrated to satisfy the SFP k-eff requirement for spent fuel pool analyses as follows:

• k-eff < 1.0 if flooded with unborated water (95% probability, 95% confidence)

In Region 2 with 4-out-of-4 storage, the requirement is met with either a) a restrictive loading curve (burnup credit curve) of minimum required assembly measured burnup versus initial fuel assembly as-built enrichment orb) credit for a control rod stored in the fuel assembly.

Region 2 analysis includes calculation and application of bias and uncertainty as well as identification of NRC administrative margin and Dominion margin to the k-eff limit. For Region 2, one 3rd order polynomial burnup credit curve is determined for fuel with initial enrichment

>2 .03 wt% U-235. The coefficients of the curve are shown in Figure 9.14.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 156 of 300 9.8 Region 3 Analysis 9.8.1 Region 3 Fuel Storage Analysis of Region 3 includes credit for burnup and decay time. A fuel assembly may be placed in Region 3 if the fuel assembly burnup and enrichment satisfy a burnup and decay time credit curve (Figure 9.20) . Analysis is performed using an infinite lattice 6x6 rack cell KENO model.

9.8.2 Region 3 Modeling Assumptions These modeling simplifications and assumptions are used in the Region 3 analysis:

a) Boraflex is not credited and is modeled as water.

b) The plenum , guide tubes above and below the active fuel, and top and bottom nozzles are modeled as water.

c) The storage cell tubes and shear wall are modeled at the nominal shear wall length (from the bottom of the fuel assembly to -5 cm above the active fuel. The rack structure below the bottom of the storage cell tube and above the shear wall is modeled as un-borated water.

d) No reduction in fuel enrichment is applied for axial blankets.

e) Fuel pin pitch growth is assumed to be the same relative magnitude as grid growth . This is reasonable because fuel pins are constrained within the grid lattice.

f) Clad OD is conservatively assumed to decrease linearly from Oto 20 GWd/MTU and remain at the minimum value at higher burnup.

g) Zr-based clad materials are modeled as Zr.

h) Certain tolerances are only calculated for fresh fuel and not for depleted fuel for efficiency. This practice is justified because results for other tolerances calculated for fresh and depleted fuel that change only modestly and because the tolerances assumed to be constant with burnup are small compared to the total uncertainty and any changes would not be seen in the RSS total.

i) Water density is determined assuming the pressure is 15 psia.

j) Grids are represented by homogenizing the grid and the water in the fuel lattice over the length of the fuel. Water inside the guide tubes are modeled as pure water.

k) Certain tolerances are only calculated for the O year decay curve and not for the decay time curves. This was done for efficiency and is justified via the same reasons listed in assumption h above.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 157 of 300 9.9 Region 3 Infinite Lattice Model Drawings and dimensions of Region 3 fuel racks are provided in Section 4.2.3. The KENO Region 3 model is a 6x6 model with void axial and periodic X-Y boundary conditions. The axial regions above and below the active fuel are large enough to be effectively neutronically infinite

(- 28 cm of water at the bottom and - 45 cm of water at the top). Each Region 3 rack is a 6x6 module with a stainless steel rack wall around the perimeter of the module. The KENO model represents an actual rack module.

The X-Y boundary conditions represent an infinite number of adjacent modules. The KENO model includes the as-installed minimum rack-to-rack spacing of 1 inch. The nominal rack spacing is 1.125 inch. The rack-to-rack adjacent cell pitch is greater than the nominal storage cell pitch as long as the rack-to-rack spacing is greater than 0.25 inch.

Figure 9.15 shows a 2-D X-Y representation of the Region 3 model. Fuel is stored in all cells of the infinite lattice model. Storage cells on the perimeter of Region 3 (a 3x3 or 3x4 cluster of rack modules as shown in Figure 4.8) do not have Boraflex or wrappers on the outer face.

Because no credit is taken for Boraflex, the model is insensitive to the presence or absence of perimeter wrappers. Figure 9.16 shows the axial view.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 158 of 300 Figure 9.15: X-Y Representation of Region 3 KENO Model Expanded

-:=

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 159 of 300 Figure 9.16: Axial Representation of Region 3 KENO Model

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 160 of 300 9.9.1 SFP Normal Operation Water Temperature and Density Rack k-eff calculations are performed at three of the four water temperatures in Table 9.1 (32°F, 68°F, and 150°F).

9.9.2 TRITON Depletion Input for Burnup Credit Depleted fuel content is calculated as described in Section 8. Uniform and NUREG/CR-6698 burnup profiles are used at 10, 20, and 30 GWd/MTU to determine the most reactive profile.

9.9.3 Asymmetric Fuel Placement Asymmetric placement of fuel in the storage cell is evaluated by co-locating fuel in the center 4x4 group of storage cells into the corner of each cell toward the center of the 6x6 model. The remaining storage locations have fuel centered in each cell. Test cases for Region 3 confirm that asymmetric placement increases rack k-eff (-0.007 dk). The Region 3 KENO base model includes the 4x4 group of asymmetrically placed fuel.

9.9.4 Fuel Geometry Changes with Burnup Region 3 depleted fuel cases use the same Zirlo-based clad creep and grid growth models as Region 2 cases. Clad creep and grid growth are evaluated as a bias.

One batch of fuel (Batch 6, introduced in Cycle 4 and discharged after Cycle 5) was built with Zircaloy-4 grids. Batch 7 (introduced in Cycle 5 and discharged after Cycle 9) was re-caged into assembly skeletons with Zircaloy-4 grids. For these 4.2 - 4.5 w/o fuel assemblies with 13.5-22.5 years decay time, the Region 3 burnup requirement (see Figure 9.20) is -40 GWd/MTU or less.

At 40 GWd/MTU, the additional grid growth of Zircaloy-4 vs. ZIRLO is [

Table 9.23 shows that the pin pitch tolerance is [ reof the nominal dimension , and Table 9.28 shows that tolerance is worth 0.001 O dk at 40 GWD/MTU . Therefore, based on fuel pitch tolerance cases, the reactivity effect of the grid growth difference is -0.001 dk. Batch 6 and 7 fuel assemblies have 6 inch natural enrichment axial blankets, which are not credited in the bounding assembly design used to determine the burnup credit requirement. A 40 GWd/MTU comparison case was run with the top and bottom 6 inches of depleted enriched fuel replaced

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 161 of 300 with depleted natural enrichment fuel. Because the nodes are 8 inches high, the top and bottom nodes were split into 2 inch and 6 inch segments with the 2 inch portion containing enriched depleted fuel. The effect of the natural enrichment blankets in the Batch 6 and 7 fuel at 40 GWd/MTU is approximately -0.03, or about 30 times the positive effect of the additional Zircaloy-4 grid expansion. Crediting the axial blanket is more than sufficient to determine that no special storage requirements are needed for Batch 6 and 7 fuel.

9.9.5 Horizontal Burnup Tilt The effect of horizontal burnup tilt is determined for Region 3 in the same manner as for Region 2 (Section 9.6.6). The two different 4x4 region burnup tilt orientations are shown in Figures 9.17 and 9.18. The lower burnup portion of each assembly is oriented towards the center of the model (Figure 9.17) or the center of each 2x2 cluster (Figure 9.18). A tilt bias is calculated using the KENO rack model k-effs from tilted and uniform burnup cases.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 162 of 300 Figure 9.17: Region 3 Horizontal Burnup Tilt KENO Model #1 Figure 9.18: Region 3 Horizontal Burnup Tilt KENO Model #2

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 163 of 300 9.9.6 Code Validation Bias and Uncertainty Validation of SCALE/KENO using ENDF/B-VII 238 group cross sections identified EALF as the primary bias and uncertainty trend. Because of the flux trap design with no credit tor Boraflex, Region 3 KENO cases have EALF ~ 0.35 eV for all analyzed fuel burnups. The applicable bias and uncertainty for fresh fuel and depleted fuel (MOX validation cases) are shown in Table 9.22.

Unlike the Region 1 and 2 analyses, the limiting temperature in Region 3 is 150 °F, therefore ,

the validation temperature bias is applicable. The validation temperature bias (dk) is calculated by multiplying the l:l. T above room temperature (K) by 1. ?E-05 dk/K. At 150 °F the bias is 0.00078 dk.

To cover gaps in the validation for minor actinides and fission products, a bias of 1.5% of the minor actinide and fission product worth is added. [13)

Table 9.22: Region 3 Validation Bias and Uncertainty Validation cases Bias (dk) Uncertainty (dk)

U02 (EALF < 0.35) 0.0034 0.0048 MOX (EALF < 0.35) 0.0017 0.0083 9.9. 7 Summary of Bias and Uncertainty Table 9.23 summarizes the biases and uncertainties incorporated in the Region 3 analysis.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 164 of 300 Table 9.23: Summary of Region 3 Biases, Uncertainties, and Conservatism ff Analysis Ad Item /Type m;w "' rit;: Explanation *' MF .  !!)ff (B=bias, C=conservatism, "ilk, di 'Willi. di U=uncertaintv), /fa. .J!lllif )( :1111; Ilk cwiimt "!~~!!it Depletion Burnable absorbers C Boundinq BPRA (24 WABA bounds IFBA).

Depletion Soluble boron C 1050 ppm bounds cycle average boron.

Depletion Fuel and Water C Bounding high node-specific. Section 8.2, 8.4, Temperature 8.5.

Depletion Specific power C Bounding high power assembly history.

Depletion Axial burnup shapes C Uniform and NUREG/CR-6801.

Depletion Grids C Maximum Zr qrids Depletion Volatile fission products C Reduced based on gap release fractions Depletion Fuel density (base case) C 95.5% net bounds all fuel batches Depletion Power history C Reduced 50% for the last 40 days of depletion Depletion lncore thimble C Included in base depletion Rack k-eff Fuel density u 95.5% [ ]a,c net bounds all fuel batches Rack k-eff Grids (base case) C Minimum volume Zr grids Rack k-eff Wrapper thickness u [ la,c Rack k-eff Rack wall thickness u r ia,c Rack k-eff Fuel pin pitch u 0.496 [ ]a,c Rack k-eff Fuel pellet OD u o.3225 r r ,c Rack k-eff Fuel clad ID u 0.329 [ ia,c Rack k-eff Fuel clad OD u 0.374 [ ia,c Rack k-eff Guide tube ID u 0.442 r la,c Rack k-eff Guide tube OD u 0.474 [ ia,c Rack k-eff Fuel stack height u [ ia,c Rack k-eff Burnup worth u 5% of burnup worth Rack k-eff Measured burnup u 4% of burnup worth Rack k-eff Enrichment u [ ]a,c Rack k-eff Cell wall thickness u [ re Rack k-eff Rack cell pitch u [ ia,c Rack k-eff Rack cell ID u r ia,c Rack k-eff Rack pitch u [ ia,c Rack k-eff Code uncertainty u Section 9.9.6 Rack k-eff KENO case uncertainty u 2 standard deviations Rack k-eff Minor actinides + FP B 1.5% of worth.

Rack k-eff Code bias B Section 9.9.6; separate bias for temperature Rack k-eff Fuel axial position B [ t *c Rack k-eff Temperature C Most reactive temp . used (150 F)

Rack k-eff Creep and grid qrowth B Modified pin pitch and clad diameter Rack k-eff Horizontal burnup tilt B Section 9.9.5 Rack k-eff Eccentric placement C Most reactive position (eccentric) used Rack k-eff NRG admin. margin B 1%~k

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 165 of 300 9.9.8 Unbounded Historical Fuel Use of 24 W ABA in the TRITON depletions leaves two batches of fuel potentially unbounded due to the burnable absorber used. In Cycle 1, some batch B and C fuel assemblies contained 24 Pyrex BPRA. Pyrex BP is somewhat more limiting than WABA because Pyrex rods displace slightly more water than WABA and have stainless steel clad instead of zirconium based clad.

The effect of Pyrex BPRA depletion vs WABA depletion is estimated to be 0.00144 dk in Table 8.6 (three cycles of depletion with BP).

The assemblies in question have more than 25 years of decay time and have an enrichment which requires between 20 and 30 GWD/MTU on the burnup curve (Figure 9.20). The Region 3 calculated differential burnup worth between 20 and 30 GWd/MTU with 25 years decay time (Table 9.33) is 0.0068 ~k / GWD/MTU. Using this sensitivity, excess burnup of 0.21 GWD/MTU is sufficient to accommodate the Pyrex BPRA depletion reactivity effect. The minimum excess burnup for Batch B and C assemblies is 1.128 GWD/MTU, therefore no special requirement is needed for storage in Region 3 (or Region 2, which requires less burnup credit).

Ten fuel assemblies from Cycles 1 and 4 were identified in Section 8.8 as having control rod insertion history that exceeds 1O steps cycle average insertion that is bounded by the standard 24 WABA depletion. Sensitivity cases in Section 8.8 estimated the reactivity effect of depletion with a control rod inserted 1O steps vs. a control rod inserted 2 nodes (bounds all MPS3 cycles) to be - 0.02 dk for 3 cycles of depletion to 54 GWd/MTU. Cycle 4 lead bank assemblies have over 1O GWd/MTU more burnup than required by the burnup credit curves (Figure 9.20). The assemblies in question have more than 18 years of decay time and have an enrichment which requires between 30 and 40 GWD/MTU on the burnup curve (Figure 9.20). The Region 3 calculated differential burnup worth between 30 and 40 GWd/MTU with 18 years decay time (Table 9.32) is 0.0058 ~k / GWD/MTU. Using this sensitivity, excess burnup of 3.42 GWD/MTU is sufficient to accommodate the Cycle 4 assemblies with unbounded control rod insertion.

Since all five Cycle 4 assemblies have more than 1O GWD/MTU of excess burnup, they can be stored in Region 3 with no restrictions.

The Cycle 1 assemblies were only depleted one cycle (-22 GWd/MTU), therefore the control rod effect cited is likely significantly overstated for them. Cycle 1 lead bank assemblies have

-30 years cooling time and a minimum of 1.6 GWd/MTU excess burnup (--0.01 dk reactivity effect) as compared to the Region 3 25 year decay burnup curve. Rather than re-calculate the

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 166 of 300 control rod effect over one cycle of depletion, the assembly with the minimum excess burnup (824) is explicitly modeled to directly confirm the as-depleted Cycle 1 fuel meets the requirements for storage in Region 3.

Table 9.24 shows the actual operating history of assembly 824 compared to the bounding model used to create the Region 3 burnup curves. To determine if the standard Region 3 modeling bounds 824, the in-rack k-eff of 824 is compared to the bounding model's 2.9 wt% , 20 GWO/T, 25 year decay time case (Table 9.33) which is used, in part, to determine the burnup curve. If assembly 824 has a lower best-estimate rack k-eff then it can be concluded that it is bounded by the standard model, because assembly 824 biases for quadrant burnup tilt and grid growth are lower than the standard model and other uncertainties (rack and fuel geometry tolerances) are the same.

Table 9.24: Depletion Conditions Comparison for Assembly 824 Parameter Bounding Model Assembly B24 Average RPO 1.44 1.160 Average Soluble Boron 1050 ppm 628 ppm Core Average Moderator 592.3°F 586.9°F Temperature Inlet Temperature 556.6°F 558.4°F Core Power 3725 MWt 3411 MWt RCS Flow 391 ,358 gpm 424,937 gpm Bypass Flow 5.6% 5.6%

Axial Shape NUREG Shapes CMS5 Shapes Axial Fuel Blankets None None Fuel Density 0.955 [ Ja,c Grids Zirconium Grids lnconel Grids Control Rod Poison Type Full Length WABA (2 Node Insertion)

Instrument Tube In-Core Thimble Water Filled G.T. and I.T. Dimensions V5H Standard Quadrant BU Tilt 1 -19.09% -2.17%

Decay Time 25 Years 29 .5 Years 3 40 days at 50% reduced 40 days at 50% reduced Low Power EOL power power Ignored in Depletion Ignored in Depletion Clad Creep Down 1 0.7% Clad Creep in SFP 0.7% Clad Creep in SFP Ignored in Depletion Ignored in Depletion Grid Growth 1*2 0.025% Growth in SFP 0% Growth in SFP

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 167 of 300 1

These effects will not be modeled differently for 824 but are highlighted to show that 824 is more benign or equal to what is modeled for the burnup curves.

2 Assembly 824 has lnconel grids. It is generally accepted that lnconel grids expand much less than zirconium based grids. At 21.736 GWD/MTU , ZIRLO experiences little growth so it is assumed that lnconel does not experience any growth .

3 Calculated 4/26/17 Table 9.25 shows that Assembly 824 with 2 nodes of control rod insertion is substantially less reactive than the standard bounding assembly. Therefore , Cycle 1 lead bank assemblies do not require special treatment and may be stored normally in Region 3.

Table 9.25: Modeling Assembly 824 with Measured Design Inputs Decay Case Burnup Enrich.

Time k-eff CJ ~k Description (GWD/MTU) (wt%)

(Yr)

Bounding Model 20 2.90 25 0.96008 0.00007 N/A Assembly 824 21 .739 2.90 29.5 0.95176 0.00007 -0.00833 Model

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 168 of 300 9.9.9 Region 3 k-eff and Margin Calculation Without Decay Time Credit With the exception of temperature changes, tolerance cases for Region 3 are performed at 150

~F with no soluble boron. The 150 °F base case has the highest k-eff of the three temperature cases. Table 9.26 shows the temperature cases as well as uniform and NUREG/CR-6801 axial profile cases for determination of the bounding profile at each burnup.

Table 9.26: Temperature and Burnup Shape Test Cases Burnup Enrich.

Case Description k-eff C1 11k (GWD/MTU) (wt%)

Temperature - 32°F 0 1.70 0.96247 0.00007 -0.00612 Temperature - 68°F 0 1.70 0.96271 0.00007 -0.00587 Temperature - 150°F 0 1.70 0.96859 0.00007 N/A Burnup Shape - NUREG 10 2.20 0.97399 0.00006 -0.00371 Burnup Shape - Uniform 10 2.20 0.97770 0.00007 N/A Burnup Shape - NUREG 20 2.70 0.97130 0.00007 N/A Burnup Shape - Uniform 20 2.70 0.96014 0.00006 -0.01116 Burnup Shape - NUREG 30 3.30 0.96414 0.00006 N/A Burnup Shape - Uniform 30 3.30 0.94629 0.00006 -0.01785 Temperature - 32°F 54 5.00 0.94337 0.00007 -0.00744 Temperature - 150°F 54 5.00 0.95081 0.00006 N/A Table 9.27 identifies the NUREG/CR-6801 Table 5 axial burnup shapes used to determine the TRITON depletion input.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 169 of 300 Table 9.27: SFP Region 3 Limiting Axial Shapes Assembly Average Limiting Shape (Uniform Burnup or NUREG/CR-6801 (GWd/MTU) Table 5 Group) 0 Uniform 10 Uniform*

20 8*

30 5*

40 3 47.8 1 53,54 1

• Uniform and NUREG shapes checked. Limiting shape listed.

Not all uncertainty terms are calculated for each burnup/enrichment combination. A full set of uncertainty cases were run for the O GWD/MTU and 54 GWD/MTU burnup steps. To save computer run time, not all uncertainty cases were run for the intermediate burnup steps (10, 20, 30, 40 and 47.8 GWD/MTU). For the uncertainty cases not specifically run for a particular burnup, the maximum uncertainty value of the next lower or higher burnup is applied to the intermediate burnup steps. Bias terms are calculated for each burnup step.

Table 9.28 shows the results of base case, uncertainty, bias, and margin calculations for Region 3 with no credit for decay time beyond 5 days. Calculated uncertainty and bias values include 2 times the root sum square of the KENO case uncertainties. Shaded values were not calculated at that burnup.

Due to the iterative nature of determining points on the burnup curve with equal margin to the limit, the final base case k-eff at each burnup (excluding 47.8 GWD/MTU) are determined by interpolating between two base cases 0.05 wt% apart. This is acceptable because only the enrichment uncertainty term is a strong function of enrichment. These two base cases (0.05 wt% enrichment difference with their own set of TRITON depletion cases) are used to calculate enrichment sensitivity. At 40 GWd/MTU the final base case k-eff is a slight extrapolation (0.004 wt%). Depletion worth was multiplied by 5% to calculate the depletion worth uncertainty. The burnup measurement uncertainty was calculated by calculating the difference between the worth of the nominal burnup and the worth of 96% of the nominal burnup. The worth of 96% of the nominal burnup is calculated by linear interpolation between adjacent calculated burnup worths.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 170 of 300 Uncertainty values are combined by root sum square. Bias values are added. NRC administrative margin of 0.01 dk is added to the total bias and uncertainty for the margin calculation. Dominion margin is 1.0 minus base case k-eff minus total bias and uncertainty.

For depleted fuel, margin is calculated using fresh fuel and depleted fuel code validation bias and uncertainty. The option with the least margin is shown. The Region 3 burnup credit curve with no decay time credit, which matches or bounds all analyzed points, is shown in Figure 9.19.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 171 of 300 Table 9.28: Region 3 Total Bias, Uncertaint,,, and Margin (0 years decay)

Burnup (GWd/MTU) 0 10 20 30 40 47.8 53.654 Enrichment (wt%) 1.703 2.095 2.599 3.244 4.004 4.555 5.00 Worths (dk)

Minor Actinides and FP 0.0000 0.0515 0.0702 0.0914 0.1120 0.1218 0.1281 Burnup 0.0000 0.0676 0.1340 0.1949 0.2511 0.2811 0.3035 Uncertainties (dk)

Fuel Stack PTO 0.0009 0.0009 0.0009 0.0009 0.0009 0.0009 0.0005 Enrichment, +0.05 0.0094 0.0055 0.0043 0.0036 0.0032 0.0027 0.0000 Pellet OD 0.0005 0.0005 . . 0.0005 ' CI 0.0005 0.0005 o.ooos r 0.0003 Active Fuel Length 0.0002 0.0002 0.0002 0.0002 0.0002 0.0003 0.0002 Clad ID 0.0002 0.0002 0.0002 0.0002 0.0002 0.0002 0.0002 Clad OD 0.0004 NIA NIA ' N/A NIA NIA N/A GTID 0.0002 0.0002 0.0002 0.0002 0.0002 0.0002 0.0002 GTOD 0.0002 o.0002 L 0.0002 0.0002 0.0002 0.0002 0.0002 Pin Pitch 0.0007 0.0009 0.0009 0.0010 0.0010 0.0011 0.0011 Cell Wall Thickness 0.0029 0.0027 0.0025 0.0026 0.0026 0.0026 0.0023 Cell l.D. 0.0000 0.0002 0.0002 0.0002 0.0002 0.0002 0.0002 Cell Pitch 0.0031 0.0033 0.0031 0.0029 0.0030 0.0030 0.0029 Wrapper Thickness 0.0027 0.0026 0.0027 0.0025 0.0025 0.0026 0.0026 Wrapper Channel Thickness 0.0002 0.0003 0.0003 0.0003 0.0003 0.0003 0.0003 Rack Wall Thickness 0.0004 0.0006 0.0006 0.0006 0.0006 0.0006 0.0006 Rack Pitch 0.0005 0.0006 0.0006 0.0006 0.0006 0.0006 0.0006 Depletion Worth Unc. N/A 0.0034 0.0067 0.0097 0.0125 0.0140 0.0152 Burnup Measurement Unc. N/A 0.0027 0.0053 0.0073 0.0090 0.0073 0.0081 Code Benchmarking Unc. 0.0048 0.0083 0.0083 0.0048 0.0048 0.0048 0.0048 KENO Case Uncertainty 0.0001 0.0001 0.0001 0.0001 0.0001 0.0001 0.0001 ASS OF UNCERTAINTIES 0.0118 0.0121 0.0136 0.0145 0.0172 0.0175 0.0185 Biases (dk)

Minor Actinides and Fission N/A 0.0008 0.0011 0.0014 0.0017 0.0018 0.0019 Clad Creep and Grid Growth N/A 0.0006 0.0011 0.0024 0.0041 0.0053 0.0061 Fuel Axial Position 0.0000 0.0000 0.0004 0.0004 0.0004 0.0004 0.0005 Radial Burnup Tilt N/A 0.0018 0.0029 0.0016 0.0020 0.0020 0.0014 Code Temperature Bias 0.0008 0.0008 0.0008 0.0008 0.0008 0.0008 0.0008 Code Benchmarking Bias 0.0034 0.0017 0.0017 0.0034 0.0034 0.0034 0.0034 SUM OF BIASES 0.0042 0.0056 0.0079 0.0099 0.0123 0.0137 0.0141 Summarv Base Case k-eff 0.9691 0.9665 0.9629 0.9606 0.9554 0.9539 0.9523 Total Bias and Uncertainty 0.0159 0.0177 0.0216 0.0244 0.0296 0.0312 0.0326 NRC Administrative Margin 0.0100 0.0100 0.0100 0.0100 0.0100 0.0100 0.0100 Maximum k-eff 0.9951 0.9942 0.9945 0.9950 0.9949 0.9951 0.9949 10CFR50.68 Limit 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 Dominion MarQin (dk) 0.0049 0.0058 0.0055 0.0050 0.0051 0.0049 0.0051

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 172 of 300 Figure 9.19: Region 3 Bounding Burnup Credit Curve (0 Years Decay) 55 --------------------~---------------------  ! .

50 o Analyzed 45 S'

I- -Poly. (Bounding) ****-- * ------------.1 ACCEPTABLE

~ 40 "C

~35 -+-----**-------------------------------------********----------******--**--**********----*********----***-----------------------------)

Q.

J E

J 30 , - - - - - r - - - - - - i r - - - - - y - --:;-;,*'----t--;:======t:::=======::t;-- - - ,

co NOT ACCEPT ABLE g:, 25 . . . . . . - - - - - - - - - - l f - - - - ------+------'----1 1-- - - -

~

~

<( 20 *+---------------------------------------------+

c

~ 15 . ,......,,-------*-*----*-*-*-*-*----*-*-*-*-*-*-!

<(

10 -i-----------~ f------------------------,-------------------------------.--------------------******--t---------------------t-*--****--------------------,

5 ; - - ---r- t - - - - - - i r - - --L!.Y~£}~0~.2~4~5~9x~4~+~4~.2~0~8x~3_-~2~6.~80~x~2 ~+~88~-*~70~.x~-~9~2-~.o~o1_ _- - j 0 ----------------- **************--**-------****--***-***-*-----f----*************--**-*-*-*-*-** -*-*---*-*-----;----*---*-*-*-----*--------------*-*---------,-----------*************---------------------+----------------------------------------- +---------------------------------------------*<

1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Max Planar Average Assembly Initial Enrichment (wt% U-235)

Satisfaction of the burnup curve requirement shall use fuel assembly characteristics as follows:

1) The fuel enrichment is the maximum planar volume averaged as-built initial enrichment in the assembly (blanket enrichment is not credited by averaging with the higher enrichment central zone).

2) The fuel burnup is the volume averaged burnup of the assembly as determined using the measured reaction rates with no reduction for measurement uncertainty.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 173 of 300 9.9.10 Region 3 k-eff and Margin Calculation With Decay Time Credit The same methodology is used to calculate the burnup curves with decay time that was used to calculate the burnup curve without decay time credit. TRITON depletions are performed with a larger decay time step at the end of the depletion. Decay times of 3, 9, 18, and 25 years are used.

Most of the uncertainty and bias values are taken from the no decay time credit analysis.

However, four of the larger uncertainty and bias terms are calculated for each specific burnup:

enrichment uncertainty, depletion worth uncertainty, burnup measurement uncertainty, and 1.5% of minor actinide and fission product worth bias. The other large uncertainty and bias terms are determined from code validation (Appendix A) and are not a function of decay time.

Enrichment uncertainty is a function of enrichment. Allowable enrichment increases with increased decay time for a given burnup on the burnup curve. Burnup worth, which includes the effects of depletion and decay time, increases with increasing decay time. When uncertainties are root sum squared together, the four large terms (enrichment, depletion worth, burnup measurement, and code benchmarking) dominate the total uncertainty.

Table 9.29 shows the base and uncertainty cases run for fuel with the four decay times. Total uncertainty, bias, and margin calculations for Region 3 with decay time credit are provided in Table 9.30 (3 years decay) , Table 9.31 (9 years decay), Table 9.32 (18 years decay) , and Table 9.33 (25 years decay). The calculations follow the same methodology used for Region 3 with no decay time. Shaded values are retained from the no decay time analysis.

The 9, 18, and 25 year decay time burnup curves do not have an analyzed burnup point at >40 GWd/MTU like the 5 day and 3 year burnup curves . The 5 day and 3 year curves show that the points at >40 GWd/MTU are more limiting for curve shape determination than the 40 GWd/MTU point. To ensure that the 9, 18 and 25 year burnup curves are conservative for burnup >40 GWd/MTU, the curvature of the 5 day decay burnup curve will be used for all of the curves.

The decay time credit burnup curves are generated by multiplying the 5 day decay coefficients by a common multiple until all the data points are bounded. This method ensures that the shape of the decay curves is the same as the 5 day decay curve. The conservatism of this approach is confirmed in two ways.

First, as shown in Figure 9.20, using this method introduces progressively higher conservatism (curve burnup higher than analyzed burnup) at 10 and 40 GWd/MTU as decay time is

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 174 of 300 increased. The second confirmation is provided directly on the 3 year burnup curve , which has a 50 GWd/MTU analyzed point that is 0.09 GWd/MTU below the curve .

Fit coefficients are provided in Table 9.34.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 175 of 300 Table 9.29: Uncertainty Cases for Region 3 Decay Time Burnup Curves Decay Burnup Enrich.

Case Description Time k-eff a ilk (GWD/MTU) (wt%)

(Yrs)

Base Case 10 2.15 3 0.96777 0.00006 N/A Interpolation Case 10 2.10 3 0.96243 0.00006 -0.00533 Burnup Worth 0 2.15 N/A 1.04166 0.00008 0.07409 Base Case 20 2.65 3 0.96097 0.00006 N/A Interpolation Case 20 2.70 3 0.96523 0.00007 0.00426 Burnup Worth 0 2.65 N/A 1.10218 0.00007 0.14140 Base Case 30 3.35 3 0.95933 0.00006 N/A Interpolation Case 30 3.40 3 0.96277 0.00006 0.00344 Burnup Worth 0 3.35 N/A 1.16391 0.00007 0.20476 Base Case 40 4.15 3 0.95300 0.00008 N/A Interpolation Case 40 4.20 3 0.95594 0.00006 0.00293 Burnup Worth 0 4.15 N/A 1.21441 0.00007 0.26163 Base Case 50 4.90 3 0.95160 0.00006 N/A Interpolation Case 50 4.95 3 0.95407 0.00006 0.00247 Burnup Worth 0 4.90 N/A 1.25019 0.00008 0.29879 Base Case 10 2.20 9 0.96644 0.00006 N/A Interpolation Case 10 2.15 9 0.96082 0.00006 -0.00562 Burnup Worth 0 2.20 N/A 1.04858 0.00007 0.08234 Base Case 20 2.75 9 0.95994 0.00006 N/A Interpolation Case 20 2.80 9 0.96421 0.00006 0.00427 Burnup Worth 0 2.75 N/A 1.11253 0.00006 0.15277 Base Case 30 3.55 9 0.95982 0.00007 N/A Interpolation Case 30 3.50 9 0.95630 0.00008 -0.00352 Burnup Worth 0 3.55 N/A 1.17811 0.00008 0.21850 Base Case 40 4.45 9 0.95401 0.00007 N/A Interpolation Case 40 4.40 9 0.95086 0.00007 -0.00314 Burnup Worth 0 4.45 N/A 1.22998 0.00007 0.27617 Base Case 10 2.25 18 0.96498 0.00007 N/A Interpolation Case 10 2.20 18 0.95937 0.00006 -0.00561 Burnup Worth 0 2.25 N/A 1.05510 0.00006 0.09030 Base Case 20 2.85 18 0.95979 0.00006 N/A Interpolation Case 20 2.90 18 0.96425 0.00007 0.00446 Burnup Worth 0 2.85 N/A 1.12216 0.00007 0.16256 Base Case 30 3.70 18 0.95820 0.00007 N/A Interpolation Case 30 3.75 18 0.96172 0.00007 0.00352 Burnup Worth 0 3.70 N/A 1.18807 0.00007 0.23006

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 176 of 300 Decay Burnup Enrich.

Case Description Time k-eff (J llk (GWD/MTU) (wt%)

(Yrs)

Base Case 40 4.70 18 0.95358 0.00007 N/A Interpolation Case 40 4.65 18 0.95075 0.00007 -0.00283 Burnup Worth 0 4.70 N/A 1.24152 0.00007 0.28814 Base Case 10 2.30 25 0.96720 0.00006 N/A Interpolation Case 10 2.25 25 0.96170 0.00006 -0.00550 Burnup Worth 0 2.30 N/A 1.06179 0.00006 0.09476 Base Case 20 2.90 25 0.96008 0.00007 N/A Interpolation Case 20 2.95 25 0.96452 0.00006 0.00443 Burnup Worth 0 2.90 N/A 1.12660 0.00007 0.16671 Base Case 30 3.80 25 0.95947 0.00006 N/A Interpolation Case 30 3.85 25 0.96294 0.00007 0.00347 Burnup Worth 0 3.80 N/A 1.19441 0.00007 0.23512 Base Case 40 4.80 25 0.95243 0.00007 N/A Interpolation Case 40 4.85 25 0.95528 0.00007 0.00285 Burnup Worth 0 4.80 N/A 1.24573 0.00007 0.29350

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 177 of 300 Table 9.30: Region 3 Total Bias, Uncertainty, and Margin (3 11ears decay)

Burnup (GWd/MTU) 0 10 20 30 40 50 Enrichment (wt%) 1.703 2.137 2.670 3.363 4.183 4.911 Worths (dk)

Minor Actinides and FP 0.0000 0.0525 0.0685 0.0890 0.1096 0.1213 Burnup 0.0000 0.0741 0.1414 0.2048 0.2616 0.2988 Uncertainties (dk)

Fuel Stack PTO 0.0009 0.0009 0.0009 0.0009 0.0009 0.0005 Enrichment, +0.05 0.0094 0.0055 0.0044 0.0036 0.0031 0.0026 Pellet OD 0.0005' 0.0005 0.0005 0.0005 0.0005 0.0003 Active Fuel Length 0.0002 0.0002 0.0002 0.0002 0.0002 0.0002 Clad ID 0.0002 0.0002 0.0002 0.0002 0.0002 0.0002 Clad OD 0.0004 NIA NIA NIA NIA NIA GTID 0.0002 0.0002 0.0002 0.0002 0.0002 !Ji 0.0002 GTOD 0.0002 0.0002 0.0002 0.0002

• 0.0002 If 0.0002 Pin Pitch 0.0007 0.0009 0.0009 0.0010 0.0010 :e, 0.0011 Cell Wall Thickness 0.0029 0.0027 0.0025 0.0026 0.0026 0.0023 Cell l.D. 0.0000 0.0002 0.0002 0.0002 0.0002 0.0002 Cell Pitch 0.0031 0.0033 0.0031 0.0029 0.0030 .y + 0.0029 Wrapper Thickness 0.0027 0.0026 0.0027 0.0025 0.0025 ' 0.0026 Wrapper Channel Thickness 0.0002 0.0003 0.0003 0.0003 0.0003 0.0003 Rack Wall Thickness 0.0004 0.0006 0.0006 0.0006 0.0006 0.0006 Rack Pitch 0.0005 0.0006 0.0006 0.0006 0.0006 0.0006 Depletion Worth Unc. NIA 0.0037 0.0071 0.0102 0.0131 0.0149 Burnup Measurement Unc. NIA 0.0030 0.0054 0.0076 0.0091 0.0074 Code Benchmarking Unc. 0.0048 0.0083 0.0083 0.0048 0.0048 0.0048 KENO Case Uncertainty 0.0001 0.0001 0.0001 0.0001 0.0002 0.0001 ASS OF UNCERTAINTIES 0.0118 0.0122 0.0139 0.0149 0.0177 0.0182 Biases (dk)

Minor Actinides and Fission Products N/A 0.0008 0.0010 0.0013 0.0016 0.0018 Clad Creep and Grid Growth N/A 0.0006 0.0011 0.0024 0.0041 0.0061 Fuel Axial Position 0.0000 0.0000 0.0004 0.0004 0.0004 0.0005 Radial Burnup Tilt N/A 0.0018 0.0029 0.0016 0.0020 0.0021 Code Temperature Bias 0.0008 0.0008 0.0008 0.0008 0.0008 0.0008 Code Benchmarking Bias 0.0034 0.0017 0.0017 0.0034 0.0034 0.0034 SUM OF BIASES 0.0042 0.0056 0.0079 0.0099 0.0123 0.0145 Summary Base Case k-eff 0.9691 0.9664 0.9627 0.9602 0.9549 0.9521 Total Bias and Uncertainty 0.0159 0.0179 0.0218 0.0248 0.0300 0.0327 NRC Administrative Margin 0.0100 0.0100 0.0100 0.0100 0.0100 0.0100 Maximum k-eff 0.9951 0.9942 0.9945 0.9950 0.9949 0.9949 10CFR50.68 Limit 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 Dominion Marqin (dk) 0.0049 0.0058 0.0055 0.0050 0.0051 0.0051

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 178 of 300 Table 9.31: Region 3 Total Bias, Uncertainty, and Margin (9 years decay)

Burnup (GWd/MTU) 0 10 20 30 40 Enrichment (wt%) 1.703 2.196 2.778 3.547 4.456 Worths (dk)

Minor Actinides and FP 0.0000 0.0526 0.0663 0.0860 0.1061 Burnup 0.0000 0.0823 0.1528 0.2185 0.2762 Uncertainties (dk)

Fuel Stack PTO 0.0009 0.0009 0.0009 0.0009 0.0009 Enrichment, +0.05 0.0094 0.0058 0.0044 0.0037 0.0033 Pellet OD 0.0005 0.0005 0.0005 0.0005 0.0005 Active Fuel Length 0.0002 0.0002 \! o.0002 n+ . 0.0002 0:0002 \

Clad ID 0.0002 0.0002 ' 0.0002 0.0002 0.0002 Clad OD 0.0004 N/A NIA N/A N/A GTID 0.0002 0.0002 0.0002 0.0002 0.0002 GTOD 0.0002 0.0002 0.0002 0.0002 0.0002 Pin Pitch 0.0007 0.0009 0.0009 0.0010 0.0010 Cell Wall Thickness 0.0029 0.0027 0.0025 0.0026 0.0026 Cell l.D. 0.0000 0.0002 0.0002 0.0002 0.0002 Cell Pitch > 0.0031 0.0033 h 0.0031 0.0029 0.0030 /

Wrapper Thickness /. 0.0027 0.0026 I 0,0027 0.0025 0.0025 Wrapper Channel Thickness 0.0002 0.0003 0.0003 0.0003 0.0003 Rack Wall Thickness 0.0004 0.0006 0.0006 0.0006 0.0006 Rack Pitch 0.0005 0.0006 0.0006 0.0006 0.0006 Depletion Worth Unc. N/A 0.0041 0.0076 0.0109 0.0138 Burnup Measurement Unc. N/A 0.0033 0.0056 0.0079 0.0092 Code Benchmarking Unc. 0.0048 0.0083 0.0083 0.0048 0.0048 KENO Case Uncertainty 0.0001 0.0001 0.0001 0.0001 0.0001 RSS OF UNCERTAINTIES 0.0118 0.0126 0.0143 0.0156 0.0183 Biases (dk)

Minor Actinides and Fission Products N/A 0.0008 0.0010 0.0013 0.0016 Clad Creep and Grid Growth NIA 0.0006 0.0011 0.0024 0.0041 Fuel Axial Position 0.0000 0.0000 0.0004 0.0004 0.0004 Radial Burnup Tilt NIA 0.0018 0.0029 0.0016 0.0020 Code Temperature Bias 0.0008 0.0008 0.0008 0.0008 0.0008 Code Benchmarking Bias 0.0034 0.0017 0.0017 0.0034 0.0034 SUM OF BIASES 0.0042 0.0056 0.0079 0.0098 0.0123 Summary Base Case k-eff 0.9691 0.9660 0.9623 0.9596 0.9544 Total Bias and Uncertainty 0.0159 0.0182 0.0222 0.0254 0.0306 NRC Administrative Margin 0.0100 0.0100 0.0100 0.0100 0.0100 Maximum k-eff 0.9951 0.9942 0.9945 0.9950 0.9949 10CFR50.68 Limit 1.0000 1.0000 1.0000 1.0000 1.0000 Dominion Margin (dk) 0.0049 0.0058 0.0055 0.0050 0.0051

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 179 of 300 Table 9.32: Region 3 Total Bias, Uncertainty, and Margin (18 years decay)

Burnup (GWd/MTU) 0 10 20 30 40 Enrichment (wt%) 1.703 2.257 2.874 3.713 4.707 Worths (dk)

Minor Actinides and FP 0.0000 0.0529 0.0640 0.0830 0.1027 Burnup 0.0000 0.0903 0.1626 0.2301 0.2881 Uncertainties (dk)

Fuel Stack PTO 0.0009 0.0009 0.0009 0.0009 0.0009 Enrichment, +0.05 0.0094 0.0058 0.0047 0.0037 0.0030 Pellet OD 0.0005 0.0005 0.0005 0.0005 0.0005 Active Fuel Length 0.0002 0.0002 0.0002 0.0002 0.0002 Clad ID 0.0002 0.0002 0.0002 0.0002 0.0002 Clad OD 0.0004 N/A NIA NIA NIA GTID 0.0002 0.0002 0.0002 0.0002 0.0002 GTOD 0.0002 0.0002 0.0002 0.0002 0.0002 Pin Pitch 0.0007 0.0009 0.0009 0.0010 0.0010 Cell Wall Thickness 0.0029 0.0027 0.0025 0.0026 0.0026 Cell l.D. 0.0000 0.0002 0.0002 0.0002 0.0002 Cell Pitch 0.0031 0.0033 0.0031 0.0029 0.0030 Wrapper Thickness 0.0027 0.0026 0.0027 0.0025 0.0025 Wrapper Channel Thickness 0.0002 0.0003 0.0003 0.0003 0.0003 Rack Wall Thickness 0.0004 0.0006 0.0006 0.0006 0.0006 Rack Pitch 0.0005 0.0006 0.0006 0.0006 0.0006 Depletion Worth Unc. N/A 0.0045 0.0081 0.0115 0.0144 Burnup Measurement Unc. N/A 0.0036 0.0058 0.0081 0.0093 Code Benchmarking Unc. 0.0048 0.0083 0.0083 0.0048 0.0048 KENO Case Uncertainty 0.0001 0.0001 0.0001 0.0001 0.0001 ASS OF UNCERTAINTIES 0.0118 0.0128 0.0147 0.0161 O.Q187 Biases (dk)

Minor Actinides and Fission Products N/A 0.0008 0.0010 0.0012 0.0015 Clad Creep and Grid Growth NIA 0.0006 0.0011 0.0024 0.0041 Fuel Axial Position 0.0000 0.0000 0.0004 0.0004 0.0004 Radial Burnup Tilt NLA 0.0018 0.0029 0.0016 Q.0020 Code Temperature Bias 0.0008 0.0008 0.0008 0.0008 0.0008 Code Benchmarking Bias 0.0034 0.0017 0.0017 0.0034 0.0034 SUM OF BIASES 0.0042 0.0056 0.0078 0.0098 0.0122 Summary Base Case k-eff 0.9691 0.9658 0.9619 0.9591 0.9540 Total Bias and Uncertainty 0.0159 0.0184 0.0225 0.0259 0.0309 NRG Administrative Margin 0.0100 0.0100 0.0100 0.0100 0.0100 Maximum k-eff 0.9951 0.9942 0.9945 0.9950 0.9949 10CFR50.68 Limit 1.0000 1.0000 1.0000 1.0000 1.0000 Dominion Margin (dk) 0.0049 0.0058 0.0055 0.0050 0.0051

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 180 of 300 Table 9.33: Region 3 Total Bias, Uncertainty, and Margin (25 years decay)

Burnup (GWd/MTU) 0 10 20 30 40 Enrichment (wt%) 1.703 2.287 2.920 3.792 4.824 Worths (dk)

Minor Actinides and FP 0.0000 0.0528 0.0629 0.0815 0.1007 Burnup 0.0000 0.0948 0.1667 0.2351 0.2935 Uncertainties (dk)

Fuel Stack PTO 0.0009 0.0009 0.0009 0.0009 0.0009 Enrichment, +0.05 0.0094 0.0057 0.0046 0.0037 0.0030 Pellet OD 0.0005 0.0005 0.0005 0.0005 0.0005 Active Fuel Length 0.0002 0.0002 0.0002 0.0002 0.0002 Clad ID 0.0002 0.0002 0.0002 0.0002 0.0002 Clad OD 0.0004 N/A NIA NIA NIA GTID 0.0002 0.0002 0.0002 0.0002 0.0002 GTOD 0.0002 0.0002 0.0002 \ ().0002 0.0002 Pin Pitch 0.0007 0.0009 0.0009 0.0010 0.0010 Cell Wall Thickness 0.0029 0.0027 0.0025 0.0026 0.0026 Cell l.D. 0.0000 0.0002 0.0002 0.0002 0.0002 Cell Pitch 0.0031 , 0.0033v 0.0031 0.0029 0.0030 Wrapper Thickness 0.0027 0.0026 . . . 0.0027 0.0025 0.0025 Wrapper Channel Thickness 0.0002 0.0003 0.0003 0.0003 0.0003 Rack Wall Thickness 0.0004 0.0006 0.0006 0.0006 0.0006 Rack Pitch 0.0005 0.0006 0.0006 0.0006 0.0006 Depletion Worth Unc. NIA 0.0047 0.0083 0.0118 0.0147 Burnup Measurement Unc. NIA 0.0038 0.0058 0.0082 0.0093 Code Benchmarking Unc. 0.0048 0.0083 0.0083 0.0048 0.0048 KENO Case Uncertainty 0.0001 0.0001 0.0001 0.0001 0.0001 ASS OF UNCERTAINTIES 0.0118 0.0129 0.0148 0.0163 0.0190 Biases (dk)

Minor Actinides and Fission Products NIA 0.0008 0.0009 0.0012 0.0015 Clad Creep and Grid Growth N/A 0.0006 0.0011 0.0024 0.0041 Fuel Axial Position 0.0000 0.0000 0.0004 0.0004 0.0004 Radial Burnup Tilt N/A 0.0018 0.0029 0.0016 0.0020 Code Temperature Bias 0.0008 0.0008 0.0008 0.0008 0.0008 Code Benchmarking Bias 0.0034 0.0017 0.0017 0.0034 0.0034 SUM OF BIASES 0.0042 0.0056 0.0078 0.0097 0.0122 Summary Base Case k-eff 0.9691 0.9658 0.9619 0.9589 0.9538 Total Bias and Uncertainty 0.0159 0.0185 0.0226 0.0261 0.0311 NRC Administrative Margin 0.0100 0.0100 0.0100 0.0100 0.0100 Maximum k-eff 0.9951 0.9943 0.9945 0.9950 0.9949 10CFR50.68 Limit 1.0000 1.0000 1.0000 1.0000 1.0000 Dominion Margin (dk) 0.0049 0.0057 0.0055 0.0050 0.0051

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 181 of 300 Figure 9.20: Region 3 Bounding Burnup Credit Curves 55 - , - - - - - ~ - - - - ~ - - - ~ - - - - ~ - - - ~ - - - - ~ - - - ~

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1.5 2.0 2.5 3.0 3.5 4 .0 4.5 5.0 Max Planar Average Assembly Initial Enrichment (wt% U-235)

The burnup curve equations have the following polynomial format:

BU [GWD/MTU] = a 4

• wt% 4 + a3

• wto/o 3 + a 2

• wt% 2 + a 1

• wt% 1 + a0

• wt% 0 wt%: maximum planar average, as-built U-235 enrichment Table 9.34: Region 3 Burnup Credit Curve Coefficients Decay T ime Credit a4 a3 a2 a1 ao No Credit -0.2459 4.208 -26.80 88.70 -92.00 3 Years -0.2338 4.001 -25.48 84.34 -87.47 9 Years -0.2153 3.684 -23.46 77.66 -80.54 18 Years -0.2020 3.458 -22.02 72.88 -75 .59 25 Years -0.1964 3.361 -21 .40 70.84 -73.47

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 182 of 300 Satisfaction of the burnup curve requirement shall use fuel assembly characteristics as follows:

1) The fuel enrichment is the maximum planar volume averaged as-built initial enrichment in the assembly (blanket enrichment is not credited by averaging with the higher enrichment central zone) .

2) The fuel burnup is the volume averaged burnup of the assembly as determined using the measured reaction rates with no reduction for measurement uncertainty.

3) The fuel assembly decay time (time elapsed since last use at power in the reactor core) is greater than or equal to the applicable burnup curve .

9.1 O Region 3 Analysis Summary Region 3 has been demonstrated to satisfy the SFP k-eff requirement for spent fuel pool analyses as follows:

• k-eff < 1.0 if flooded with unborated water (95% probability, 95% confidence)

In Region 3 with 4-out-of-4 storage, the requirement is met with either a restrictive loading curve (burnup credit curve) of minimum required assembly measured burnup versus initial fuel assembly as-built enrichment (maximum planar volume averaged as-built initial enrichment in the assembly) and with minimum decay time.

Region 3 analysis includes calculation and application of bias and uncertainty as well as identification of NRC administrative margin and Dominion margin to the k-eff limit. For Region 3, five 4th order polynomial burnup credit curves are determined for fuel with initial enrichment

>1 .703 wt% U-235 (Table 9.34).

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 183 of 300 10 Interface Analysis 10.1 Interfaces Between Storage Configurations Within a Rack Only Region 1 has two different sub-regions with different storage requirements within the same rack. The Region 1A/ 1B interface is directly modeled and dispositioned in Section 9.3.

10.2 Interfaces Between Dissimilar Racks Analysis of Region interfaces will determine if the proximity of one Region to another affects the conclusions of the individual Region infinite lattice analyses. The MPS3 interface analysis confirms that interface effects do not reduce the minimum margin to the k-eff limit determined for Regions 1, 2, and 3. Part of that confirmation is justification of appropriate interface model bias and uncertainty. Reference 4 provides this guidance for analysis of storage configuration interfaces:

Interfaces: For applications that contain more than a single storage configuration, in order to ensure that the regulatory requirement for keff to be known with a 95 percent probability at a 95 percent confidence level is met the NCS analysis should consider the interface between storage configurations. Given all the combinations that are in existence, it is impossible to predict all of the combinations that could be proposed.

Therefore, the staff should verify that each application includes a portion of the analysis that demonstrates that the interface analysis used is appropriate for its specific conditions.

i. Absent a determination of a set of biases and uncertainties specifically for the combined interface model, use of the maximum biases and uncertainties from the individual storage configurations should be acceptable in determining whether the keff of the combined interface model meets the regulatory requirements.

For some interfaces, use of the maximum total bias and uncertainty from the individual Regions is adequate. This is the simplest approach. For some interfaces, a set of biases and uncertainties are developed specifically for the interface configuration.

In the interface analysis, the KENO best estimate k-eff for a model representing the nominal fuel and rack design is referred to as "base case k-eff". The k-eff that is compared to the regulatory limit that includes allowance for total bias and uncertainty is referred to as the 95/95 k-eff.

KENO rack models are developed using Region 1, 2, and 3 infinite lattice models. Depleted fuel isotopic content is the same as used for Region 1, 2, and 3 infinite lattice analyses.

L_ _

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 184 of 300 10.2.1 MPS3 Region Interfaces Region interfaces can be seen in Figure 4.3.

1) Region 2 meets Region 1

2) Region 2 meets Region 3 10.2.2 Interface Analysis Method Interface analysis methodology is summarized in the steps below. Refer to Figures 10.1 and 10.3 for an illustration of the KENO interface models.

• Develop a KENO interface model with two Regions (one on each side of the interface).

• Choose the number of storage cells in each Region along the interface to minimize model gaps at the model boundary due to different cell pitch on either side of the interface. This is intended to maintain fidelity to the infinite lattice analysis as closely as possible.

• Use reflective or periodic boundary conditions as appropriate to maintain fidelity with infinite lattice analysis.

• Compare the interface model k-eff to the infinite lattice analysis "base case" k-eff for each Region.

• For a Region with burnup credit, perform interface calculations with fresh fuel and with maximally depleted fuel to capture potential axial shape effects at the interface.

• For two Regions with different bounding temperature , model the interface at both bounding temperatures.

There are several possible results in the interface analysis:

1) Interface model k-eff is greater than both infinite lattice Region base case k-effs. The interface is "adverse" because it increases k-eff as compared to both infinite lattice analyses. The infinite lattice 95/95 k-eff margin calculations are not bounding for the interface model. The MPS3 analysis has no adverse interfaces.

2) Interface model k-eff is less than both infinite lattice Region base case k-effs. The interface is "benign", meaning the interface reduces k-eff as compared to both infinite lattice analyses. The infinite lattice margin calculations are bounding. No additional analysis is needed. If the highest Region total bias and uncertainty is added to the interface model k-eff, the 95/95 k-eff is lower than the infinite lattice Region 95/95 k-eff.

3) Interface model k-eff is less than one infinite lattice Region base case k-eff and greater than the other. This is an "indeterminate" interface.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 185 of 300 With an indeterminate interface, neutron leakage across the interface from the higher reactivity Region (Region "A") into the lower reactivity Region (Region "B") results in a base case k-eff less than Region A and greater than Region B. It is not clear what total bias and uncertainty value should be applied to the interface model to determine the interface model 95/95 k-eff.

One way to determine the 95/95 k-eff for an indeterminate interface is to use the higher total bias and uncertainty of Region A or Region B for the interface model. This typically results in an interface model 95/95 k-eff higher than either infinite lattice Region analysis. This approach may be unnecessarily conservative and result in apparent margin to the limit that may be insufficient.

Alternatively, the total bias and uncertainty of the interface model can be directly calculated. In the calculation of total bias and uncertainty, there are typically a few dominant terms. For the purpose of determining interface model bias and uncertainty, it is not necessary to calculate all bias and uncertainty items. MPS3 indeterminate interfaces are evaluated by calculating dominant k-eff tolerance values for each Region in the interface model independently and comparing them to the same tolerance values calculated for the infinite lattice Region analyses.

This comparison is used to determine Region weighting factors, which are then used to calculate a weighted total bias and uncertainty value for the interface model.

10.3 Region 1-2 Interface Region 1 and Region 2 are both most reactive at 32 °F. Interface cases are run at 32 °F.

Region 1 is represented by the infinite lattice model (4.75 wt% fresh fuel with no SFP wall neutron leakage boundary), which bounds the k-eff of the wall credit model. Due to the rack placement in the SFP, only the 4.75 w/o fresh fuel side of Region 1 has an interface with Region

2. Region 1 with 5.0 wt% fuel (Table 9.6) requires burnup credit of 2 GWd/MTU, which is well below the burnup at which non-uniform axial burnup shapes become more reactive than uniform shapes. The 5.0 wt% 2 GWd/MTU condition has over 0.01 dk more margin than the fresh fuel case and the 12 IFBA rods condition has over 0.02 dk more margin than the fresh fuel case.

Because of these features, only fresh 4.75 wt% fuel is used to represent Region 1 in the interface models.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 186 of 300 Region 2 burnup credit varies from O MWd/MTU at 2.03 wt% to 40.25 MWd/MTU at 5.0 wt% .

To investigate possible interface axial burnup shape interference effects, Region 1-2 interface models include 2.03 wt% fresh fuel and 40 GWd/MTU 5.0 wt% depleted fuel (NUREG axial burnup shape).

The Region 1-2 interface model is formed from a 9x9 cell Region 1 adjacent to a 1Ox10 cell Region 2 (Figure 10.1 ). These array sizes are chosen to closely match the overall dimension of each Region along the common interface (prevents non-physical water gaps across the periodic boundary of the model). Note that the Region 1 cell pitch in the x-coordinate direction is different than cell pitch in they-coordinate direction. The longer pitch is perpendicular to the interface.

Figure 10.1: X-Y Representation of the Region 1-2 Interface Model With periodic boundary conditions, the Region 1-2 model represents infinite "columns" of Region 1 and 2 cells in the Y dimension with a very small 0.17 inch water gap between groupings of 9 Region 1 cells (this occurs because the Region 2 1Ox1 O group is 0.17 inches longer in the y-direction than the Region 1 9x9 group). The gap is kept as small as possible to maintain fidelity between the infinite lattice Region analysis and the interface models. Only the intra-rack gap resulting from the water at the edge of the infinite lattice models (a result of modeling normal cell pitch) is credited. No additional intra-rack spacing is added for the nominal 1.5 inch spacing between Region 1 and 2.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 187 of 300 The interface model has 0.62 inches between the Region 1 and Region 2 wrappers (Figure 1O.

2). The rack baseplates extend a total of 1.5 inches outside the faces of the outer storage cell walls. Accounting for the extension of the Boral wrappers beyond the cell wall leads to a minimum design spacing between adjacent wrappers across the Region 1-2 interface of 1.2 inches. The model is conservative because it has rack cells - 0.6 inch closer together than is physically possible and the flux trap between Region 1 and 2 has less water gap than actual.

Periodic boundary conditions are used in the X and Y direction. Reflective boundary conditions are used in the Z direction.

Note that depleted fuel base case k-eff for the Region 2 infinite lattice model in this section does not match the k-eff in Table 9.19. Region 2 interface model depleted fuel cases are modeled with Zirc2 rather than Zr clad material. To maintain modeling consistency in the comparison of infinite lattice k-eff and interface model k-eff, the Region 2 infinite lattice k-eff with Zirc2 clad modeling is reported. Region 2 sensitivity cases were also run with Zirc-2 cladding , therefore the sensitivity comparison is also consistent.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 188 of 300 Figure 10.2: X-Y Representation of the Region 1-2 Interface Showing Flux Traps

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 189 of 300 10.3.1 Region 1-2 Interface with Region 2 Fresh Fuel Table 10.1 summarizes infinite lattice and interface model results with fresh fuel in both Regions. All cases have fuel centered in the rack cells which maximizes k-eff in Regions 1 and

2. The interface model k-eff is bounded by the infinite lattice analysis k-eff of Region 1 and Region 2. The final case in Table 10.1 confirms that modestly increasing intra-rack spacing reduces k-eff. No additional Region 1-2 interface analysis is needed for this condition.

Table 10.1: Region 1-2 Interface with Fresh Fuel Enrichment Region k-eff Uncert. Note (wt% U235) 4.75 I NIA 1 0.96745 0.00008 lnifinite lattice Reg. 1 NIA I 2.03 2 0.96872 0.00006 lnifinite lattice Reg. 2 4.75 I 2.03 1-2 0.96661 0.00007 Region 1-2 interface 4.75 I 2.03 1-2 0.96426 0.00007 Increase intra-rack spacing 2 cm 10.3.2 Region 1-2 Interface with Region 2 Depleted Fuel Table 10.2 summarizes infinite lattice and interface model results with depleted fuel (5.0 wt%,

40 GWdlMTU) in Region 2. All cases have fuel centered in the rack cells which maximizes k-eff in Regions 1 and 2. The first two cases are infinite lattice cases for comparison. The interface model k-eff is not bounded by the Region 2 infinite lattice k-eff. This result is expected because of the relatively large infinite lattice reactivity difference between Region 1 and 2.

Table 10.2: Region 1-2 Interface (Region 1 Fresh Fuel and Region 2 Depleted Fuel)

Enrichment and Burnup Region k-eff Uncert. Note 4.75 OG I NIA 1 0.96745 0.00008 lnifinite lattice Reg. 1 NIA I 5.0 40G 2 0.95678 0.00006 lnifinite lattice Reg. 2 4.75 I 5.0 40G 1-2 0.96384 0.00007 Reqion 1-2 interface Because the interface result is not bounded by the infinite lattice analysis k-eff of Region 2, there is a possibility that the interface model has less margin to the k-eff limit than indicated in the Region 2 infinite lattice analysis. Table 10.3 documents interface model tolerance cases.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 190 of 300 These will be used to compare geometric and fuel content sensitivity of the interface model to analogous infinite lattice Region sensitivities in Table 10.4. The tolerances in Table 10.3 are selected from the most significant tolerances identified in the infinite lattice analysis.

Table 10.3: Region 1-2 Interface Model Selected Tolerances (Fresh/Depleted)

Enrichment and Burnup Region k-eff Uncert. dk Note 4.75 I 5.0 40G 1-2 0.96384 0.00007 N/A Base case 4.75 I 5.0 41G 1-2 0.96366 0.00008 -0.0002 Reg. 2 burnup increase 4.75 I 5.0 40G 1-2 0.96377 0.00007 -0.0001 Reg . 2 narrow BORAL 4.75 I 5.0 40G 1-2 0.96377 0.00007 -0.0001 Reg. 2 growth and creep Reg. 1 enrichment 4.80 I 5.0 40G 1-2 0.96561 0.00008 0.0018 increase +0.05 wt%

4.75 I 5.0 40G 1-2 0.95794 0.00008 -0.0059 Reg. 1 cell ID change Region 1 clad OD 4.75 I 5.0 40G 1-2 0.96563 0.00008 0.0018 chanae Table 10.4 compares Region 1 and Region 2 infinite lattice tolerance sensitivity and tolerance sensitivity of the interface model. All tolerance values are best estimate (t.k-eff). The relative sensitivity of the interface model to perturbations in each Region is calculated as (Interface tolerance t.k / Infinite lattice t.k, %). The weighted average sensitivity of the interface model for each Region is calculated using as follows (infinite lattice tolerance is the absolute value):

Weighted Importance= L(infinite lattice tolerance x relative sensitivity)i / rinfinite lattice tolerancei It is clear for this interface condition (fresh fuel Region 1, depleted fuel Region 2) that the interface model is dominated by Region 1 and is insensitive to perturbations in Region 2.

Therefore , the Region 1 total bias and uncertainty may appropriately be applied to the interface model. Table 10.5 presents the interface model margin calculation.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 191 of 300 Table 10.4: Region 1-2 Interface Model Tolerance Comparison (Fresh/ Depleted)

Reg.1 Reg.2 Tolerance Interface tol. I Tolerance case Model Enrich./ Enrich./

(k - k base)* Inf. lattice tol.

Burnup** Burnup**

Increase enrich Interface 1-2 4.80 IO 5.0 I 40 0.0018 89%

0.05 Reg1 Reqion 1 4.80 I 0 N/A 0.0020 Reduce cell ID Interface 1-2 4.75 IO 5.0 I 40 -0.0059 100%

Reg1 Reqion 1 4.75 IO N/A -0.0059 Reduce clad OD Interface 1-2 4.75 IO 5.0 I 40 0.0018 97%

Reg1 Region 1 4.75 I 0 N/A 0.0019 Weighted average Region 1 importance 97%

Increase BU 1 Interface 1-2 4.75 IO 5.0 I 41 -0.0002 4%

GWd/MTU Reg2 Region 2 N/A 5.0 I 41 -0.0041 Narrow BORAL Interface 1-2 4.75 IO 5.0 I 40 -0.0001

-3%

Reg 2 Reqion 2 N/A 5.0 I 40 0.0022 Interface 1-2 4.75 I 0 5.0 I 40 -0.0001 Growth Reg 2 -1%

Region 2 N/A 5.0 I 40 0.0048 Weighted average Region 2 importance 0%

• Best estimate dk (does not include statistical uncertainty)

• • Enrichment is wt% U-235, Burnup is GWD/MTU L

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 192 of 300 Table 10.5: Region 1-2 Interface Model Margin (Fresh/ Depleted)

Region 1 total bias and uncertainty 0.0247 (includes 0.01 dk NRC admin marain)

Region 2 total bias and uncertainty 0.0380 (includes 0.01 dk NRC admin marqin)

Region 1 importance 100%

Reqion 2 importance 0%

Interface model base k-eff 0.9638 Interface model bias and uncertainty 0.0247 (weiohted total bias and uncertainty)

Interface model 95/95 k-eff 0.9885 (interface k-eff + wtd. total bias+ uncert.)

(greater than infinite lattice Reg 1 and Interface model marqin to k-eff limit 0.0115 Req 2)

Adequate margin to k-eff limit? YES (bounded bY infinite lattice analysis)

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 193 of 300 10.4 Region 2-3 Interface Region 2 is most reactive at 32 F. Region 3 is most reactive at 150 F. Interface cases are run at both temperatures. Both Regions have substantial burnup requirements for all but very low enrichment fuel, so fresh fuel and high burnup fuel is considered in each Region.

The Region 2-3 interface models are formed from a 1Ox14 or 14x14 cell Region 2 and two 6x6 cell Region 3 racks (Figure 10.3). Region 3 racks are 6x6 enclosed by stainless steel on all four sides and are modeled as installed with a small (-1 inch) gap between adjacent modules. The 14 Region 2 cells match the 2 Region 3 racks along the interface within 0.5 inch.

Figure 10.3: X-Y Representation of the Region 2-3 Interface Model With periodic X-Y boundary conditions, the Region 2-3 model represents infinite columns of Region 3 racks and Region 2 cells in the Y dimension.

In the X dimension at the rack interface, only the intra-rack gap resulting from the water at the edge of the infinite lattice models (a result of modeling normal cell pitch) is credited, resulting in

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 194 of 300 a 0.565 inch gap. No additional intra-rack spacing is added to credit the 1.28 inch spacing between Region 2 and 3 racks. Reflective boundary conditions are used in the Z direction.

Note that the depleted fuel base case k-eff for the Region 2 infinite lattice model in this section does not match the k-eff in Table 9.19. Region 2 interface model depleted fuel cases are modeled with Zirc2 rather than Zr clad material. To maintain modeling consistency in the comparison of infinite lattice k-eff and interface model k-eff, the Region 2 infinite lattice k-eff with Zirc2 clad modeling is reported. Region 2 sensitivity cases were also run with Zirc-2 cladding, therefore the sensitivity comparison is also consistent.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 195 of 300 10.4.1 Region 2-3 Interface (Fresh / Fresh)

Table 10.6 summarizes infinite lattice and interface model results with fresh fuel in both Regions. Region 2 (1Ox14) has fuel centered in the rack cells to maximize k-eff. Region 3 has a 4x4 cell section of each rack module asymmetrically loaded towards the rack center to maximize k-eff. The interface model k-eff is bounded by the maximum infinite lattice analysis k-eff of Region 1 and Region 2. The final case in Table 10.6 confirms that modestly increasing intra-rack spacing reduces k-eff. No additional Region 2-3 interface analysis is needed for this condition.

Table 10.6: Region 2-3 Interface (Fresh/ Fresh)

Enrichment Region k-eff Uncert. Note (wt% U235) 2.03 I NIA 32F Reg 2 0.96872 0.00006 Infinite lattice Reg. 2 2.03 IN/A 150FReg2 0.95676 0.00006 Infinite lattice Reg. 2 N/A/1.70 32F Reg 3 0.96247 0.00007 Infinite lattice Reg. 3 N/A/1 .70 150FReg3 0.96859 0.00007 Infinite lattice Reg. 3 2.03 I 1.70 32F Reg 2-3 0.96420 0.00006 Region 2-3 Interface 2.03 I 1.70 150F Reg 2-3 0.96686 0.00007 Region 2-3 Interface 2.03 I 1.70 150F Reg 2-3 0.96628 0.00007 2cm larger intra-rack Reg 2-3 gap 10.4.2 Region 2-3 Interface (Fresh / Depleted)

Table 10.7 summarizes infinite lattice and interface model results with depleted fuel (5.0 w/o, 54 GWd/MTU) in Region 3. Region 2 (14x14) has fuel centered in the rack cells to maximize k-eff.

Region 3 has a 4x4 cell section of each rack module asymmetrically loaded to maximize k-eff.

The first two cases are infinite lattice cases for comparison. The interface model k-eff is higher than the Region 3 infinite lattice k-eff. This result is expected because of the relatively large infinite lattice reactivity difference between Region 2 and 3.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 196 of 300 Table 10.7: Region 2-3 Interface (Fresh/ Depleted)

Enrich. and Region k-eff Uncert. Note Burnup*

2.03 OG IN/A 32F Reg 2 0.96872 0.00006 Infinite lattice Req. 2 2.03 OG IN/A 150FReg2 0.95676 0.00006 Infinite lattice Reg. 2 N/A/ 5.0 54G 32F Reg 3 0.94337 0.00007 Infinite lattice Req. 3 N/A/ 5.0 54G 150F Reg 3 0.95081 0.00006 Infinite lattice Reg. 3 2.03 OG I 5.0 54G 32F Reg 2-3 0.96589 0.00007 Fresh Reg. 2 / Max BU Req 3 2.03 OG I 5.0 54G 150F Reg 2-3 0.95386 0.00006 Fresh Req. 2 / Max BU Req 3

• Enrichment is w/o U-235; burnup is GWd/MTU The interface model is most reactive at 32 F. Because the interface k-eff (0.96589) is greater than the infinite lattice analysis k-eff of Region 3 (0.95081 ), there is a possibility that the interface model has less margin to the k-eff limit than indicated in the Region 3 infinite lattice analysis . Table 10.8 documents interface model tolerance cases. These will be used to compare geometric and fuel content sensitivity of the interface model to infinite lattice Region sensitivities in Table 10.9. The tolerances in Table 10.8 are selected from the most significant tolerances identified in the infinite lattice analysis. All cases are run at the limiting interface model temperature (32 °F) .

Table 10.8: Region 2-3 Interface Model Selected Tolerances (Fresh/ Depleted)

Enrichment and Burnup* Region k-eff Uncert. dk Note 2.03 OG I 5.0 54G 2-3 0.96589 0.00006 N/A Base case Reg 3 reduce burnup 1 2.03 OG I 5.0 53G 2-3 0.96584 0.00007 -0.0001 GWd/MTU Reg 3 creep and grid 2.03 OG I 5.0 54G 2-3 0.96588 0.00006 0.0000 growth Reg 2 reduce enrichment 2.0 OG I 5.0 54G 2-3 0.96153 0.00006 -0 .0044 0.03 wt%

• Enrichment is w/o U-235; burnup is GWd/MTU

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 197 of 300 Table 10.9 compares infinite lattice tolerance sensitivity from the infinite lattice analyses (Region 2 and Region 3) and tolerance sensitivity of the interface model. All infinite lattice values are calculated at the bounding temperature for that Region. Interface model tolerances are calculated at the limiting interface model temperature. All tolerance values are best estimate

(~k-eff) .

Table 10.9: Region 2-3 Interface Model Tolerance Comparison (Fresh / Depleted)

Reg.2 Reg. 3 Tolerance Tolerance Interface tol. /

Model Enrich./ Enrich./ (k- k case Inf. lattice tol.

Burnup** Burnup** ' base)*

Reduce Interface 2-3 2.0 IO 5.0 I 54 -0.0044 99%

enrich 0.03 Reqion 2 2.0 I 0 N/A -0.0044 Weighted average Region 2 importance 99%

Reduce BU 1 Interface 2-3 2.03 I 0 5.0 I 53 -0.0001

-1 %

GWd/MTU Region 3 N/A 5.0 I 53 0.0044 Interface 2-3 2.03 IO 5.0 I 54 0.0000 Growth 0%

Reqion 3 N/A 5.0 I 54 0.0061 Weighted average Region 3 importance -1%

• Best estimate dk (does not include statistical uncertainty)

• • Enrichment is wt% U-235, Burnup is GWD/MTU At this interface condition (fresh fuel Region 2, depleted fuel Region 3) the interface model is dominated by Region 2 and is insensitive to perturbations in Region 3. Therefore , it is appropriate to apply the Region 2 infinite lattice analysis total bias and uncertainty to the interface model. Table 10.1 O presents the interface model margin calculation.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 198 of 300 Table 10.10: Region 2-3 Interface Model Margin (Fresh / Depleted)

Region 2 total bias and uncertainty 0.0233 (includes 0.01 dk NRG admin margin)

Region 3 total bias and uncertainty 0.0426 (includes 0.01 dk NRG admin marqin)

Region 2 importance 100%

Reqion 3 importance 0%

Interface model base k-eff 0.9659 Interface model bias and uncertainty 0.0233 (weiqhted total bias and uncertainty)

Interface model 95/95 k-eff 0.9892 (interface k-eff + wtd. total bias+ uncert.)

(greater than infinite lattice Reg 2 and Interface model margin to k-eff limit 0.0108 Reg 3)

Adequate margin to k-eff limit? YES (bounded by infinite lattice analysis) 10.4.3 Region 2-3 Interface (Depleted / Fresh)

Table 10.11 summarizes infinite lattice and interface model results with depleted fuel (5 .0 w/o ,

40 GWd/MTU) in Region 2. Region 2 (14x14) has fuel centered in the rack cells to maximize k-eff. Region 3 has a 4x4 cell section of each rack module asymmetrically loaded to maximize k-eff. The first four cases are infinite lattice cases for comparison. The interface model k-eff is not bounded by the Region 2 infinite lattice k-eff. This result is expected because of the relatively large infinite lattice reactivity difference between Region 2 and 3.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 199 of 300 Table 10.11: Region 2-3 Interface (Depleted/ Fresh)

Enrich. and Region k-eff Uncert. Note Burnup*

5.0 40G I NIA 32F Reg 2 0.95678 0.00006 Infinite lattice Reg. 2 5.0 40G I NIA 150F Reg 2 0.94838 0.00006 Infinite lattice Reg. 2 NIAl 1.700G 32F Req 3 0.96247 0.00007 infinite lattice Req. 3 NIAl1.700G 150F Reg 3 0.96859 0.00007 infinite lattice Reg. 3 5.0 40G I Interface Max BU Reg. 2 I Fresh 1.70 OG 32F Req 2-3 0.96081 0.00007 Req 3 5.0 40G I Interface Max BU Reg. 2 I Fresh 1.70 OG 150F Reg 2-3 0.96650 0.00006 Req 3

• Enrichment is w/o U-235; burnup is GWd/MTU The interface model is most reactive at 150 F. Because the interface k-eff (0.96650) is greater than the infinite lattice analysis k-eff of Region 2 (0.95678), there is a possibility that the interface model has less margin to the k-eff limit than indicated in the Region 2 infinite lattice analysis. Table 10.12 documents interface model tolerance cases. They will be used to compare geometric and fuel content sensitivity of the interface model to infinite lattice Region sensitivities in Table 10.13. The tolerances in Table 10.14 are selected from the most significant tolerances identified in the infinite lattice analysis. All cases are run at the limiting interface model temperature ( 150 F) .

Table 10.12: Region 2-3 Interface Model Selected Tolerances (Depleted / Fresh)

Enrichment and Burnup* Region k-eff Uncert. dk Note 5.0 40G I 1.70 OG 150F Req 2-3 0.96650 0.00006 NIA Base case Reg 3 increase enrichment 0.05 5.0 40G I 1.75 OG 150F Req 2-3 0.97574 0.00006 0.0092 wt%

Reg 2 Increase burnup 1 5.041G l 1.700G 150F Req 2-3 0.96644 0.00007 -0.0001 GWdlMTU 5.0 40G I 1.70 OG 150F Reg 2-3 0.96661 0.00007 0.0001 Reg 2 Creep and grid growth

• Enrichment is w/o U-235; burnup is GWd/MTU

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 200 of 300 Table 10.13 compares infinite lattice tolerance sensitivity from the infinite lattice analyses (Region 2 and Region 3) and tolerance sensitivity of the interface model. All infinite lattice values are calculated at the bounding temperature for that Region. Interface model tolerances are calculated at the limiting interface model temperature. All tolerance values are best estimate (Llk-eff).

Table 10.13: Region 2-3 Interface Model Tolerance Comparison (Depleted/ Fresh)

Reg.2 Reg.3 Tolerance Tolerance Interface tol. I Model Enrich./ Enrich./ (k - k case Inf. lattice tol.

Burnup** Burnup** base)*

Increase BU 1 Interface 2-3 5.0 I 41 1.70 I 0 -0.0001 1%

GWd/MTU Region 2 5.0 I 41 N/A -0.0041 Grid growth Interface 2-3 5.0 I 40 1.70 I 0 0.0001 2%

and clad creep Reqion 2 5.0 I 40 N/A 0.0048 Weighted average Region 2 importance 2%

Increase enrich Interface 2-3 5.0 I 40 1.75 / 0 0.0092 101%

0.05 Reqion 3 N/A 1.75 / 0 0.0092 Weighted average Region 3 importance 101%

• Best estimate dk (does not include statistical uncertainty)

• • Enrichment is wt% U-235, Burnup is GWD/MTU At this interface condition (depleted fuel Region 2, fresh fuel Region 3) the interface model is dominated by Region 3 and is insensitive to perturbations in Region 2. The Region 2 tolerance values (dk) are essentially zero (approximately one RSS sigma). Therefore, it is appropriate to apply the Region 3 infinite lattice analysis total bias and uncertainty to the interface model.

Table 10.14 presents the interface model margin calculation.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 201 of 300 Table 10.14: Region 2-3 Interface Model Margin (Depleted / Fresh)

Region 2 total bias and uncertainty 0.0380 (includes 0.01 dk NRC admin margin)

Region 3 total bias and uncertainty 0.0259 (includes 0.01 dk NRC admin margin)

Region 2 importance 0%

Region 3 importance 100%

Interface model base k-eff 0.9665*

Interface model bias and uncertainty 0.0259 (weighted total bias and uncertainty)

Interface model 95/95 k-eff 0.9924 (interface k-eff + wtd. total bias + uncert:)

(greater than infinite lattice Reg 2 and Interface model margin to k-eff limit 0.0076* Reg 3)

Adequate margin to k-eff limit? YES (bounded by infinite lattice analysis)

• Region 3 Infinite lattice analysis uses 1.703 wt%, which is 0.0006 more reactive than the 1.70 wt% used for the interface analysis. Interface model margin is 0.007 dk accounting for 0.003 wt% enrichment increase.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 202 of 300 10.4.4 Region 2-3 Interface (Depleted / Depleted)

Table 10.15 summarizes infinite lattice and interface model results with depleted fuel in both Regions. Region 2 (14x14) has fuel centered in the rack cells to maximize k-eff. Region 3 has a 4x4 cell section of each rack module asymmetrically loaded to maximize k-eff. T he first four cases are infinite lattice cases for comparison. The interface model k-eff is not bounded by the Region 3 infinite lattice k-eff. Th is result is expected because of the relatively large infinite lattice reactivity difference between Region 2 and 3.

Table 10.15: Rec ion 2-3 Interface (Depleted/ Depleted)

Enrich. and Region k-eff Uncert. Note Burnup*

5.0 40G I NIA 32F Reg 2 0.95678 0.00006 infinite lattice Reg. 2 5.0 40G I NIA 150FReg2 0.94838 0.00006 infinite lattice Reg. 2 NIA I 5.0 54G 32F Reg 3 0.94337 0.00007 infinite lattice Reg. 3 NIA I 5.0 54G 150F Reg 3 0.95081 0.00006 infinite lattice Reg. 3 5.0 40G I Interface Max BU Reg. 2 I Max BU 5.0 54G 32F Reg 2-3 0.95423 0.00006 Reg 3 5.0 40G I Interface Max BU Reg. 2 I Max BU 5.0 54G 150F Reg 2-3 0.94939 0.00007 Reg 3

• Enrichment is w/o U-235; burnup is GWd/MTU The interface model is most reactive at 32 °F. Because the interface k-eff (0.95423) is greater than the infinite lattice analysis k-eff of Region 3 (0.95081 ), there is a possibility that the interface model has less margin to the k-eff limit than indicated in the Region 3 infinite lattice analysis. Table 10.16 documents interface model tolerance cases. They will be used to compare geometric and fuel content sensitivity of the interface model to infinite lattice Region sensitivities in Table 10.17 . All cases are run at the limiting interface model temperature (32 OF).

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 203 of 300 Table 10.16: Region 2-3 Interface Model Selected Tolerances (Depleted/ Deplete d)

Enrichment and Burnup* Region k-eff Uncert. dk Note 5.0 40G I 5.0 54G 32F Req 2-3 0.95423 0.00006 N/A Base case Region 2 increase burnup 1 5.0 41G / 5.0 54G 32F Reg 2-3 0.95023 0.00006 -0.0040 GWd/MTU 5.0 40G I 5.0 54G 32F Req 2-3 0.95903 0.00005 0.0048 Reqion 2 creep and qrowth Region 3 reduce burnup 1 5.0 40G I 5.0 53G 32F Reg 2-3 0.95429 0.00005 0.0001 GWd/MTU 5.0 40G I 5.0 54G 32F Reg 2-3 0.95434 0.00006 0.0001 Reaion 3 creeo and arowth

• Enrichment is w/o U-235; burnup is GWd/MTU Table 10.17 compares infinite lattice tolerance sensitivity from the infinite lattice analyses (Region 2 and Region 3) and tolerance sensitivity of the interface model. All infinite lattice values are calculated at the bounding temperature for that Region. Interface model tolerances are calculated at the limiting interface model temperature. All tolerance values are best estimate (L\k-eff).

Table 10.17: Region 2-3 Interface Model Tolerance Comparison (Depleted/

Depleted)

Reg.1 Reg.2 Tolerance Interface tol.

Tolerance (k- k Model Enrich./ Enrich./ / Inf. lattice case Burnup** Burnup** base)* tol.

Increase BU 1 Interface 2-3 5.0 I 41 5.0 I 54 -0.0040 98%

GWd/MTU Region 2 5.0 I 41 N/A -0.0041 Creep and grid Interface 2-3 5.0 I 40 5.0 I 54 0.0048 100%

growth Region 2 5.0 I 40 N/A 0.0048 Weighted avg. Region 2 importance 99%

Reduce BU 1 Interface 2-3 5.0 I 40 5.0 I 53 0.0001 1%

GWd/MTU Region 3 N/A 5.0 I 53 0.0044 Creep and grid Interface 2-3 5.0 I 40 5.0 I 54 0.0001 2%

growth Region 3 N/A 5.0 I 54 0.0061 Weighted avg. Region 3 importance 2%

• Best estimate dk (does not include statistical uncertainty)

• • Enrichment is wt% U-235, Burnup is GWD/MTU At this interface condition (depleted fuel Region 2, depleted fuel Region 3) the interface model is dominated by Region 2 and is less sensitive to perturbations in Region 3. A blended total bias

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 204 of 300 and uncertainty may be applied to the interface model composed of 99% Region 2 and 2%

Region 3.

Although this interface condition is clearly dominated by Region 2, it is used here as an example of a mixed importance interface calculation. For cases with mixed importance (significant importance in both Regions), a more detailed total bias and uncertainty calculation is appropriate for the following reasons:

1) The NRC administrative margin (0.01 dk) should be removed from the total prior to the weighting and added back to the final result. If not, for a case with 101 % total importance, the NRC administrative margin will have inadvertently increased from 0.01 dk to 0.0101 dk.

2) The code bias and code uncertainty should be removed from the total prior to the weighting and incorporated in the final result (bias added back and uncertainty included by RSS).

3) Weighted uncertainty components for each Region could be statistically combined to obtain the total rather than being added together because they are independent (different fuel in different racks, different rack manufacturers, etc) .

For simplicity, only the first item will be included in the interface margin calculation . Table 10.18 presents the interface model margin calculation.

Table 10.18: Region 2-3 Interface Model Margin (Depleted / Depleted)

Reqion 2 total bias and uncertainty 0.0380 (includes 0.01 dk NRG admin marqin)

Reqion 3 total bias and uncertainty 0.0426 (includes 0.01 dk NRG admin marqin)

Reqion 2 importance 99%

Region 3 importance 2%

Interface model base k-eff 0.9542 Interface model bias and uncertainty 0.0383 (weighted total bias and uncertainty)*

Interface model 95/95 k-eff 0.9925 (interface k-eff + wtd. total bias+ uncert.)

(greater than infinite lattice Reg 2 and Interface model marqin to k-eff limit 0.0075** Req 3)

Adequate marqin to k-eff limit? YES (bounded by infinite lattice analysis)

• NRC administrative margin removed prior to weighting and added back to final result.

• • Region 3 infinite lattice margin is calculated with a burnup 0.35 GWd/MTU lower than the interface model (0.0015 dk effect). The interface model Region 3 burnup tolerance case shows that the interface model is insensitive to small Region 3 burnup changes.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 205 of 300 10.5 Interface Analysis Summary There are two MPS3 SFP Region interfaces:

1) Region 2 meets Region 1

2) Region 2 meets Region 3 The Region 2 interfaces with Regions 1 and 3 are either benign or indeterminate. For each indeterminate interface, interface model margin calculations demonstrate that the interface model has adequate margin to the k-eff limit. In addition , the interface analysis margin to the limit is greater than the comparable infinite lattice Region margin.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 206 of 300 11 Normal Conditions Normal operation pool water temperature ranges up to 150 °F. Temperatures ranging from 32

°F to 150 °F are considered in Section 9 Region analyses. Normal fuel storage may include spent burnable absorbers, source rods, or control rods. Inserting these in the assembly decreases k-eff. Tolerance calculations for each Region (fuel pin pitch, guide tube OD, grid size) confirm that displacing water in the fuel lattice reduces k-eff with no soluble boron present, so the normal operation of storing and moving these inserts is acceptable.

11.1 Fuel Handling MPS3 fuel handling procedures for fresh and spent fuel place the following limits (among others) on fuel movement:

1) Only one fuel assembly at a time is allowed in the transfer canal

2) At least 12 inches of separation is maintained between any two fuel assemblies outside of the racks

3) Fuel may be lowered into an approved storage location

4) Fuel being lowered must remain at least 12 inches away from the nearest fuel assembly if being lowered outside an approved storage location

5) Fuel may be lowered into a Dry Shielded Canister in the Cask Pit Two fresh 5.0 wt% fuel assemblies 30.48 cm apart (12 inches) produce a k-eff less than 0.94 (Figure 11.1 ), which is less than the base case k-eff for Regions 1-3 (Section 9). Closer proximity of fuel being normally handled outside the storage racks is precluded by handling procedures. Figure 11.2 confirms that the fuel elevator, transfer track, and upender are not in close proximity to fuel racks. Normal fuel handling procedures preclude lowering fuel adjacent to a fuel rack (not an approved storage location), so proximity of a fuel assembly outside the SFP racks within 12 inches of fuel in SFP racks does not need to be considered a normal configuration.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 207 of 300 Figure 11.1: k-eff of Two Fresh 5 wt% Fuel Assemblies in Unborated Water 1.10000 1 .08000 1 .06000 1.04000

t: 1.02000 QI I

..ll:

1.00000 0 .98000 0 .96000 0 .94000 0 .92000 0 5 10 15 20 25 30 35 Face-to-Face Seperation of Assemblies (cm)

Figure 11.2: Location of MPS3 Fuel Handling Equipment STORAGE JIBE.A./

J-,--,-- SPE.,,., rUELSHifl'ING CASK STORAGEAR.EA Upender _,.--,-- SPENT FUEL SHil'FL'IG CASK LOADING AR.EA Unloadi Position SPE\,FUEL STORAGE RACKS SPD-1 FUEL POOL

.,.__- +---+- SPENT F<;""ELPOOL NEWFGEL ELEVATOR FUEL TR"-1\'SFER TIJBE

• .~ *..' FUELTRA>'ISFER TRACK FUEL UPENDER 3:FNT- R l CONTAINMENT STRUCTURE

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 208 of 300 11.2 Fuel Inspection Per fuel handling procedure, fuel inspections are performed on only one fuel assembly at a time and is governed by normal fuel handling restrictions. No additional analysis is needed.

11.3 Non-standard Fuel 11.3.1 Fuel Rod Storage Canister A fuel rod storage canister (FRSC) in the MPS3 SFP allows for storage of up to 52 fuel rods.

The key design features of the FRSC are:

• Square lattice

• 0.937 inch storage tube pitch

• 52 fuel rod storage tubes

• Storage tube OD 0.625 inch (48 tubes) and 0.750 inch (4 tubes)

• Storage tube wall thickness 0.035 inch (48 tubes) and 0.049 inch (4 tubes)

The FRSC is analyzed assuming it contains 52 fresh 5.0 w/o fuel rods. Figure 11.3 depicts the FRSC in a Region 2 KENO model. Figure 11.4 shows the Region 3 FRSC model. Table 11.1 documents Region 2 and Region 3 infinite lattice KENO model results. The FRSC was not modeled in Region 1 because Regions 2 and 3 are far more reactive rack designs. Uncertainty and bias items are not modeled due to the large amount of margin to the k-eff limit. Cases include centered and eccentric placement of the FRSC in the storage cells.

The cross section treatment for the FRSC assumes that it is an infinite array of fuel pins using the FRSC fuel pin pitch. Several attempts of improved modeling were made but none of the various modeling attempts increased k-eff above 0.7 with O ppm soluble boron. Special modeling is not needed.

The FRSC may be stored in any storage location in the MP3 SFP in which a fuel assembly is permitted.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 209 of 300 Figure 11.3: Region 2 Fuel Rod Storage Canister Model

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Figure 11.4: Region 3 Fuel Rod Storage Canister Model

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 210 of 300 Table 11.1: Region 2 and 3 Fuel Rod Storage Canister Model Results Temperature Region (F) K-eff Uncert. EALF 2 32 0.68699 0.00007 0.087 3 150 0.68923 0.00006 0.088 11.3.2 Reconstituted Fuel Fuel pins in assemblies are sometimes removed and replaced to create reconstituted assemblies. There are two reconstitution scenarios evaluated:

• Remove a fuel pin(s) and replace it with a stainless steel pin(s).

• Remove a fuel pin(s) and do not replace the missing fuel pin(s) with anything. This scenario leaves an open water hole in the fuel lattice.

Because the fuel assembly pin lattice is usually under-moderated, an open water hole in a fuel assembly can increase the reactivity of the assembly. Therefore if an assembly is being reconstituted , storage requirements are established to accommodate fuel pin removal. Figures 11 .5 (Region 2) and 11 .6 (Region 3) represent modified versions of the infinite lattice Region 2 and 3 KENO models in which the assembly to be reconstituted has empty cells on all four faces.

Region 1 is not modeled because Regions 2 and 3 are far more reactive rack designs.

Table 11 .2 documents the results of three fuel reconstitution cases for each Region with no soluble boron. The first case has low enrichment fresh fuel in centered in each storage location (2.0 w/o in Region 2, 1.7 w/o in Region 3). The second case has a fresh 5.0 w/o fuel assembly with four empty face neighbors. The third case replaces the 5.0 w/o assembly with a low enrichment assembly. These cases show that the model k-eff is insensitive to the enrichment of the reconstitution assembly (up to 5.0 wt%) and that the small amount of enrichment sensitivity exhibited is two orders of magnitude less than the k-eff reduction vs the base case. These results indicate that with four face-adjacent empty cells , the reconstitution assembly is effectively neutronically isolated and does not need further analysis to support the reconstitution process in Region 1, 2, or 3 in the MP3 SFP.

Replacement of a fuel rod with a stainless steel rod reduces k-eff because there is no change in the amount of moderator present and there is a replacement of fissile material with a weak neutron absorber. Region 2 and Region 3 stainless steel replacement rod cases in Table 11.3 L

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 211 of 300 confirm this reduction in fuel reactivity regardless of fuel pin position . Cases in Table 11.3 have no soluble boron.

MP3 fuel assembly MR71, shown in Figure 11 .7, has two peripheral fuel pin locations that are empty (locations H01 and J01 ). Region 2 and Region 3 cases in Table 11.3 confirm that removal of fuel pins H01 and J01 reduces k-eff. MR71 may be stored as a normal fuel assembly.

Figure 11.5: Region 2 Reconstitution Model

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 212 of 300 Figure 11.6: Region 3 Reconstitution Model

• • • o**

• • o*o*

• • • o**

Table 11.2: Fuel Reconstitution Assembly Isolation Temp.

Region (F) k-eff Uncert. dk Note 2 32 0.96422 0.00006 Base Base case all 2.0 w/o Base case with 1 5.0 w/o fresh and 2 32 0.94661 0.00007 -0.0176 face neiqhbors empty Base case with 1 2.0 w/o fresh and 2 32 0.94648 0.00006 -0.0177 face neighbors empty 3 150 0.96154 0.00006 Base Base case all 1. 7 w/o Base case with 1 5.0 w/o fresh and 3 150 0.93573 0.00006 -0.0258 face neighbors empty Base case with 1 1.7 w/o fresh and 3 150 0.93559 0.00006 -0.0260 face neiqhbors empty

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 213 of 300 Table 11.3: Fuel Reconstitution Fuel Rod Replacement Enrich. Temp.

Region (w/o) (F} k-eff Uncert. EALF Note 2 2.0 32 0.96422 0.00006 0.189 Base case all 2.0 w/o Replace 1 perimeter rod with 2 2.0 32 0.96238 0.00006 0.189 SS rod Replace 1 center rod with SS 2 2.0 32 0.96001 0.00007 0.189 rod Replace 1 interior rod with SS 2 2.0 32 0.96074 0.00006 0.189 rod Replace 1 corner rod with SS 2 2.0 32 0.96213 0.00006 0.189 rod Replace 2 rods on 1 face with 2 2.0 32 0.96194 0.00006 0.187 water n 2 5.0 32 1.20103 0.00001 0.312 Base case all 5.0 w/o <V Replace 1 perimeter rod with 2 5.0 32 1.19923 0.00007 0.311 #ifa SS rod 3 1.7 150 0.96154 0.00006 0.134 Base case all 1.7 w/o Replace 1 perimeter rod with 3 1.7 150 0.95752 0.00006 0.134 SS rod Replace 1 center rod with SS 3 1.7 150 0.95798 0.00006 0.134 rod Replace 2 rods on 1 face with 3 1.7 150 0.95816 0.00006 0.1 34 water L .

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 214 of 300 Figure 11.7: Region 3 Fuel Assembly MR71 Model Removed fuel pins

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 215 of 300 11.4 Normal Condition Boron Credit The Code of Federal Regulations Title 1O Part 50 Section 68 (b).4 states:

"If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0. 95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. "

The MPS3 SFP criticality analysis credits soluble boron. Rather than searching for the soluble boron required for k<0.95 for all normal configurations, KENO model calculations are performed with 600 ppm soluble boron, which is shown to be more than sufficient to meet the criticality requirement. The normal storage soluble boron calculation scope includes:

• Region 1 6x6 infinite lattice model (4 out of 4 storage) o 4.75 w/o, 0 GWd/MTU

• Important bias and uncertainty items calculated o 5.0 w/o, 2 GWd/MTU

• Region 2 2x2 infinite lattice model (4 out of 4 storage) o 2.03 w/o, O GWd/MTU

• Bias and uncertainty calculated o 5.0 w/o, 40 GWd/MTU

• Important bias and uncertainty items calculated o 5.0 w/o with a control rod

• Region 3 6x6 infinite lattice model (4 out of 4 storage, 4x4 asymmetric placement) o 1.7 w/o, O GWd/MTU o 5.0 w/o 53 GWd/MTU The Region 1 IFBA credit scenario is not evaluated because it is the least reactive of the Region 1 storage options and because Region 1 is not limiting with 600 ppm soluble boron. No borated Region interface cases are required because the interface configuration reactivity is bounded by infinite lattice model reactivity. No borated non-standard fuel cases are required because non-standard fuel bearing components in the MP3 SFP are less reactive than normal fuel assemblies. No specific analysis is needed for normal fuel handling because fuel handling k-eff is lower than infinite lattice Region k-eff.

Infinite lattice KENO models from Section 9 analyses are modified to include 600 ppm soluble boron. The bounding Region is determined by adding the appropriate total bias and uncertainty

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 216 of 300 for the un-borated condition to the 600 ppm k-eff. To conservatively account for fuel assembly structure that displaces borated water, a portion of the top and bottom reflectors are conservatively modeled as unborated water.

Table 11.4 contains normal condition base case results for Regions 1, 2, and 3 with 600 ppm soluble boron.

For Region 1, the most limiting normal condition is fresh 4.75 w/o fuel at 32 °F. Table 11.5 shows the calculation of total bias and uncertainty for the Region 1 600 ppm condition along with no soluble boron values for comparison. Shaded values (low significance tolerances) are retained from the no soluble boron calculation. Total bias and uncertainty with 600 ppm boron is slightly lower than with no boron. The margin calculation shows that 600 ppm soluble boron is more than enough to meet the requirement of k < 0.95.

A full set of 600 ppm tolerance cases (bias and uncertainty) are run for Region 2 for the calculation of Dominion administrative margin for the normal storage condition . Table 11 .6 shows the calculation of fresh fuel total bias and uncertainty for the Region 2 600 ppm condition along with no boron values for comparison. Total bias and uncertainty with 600 ppm boron is slightly higher than with no boron. The margin calculation shows that 600 ppm soluble boron is more than enough to meet the requirement of k < 0.95. A water displacement tolerance case (Table 11.4) confirms that displacement of water by inert (Zr) rods decreases fuel reactivity.

This case allows for storage of depleted BPRA or similar water displacing materials in the fuel assembly guide tubes.

For Region 2, the most limiting normal condition is 5.0 w/o fuel with 40 GWd/MTU burnup at 32

°F. Table 11.7 shows the calculation of depleted fuel total bias and uncertainty for the Region 2 600 ppm condition along with no boron values for comparison. Burnup measurement tolerance is approximated by multiplying the O ppm tolerance by the relative total burnup worth (600 ppm burnup worth divided by the O ppm burnup worth). This approximation is used because the measurement uncertainty was calculated for the no boron condition by interpolating between burnup steps, but for the 600 ppm condition only one burnup step is available. Code uncertainty and bias values are based on MOX results using a maximum EALF of 0.5 eV. The MOX values result in less margin than 0.5 eV fresh fuel values. Shaded values (low significance tolerances) are retained from the no soluble boron calculation. Total bias and uncertainty with 600 ppm

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 217 of 300 boron is slightly lower than with no boron. The margin calculation shows that 600 ppm soluble boron is more than enough to meet the requirement of k < 0.95.

The highest Region 3 600 ppm base case k-eff is 0.844. Region 1 and 2 bias and uncertainty results show that total bias and uncertainty with 600 ppm boron is sim ilar to total bias and uncertainty with O ppm . Maximum no boron Region 3 total bias and uncertainty (5 day decay time) is 0.043 including 0.01 NRG administrative margin. The estimated Reg ion 3 95/95 k-eff with 600 ppm soluble boron is 0.887, which is clearly non-limiting.

Table 11.4: Region 1-3 Normal Condition Boron Requirement Cases (600 ppm)

Enrich. Burnup Temp. Tolerance Region (w/o) (GWd/MTU) (OF) k-eff Uncert. EALF (dk)* Notes 1 4.75 0 32 0.90300 0.00007 0.372 N/A Base case 1 5.0 2 32 0.89452 0.00007 0.412 N/A Base case 1 4.75 0 150 0.89680 0.00007 0.411 -0.0060 Increase temp 2 2.03 0 32 0.86858 0.00006 0.236 N/A Base case 2 2.03 0 32 0.85816 0.00006 0.258 -0.0103 Zr rods in quide tubes 2 5.0 0 32 0.87641 0.00007 0.569 N/A Control rod credit base 2 5.0 40 32 0.90059 0.00006 0.478 N/A Burnup credit base case 3 1.7 0 150 0.80516 0.00005 0.180 N/A Base case 3 5.0 53 150 0.84428 0.00007 0.334 N/A Base case

• Includes 2xRSS uncertainty for comparison with O ppm tolerance

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 218 of 300 Table 11.5: Region 1 Bias, Uncertainty, and Margin Comparison Soluble Boron (ppm) 0 600 Burnup (GWd/MTU) 0 0 Enrichment (wt%) 4.75 4.75 Uncertainties (dk)

Rack wall thickness 0.0023 0.0021 Rack cell pitch 0.0054 0.0050 Rack cell ID 0.0061 0.0054 Fuel density 0.0007 0.0007 Enrichment 0.0022 0.0022 Wrapper thickness 0.0005 * *** 0.0005 Wrapper width 0.0004 0.0004 Fuel OD 0.0004 0.0004 Clad OD 0.0021 0.0015 Clad ID 0.0002 0.0002 Fuel pin pitch 0.0009 + 0.0009 GTOD 0.0005 0.0005 GTID 0.0006 .* . 0.0006 Poison width 0.0013 0.0013 Poison thickness 0.0010 0.0010 Fuel stack lenQth 0.0002 0.0002 2 x KENO std. dev. 0.0002 0.0002 Code uncertainty 0.0048 0.0048 RSS OF UNCERTAINTIES 0.0104 0.0097 Biases (dk)

Code bias 0.0034 0.0034 Blisters 0.0009 0.0009 SUM OF BIASES 0.0043 0.0043 Summary Base Case k-eff 0.9675 0.9030 Total Bias and Uncertainty 0.0147 0.0139 NRC Administrative Margin 0.0100 0.0100 Maximum k-eff 0.9921 0.9269 10CFR50.68 Limit 1.0000 0.9500 Dominion Margin (dk) 0.0079 0.0231

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 219 of 300 Table 11.6: Region 2 Bias, Uncertainty, and Margin Comparison (fresh fuel)

Soluble Boron loom) 0 600 Burnup (GWd/MTU) 0 0 Enrichment (wt%) 2.03 2.03 Uncertainties (dk)

Rack wall thickness 0.0005 0.0005 Rack cell pitch 0.0006 0.0011 Rack cell ID 0.0003 0.0002 Fuel density 0.0009 0.0011 Enrichment 0.0076 0.0079 Wrapper thickness 0.0005 0.0002 Wrapper width 0.0003 0.0002 Fuel OD 0.0006 0.0007 Clad OD 0.0014 0.0008 Clad ID 0.0003 0.0004 Fuel pin pitch 0.0007 0.0005 GTOD 0.0006 0.0003 GTID 0.0005 0.0003 Poison width 0.0025 0.0018 Poison thickness 0.0003 0.0003 Fuel stack lenqth 0.0003 0.0002 2 x KENO std. dev. 0.0001 0.0001 Code uncertainty 0.0048 0.0048 ASS OF UNCERTAINTIES 0.0096 0.0097 Biases (dk)

Code bias 0.0034 0.0034 Blisters 0.0004 0.0004 SUM OF BIASES 0.0038 0.0038 Summary Base Case k-eff 0.9687 0.8686 Total Bias and Uncertainty 0.0133 0.0134 NRC Administrative Margin 0.0100 0.0100 Maximum k-eff 0.9920 0.8920 10CFR50.68 Limit 1.0000 0.9500 Dominion Margin (dk) 0.0080 0.0580

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 220 of 300 Table11.7:R eg1on

• 20*1as, Uncertamty, an dM argm . Com:>arison (de pleted fuel)

Soluble Boron (porn) 0 600 Burnup (GWd/MTU) 40.25 40 Enrichment (wt%) 5.0 5.0 Worths (dk)

Minor Actinides and FP 0.108 0.095 Burnup worth 0.244 0.226 Uncertainties (dk)

Rack wall thickness 0.0004 0.0005 Rack cell pitch 0.0015 0.0015 Rack cell ID 0.0002 0.0002 Fuel density 0.0009 Ji 0.0011 Enrichment 0.0000 0.0000 Wraooer thickness 0.0004 * * *

• 0.0004 Wraooer width 0.0003 0.0003 Burnup worth (5%) 0.0122 0.0113 Measured Burnup (4.0%) 0.0099 0.0091 Fuel OD 0.0002 0.0007 Clad OD 0.0000 0.0000 Clad ID 0.0003 0.0004 Fuel pin pitch 0.0012 0.0012 GTOD 0.0007 0.0007 GTID 0.0004 0.0004 Poison width 0.0024  ;; 0.0024 Poison thickness 0.0003 0.0003 Fuel stack length 0.0003 0.0003 2 x KENO std. dev. 0.0001 0.0001 Code uncertainty 0.0094 0.0099 RSS OF UNCERTAINTIES 0.0186 0.0179 Biases (dk)

Code bias 0.0019 0.0021 Blisters 0.0004 MW 0.0004 Growth,creep 0.0050 0.0038 Tilt 0.0005 .11[ 0.0005 1.5% minor actinides and FP 0.0016 0.0014 SUM OF BIASES 0.0093 0.0082 Summary Base Case k-eff 0.9568 0.9006 Total Bias and Uncertainty 0.0280 0.0261 NRC Administrative Margin 0.0100 0.0100 Maximum k-eff 0.9948 0.9367 10CFR50.68 Limit 1.0000 0.9500 Dominion Margin (dk) 0.0052 0.0133

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 221 of 300 12 Accident Analysis Accident scenarios include boron dilution to a minimum boron with normal storage, single fuel assembly mis-load in the fuel racks, multiple fuel assembly mis-load, loss of cooling (including partial voiding), dropped assembly into the racks with grid damage (optimum fuel pin pitch) , fuel handling error (two fresh fuel assemblies out of rack in close proximity) , and mis-placement of an assembly between fuel racks. An evaluation of boron dilution time and mitigation is included in Attachment 7.

The multiple mis-load accident is the bounding accident scenario. Multiple mis-load cases for non-limiting Regions 1 and 3 use infinite lattice KENO models with fresh 5.0 w/o U-235 fuel in all storage locations. The Region 2 multiple mis-load uses a larger KENO model to simulate the most reactive batch of fuel expected to be present in the MPS3 SFP. A full set of tolerance and bias cases are calculated for the multiple mis-load condition with 2550 ppm soluble boron.

12.1 Accident Condition Soluble Boron Requirement 12.1.1 Loss of Cooling A postulated loss of cooling event is simulated for each Region at 212 °F with 2550 ppm soluble boron and with optimum voiding. Infinite lattice Region models are used.

Region 1 loss of cooling models have un-poisoned 5.0 w/o U-235 fresh fuel This model is conservative because 5.0 w/o fresh fuel requires 12 I FBA rods to be stored in Region 1's infinite lattice model. Region 2 and 3 cases are run with high enrichment depleted fuel. As shown in the normal storage 600 ppm cases, boron worth is lower for depleted high enrichment fuel than for fresh low enrichment fuel. Isotopic content in the Region 2 and 3 infinite lattice analysis credits all TRITON isotopes except for partial removal of volatile fission products. However, because the loss of cooling is not the limiting accident condition, the fuel is conservatively modeled for convenience using only major actinides.

Results of the loss of cooling cases are shown in Table 12.1. The highest k-eff for the loss of cooling event is 0.885 (Region 3) without credit for minor actinides and fission products (-0.128 dk at 5.0 wt%, 54 GWD/MTU). Compared to the multiple mislead accident, the loss of cooling event is non-limiting.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 222 of 300 Table 12.1: Loss of Cooling Event Enrich. Burnup Temp.

Region (wt%) (GWd/MTU) Boron (F) K-eff Uncert. Notes 1 5.0 0 2550 150 0.76221 0.00006 Max temp nominal 1 5.0 0 2550 212 0.76067 0.00006 0% void 1 5.0 0 2550 212 0.75624 0.00006 10% void 2 5.0 40* 2550 150 0.83291 0.00006 Max temp nominal 2 5.0 40* 2550 212 0.83350 0.00006 0% void 2 5.0 40* 2550 212 0.83295 0.00006 10% void 3 5.0 53* 2550 150 0.72689 0.00006 Max temp nominal 3 5.0 53* 2550 212 0.73413 0.00006 0% void 3 5.0 53* 2550 212 0.74945 0.00006 10% void 3 5.0 53* 2550 212 0.76735 0.00006 20% void 3 5.0 53* 2550 212 0.83934 0.00006 50% void 3 5.0 53* 2550 212 0.86568 0.00006 60% void 3 5.0 53* 2550 212 0.88504 0.00006 70% void 3 5.0 53* 2550 212 0.87920 0.00005 80% void

• Major actinides 12.1.2 Single Mis-placement A single assembly mis-placement between fuel racks could produce a higher k-eff than an in-rack mis-load. The between rack mis-load will be considered using modified Region interface models. Rather than including soluble boron in those models, the un-borated between-rack mis-load k-eff will be compared to the un-borated in-rack mis-load k-eff. If the between rack k-eff is less than the in-rack k-eff, the between-rack mis-load event is non-l imiting. Because only the relative magnitude of the mis-load effect is of interest, fresh fuel of the enrichment analyzed in Section 9 will be used to represent the normally loaded fue l in the racks.

Figure 12.1 shows a portion of the modified KENO Region 1-2 interface model with a mis-loaded assembly where the racks meet. The full model includes a 1Ox1 O section of Region 2 rack cells and an 18x9 section of Region 1 rack cells. Periodic X-Y boundary conditions are used.

Figure 12.2 shows an X-Y view of the modified KENO Region 2-3 interface model with a mis-loaded assembly where the racks meet. The actual water space between racks is much larger than modeled (see Figure 4.3 for comparison). Periodic X-Y boundary conditions a re used.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 223 of 300 Results in Table 12.2 show that mis-placement of a single fresh 5.0 w/o fuel assembly between fuel racks increases k-eff less than a mis-load in the fuel rack. Single assembly mis-load in a fuel rack is bounded by multiple mis-load in a fuel rack, therefore the single mis-placement accident is not limiting.

Figure 12.1: Region 1-2 Interface Model for Between-rack Mis-placement 11111111111111111111111 11111111111111111111111 11111111111111111111111 11111111111111111111111 Ill

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 224 of 300 Figure 12.2: Region 2-3 Interface Model for Between-rack Mis-placement

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 225 of 300 Table 12.2: Single Fuel Assembly Mis-placement Enrich. Burnup* Boron Temp. Uncert.

Region (wt%)* (GWd/MTU) loom) (OF) K-eff (dk) EALF Notes Modified interface model with one fresh 5.0 w/o 1/2 4.75/2.03 0/0 0 32 0.96696 0.00010 0.274 assembly between racks Modified interface model with one fresh 5.0 w/o in 1/2 4.75/2.03 0/0 0 32 0.98425 0.00009 0.203 ReQion 2 rack Modified interface model with one fresh 5.0 w/o 2/3 2.03/1.70 0/0 0 150 0.97878 0.00010 0.193 assembly between racks Modified interface model with one fresh 5.0 w/o in 2/3 2.03/1.70 0/0 0 150 1.04389 0.00010 0.151 Region 3 rack

• Case has 2 Regions of fuel

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 226 of 300 12.1.3 Multiple Mis-load Analysis of the multiple mis-load is performed with KENO Region 1, 2, and 3 infinite lattice models with 2550 or 2600 ppm soluble boron. A very conservative bounding approach is to presume an un-poisoned fresh fuel batch so large that all storage locations in the model are loaded with un-poisoned fresh 5.0 w/o fuel. This approach is sufficient to demonstrate that k-eff is much less than 0.95 for Regions 1 and 3 and that Region 2 is the most limiting Region for the multiple mis-load condition. Region 2 has the smallest cell pitch of the three rack designs, has the least room for borated water, and contains borated panels which reduced soluble boron efficiency by raising EALF. These features would require more soluble boron using this bounding approach, therefore a more realistic approach is used.

Table 12.3 shows the fu!3I batch history of MPS3. Batch characteristics of interest are as follows:

• The largest batch size to date is 88 fuel assemblies (maximum enrichment 4.5 w/o).

• The highest batch average enrichment is 4.9 w/o (72 assemblies).

• Removable burnable poisons were used in Cycles 1 and 2. Thereafter, the minimum assembly average IFBA loading in any sub-batch has been 55 rods.

• IFBA loadings in Cycles 5 and 6 used "1.0x" IFBA enrichment. All cycles thereafter have used "1.5x".

• For all lFBA cycles, the largest number of assemblies with no IFBA is 20. Since Cycle 7, every fuel assembly has contained at least 32 1.5x IFBA rods with 12 inch maximum cutback.

Based on this history, a composite most reactive anticipated fuel batch can be developed:

• 24 fresh un-poisoned 5.0 w/o fuel assemblies o Maximum allowable fuel enrichment o More un-poisoned fuel assemblies than in any IFBA cycle

• All other fuel assemblies 5.0 w/o with 32 1.5x IFBA rods and 12 inch cutback. IFBA B-1O content is conservatively reduced 10%.

o Higher fuel enrichment than any historical batch average o Minimum number of IFBA rods used in any assembly with IFBA to date o Lower number of IFBA rods than in any IFBA fuel batch to date o More fuel assemblies than any batch A modified infinite lattice 20x20 storage cell KENO model is used for the Region 2 multiple mis-load evaluation. Figure 12.4 is a KEN03D representation of the model. The most reactive fuel

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 227 of 300 (un-poisoned) is conservatively grouped together in a 6x4 block. Periodic boundary conditions are used.

Table 12.3: MPS3 Fuel Batch History Enrich. Avg. Average FAw/o Minimum Cycle Batch w/o #FA Enrich #IFBA IFBA/ FA IFBA IFBA/FA 1 2.4 65 O* 0 65 0 1 2.90 2 2.9 64 O* 0 64 0 3 3.4 64 O* 0 64 0 2 4A 3.50 56 3.60 O* 0 56 0 48 3.80 28 O* 0 28 0 3 SA 4.10 32 4.33 2448 77 0 32 58 4.50 44 2656 60 12 0 4 6A 4.20 32 4.39 4096 128 0 128 68 4.50 56 3968 71 16 0 5 7 4.40 84 4.40 4608 55 20 0 6 8 4.60 84 4.60 8896 106 0 32 7 9A 4.40 21 4.70 2400 114 0 104 98 4.80 60 5184 86 0 48 8 10A 4.40 36 4.61 4608 128 0 128 108 4.80 40 2624 66 0 48 9 11A 4.20 41 4.45 5248 128 0 128 118 4.70 40 2688 67 0 64 10 12A 4.70 16 4.89 2048 128 0 128 128 4.95 48 4512 94 0 32 12C 4.95 8 832 104 0 104 11 13A 4.00 37 4.47 4544 123 0 104 138 4.95 36 2976 83 0 32 12 14A 4.70 36 4.83 4224 117 0 80 148 4.95 40 3360 84 0 32 13 15A 4.10 56 4.34 6144 110 0 64 158 4.90 24 2048 85 0 48 14 16A 4.10 37 4.58 4352 118 0 104 168 4.95 48 - 4544 95 0 48 15 17A 4.10 56 4.38 6464 115 0 48 178 4.95 28 2112 75 0 48 16 18A 4.10 53 4.42 5824 110 0 80 188 4.95 32 2592 81 0 48 17 19A 4.10 52 4.42 5312 102 0 80 198 4.95 32 2784 87 0 48

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 228 of 300 Figure 12.4: Region 2 KENO Multiple Mis-load Model (partial X-Y view)

Multiple fresh fuel batch mis-load results are shown in Table 12.4. Loading Region 1 with un-poisoned fresh 5.0 w/o fuel produces a very low k-eff (-0. 76) with 2550 soluble boron in the SFP water. Region 1 is non-limiting for the multiple mis-load accident. Loading Region 3 with un-poisoned fresh 5.0 w/o fuel produces a low k-eff (-0.86) with 2550 soluble boron in the SFP water. Region 3 is non-limiting for the multiple mis-load accident.

Soluble boron is less effective in reducing k-eff in Region 2 because the Region 2 racks have the smallest cell pitch of the three rack designs in the MPS3 SFP. Region 2 cases are modeled using the most reactive anticipated fuel batch. Region 2 multiple mislead results in Table 12.4

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 229 of 300 with 2550 ppm soluble boron show essentially the same k-eff at 32°F and 150°F when the benchmarking temperature bias of 0.008 is added to the 150°F case. Two 2600 ppm cases are provided: a base case, and a 12 inch IFBA cutback case. The base case (and all tolerance cases in Table 12.5 modeled 9 inch IFBA cutback at the top and bottom of each IFBA pin rather than 12 inches. The 12 inch cutback increases k-eff by only 0.0001 dk. This case k-eff is used as the "Base Case k-eff" in the Table 12.6 calculation of bias, uncertainty and margin.

Table 12.5 contains the results of Region 2 multiple mis-load tolerance cases with 2550 ppm soluble boron. Total bias and uncertainty, including 0.01 dk NRG administrative margin, is nearly the same as for 600 ppm boron normal storage. This illustrates that it is acceptable to apply tolerance cases run at 2550 ppm to the 2600 ppm 95/95 k-eff. Table 12.6 shows the calculation of total bias, uncertainty, and margin. Code uncertainty and bias values are for U02 fuel using a maximum EALF of 0.65 eV. With 2600 ppm soluble boron, the k-eff limit is met with Dominion administrative margin over 0.006 dk.

Table 12.4: Multiple Mis-load Base Case Results Enrich. Burnup Boron Temp.

Region (w/o) (MWd/MTU) (ppm) (F) K-eff Uncert. EALF Notes 1 5.0 0 2550 32 0.76359 0.00006 0.62 All 5.0 fresh, no IFBA 1 5.0 0 2550 150 0.76221 0.00006 0.675 All 5.0 fresh, no IFBA 24 fresh 5.0 no IFBA, 376 fresh 5.0 32 IFBA in 2 5.0/5.0* 0/0* 2550 32 0.92495 0.00007 0.598 20x20**

24 fresh 5.0 no IFBA, 376 fresh 5.0 32 IFBA in 2 5.0/5.0* 0/0* 2550 150 0.92413 0.00006 0.651 20x20**

24 fresh 5.0 no IFBA, 376 fresh 5.0 32 IFBA in 2 5.0/5.0* 0/0* 2600 32 0.92133 0.00006 0.605 20x20**

24 fresh 5.0 no IFBA, 376 fresh 5.0 32 IFBA in 20x20, 12inch IFBA 2 5.0/5.0* 0/0* 2600 32 0.92142 0.00006 0.604 cutback 3 5.0 0 2550 32 0.85038 0.00006 0.426 All 5.0 fresh 3 5.0 0 2550 150 0.85760 0.00007 0.451 All 5.0 fresh

• Case has 2 Regions of fuel

• • Case has 9 inch I FBA cutback

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 230 of 300 Table 12.5: Multiple Mis-load Tolerance Results (2550 ppm boron)

Uncert. Tolerance Case k-eff (dk) EALF (dk)* Notes Base case 20x20, 32F, 5.0 w/o, 24 unpoisoned, 376 with Base 0.92495 0.00007 0.598 N/A 32 IFBA. 2550 ppm boron Rack wall thickness 0.92557 0.00006 0.599 0.00079 Increase cell wall thickness Rack cell pitch 0.92748 0.00006 0.608 0.00271 Decrease rack cell pitch Rack cell ID** 0.92482 0.00007 0.598 0.00033 Decrease rack cell ID Fuel density 0.92614 0.00006 0.603 0.00137 Increase density 0.51 %

Wrapper thickness 0.92524 0.00006 0.598 0.00047 Increase wrapper thickness Wrapper width** 0.92481 0.00006 0.598 0.00032 Decrease wrapper width Fuel OD 0.92547 0.00007 0.601 0.00070 Increase fuel OD Clad OD 0.92506 0.00006 0.594 0.00029 Decrease clad OD Clad ID 0.92504 0.00006 0.599 0.00026 Increase clad ID Fuel pin pitch 0.92514 0.00007 0.597 0.00038 Increase fuel pin pitch GTOD 0.92500 0.00006 0.597 0.00023 Decrease GT OD GTID 0.92503 0.00007 0.597 0.00026 Increase GT ID Poison width 0.92665 0.00007 0.595 0.00189 Decrease BORAL width Poison thickness 0.92545 0.00007 0.600 0.00069 Increase BORAL thickness Fuel stack length 0.92497 0.00006 0.598 0.00019 Increase stack heiQht Blisters 0.92546 0.00006 0.600 0.00068 Void in wraooer water space Maximum Qrids 0.92492 0.00006 0.602 0.00014 Maximum Zr Qrid Larqe GT 0.92512 0.00006 0.601 0.00034 Larger GT

• Includes 2xRSS uncertainty

• • Tolerance direction reversed vs. O ppm conditions, tolerance calculation reversed

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 231 of 300 Table 12.6: Multiple Mis-load Bias, Uncertainty and Margin Soluble Boron (ppm) 2600 Burnup (GWd/MTU) 0 Enrichment (wt%) 5.0 Uncertainties (dk)

Rack wall thickness 0.0008 Rack cell pitch 0.0027 Rack cell ID 0.0003 Fuel density 0.0014 Enrichment 0.0000 Wrapper thickness 0.0005 Wrapper width 0.0003 Fuel OD 0.0007 Clad OD 0.0003 Clad ID 0.0003 Fuel pin pitch 0.0004 GTOD 0.0002 GTID 0.0003 Poison width 0.0019 Poison thickness 0.0007 Fuel stack lenQth 0.0002 2 x KENO std. dev. 0.0001 Code uncertainty 0.0051 RSS OF UNCERTAINTIES 0.0064 Biases (dk)

Code bias 0.0047 Blisters 0.0007 Guide tube desiQn 0.0003 SUM OF BIASES 0.0057 Summary Base Case k-eff 0.9214 Total Bias and Uncertainty 0.0122 NRG Administrative Margin 0.0100 Maximum k-eff 0.9436 10CFR50.68 Limit 0.9500 Dominion Margin (dk) 0.0064

• Uncertainty and bias values were calculated with 2550 ppm

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 232 of 300 12.1.4 Dropped Assembly To simulate a fuel assembly damaged as a result of being dropped into a rack cell , the fuel rod pitch is increased to represent grid damage in order to maximize damaged fuel reactivity. A fuel assembly with increased pin pitch is placed into every rack cell. This approach is conservative because all fuel in the rack is modeled as damaged and because depleted fuel is represented using only major actinides. Figure 12.5 shows an X-Y plane view of the Region 2 KENO dropped assembly model.

Figure 12.5: Region 2 Dropped Assembly Model Dropped assembly accident results in Table 12.7 confirm that the dropped assembly accident is not limiting. These cases conservatively assume all fuel in the SFP is damaged such that the fuel pin pitch increases to an optimal value within the storage cell. Maximum k-eff for any region is 0.84. The dropped assembly event is not limiting.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 233 of 300 Table 12. 7: Dropped Assembly Results Enrich. Burnup Temp.

Region (w/o) (GWd/MTU) Boron (F) K-eff Uncert. EALF Notes 1 5.0 0 2550 32 0.76359 0.00006 0.620 All 5.0 fresh, no IFBA 1 5.0 0 2550 32 0.76603 0.00006 0.592 Pin pitch +0.02 cm 1 5.0 0 2550 32 0.76741 0.00006 0.567 Pin pitch +0.04 cm 2 5.0 40 2550 32 0.83245 0.00006 0.616 Nominal pitch 2 5.0 40 2550 32 0.83731 0.00005 0.596 Pin pitch +0.02 cm 2 5.0 40 2550 32 0.84005 0.00006 0.581 Pin pitch +0.04 cm 3 5.0 53 2550 150 0.72689 0.00006 0.498 Nominal pitch 3 5.0 53 2550 150 0.73586 0.00005 0.467 Pin pitch +0.02 cm 3 5.0 53 2550 150 0.74532 0.00006 0.439 Pin pitch +0.04 cm 3 5.0 53 2550 150 0.74898 0.00006 0.430 Pin pitch +0.047 cm 12.1.5 Handling Error A fuel handling error is simulated by modeling two adjacent unpoisoned fresh 5.0 w/o fuel assemblies in water with 2550 ppm soluble boron. Fuel assembly separation of 0, 2 and 4 cm are considered. Fuel handling error results in Table 12.8 confirm that the handling error accident is not limiting. Maximum k-eff is 0.77 with no separation between fuel assemblies.

Table 12.8: Dropped Assembly Results Enrich. Burnup Boron Temp.

(wt%) (GWd/MTU) (ppm) (F) k-eff Uncert. EALF Notes 2 FA in water with O 5.0 0 2550 150 0.77425 0.00007 0.543 cm separation 2 FA in water with 2 5.0 0 2550 150 0.75303 0.00006 0.501 cm separation 2 FA in water with 4 5.0 0 2550 150 0.71605 0.00007 0.513 cm separation

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 234 of -300 12.1.6 Seismic Event The Region 1 and 2 rack design provides more flux trap space between storage cells of adjacent racks than the nominal interior cell spacing (details in Sections 4.2.1 and 4.2.2 for Regions 1 and 2, respectively). This occurs because minimum rack to rack spacing is limited by the protrusion of the rack baseplate beyond the storage cell envelope. Therefore, seismic activity cannot result in Region 1 or 2 storage cells being closer together than in the infinite lattice model.

The Region 3 racks have no physical barriers preventing them from moving closer together during a seismic event. The Region 3 KENO model includes the as-installed minimum rack-to-rack spacing of 1 inch. The rack-to-rack separation uncertainty case of -0.125 inches was worth 0.0006 b.. k (shown in Table 9.28). Multiplying this tolerance by 8 to simulate no space in-between rack walls results in a worth of -0.005 b.. k. Table 12.1 shows that Region 3 with 2550 ppm of boron has a k-eff of 0.72689. There is more than enough margin to accommodate an increase of 0.005 b.. k. Therefore, the seismic event is bounded by the multiple mis-load accident.

12.2 Accident Analysis Summary Evaluation of accident scenarios including single fuel assembly mis-load in the fuel racks, multiple fuel assembly mis-load in the fuel racks, loss of cooling including partial voiding, dropped assembly into the racks with grid damage (optimum fuel pin pitch), mis-placement of an assembly between fuel racks, and a fuel handling error confirms that 2600 ppm is more than sufficient to ensure k-eff < 0.95. The multiple mis-load accident results in the highest k-eff.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 235 of 300 13 Summary and Conclusions A SFP ~riticality analysis is presented for MPS3 to allow for a measurement uncertainty recapture power uprate, to incorporate spent fuel pool (SFP) soluble boron credit, to update computer codes and analysis methods, to eliminate the need for empty storage cells and cell blockers, and to increase identified margin to k-eff limits. A new fuel storage area (NFSA) criticality analysis is performed for MPS3 to update computer codes and analysis methods and to increase identified margin to k-eff limits. The current enrichment limit for the MPS3 NFSA and SFP is 5.0 wt% U-235.

As part of the criticality analyses, computer code validation is described in Appendix A. In support of the use of boron credit in the SFP criticality analysis, a boron dilution analysis is provided in Attachment 7 that identifies potential SFP dilution events, dilution water source flow rates and volumes, means of detection, time required for detection and mitigation, and minimum dilution event soluble boron concentration.

13.1 New Fuel Storage Area The criticality analysis demonstrates adequate margin to the k-eff limits with the NFSA fully loaded with a fresh 5.0 wt% fuel assembly in each storage cell. A composite bounding fuel assembly design is used (Table 13.1 ).

Table 13.1: Bounding Fuel Design Values Enrichment s 5.0 wt% U-235 Pellet Diameter 0.3225 inch Clad Inner Diameter 0.329 inch Clad Outer Diameter 0.374 inch Clad Material Zirconium alloy Rod Pitch 0.496 inch Grid Volume ]8' 0 , excluding lnconel grids Grid Material Zirconium alloy Pellet Stack Net Density s 95.5% of theoretical

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 236 of 300 13.2 Spent Fuel Pool The MPS3 SFP contains three unique storage rack designs (Regions 1, 2, and 3). The criticality analysis demonstrates adequate margin to the k-eff limits in each of the three SFP Regions assuming four-out-of-four storage in each Region. Regions 1 and 2 credit installed BORAL neutron absorber. Region 3 contains but does not credit BORAFLEX neutron absorber.

A composite bounding fuel assembly design is used (Table 13.2).

13.3 Bounding Fuel Design Values Table 13.2 summarizes the bounding fuel design used for this analysis.

Table 13.2: Bounding Fuel Design Values Enrichment s; 5.0 wt% U-235 Pellet Diameter 0.3225 inch Clad Inner Diameter 0.329 inch Clad Outer Diameter 0.374 inch Clad Material Zirconium alloy Rod Pitch 0.496 inch Grid Volume Grid Material Zirconium alloy Pellet Stack Net Density s; 95.5% of theoretical Grid Material A low growth zirconium alloy such as ZIRLO.

Zircalo 4 is covered for existin invento .

Axial blankets are not credited in the analysis and are permitted as well as annular pellets.

Burnable absorbers are credited in portions of the analysis and affect the fuel depletion.

Depletion related limits for burnable absorbers follow in the next section.

13.4 Bounding Depletion Condition Input Depletion parameters were selected to cover past and anticipated future operation. Table 13.3 lists key depletion condition input selected to bound actual fuel depletion conditions. The temperature and soluble boron assumptions are averages over the total burnup (multi-cycle) for a given assembly. Exceptions to these parameters for existing inventory are justified in the analysis.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 237 of 300 Maximum Burnup This is an average for all cycles in which Averaged Soluble .:s. 1050 ppm the assembly was depleted.

Boron Maximum Core This is not a value used directly in the Average

.:s._623.19 °F analysis but is a proxy for the nodal Moderator Exit moderator and fuel temperatures.

Temperature 0 to 30 GWd/MTU BARAP S1.44 Maximum Burnup Linearly decreases to 1.30 at 52 Averaged Relative .:s._Figure 8.1 GWd/MTU Assembly Power Then linearly decreases to 1.0 at 60 GWd/MTU Maximum .:s._8 fingered WABA with up to Removable 200 IFBA, Analysis assumed 24 WABA as well as Burnable

.:s._24 fingered WABA with no some combinations of IFBA and WABA.

Absorbers with IFBA IFBA 13.5 Summary of Loading Constraints Storage geometries for each Region of the SFP are listed in Table 13.4. Region 1 is composed of subregions 1A and 1B. Region 1B occupies the 2 rows of storage cells in each Region 1 rack module closest to the West SFP wall. Figure 13.1 shows the Region 1A/1 B configuration which credits neutron leakage at the SFP wall. Maximum fuel enrichment for storage in the NFSA and the SFP is 5.0 wt% U-235. The Region 2 and Region 3 burnup credit curves are shown in Figures 13.2 and 13.3, respectively.

• Polynomial coefficients for each curve are shown in Table 13.5.

Satisfaction of the burnup curve requirement shall use fuel assembly characteristics as follows:

1) The fuel enrichment is the maximum planar volume averaged as-built initial enrichment in the assembly (blanket enrichment is not credited by averaging with the higher enrichment central zone).

2) The fuel burnup is the volume averaged burnup of the assembly as determined using the measured reaction rates with no reduction for measurement uncertainty.

3) The fuel assembly decay time is the time elapsed since last use at power in the reactor core.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 238 of 300 T abl e 13.4. MPS3 SFP Storage Constramt summary Region Geometrv Credits Descriotion 1A 4 out of 4 None s 4. 75 wt% U-235 1A 4 out of 4 .:: 12 IFBA s 5.00 wt% U-235

.::2 GWc:1/MTU, 1A 4 out of 4 . Burnup s 5.00 wt% U-235 s 5.0 wt% U-235, two rows of 18 4 out of 4 SFP wall rack cells nearest the West SFP wall (FiQure 13.1 ).

2 4 out of 4 Burnup Table 13.5 2 4 out of 4 Control rod s 5.00 wt% U-235 Burnup, decay 3 4 out of 4 Table 13.5 time Assemblies reconstituted with stainless steel rods can be loaded using their burnup and enrichment just as any non-reconstituted assembly. During reconstitution, the assembly must be isolated by vacating the four face-adjacent cells or by ensuring there is at least 12 inches of separation between the assembly being handled and any other assembly. The Fuel Rod Storage Canister and any non-fuel items are allowed to be stored in any storage location that an assembly can be stored. Depleted discrete burnable poisons can be stored in assembly guide tubes.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 239 of 300 Figure 13.1: Region 1A/1 B Orientation

• • • • • • • ~

• • • • • • • Region1A

• • • • • • • . Region18

~ - - - - - - '~ West SFP Wall I

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 240 of 300 Figure 13.2: Region 2 Burnup Credit Curve 45 T. . .

40 I I l I i

l 5' 35 -;-----------+---+-<

~~ 30 r-"'.Ci---,-,-c--!--:-=~=,\..

. I

+-*********-******+*******************j********** ...........I............ .

~ I

...~

I 25 NOT ACCEPTABLE CD a,

~20 -+----i---i----t-----+---t---,;,,o--t----;-----;----;---;---+---t---!i-----i 1i15 +.....................................,. . . . . . . . . . . ,............

Ea, Ill

~ 10 ---****r*********--**-* *****-1I - ............. ********-*--+--+*--******-**-'¢.!.'

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!i 2.8 3.0 3.2 I

r 3.4 3.6 3.8 4.0 4.2 l 4.4 4.6 4.8

' ............................................................................. ...1-....................

5.0 Max Planar Average Assembly Initial Enrichment (wt% U-235)

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 241 of 300 Figure 13.3: Region 3 Burnup Credit Curves 55 so i  !

45 ****-****-****--*-***-----..-*t-*--**-*--------- ***--*-**--***--**-*---+----*******-****

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~ 40 ACCEPTABLE "C

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5 0

1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Max Planar Average Assembly Initial Enrichment (wt% U-235)

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 242 of 300 Table 13.5: Burnup Credit Curve Polynomial Coefficients Region Decay Time a4 a3 a2 a1 ao (Years) 2 No Credit N/A -0.16936 -0.04949 20.551 -40.098 3 No Credit -0.2459 4.208 -26.80 88.70 -92.00 3 3 Years -0.2338 4.001 -25.48 84.34 -87.47 3 9 Years -0.2153 3.684 -23.46 77.66 -80.54 3 18 Years -0.2020 3.458 -22.02 72.88 -75.59 3 25 Years -0.1964 3.361 -21.40 70.84 -73.47 The burnup curve equations have the following polynomial format:

BU [GWD/MTU] = a 4

• wt% 4 + a 3

• wt% 3 + a 2

• wt% 2 + a 1

• wt% 1 + a 0

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 243 of 300 14 References

[1] Code of Federal Regulations, Title 1O, Part 50, Section 68, "Criticality Accident Requirements."

[2] Code of Federal Regulations, Title 10, Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," Criterion 62, "Prevention of Criticality in Fuel Storage and Handling."

[3] NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 9.1.1 Revision 3 (Criticality Safety of Fresh and Spent Fuel Storage and Handling), U.S. Nuclear Regulatory Commission, Washington, DC, March 2007.

[4] Final Division of Safety Systems Interim Staff Guidance, DSS-ISG-2010-01 Revision 0, Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools,"

Nuclear Regulatory Commission, October 2011. (ADAMS Accession Number ML110620086.)

[5] Dominion Resources Services, Inc. Letter to NRC, "Response to Generic Letter (GL) 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools," 11/1/2016 (Serial No.16-161 ).

[6] Dominion Letter, "DOMINION NUCLEAR CONNECTICUT, INC. MILLSTONE POWER STATION UNIT 3 RESPONSE TO REQUEST FOR SUPPLEMENTAL INFORMATION REGARDING GENERIC LETTER 2016-01, MONITORING OF NEUTRON-ABSORBING MATERIALS IN SPENT FUEL POOLS", 11/22/2017 (Serial No.17-447, ADAMS Accession No. ML 17338A05}.

[7] "SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation," ORNL/TM-2005/39, Version 6, Volumes 1-3, Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA, January 2009.

[8] S. Goluoglu, N. F. Landers, L. M. Petrie, and D. F. Hollenbach,"CSAS5: Control Module For Enhanced Criticality Safety Analysis Sequences with KENO V.A," ORNL/TM-2005/39, Version 6, Vol. 1, Section CS, Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA, January 2009. -

[9] J.C. Wagner, M. D. DeHart, and C. V. Parks, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses," US Nuclear Regulatory Commission, NUREG/CR-6801 Oak Ridge National Laboratory, Oak Ridge, Tenn. (2003).

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 244 of 300

[1 O] Letter from U.S.N.R.C (Richard Guzman) to Dominion (D. Heacock), "Millstone Power Station, Unit No. 3 - Issuance of Amendment Adopting Dominion Core Design And Safety Analysis Methods and Addressing The Issues Identified in Three Westinghouse Communication Documents (CAC NO. MF6251 ).", July 28, 2016. Dominion Energy Nuclear Licensing Serial Number 16-317 (Received August 2, 2016). NRG Accession Number: ML16131A728

[11] M.A. Jessee, and M. D. DeHart, "TRITON: A Two-Dimensional Transport and Depletion Module tor Characterization of Spent Nuclear Fuel," ORNUTM-2005/39, Version 6 Vol. 1, Section T1, Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA, January 2009.

[12] J.C. Dean and R.W. Tayloe, Jr., Guide for Validation of Nuclear Criticality Safety Ca/culational Methodology, NUREG/CR-6698, Nuclear Regulatory Commission, Washington, DC January 2001

[13] J. M. Scaglione, D. E. Mueller, J.C. Wagner and W. J. Marshall, An Approach for Validating. Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality (keff) Predictions, US Nuclear Regulatory Commission, NUREG/CR-7109, Oak Ridge National Laboratory, Oak Ridge, Tenn. (2012).

[14] Letter from U.S.N.R.C (Richard Guzman) to Dominion (D. Heacock), "Millstone Power Station, Unit No. 2 - Issuance of Amendment RE: Technical Specification Changes for Spent Fuel Storage (TAC NO. MF0435)", June 23, 2016. NRG ADAMS Accession Number: ML16003A008

[15] USNRC Reg. Guide 1.183, "Alternative Radiological Source Terms tor Evaluating Design Basis Accidents at Nuclear Power Reactors', US Nuclear Regulatory Commission, Washington, DC, July 2000.

[16] C. E. Beyer, and P. M. Clifford, "Update of Gap Release Fractions tor Non-LOCA Events Utilizing the Revised ANS 5.4 Standard', PNNL-18212 Rev. 1, Pacific Northwest National Laboratory, Richland, Washington, June 2011. ADAMS Accession Number ML112070118.

[17] J. 0. Barner, "Characterization Of LWR Spent Fuel MCC-Approved Testing Materia/--

ATM-101," PNL-5109 Rev. 1, Pacific Northwest National Laboratory, Richland, Washington (1985).

[18] R. J. Guenther, et al, "Characterization Of Spent Fuel Approved Testing Material--ATM-103," PNL-5109-103, Pacific Northwest National Laboratory, Richland, Washington (1988).

[19] R. J. Guenther, et al, "Characterization Of Spent Fuel Approved Testing Material--ATM-104," PNL-5109-104, Pacific Northwest National Laboratory, Richland, Washington (1991 ).

[20] R. J. Guenther, et al, "Characterization Of Spent Fuel Approved Testing Materia/--ATM-105," PNL-5109-105, Pacific Northwest National Laboratory, Richland, Washington (1991 ).

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 245 of 300

[21] R. J. Guenther, et al, "Characterization Of Spent Fuel Approved Testing Material--ATM-106," PNL-5109-106, Pacific Northwest National Laboratory, Richland, Washington (1988).

[22] A. Simpson, T. Marlow, J. Franco, M. Clapham and A. Chesterman, "Operational Experience in Radiometric Instrumentation for Spent Fuel Monitoring," Proceedings of the Institute of Nuclear Materials Management: INMM 44th Annual Meeting, Phoenix, Arizona, July 13-17, 2003, Institute of Nuclear Materials Management, Oakbrook Terrace, IL (2003).

[23] Letter from Dominion (M. Sartain) to U.S.N.R.C, "Virginia Electric and Power Company (DOMINION) North Anna Power Station Units 1 and 2 Proposed Licensing Amendment Request for Spent Fuel Storage and New Fuel Storage (Serial No.16-383)", May 2, 2017.

NRC ADAMS Accession Number: ML17128A451

[24] "Studsvik Scandpower Inc. Approved Version Submittal for TOPICAL REPORT SSP P01/028-TR, "GENERIC APPLICATION OF THE STUDSVIK SCANDPOWER CORE MANAGEMENT SYSTEM TO PRESSURIZED WATER REACTORS" (CAC NO.

MF7273)", October 6, 2017. NRC ADAMS Accession Number: ML17279A983

[25] ZIRAT14 Special Topical Report, "In-Reactor Creep of Zirconium Alloys," September 2009, Advanced Nuclear Technology International.

[26] International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)3, Volumes IV and VI, Nuclear Energy Agency, OECD, Paris, September, 2014.

[27] D. E. Mueller, K. R. Elam, and P. B. Fox, Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data, NUREG/CR-6979 (ORNL/TM-2007/083),

prepared for the US Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., September 2008

[28] U.S. Nuclear Regulatory Commission, Spent Fuel Project Office Interim Staff Guidance -

8, Rev. 3 - Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks, U.S. Nuclear Regulatory Commission, April, 2012.

[29] A. Tsiboulia, et al., 'Water-Moderated U(4.92)0 2 Fuel Rods in 1.29, 1.09, and 1.01 cm Pitch Hexagonal Lattices at Different Temperatures," LEU-COMP-THERM-026, International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)3, Volume IV, Nuclear Energy Agency, OECD, Paris, September, 2014.

[30] A. D. Santos, et al., "Critical Loading Configurations of the IPEN/MB-01 Reactor Considering Temperature Variation From 14° C to 85° C," LEU-COMP-THERM-046, International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC{95)3, Volume IV, Nuclear Energy Agency, OECD, Paris, September, 2014.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 246 of 300

[31] D. Lancaster, Sensitivity Analyses for Spent Fuel Pool Criticality, Technical Report EPRI Doc No. 3002003073,December 2014.

[32] ANS-8.1/N16.1-1975, 'Criticality safety in operations with fissionable material".

[33] YAEC-1937, "Axial Burnup Profile Database for Pressurized Water Reactors", 05/1997.

[34] EPRI Handbook of Neutron Absorber Materials for Spent Nuclear Fuel Transportation and Storage Applications, 11/25/2009.

[35] NRG Letter, "

SUMMARY

OF APRIL 21, 2004, MEETING WITH PACIFIC GAS &

ELECTRIC AND AAR TO DISCUSS BLISTERING ON BORAL MATERIAL USED IN THE SPENT FUEL POOL AT THE DECOMMISSIONING HUMBOLDT BAY PLANT", April 21 2004.

[36] Regulatory Information Conference, "Degradation of Boral in Spent Fuel Pool at Taiwan's BWR NPPs", 3/13/2014.

[37] WCAP-15063-P-A, Westinghouse Electric Company LLC., 'Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)", July 2000

[38] Dennis Gottuso, Jean-Noel Canat, and Pierre Mellard, "A Family of Upgraded Fuel Assemblies for PWR," Top Fuel 2006, 2006 International Meeting on LWR Fuel Performance, October 22-26, 2006, Salamanca, Spain, European Nuclear Society.

[39] Not Used.

[40] Not Used.

[41] NotUsed.

[42] David Mitchell, Anand Garde, and Dennis Davis, "Optimized ZIRLO' Fuel Performance in Westinghouse PWRs," Proceedings of the 2010 LWR Fuel Performance Meeting/Top Fuel!WRFPM, September 26-29, 2010, Orlando, Florida, USA, American Nuclear Society, La Grange Park, Illinois.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 247 of 300 Appendix A: Validation for Criticality Analysis Using Laboratory Critical Experiments A.1 . Overview This appendix determines the computer code and cross-section library bias and uncertainty in the k's calculated for the MPS3 spent fuel pool and new fuel storage area when using the CSAS5 module of SCALE 6.0 and the 238 Group ENDF/B-VII.O cross section library. [A.1] The CSAS5 module executes the CENTRM and BONAMI programs for the resonance self-shielding calculations and KENO V.a for the Monte Carlo calculation of k. All the computer runs use a large Monte Carlo sampling of at least 1500 generations and 6000 neutrons per generation.

The bias and uncertainties determined in this Appendix covers the major actinides plus structural and absorber materials.

This validation follows the direction of NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology [A.2]. The guide establishes the following steps for performing the validation:

1. Define operation/process to identify the range of parameters to be validated

2. Select critical experiment data

3. Model the experiments

4. Analyze the data

5. Define the area of applicability of the validation and limitations It further defines the steps of "Analyze the data" as:

1. Determine the Bias and Bias Uncertainty

2. Identify Trends in Data, Including Discussion of Methods for Establishing Bias Trends

3. Test for Normal or Other Distributions

4. Select the Statistical Method for Treatment of Data

5. Identify and Support Subcritical Margin

6. Calculate the Upper Safety Limit

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 248 of 300 The validation consists of modeling 232 U02 and 136 MOX critical experiments for the determination of the bias and the uncertainty in the calculation of k for fresh fuel and spent fuel.

In addition 17 critical experiments are modeled to determine a temperature bias.

A.2. Definition of the Range of Parameters to Be Validated The validation guidance document [A.2] states:

"Prior to the initiation of the validation activity, the operating conditions and parameters for which the validation is to apply must be identified. The fissile isotope, enrichment offissile isotope, fuel density, fuel chemical form, types of neutron moderators and reflectors, range of moderator to fissile isotope, neutron absorbers, and physical configurations are among the parameters to specify. These parameters will come to define the area of applicability for the validation effort."

The fuel is low enriched uranium dioxide (less than or equal to 5.0 wt% U-235). The fuel is in pellets with a density of greater than 94% of the theoretical density. The only significant neutron moderators are water and the oxygen in the fuel pellet. The neutron absorbers credited are boron (in Boral panels, IFBA coatings, or in solution) and Ag-In-Cd or Hf control rods which may be credited. The reflectors are water, steel, or concrete. The fuel is in assemblies with a rectangular pitch. The assemblies are arranged in cells with space between the cells. The

-(

assemblies and cells are in water with varying density and temperature.

A.3. Selection of the Critical Benchmark Experiments A.3. 1 Selection of the Fresh U02 Critical Benchmark Experiments The U02 benchmarks that were selected met the following criteria:

• Low enriched (5 wt% U-235 or less) U02 to cover the principle isotopes of concern.

• Fuel in rods to assure that the heterogeneous analysis used in SCALE also is applied in the benchmark analysis.

• Square lattices to assure the lattice features of SCALE used in the rack analysis are also modeled in the critical benchmarks selected.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 249 of 300

• Presence of soluble boron, boron bearing rods or plates and Cd to cover most of the control rod absorption.

• No emphasis on a feature or material not of importance to the rack analysis such as lead or copper.

The OECD/NEA International Handbook of Evaluated Criticality Safety Benchmarks Experiments [A.3] is the appropriate reference for criticality safety benchmarks. This handbook has reviewed the available benchmarks and evaluated the uncertainties in the experiments.

The appropriate modeling is presented. All of the experiments used in this validation were

'taken from this handbook. Volume IV of the handbook is for low enriched uranium systems.

The section of Volume IV of interest to this validation is the "Thermal Compound Systems." All of the experiments selected are numbered LEU-COMP-THERM-OXX. This validation will refer to the experiments LEU-COMP-THERM-OXX as just XX where any leading zero is not included.

(Experiments are also referred to as LCT-XX.)

There are more critical experiments in the handbook that meet the requirements for this validation than would be necessary to use. However, most of the applicable available benchmarks were used. There are 92 sets of benchmarks in the September 2014 version of the handbook. Of these, 24 sets were excluded since they used hexagonal arrays; 4 sets were excluded due to enrichments of 6.9 wt% U-235 or higher; 8 sets were excluded due to not being water moderated fuel rods; 4 sets were excluded due to large experimental uncertainties; 8 experiment sets were excluded due to using materials that are not in spent fuel pool racks, such as copper tubes, Gadolinium rods or solution, Titanium screens, or lead reflectors. This leaves 44 benchmark sets of which 27 sets were used for this validation. The 17 unused benchmark sets were reviewed to be sure that there was no feature of the experimental set that was missing in the selected 25 sets. LCT-COMP-THERM-46 is used for the temperature bias is covered in Section A.4 but not included in this set.

The selected 27 benchmark sets include critical experiments from six different critical experiment facilities. The fuel was mainly clad in aluminum, but experiments with stainless steel and zirconium cladding were also in the set.

The critical benchmark sets generally contained multiple experiments, but not all cases from each critical benchmark set is used. In some sets there are experiments that utilize materials

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 250 of 300 that are outside the area of applicability for this validation, such as lead or copper reflectors.

The 27 selected benchmark sets resulted in 232 experiments that are used for the statistical analysis.

A later section will evaluate the area of applicability provided by this selection of critical benchmarks.

Table A.3.1 provides a summary of all the low enriched thermal experiments (non-U metal) from the OECD/NEA handbook [A.3] and why some experiments were not used.

Table A.3.1: Selection Review of OECD/NEA Criticality Benchmarks (All Experiments Start With LEU-COMP-THERM-)

Benchmark : b~scriptio~ / * *

• • • .,< *..., ,* . ....... * <. 1 Selecteij L **

, L.ab > .*

Number ... I>****

WATER-MODERATED U(2.35)0 2 FUEL RODS IN 1 PNL All 8 2.032-CM SQUARE-PITCHED ARRAYS WATER-MODERATED U(4.31)0 2 FUEL RODS IN 2 PNL All5 2.54-CM SQUARE-PITCHED ARRAYS WATER-MODERATED U(2.35)0 2 FUEL RODS IN None. Gd impurity not 3 1.684-CM SQUARE-PITCHED ARRAYS PNL well known. Not (GADOLINIUM WATER IMPURITY) benchmark quality.

WATER-MODERATED U(4.31)0 2 FUEL RODS IN None. Gd impurity not 4 1.892-CM SQUARE-PITCHED ARRAYS PNL well known. Not (GADOLINIUM WATER IMPURITY) benchmark quality.

CRITICAL EXPERIMENTS WITH LOW-ENRICHED URANIUM DIOXIDE FUEL RODS IN None. Soluble Gd not 5 PNL WATER CONTAINING DISSOLVED used in pools.

GADOLINIUM CRITICAL ARRAYS OF LOW-ENRICHED U0 2 6 FUEL RODS WITH WATER-TO-FUEL VOLUME JAEA All 18 RATIOS RANGING FROM 1.5 TO 3.0 Only 4 cases used rest WATER-REFLECTED 4.738-WT.%-ENRICHED 7 Valduc are in hexagonal URANIUM DIOXIDE FUEL-ROD ARRAYS arrays.

CRITICAL LATTICES OF U02 FUEL RODS AND 8 B&W All 17 PERTURBING RODS IN BORATED WATER WATER-MODERATED RECTANGULAR CLUSTERS OF U(4.31 )0 2 FUEL RODS (2.54-CM 21 cases used. Did not 9 PITCH) SEPARATED BY STEEL, BORAL, PNL include Copper cases COPPER, CADMIUM, ALUMINUM, OR ZIRCALOY-4 PLATES

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 251 of 300

.**eenchmark . t* . ...

Lab Selectecl

/ >*

N b Oescnp 10n

• um !r .* / ****WATER-MODERATED

. . . .* . * . * ** *. * .* *U(4.31)0 I ,*.*.**** *.. * / **. ) . * . . . . * .* < .*. ***.***.*. * *>

2 FUEL RODS 22 cases used. Did not 10 REFLECTED BY TWO LEAD, URANIUM, OR PNL use lead cases since no STEEL WALLS lead in pools.

CRITICAL EXPERIMENTS SUPPORTING CLOSE PROXIMITY WATER STORAGE OF 11 B&W All 15 POWER REACTOR FUEL (PART I - ABSORBER RODS)

WATER-MODERATED RECTANGULAR CLUSTERS OF U(2.35)02 FUEL RODS(1.684-CM None. Gd impurity not 12 PITCH) SEPARATED BY STEEL, BORAL, PNL well known. Not BOROFLEX, CADMIUM.OR COPPER PLATES benchmark quality.

(GADOLINIUM WATER IMPURITY)

WATER-MODERATED RECTANGULAR CLUSTERS OF U(4.31 )0 2 FUEL RODS (1.892- 5 cases used. Did not 13 CM PITCH) SEPARATED BY STEEL, BORAL, PNL use the cases with BOROFLEX, CADMIUM, OR COPPER PLATES, copper.

WITH STEEL REFLECTING WALLS WATER-REFLECTED ARRAYS OF U(4.31)0 2 None used. High boron 14 FUEL RODS (1.890-CM AND 1.715-CM SQUARE PNL content uncertainty.

PITCH) IN BORATED WATER Not benchmark quality.

THE VVER EXPERIMENTS: REGULAR AND None used due to hex 15 PERTURBED HEXAGONAL LATTICES OF LOW- KFKI arrays.

ENRICHED U0 2 FUEL RODS IN LIGHT WATER WATER-MODERATED RECTANGULAR CLUSTERS OF U(2.35)02 FUEL RODS (2.032- 26 cases used. Did not 16 CM PITCH) SEPARATED BY STEEL, BORAL, PNL use the copper or COPPER, CADMIUM, ALUMINUM, OR borated panel cases ZIRCALOY-4 PLATES WATER-MODERATED U(2.35)0 2 FUEL RODS 23 cases used. Did not 17 REFLECTED BY TWO LEAD, URANIUM, OR PNL use the 6 cases with a STEEL WALLS lead reflector.

LIGHT WATER MODERATED AND REFLECTED None used. Only 1 18 LOW ENRICHED URANIUM DIOXIDE (7 WT.%) Wintrith case Complex system.

ROD LATTICE WATER-MODERATED HEXAGONALLY Kurchatov None used due to hex 19 PITCHED LATTICES OF U(5%)0 2 STAINLESS Institute arrays.

STEEL CLAD FUEL RODS WATER-MODERATED HEXAGONALLY PITCHED PARTIALLY FLOODED LATTICES OF Kurchatov None used due to hex 20 U(5%)0 2 ZIRCONIUM CLAD FUEL RODS, 1.3- Institute arrays.

CM PITCH

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 252 of 300 Benchmark

. * . ... .....*. ..... . **.*** * *:**: *i* *. ... .*.... *.

Number Description tab . *. S~lecte~ .L \ .....

• . .* ** .* ...  : .. : .* .. ...... ....... **.* >/ \.. < < * ....*.

HEXAGONALLY PITCHED PARTIALLY FLOODED LATTICES OF U{5%)0 2 ZIRCONIUM Kurchatov None used due to hex 21 CLAD FUEL RODS MODERATED BY WATER Institute arrays.

WITH BORIC ACID UNIFORM WATER-MODERATED Kurchatov None used due to hex 22 HEXAGONALLY PITCHED LATTICES OF RODS Institute arrays.

WITH U(10%)02 FUEL PARTIALLY FLOODED UNIFORM LATTICES OF Kurchatov None used due to hex 23 RODS WITH U{10%)02 FUEL Institute arrays.

WATER-MODERATED SQUARE-PITCHED Did not use either case Kurchatov 24 UNIFORM LATTICES OF RODS WITH U{10%)02 due to 10 wt% U-235 Institute FUEL enrichment WATER-MODERATED HEXAGONALLY Kurchatov None used due to hex 25 PITCHED LATTICES OF U(7.5%)0 2 STAINLESS-Institute arrays.

STEE~CLADFUELRODS WATER-MODERATED U(4.92)0 2 FUEL RODS IN None used due to hex 26 1.29, 1.09, AND 1.01 CM PITCH HEXAGONAL IPPE arrays.

LATTICES AT DIFFERENT TEMPERATURES WATER-MODERATED AND LEAD-REFLECTED None used due to lead 27 4.738% ENRICHED URANIUM DIOXIDE ROD Valduc reflector.

ARRAYS WATER-MODERATED U(4.31)0 2 FUEL RODS IN TRIANGULAR LATTICES WITH BORON, None used due to hex 28 PNL CADMIUM AND GADOLINIUM AS SOLUBLE arrays.

POISONS None used. No SCALE sample decks. hf WATER MODERATED AND WATER plates cases without hf REFLECTED 4.74% ENRICHED URANIUM 29 Valduc have the same pitch DIOXIDE ROD ARRAYS SURROUNDED BY and pin as benchmark 7 HAFNIUM PLATES above. No significant additional value VVER PHYSICS EXPERIMENTS: REGULAR HEXAGONAL (1.27-CM PITCH) LATTICES OF Kurchatov None used due to hex 30 LOW-ENRICHED U{3.5 WT.% 235U)0 2 FUEL Institute arrays.

RODS IN LIGHT WATER AT DIFFERENT CORE CRITICAL DIMENSIONS WATER-MODERATED HEXAGONALLY PITCHED PARTIALLY FLOODED LATTICES OF Kurchatov None used due to hex 31 U{5%)02 ZIRCONIUM-CLAD FUEL RODS, 0.8- Institute arrays.

CM PITCH UNIFORM WATER-MODERATED LATTICES OF Kurchatov None used due to hex 32 RODS WITH U(10%)0 2 FUEL IN RANGE FROM Institute arrays.

20°C TO 274 °C

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 253 of 300 Benchmark I Description .. )

• Lab Selected.....

1 Number ..

...* *.:* .::.::ii ::*> :.< .. *:. ..

REFLECTED AND UNREFLECTED

• ..... :..... >*:i. :* r >/ .. *.... ::: :**

I 33 ASSEMBLIES OF 2 AND 3%-ENRICHED ORNL None used. Not U0 2 URANIUM FLUORIDE IN PARAFFIN FOUR 4.738-WT.%-ENRICHED URANIUM 6 borated plate cases DIOXIDE ROD ASSEMBLIES CONTAINED IN used. Rejected cases 34 CADMIUM, BORATED STAINLESS STEEL, OR Valduc with less than 2.5 cm BORAL SQUARE CANISTERS, WATER- separation due to high MODERATED AND -REFLECTED uncertainty.

Used 2 cases. Did not CRITICAL ARRAYS OF LOW-ENRICHED U0 2 use the case with 35 FUEL RODS IN WATER WITH SOLUBLE JAEA dissolved Gd. (not like GADOLINIUM OR BORON POISON pool).

THE VVER EXPERIMENTS: REGULAR AND PERTURBED HEXAGONAL LATTICES OF LOW- None used due to hex 36 KFKI ENRICHED U0 2 FUEL RODS IN LIGHT WATER - arrays.

Part 2 WATER-MODERATED AND PARTIALLY None used. No SCALE 37 CONCRETE-REFLECTED 4.738-WT.%- Valduc sample decks.

ENRICHED URANIUM DIOXIDE ROD ARRAYS None used. No SCALE WATER-MODERATED 4.738-WT.%-ENRICHED sample decks. Used a 38 URANIUM DIOXIDE ROD ARRAYS NEXT TO A Valduc borated concrete BORATED CONCRETE SCREEN reflector (not like pool).

INCOMPLETE ARRAYS OF WATER-39 REFLECTED 4.738-WT.%-ENRICHED URANIUM Valduc Used all 17 cases.

DIOXIDE FUEL-ROD ARRAYS FOUR 4.738-WT.%-ENRICHED URANIUM DIOXIDE ROD ASSEMBLIES CONTAINED IN 4 cases used. Did not 40 BORATED STAINLESS STEEL OR BORAL Valduc use lead reflector SQUARE CANISTERS, WATER MODERATED cases.

AND REFLECTED BY LEAD OR STEEL Did not use the 5 cases STORAGE ARRAYS OF 3%-ENRICHED LWR due to complex 41 ASSEMBLIES: THE CRISTO II EXPERIMENT IN Cadarache geometry and no THE EOLE REACTOR SCALE sample deck.

WATER-MODERATED RECTANGULAR CLUSTERS OF U(2.35)02 FUEL RODS (1.684-Used 5 cases. Did not 42 CM PITCH) SEPARATED BY STEEL, BORAL, PNL use copper cases.

BOROFLEX, CADMIUM, OR COPPER PLATES, WITH STEEL REFLECTING WALLS CRITICAL LOADING CONFIGURATIONS OF None used due to Gd 43 THE IPEN/MB-01 REACTOR WITH A HEAVY SS- IPEN rods.

304 REFLECTOR

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 254 of 300 eenc:farii~rlc Number

. ... \*

CRITICAL LOADING CONFIGURATIONS OF None used due to 44 THE IPEN/MB-01 REACTOR WITH U0 2 , IPEN copper rods.

STAINLESS STEEL AND COPPER RODS PLEXIGLAS OR CONCRETE-REFLECTED Rocky None used since not 45 U(4.46)30 8 WITH H/U=0.77 AND INTERSTITIAL Flats pin geometry.

MODERATION CRITICAL LOADING CONFIGURATIONS Used only for OF THE IPEN/MB-01 REACTOR CONSIDERING 46 IPEN temperature bias-. See TEMPERATURE Reference 9.

VARIATION FROM 14°C TO 85°C FUEL TRANSPORT FLASK CRITICAL None used. 3 complex 47 BENCHMARK EXPERIMENTS WITH LOW- Winfrith cases. No SCALE ENRICHED URANIUM DIOXIDE FUEL sample decks.

LIGHT WATER MODERATED AND REFLECTED 48 LOW-ENRICHED (3 WT.% 235U) URANIUM Winfrith All 5 cases used DIOXIDE ROD LATTICES MARACAS PROGRAMME: POLYTHENE-None used. Powder REFLECTED CRITICAL CONFIGURATIONS 49 Valduc rather than pellets. Not WITH LOW-ENRICHED AND LOW-MODERATED similar to pools.

URANIUM DIOXIDE POWDER, U(5)02 7 cases used. Did not 149SM SOLUTION TANK IN THE MIDDLE OF use cases with 50 WATER-MODERATED 4. 738-WT. %-ENRICHED Valduc dissolved Sm. This is URANIUM DIOXIDE ROD ARRAYS not typical of pools.

9 cases used. Did not CRITICAL EXPERIMENTS SUPPORTING use cases with the CLOSE PROXIMITY WATER STORAGE OF borated Al plates since 51 B&W POWER REACTOR FUEL (PART II - ISOLATING primary source listed a PLATES) high uncertainty in the boron content.

URANIUM DIOXIDE (4.738-WT.%-ENRICHED)

FUEL ROD ARRAYS MODERATED AND None used due to hex 52 Valduc REFLECTED BY GADOLINIUM NITRATE arrays.

SOLUTION VVER PHYSICS EXPERIMENTS: REGULAR HEXAGONAL (1.27 CM PITCH) LATIICES OF Kurchatov None used due to hex 53 LOW-ENRICHED U(4.4 WT.% 235U)0 2 FUEL Institute arrays.

RODS IN LIGHT WATER AT DIFFERENT CORE CRITICAL DIMENSIONS CRITICAL LOADING CONFIGURATIONS OF None used due to Gd 54 THE IPEN/MB-01 REACTOR WITH U02 , AND IPEN rods.

UOrGd 2 0 3 RODS

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 255 of 300 LIGHT-WATER MODERATED AND REFLECTED Neither case used.

55 LOW-ENRICHED URANIUM (3 wt.% 235U} Winfrith Complex geometry no DIOXIDE ROD LATTICES KENO-V.a sample deck None used. No sample CRITICAL EXPERIMENT WITH BORAX-V SCALE decks.

56 BOILING WATER REACTOR TYPE FUEL INL Complex BWR ASSEMBLIES geometry.

4.738-WT.%-ENRICHED URANIUM DIOXIDE None used. No sample 57 FUEL ROD ARRAYS REFLECTED BYWATER IN Valduc SCALE decks.

A DRY STORAGE CONFIGURATION CRITICAL LOADING CONFIGURATIONS OF None used. No sample 58 THE IPEN/MB-01 REACTOR WITH LARGE VOID IPEN SCALE decks.

IN THE REFLECTOR 59 Not included in 2014 Handbook RBMK GRAPHITE REACTOR: UNIFORM CONFIGURATIONS OF U(1.8, 2.0, or 2.4%

235U)02 FUEL ASSEMBLIES, AND CONFIGURATIONS OF U(2.0% 235U}0 2 Kurchatov None used. RBMK -

60 ASSEMBLIES WITH EMPTY CHANNELS, Institute not typical of LWRs WATER COLUMNS, AND BORON OR THORIUM ABSORBERS, WITH OR WITHOUT WATER IN CHANNELS VVER PHYSICS EXPERIMENTS: HEXAGONAL (1.27-CM PITCH) LATTICES OF U(4.4 WT.%

235U)0 2 FUEL RODS IN LIGHT WATER, Kurchatov None used due to hex 61 PERTURBED BY BORON, HAFNIUM, OR Institute arrays.

DYSPROSIUM ABSORBER RODS, OR BY WATER GAP WITH/WITHOUT EMPTY ALUMINIUM TUBES 2.6%-ENRICHED U02 RODS IN LIGHT-WATER None used. No SCALE 62 MODERATOR WITH BORATED STAINLESS JAEA sample decks.

STEEL PLATE: SINGLE ARRAYS LIGHT-WATER MODERATED AND REFLECTED LOW-ENRICHED URANIUM (3 wt.% 235U} None used. No SCALE 63 Winfrith DIOXIDE ROD LATTICES WITH DISCRETE sample decks.

POISON-ROD ARRAYS VVER PHYSICS EXPERIMENTS: REGULAR HEXAGONAL (1.27 CM PITCH) LATTICES OF LOW-ENRICHED Kurchatov None used since hex 64 Institute geometry.

U(2.4 WT.% 235U)02 FUEL RODS IN LIGHT WATER AT DIFFERENT CORE CRITICAL DIMENSIONS

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 256 of 300

=::*,,::-= .. : "' .  :> . * :.*** .. .*.  :,****

Benchmark

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• }'******

Number

• Description Lab Selected I:.* .. .... .: .. * *. '*. :**

CRITICAL CONFIGURATIONS OF 2.6%-

ENRICHED U0 2 ROD ARRAYS IN LIGHT- None used. No SCALE 65 JAEA WATER MODERATOR WITH BORATED sample decks.

STAINLESS STEEL PLATE: COUPLED ARRAYS PLEXIGLAS-REFLECTED, CONCRETE-Rocky None used. Not an 66 REFLECTED, OR THIN STEEL-REFLECTED Flats array of rods.

U(4.46)30 8 WITH H/U=0.77 AND HEU DRIVERS 67 Not included in 2014 Handbook PLEXIGLAS-REFLECTED, CONCRETE-REFLECTED, OR THIN STEEL-REFLECTED Rocky None used. Not an 68 U(4.48)30 8 WITH H/U=1.25 OR H/U=2.03 AND Flats array of rods.

HEU DRIVERS PLEXIGLAS-REFLECTED U(4.48)30 8 WITH Rocky None used. Not an 69 H/U=1.25 OR H/U=2.03 AND INTERSTITIAL Flats array of rods.

MODERATION VVER PHYSICS EXPERIMENTS: REGULAR HEXAGONAL (1.10-CM PITCH) LATTICES OF Kurchatov None used due to hex 70 LOW-ENRICHED U(6.5 WT.% 235U)0 2 FUEL Institute arrays.

RODS IN LIGHT WATER AT DIFFERENT CORE CRITICAL DIMENSIONS LOW MODERATED 4.738-WT.%-ENRICHED 71 Valduc All 4 cases used.

URANIUM DIOXIDE FUEL ROD ARRAYS UNDER-MODERATED 4.738-WT.%-ENRICHED Used 3 cases. Did not 72 URANIUM DIOXIDE FUEL ROD ARRAYS Valduc use Polyethylene REFLECTED BY WATER OR POLYETHYLENE reflector cases.

UNDER-MODERATED 4.738-WT.%-ENRICHED URANIUM DIOXIDE FUEL ROD ARRAYS None used. No SCALE 73 Valduc REFLECTED BY WATER WITH sample decks.

HETEROGENEITIES MIRTE PROGRAM FOUR 4.738-WT.%-ENRICHED URANIUM-DIOXIDE FUEL-ROD ARRAYS Not used due to 74 Valduc IN WATER SEPARATED BY A CROSS-SHAPED Titanium screens SCREEN OF TITANIUM (5 MM AND 10 MM THICK)

VVER PHYSICS EXPERIMENTS: HEXAGONAL (1.10 CM PITCH) LATTICES OF LOW-Kurchatov None used due to hex 75 ENRICHED U(6.5 WT.% 235U)0 2 FUEL RODS IN Institute arrays.

LIGHT WATER, PERTURBED BY BORON ABSORBER RODS AND WATER HOLES

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 257 of 300

.... *.. . . . . .. //i ... ..

. Benchmark Description . *< *.* Lab e *Select~d <**

Number 1 LIGHT WATER MODERATED AND REFLECTED LOW ENRICHED URANIUM (3 WT.% 235U) None used. No KENO 76 Winfrith DIOXIDE ROD LATTICES WITH EX-CORE Va sample decks.

DETECTOR FEATURE Only one case used.

Rest of cases same CRITICAL LOADING CONFIGURATIONS OF materials with small 77 IPEN THE IPEN/MB-01 REACTOR modification of arrays.

Not sufficiently independent.

WATER-MODERATED SQUARE-PITCHED U(6.90)02 FUEL ROD Not used due to high 78 LATTICES WITH 0.52 FUEL-TO-WATER SNL enrichment.

VOLUME RATIO (0.855 CM PITCH)

WATER-MODERATED U(4.31)0 2 FUEL ROD None used due to hex 79 SNL LATTICES CONTAINING RHODIUM FOILS arrays ..

WATER-MODERATED SQUARE PITCHED 0 U(6.90)02 FUEL ROD Not used due to high 80 SNL LATTICES WITH 0.67 FUEL TO WATER enrichment.

VOLUME RATIO PWR TYPE U0 2 FUEL RODS WITH Single case not use.

ENRICHMENTS OF 3.5 AND 6.6 WT.% WITH 81 BURNABLE ABSORBER ("OTIO HAHN" ANEX No sample SCALE deck. Unusual case.

NUCLEAR SHIP PROGRAM, SECOND CORE)

CRITICAL LOADING CONFIGURATIONS OF Used only one case.

THE IPEN/MB-01 REACTOR WITH LOW 82 IPEN Rest of cases were not ENRICHED FUEL AND BURNABLE POISON significantly different.

RODS CRITICAL LOADING CONFIGURATIONS OF Used only one case.

83 THE IPEN/MB-01 REACTOR WITH A BIG IPEN Rest of cases were not CENTRAL VOID significantly different.

CRITICAL LOADING CONFIGURATIONS OF 84 THE IPEN/MB-01 REACTOR WITH A CENTRAL IPEN Used the single case ..

CRUCIFORM ROD VVER PHYSICS EXPERIMENTS: REGULAR HEXAGONAL (1.27 CM PITCH) LATIICES OF Kurchatov None used due to hex 85 LOW-ENRICHED U(6.5 WT.% 235U)02 FUEL Institute arrays.

RODS IN LIGHT WATER AT DIFFERENT CORE CRITICAL DIMENSIONS

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 258 of 300 Benchmark Description Lab ..

• Selected> ....

Number

• . . * *.. >< < *.*. / .*.* * * * >... / . . ...... .. . . . <********* .......

............. I **

VVER PHYSICS EXPERIMENTS: HEXAGONAL LATTICES (1.275 CM PITCH) OF LOW None used due to hex 86 NRI ENRICHED U(3.6, 4.4 WT.% 235U)0 2 FUEL arrays.

ASSEMBLIES IN LIGHT WATER WITH H3B03 VVER PHYSICS EXPERIMENTS: HEXAGONAL LATTICES (1.22-CM PITCH) OF LOW-None used due to hex 87 ENRICHED U(3.6, 4.4 WT.% U235)02 FUEL NRI arrays.

ASSEMBLIES IN LIGHT WATER WITH VARIABLE FUEL-ASSEMBLY PITCH CRITICAL LOADING CONFIGURATIONS OF THE IPEN/MB-01 Not used due to no same SCALE decks 88 REACTOR WITH HEAVY REFLECTORS IPEN and no significant COMPOSED OF CARBON contribution.

STEEL AND NICKEL CRITICAL LOADING CONFIGURATIONS OF Used only one case.

89 THE IPEN/MB-01 REACTOR WITH U02 AND  !PEN Rest of cases were not BORATED STAINLESS STEEL PLATES significantly different.

CRITICAL LOADING CONFIGURATIONS OF Used only one case.

90 THE IPEN/MB-01 REACTOR WITH U0 2 AND IPEN Rest of cases were not STAINLESS STEEL RODS significantly different.

CRITICAL LOADING CONFIGURATIONS OF Not used due to Gd 91 THE IPEN/MB-01 REACTOR WITH U0 2, IPEN rods.

STAINLESS STEEL AND GD203 RODS CRITICAL LOADING CONFIGURATIONS OF Not used due to THE IPEN/MB-01 sufficient boron cases 92  !PEN already and no SCALE REACTOR WITH SOLUBLE BORON sample input.

DEUTERIUM CRITICAL ASSEMBLY WITH 1.2% Not used since cases 93

• ENRICHED URANIUM VARYING COOLANT PNC use D20 rather than VOID FRACTION AND LATTICE PITCH H20 VVER PHYSICS EXPERIMENTS: REGULAR HEXAGONAL (1.10 CM PITCH) TWO-REGION Kurchatov None used due to hex 94 LATTICES OF LOW-ENRICHED U(6.5 AND 4.4 Institute arrays.

WT.% 235U)0 2 FUEL RODS IN LIGHT WATER AT DIFFERENT CORE CRITICAL DIMENSIONS A.3.2 Selection of MOX Critical Experiments Burned fuel contains a low concentration of plutonium (less than 2 wt%) as well as uranium, and thus is actually Mixed Oxide (MOX) fuel. Most classical MOX experiments have plutonium concentrations at least twice as high as that contained in burned fuel. A series of experiments

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 259 of 300 were performed in France, Haut Taux de Combustion (HTC) critical experiments, and purchased by the US for domestic use. These experiments model the uranium and plutonium concentration of 4.5 wt% U-235 fuel burned to 37.5 GWd/T [A.4]. This fuel has 1.1 wt%

plutonium and 1.57 wt% U-235. All of the HTC critical experiments used the same fuel pins.

The criticality of these experiments was controlled by adjusting the critical water height. The fuel pins were used in 156 critical arrangements. The experiments were performed in four phases.

HTC Phase 1 [A.5] consists of 17 cases where the pin pitch was varied from 1.3 cm to 2.3 cm and different quantities of pins were used to change the critical height. An 181h case was done where the array was moved to the edge of the tank, so the boundary was the steel tank followed by void. This condition is not typical of a spent fuel pool, so this case was not analyzed. HTC Phase 2 [A.6] consisted of 20 cases where gadolinium of various concentrations was dissolved in the water (Phase 2a) and 21 cases where boron was dissolved in the water (Phase 2b).

Since Gadolinium is not credited except as a fission product, Phase 2a cases are not selected for analysis. Phase 3 [A.7] consists of 26 experiments where the pins were arranged as 4 assemblies. Each assembly used a 1.6 cm pin pitch. The assembly separation was varied, as well as the number of pins in each assembly. Eleven cases boxed the assemblies with an absorber {borated steel, Boral, or cadmium). All boxed cases are selected including the cadmium boxed assembly cases, as Cadmium is credited as part of the control rod credit.

Finally, Phase 4 [A.8] consisted of redoing the same type of experiments as Phase 3, except with lead and steel reflector screens. All the cases with steel reflectors were selected. The lead reflector cases were not used. In review, a total of 97 HTC critical experiments are included.

Since the burnup requirements may exceed 37.5 GWd/MTU, and so that the MOX case set may include benchmarks from more than the HTC experiments alone , MOX benchmarks from the OECD/NEA handbook [A.3] were reviewed for inclusion. There are only 63 low enriched MOX pin critical experiments documented in the OECD/NEA handbook. Since spent nuclear fuel never reaches greater than 2 wt% Pu, 24 cases with above 2 wt% Pu are not used in the analysis. Thus only 39 non-HTC MOX critical experiments are used. The total MOX set is 39 (OECD/NEA)+97 (HTC) or 136 critical experiments. The 24 MOX cases with greater than 2 wt% Pu were analyzed but are only used to show the dependence of the bias on plutonium content.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 260 of 300 A.4. Modeling and Calculating k of the Critical Experiments For most cases, input decks exist on the OECD/NEA handbook [A.3] disc. In general, these input decks were used with minor modifications. For example, none of the decks were for SCALE 6.0 or the ENDF/B-VII library, and the number of neutrons per generation and the number of generations were, in general, too low. All of the decks were modified to 6000 neutrons per generation and 1500 generations. This was sufficient to make the Monte Carlo uncertainty to be 0.0002 or about one tenth the experimental uncertainty. It was confirmed that the input decks matched the isotopic content given in the handbook. The geometric modeling in the decks also matched the descriptions in the handbook. In short, although there was considerable help by starting with the input files given in the handbook, the ownership of the files was taken, as required by NUREG/CR-6698 [A.2] and as stated in section 2.3:

For specific critic.al experiments, the facility or site may choose to use input files generated elsewhere to expedite the validation process. The site has the responsibility for ensuring that input files and the options selected are appropriate for use. Regardless of the source of the input file, the site must have reviewed the description of each critical experiment and determined that the representation of the experiment, including simplifying assumptions and options, are consistent with the intended use. In other words, the site must assume ownership of the input file.

KENO case k convergence was verified using two techniques; checking for satisfaction of the chi-squared test in the output, and performing a statistical test that compares the average k of the first half of generations with the average k of the second half of generations within their respective uncertainties. This second technique is considered equivalent of viewing the plotted output and looking for a variation or trend which would indicate a lack of convergence. If either of these tests failed and the output looked suspect, then the cases were rerun with more generations so that the tests succeeded.

Table A.4.1 shows the results of the analysis of the 232 U02 critical experiments, along with parameters that are used to check for trends in the results. The spectral index, the Energy of the Average Lethargy of the neutrons causing Fission (EALF) is a calculated value from the SCALE output. Note that some of the critical experiments were actually slightly supercritical.

For the supercritical experiments the calculated k's were divided by the measured k before being placed on Table A.4.1.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 261 of 300 m

LCT-1 1 2.350 1.270 2.032 0.0960 0.003 0.9981 2 2.350 1.270 2.032 0.0955 0.003 0.9980 3 2.350 1.270 2.032 0.0945 0.003 0.9972 4 2.350 1.270 2.032 0.0952 0.003 0.9976 5 2.350 1.270 2.032 0.0939 0.003 0.9958 6 2.350 1.270 2.032 0.0947 0.003 0.9980 7 2.350 1.270 2.032 0.0931 0.0031 0.9977 8 2.350 1.270 2.032 0.0941 0.003 0.9966 LCT-2 1 4.310 1.415 2.540 0.1129 0.002 0.9974 2 4.310 1.415 2.540 0.1128 0.002 0.9994 3 4.310 1.415 2.540 0.1128 0.002 0.9982 4 4.310 1.415 2.540 0.1117 0.0018 0.9978 5 4.310 1.415 2.540 0.1101 0.0019 0.9963 LCT-6 1 2.596 1.417 1.849 0.2351 0.002 0.9977 2 2.596 1.417 1.849 0.2420 0.002 0.9981 3 2.596 1.417 1.849 0.2484 0.002 0.9985 4 2.596 1.417 1.956 0.1812 0.002 0.9983 5 2.596 1.417 1.956 0.1866 0.002 0.9979 6 2.596 1.417 1.956 0.1913 0.002 0.9992 7 2.596 1.417 1.956 0.1963 0.002 0.9988 8 2.596 1.417 1.956 0.2018 0.002 0.9990 9 2.596 1.417 2.150 0.1352 0.002 0.9987 10 2.596 1.417 2.150 0.1388 0.002 0.9983 11 2.596 1.417 2.150 0.1421 0.002 0.9985 12 2.596 1.417 2.150 0.1456 0.002 0.9979 13 2.596 1.417 2.150 0.1486 0.002 0.9986 14 2.596 1.417 2.293 0.1142 0.002 0.9991 15 2.596 1.417 2.293 0.1171 0.002 0.9991 16 2.596 1.417 2.293 0.1196 0.002 0.9991 17 2.596 1.417 2.293 0.1223 0.002 0.9990 18 2.596 1.417 2.293 0.1249 0.002 0.9987 LCT-7 1 4.738 0.940 1.260 0.2406 0.0014 0.9958 2 4.738 0.940 1.600 0.1089 0.0008 0.9986 3 4.738 0.940 2.100 0.0707 0.0007 0.9976 4 4.738 0.940 2.520 0.0605 0.0008 0.9980 LCT-8 1 2.459 1.206 1.636 0.2780 0.0012 0.9965 2 2.459 1.206 1.636 0.2452 0.0012 0.9969 3 2.459 1.206 1.636 0.2450 0.0012 0.9974 4 2.459 1.206 1.636 0.2458 0.0012 0.9969 5 2.459 1.206 1.636 0.2454 0.0012 0.9962 6 2.459 1.206 1.636 0.2445 0.0012 0.9966 7 2.459 1.206 1.636 0.2445 0.0012 0.9965 8 2.459 1.206 1.636 0.2426 0.0012 0.9960

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 262 of 300

[n~hinlir 1cm. ****

. . .* * *.c~se*

• Enrichment No~** . . .(\/Vt°loU'"

.*. . . F'.l1elPin 1

fll~IPin <EALf / ***IVleasurrraepy C:alcLll~te .

Diameter \.* Pitch .. (eV) ** Uhcertainty l<. dk.

f .

• ......... .* +2asr** .* * *.*. Cc;m> ., Ccm> . *.* T. Cdetta 1c5 + {

• 9 2.459 1.206 1.636 0.2419 0.0012 0.9964 10 2.459 1.206 1.636 0.2481 0.0012 0.9966 11 2.459 1.206 1.636 0.2534 0.0012 0.9970 12 2.459 1.206 1.636 0.2470 0.0012 0.9970 13 2.459 1.206 1.636 0.2474 0.0012 0.9969 14 2.459 1.206 1.636 0.2492 0.0012 0.9967 15 2.459 1.206 1.636 0.2496 0.0012 0.9962 16 2.459 1.206 1.636 0.2272 0.0012 0.9973 17 2.459 1.206 1.636 0.1982 0.0012 0.9963 LCT-9 1 4.310 1.415 2.540 0.1126 0.0021 0.9983 2 4.310 1.415 2.540 0.1119 0.0021 0.9978 3 4.310 1.415 2.540 0.1123 0.0021 0.9980 4 4.310 1.415 2.540 0.1121 0.0021 0.9982 5 4.310 1.415 2.540 0.1135 0.0021 0.9987 6 4.310 1.415 2.540 0.1125 0.0021 0.9981 7 4.310 1.415 2.540 0.1136 0.0021 0.9993 8 4.310 1.415 2.540 0.1126 0.0021 0.9985 9 4.310 1.415 2.540 0.1134 0.0021 0.9986 16 4.310 1.415 2.540 0.1135 0.0021 0.9981 17 4.310 1.415 2.540 0.1128 0.0021 0.9986 18 4.310 1.415 2.540 0.1136 0.0021 0.9981 19 4.310 1.415 2.540 0.1129 0.0021 0.9991 20 4.310 1.415 2.540 0.1136 0.0021 0.9979 21 4.310 1.415 2.540 0.1128 0.0021 0.9988 22 4.310 1.415 2.540 0.1136 0.0021 0.9989 23 4.310 1.415 2.540 0.1128 0.0021 0.9989 24 4.310 1.415 2.540 0.1120 0.0021 0.9982 25 4.310 1.415 2.540 0.1118 0.0021 0.9984 26 4.310 1.415 2.540 0.1119 0.0021 0.9986 27 4.310 1.415 2.540 0.1117 0.0021 0.9982 LCT-10 5 4.310 1.415 2.540 0.3478 0.0021 0.9995 6 4.310 1.415 2.540 0.2567 0.0021 1.0001 7 4.310 1.415 2.540 0.2058 0.0021 1.0003 8 4.310 1.415 2.540 0.1819 0.0021 0.9972 9 4.310 1.415 2.540 0.1219 0.0021 1.0008 10 4.310 1.415 2.540 0.1179 0.0021 1.0004 11 4.310 1.415 2.540 0.1152 0.0021 1.0009 12 4.310 1.415 2.540 0.1121 0.0021 0.9992 13 4.310 1.415 2.540 0.1104 0.0021 0.9971 14 4.310 1.415 1.892 0.3064 0.0028 1.0008 15 4.310 1.415 1.892 0.2941 0.0028 1.0014 16 4.310 1.415 1.892 0.2845 0.0028 1.0022 17 4.310 1.415 1.892 0.2786 0.0028 1.0013 18 4.310 1.415 1.892 0.2736 0.0028 1.0016

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 263 of 300 r:;nchffla~il~~::c:;

IT> ** t * / 1 Enrichment FUelPin**** Fuel Pin** <EALF ,. Measurm.E!nt Calculate 1

{CYJf;)u- >* * *~ir~;er11>ri 9~*.** I +(eV) } / Uncer"tainfv:. ) d~ff< .

cmr* .... ** (delta k) ... ..1.0007 ..

19 4.310 1.415 1.892 0.2668 0.0028 24 4.310 1.415 1.892 0.5905 0.0028 0.9988 25 4.310 1.415 1.892 0.5448 0.0028 1.0001 26 4.310 1.415 1.892 0.5056 0.0028 1.0007 27 4.310 1.415 1.892 0.4722 0.0028 1.0010 28 4.310 1.415 1.892 0.4419 0.0028 1.0016 29 4.310 1.415 1.892 0.4177 0.0028 1.0015 30 4.310 1.415 1.892 0.3642 0.0028 0.9987 LCT-11 1 2.459 1.206 1.636 0.1677 0.0018 0.9970 2 2.459 1.206 1.636 0.2436 0.0032 0.9963 3 2.459 1.206 1.636 0.1915 0.0032 0.9967 4 2.459 1.206 1.636 0.1917 0.0032 0.9974 5 2.459 1.206 1.636 0.1926 0.0032 0.9967 6 2.459 1.206 1.636 0.1937 0.0032 0.9968 7 2.459 1.206 1.636 0.1949 0.0032 0.9975 8 2.459 1.206 1.636 0.1962 0.0032 0.9968 9 2.459 1.206 1.636 0.1971 0.0032 0.9969 10 2.459 1.206 1.636 0.1854 0.0017 0.9942 11 2.459 1.206 1.636 0.1622 0.0017 0.9942 12 2.459 1.206 1.636 0.1664 0.0017 0.9939 13 2.459 1.206 1.636 0.1466 0.0017 0.9945 14 2.459 1.206 1.636 0.1498 0.0017 0.9949 15 2.459 1.206 1.636 0.1382 0.0018 0.9953 LCT-13 1 4.310 1.415 1.892 0.2836 0.0018 1.0000 2 4.310 1.415 1.892 0.2935 0.0018 1.0001 3 4.310 1.415 1.892 0.2969 0.0018 0.9998 4 4.310 1.415 1.892 0.2958 0.0018 1.0014 5 4.310 1.415 1.892 0.2953 0.0032 1.0001 LCT-16 1 2.350 1.270 2.032 0.0951 0.0031 0.9973 2 2.350 1.270 2.032 0.0948 0.0031 0.9958 3 2.350 1.270 2.032 0.0947 0.0031 0.9974 4 2.350 1.270 2.032 0.0949 0.0031 0.9964 5 2.350 1.270 2.032 0.0945 0.0031 0.9966 6 2.350 1.270 2.032 0.0955 0.0031 0.9970 7 2.350 1.270 2.032 0.0953 0.0031 0.9970 8 2.350 1.270 2.032 0.0966 0.0031 0.9974 9 2.350 1.270 2.032 0.0958 0.0031 0.9978 10 2.350 1.270 2.032 0.0966 0.0031 0.9968 11 2.350 1.270 2.032 0.0959 0.0031 0.9979 12 2.350 1.270 2.032 0.0971 0.0031 0.9975 13 2.350 1.270 2.032 0.0961 0.0031 0.9977 14 2.350 1.270 2.032 0.0971 0.0031 0.9976 21 2.350 1.270 2.032 0.0967 0.0031 0.9978 22 2.350 1.270 2.032 0.0964 0.0031 0.9976

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 264 of 300 Serichmar .case * *L §f'frichrfienf Fuel Pin Fuel Pin . \.*.E.JXLF\ .IVI ea~1.1r1n~nf /Calculate'**

~(). ...

{.Jwt%U-klD

... ... ... ************* ..... Y 235)

Diameter

.... (cm) . ***** ~~i,. \ } (eV)> Un9~rtainty <

.*.*........<(d~ltak

........* :) : . . d keff 23 2.350 1.270 2.032 0.0960 0.0031 0.9974 24 2.350 1.270 2.032 0.0964 0.0031 0.9974 25 2.350 1.270 2.032 0.0960 0.0031 0.9977 26 2.350 1.270 2.032 0.0965 0.0031 0.9976 27 2.350 1.270 2.032 0.0959 0.0031 0.9975 28 2.350 1.270 2.032 0.0943 0.0031 0.9977 29 2.350 1.270 2.032 0.0942 0.0031 0.9966 30 2.350 1.270 2.032 0.0943 0.0031 0.9967 31 2.350 1.270 2.032 0.0943 0.0031 0.9978 32 2.350 1.270 2.032 0.0942 0.0031 0.9970 LCT-17 4 2.350 1.270 2.032 0.1979 0.0031 0.9979 5 2.350 1.270 2.032 0.1749 0.0031 0.9991 6 2.350 1.270 2.032 0.1652 0.0031 0.9996 7 2.350 1.270 2.032 0.1575 0.0031 0.9990 8 2.350 1.270 2.032 0.1316 0.0031 0.9973 9 2.350 1.270 2.032 0.1084 0.0031 0.9970 10 2.350 1.270 2.032 0.0993 0.0031 0.9980 11 2.350 1.270 2.032 0.0975 0.0031 0.9979 12 2.350 1.270 2.032 0.0963 0.0031 0.9977 13 2.350 1.270 2.032 0.0950 0.0031 0.9975 14 2.350 1.270 2.032 0.0942 0.0031 0.9983 15 2.350 1.270 1.684 0.1763 0.0028 0.9974 16 2.350 1.270 1.684 0.1705 0.0028 0.9974 17 2.350 1.270 1.684 0.1656 0.0028 0.9991 18 2.350 1.270 1.684 0.1640 0.0028 0.9972 19 2.350 1.270 1.684 0.1615 0.0028 0.9972 20 2.350 1.270 1.684 0.1600 0.0028 0.9964 21 2.350 1.270 1.684 0.1587 0.0028 0.9969 22 2.350 1.270 1.684 0.1575 0.0028 0.9956 26 2.350 1.270 1.684 0.3652 0.0028 0.9950 27 2.350 1.270 1.684 0.3144 0.0028 0.9971 28 2.350 1.270 1.684 0.2748 0.0028 0.9979 29 2.350 1.270 1.684 0.2463 0.0028 0.9981 LCT-34 4 4.740 0.940 1.600 0.1363 0.0039 1.0010 5 4.740 0.940 1.600 0.1329 0.0039 0.9995 6 4.740 0.940 1.600 0.1297 0.0039 1.0012 7 4.740 0.940 1.600 0.1278 0.0039 0.9993 8 4.740 0.940 1.600 0.1256 0.0039 0.9992 15 4.740 0.940 1.600 0.1348 0.0043 0.9945 LCT-35 1 2.596 1.417 1.956 0.2073 0.0018 0.9981 2 2.596 1.417 1.956 0.2111 0.0019 0.9971 LCT-39 1 4.738 0.940 1.260 0.2216 0.0014 0.9953 2 4.738 0.940 1.260 0.2112 0.0014 0.9968 3 4.738 0.940 1.260 0.1920 0.0014 0.9964

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 265 of 300 Benchmar : Case** Enrichment Fuel Pin. : FuefPin .* :/EA~F . ****.Measur111~nt Calculate 1

..*... . o u.:.L. . * **.. Diameter; k1D I\ No. (wto/c I\\Pi,ch *** 1:r (eV) *'*.* I) l.Jrtc~rt;;ltn~y < d k~ff:

• .* **I /235) .)< (cm).<* ... (cm) .***** *

(deltak) 4 4.738 0.940 1.260 0.1834 0.0014 0.9955 5 4.738 0.940 1.260 0.1391 0.0009 0.9980 6 4.738 0.940 1.260 0.1452 *o.0009 0.9979 7 4.738 0.940 1.260 0.2124 0.0012 0.9964 8 4.738 0.940 1.260 0.2026 0.0012 0.9958 9 4.738 0.940 1.260 0.1970 0.0012 0.9967 10 4.738 0.940 1.260 0.1727 0.0012 0.9973 11 4.738 0.940 1.260 0.2214 0.0013 0.9950 12 4.738 0.940 1.260 0.2159 0.0013 0.9956 13 4.738 0.940 1.260 0.2140 0.0013 0.9953 14 4.738 0.940 1.260 0.2120 0.0013 0.9957 15 4.738 0.940 1.260 0.2109 0.0013 0.9958 16 4.738 0.940 1.260 0.2099 0.0013 0.9963 17 4.738 0.940 1.260 0.2096 0.0013 0.9965 LCT-40 1 4.740 0.940 1.600 0.1426 0.0039 0.9968 5 4.740 0.940 1.600 0.1375 0.0042 0.9948 9 4.740 0.940 1.600 0.1469 0.0046 0.9983 10 4.740 0.940 1.600 0.1416 0.0046 0.9928 LCT-42 1 2.350 1.270 1.684 0.1680 0.0016 0.9972 2 2.350 1.270 1.684 0.1743 0.0016 0.9970 3 2.350 1.270 1.684 0.1809 0.0016 0.9978 4 2.350 1.270 1.684 0.1796 0.0017 0.9982 5 2.350 1.270 1.684 0.1765 0.0033 0.9983 LCT-48 1 3.005 1.094 1.320 0.6740 0.0025 0.9978 2 3.005 1.094 1.320 0.6467 0.0025 0.9984 3 3.005 1.094 1.320 0.6771 0.0025 0.9977 4 3.005 1.094 1.320 0.6788 0.0025 0.9983 5 3.005 1.094 1.320 0.6691 0.0025 0.9977 LCT-50 1 4.738 0.940 1.300 0.1992 0.0010 0.9976 2 4.738 0.940 1.300 0.1906 0.0010 0.9972 3 4.738 0.940 1.300 0.2072 0.0010 0.9970 4 4.738 0.940 1.300 0.1976 0.0010 0.9967 5 4.738 0.940 1.300 0.2218 0.0010 0.9983 6 4.738 0.940 1.300 0.2134 0.0010 0.9986 7 4.738 0.940 1.300 0.2094 0.0010 0.9988 LCT-51 1 C10 2.459 1.206 1.636 0.1468 0.0020 0.9960 2 c11a 2.459 1.206 1.636 0.1953 0.0024 0.9979 3 c11b 2.459 1.206 1.636 0.1951 0.0024 0.9973 4 c11c 2.459 1.206 1.636 0.1968 0.0024 0.9970 5 c11d 2.459 1.206 1.636 0.1974 0.0024 0.9974 6 c11e 2.459 1.206 1.636 0.1989 0.0024 0.9967 7 c11f 2.459 1.206 1.636 0.1991 0.0024 0.9972 8 c11Q 2.459 1.206 1.636 0.2000 0.0024 0.9970 9 c12 2.459 1.206 1.636 0.1660 0.0019 0.9968

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 266 of 300 Benchmar ******case .. Enrichment Fuel Pin fuel Pin ***.* EALF ***Measurment .~alculate>

klD . N *. .**.... (wt%U'.'.' Diameter *** Pitch (eV). *** .... *unc~rtainty dkett

..* ********* .. I t**.*. o... ..... >' 235) * . .*

. \ * . *. . {delta kl t Li.*...... > .

. * (cm) ***********.**(cm) ** i * .*

LCT-71 1 4.738 0.949 1.100 0.7553 0.00076 0.9943 2 4.738 0.949 1.100 0.6915 0.00076 0.9945 3 4.738 0.949 1.100 0.6563 0.00076 0.9943 4 4.738 0.949 1.075 0.8432 0.0008 0.9938 LCT-72 1 4.738 0.949 1.600 0.1101 0.0012 0.9985 2 4.738 0.949 1.600 0.1062 0.0012 0.9973 3 4.738 0.949 1.600 0.1083 0.0012 0.9980 LCT-77 3 4.349 0.980 1.500 0.1618 0.0010 1.0005 LCT-82 3 4.349 0.980 1.500 0.1494 0.0010 1.0005 LCT-83 1 4.349 0.980 1.500 0.1512 0.0010 0.9999 LCT-84 1 4.349 0.980 1.500 0.1541 0.0010 0.9997 LCT-89 1 4.349 0.980 1.500 0.1529 0.0010 1.0000 LCT-90 1 4.349 0.980 1.500 0.1458 0.0010 0.9997 The HTC modeling utilized References A.5 through A.8 for all the details for the analysis. The references include detailed experiment setup materials and geometry, which was used to create detailed SCALE models. Table A.4.2 shows the results of the SCALE calculations of the HTC experiments. The fuel pins for all the HTC cases are the same. The plutonium weight % is always 1.1 wt% Pu.

Table A.4.3 is the results of the SCALE calculations of the MOX critical experiments from the OECD/NEA handbook. [A.3]

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 267 of 300 Table A.4.2: HTC Critical Experiment Results with SCALE 6.0 and ENDF/8-VII HTC-P1-C01 0.99924 0.00020 0.00182 0.0691 2.3 HTC-P1-C02 0.99915 0.00019 0.00182 0.0662 2.3 HTC-P1-C03 0.99929 0.00019 0.00182 0.0661 2.3 HTC-P1-C04 1.00021 0.00024 0.00182 0.0845 1.9 HTC-P1-C05 1.00028 0.00023 0.00182 0.0823 1.9 HTC-P1-C06 0.99974 0.00022 0.00182 0.0817 1.9 HTC-P1-C07 0.99992 0.00024 0.00182 0.1019 1.7 HTC-P1-C08 0.99958 0.00024 0.00182 0.1002 1.7 HTC-P1-C09 0.99917 0.00024 0.00182 0.0993 1.7 HTC-P1-C1 O 1.00017 0.00025 0.00182 0.1397 1.5 HTC-P1-C11 0.99882 0.00023 0.00182 0.1350 1.5 HTC-P1-C12 0.99864 0.00023 0.00182 0.1331 1.5 HTC-P1-C13 0.99834 0.00027 0.00182 0.2542 1.3 HTC-P1-C14 0.99813 0.00024 0.00182 0.2322 1.3 HTC-P1-C15 0.99766 0.00025 0.00182 0.2286 1.3 HTC-P1-C16 1.00008 0.00023 0.00182 0.1010 1.7 HTC-P1-C17 0.99937 0.00021 0.00182 0.0989 1.7 Boron ppm HTC-P2-B0R-C01 0.99878 0.00024 0.00247 0.2451 1.3 100 HTC-P2-B0R-C02 0.99783 0.00026 0.00247 0.2426 1.3 106 HTC-P2-B0R-C03 0.99790 0.00024 0.00247 0.2530 1.3 205 HTC-P2-B0R-C04 0.99880 0.00024 0.00247 0.2612 1.3 299 HTC-P2-B0R-C05 0.99855 0.00022 0.00247 0.2721 1.3 400 HTC-P2-B0R-C06 0.99823 0.00023 0.00247 0.2688 1.3 399 HTC-P2-B0R-C07 0.99934 0.00027 0.00247 0.2776 1.3 486 HTC-P2-B0R-C08 0.99847 0.00022 0.00247 0.2847 1.3 587 HTC-P2-B0R-C09 0.99930 0.00022 0.00247 0.1652 1.5 595 HTC-P2-B0R-C10 0.99789 0.00022 0.00247 0.1600 1.5 499 HTC-P2-B0R-C11 0.99959 0.00023 0.00247 0.1555 1.5 393 HTC-P2-B0R-C12 0.99963 0.00021 0.00247 0.1492 1.5 295 HTC-P2-B0R-C13 0.99893 0.00024 0.00247 0.1445 1.5 200 HTC-P2-B0R-C14 1.00255 0.00026 0.00247 0.1391 1.5 89 HTC-P2-B0R-C15 1.00337 0.00024 0.00247 0.1026 1.7 90 HTC-P2-B0R-C16 1.00162 0.00024 0.00247 0.1066 1.7 194 HTC-P2-B0R-C17 1.00309 0.00021 0.00247 0.1098 1.7 286 HTC-P2-B0R-C18 0.99343 0.00020 0.00247 0.1152 1.7 415 HTC-P2-B0R-C19 1.00041 0.00023 0.00247 0.1041 1.7 100

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 268 of 300 HTC-P2-BOR-C20 0.99279 0.00020 0.00247 0.0892 1.9 220 HTC-P2-BOR-C21 0.99689 0.00026 0.00247 0.0857 1.9 110 HTC-P3-C1 0.99816 0.00019 0.00322 0.1230 1.6 Borated SS HTC-P3-C2 0.99992 0.00018 0.00355 0.1399 1.6 Borated SS HTC-P3-C3 0.99713 0.00019 0.00322 0.1285 1.6 Borated SS HTC-P3-C4 0.99721 0.00019 0.00322 0.1232 1.6 Borated SS HTC-P3-C5 0.99697 0.00019 0.00322 0.1340 1.6 Borated SS HTC-P3-C6 1.00014 0.00018 0.00489 0.1288 1.6 Bora!

HTC-P3-C7 0.99594 0.00018 0.00322 0.1278 1.6 Cd HTC-P3-C8 1.00361 0.00017 0.00322 0.1381 1.6 Cd HTC-P3-C9 0.99672 0.00019 0.00322 0.1325 1.6 Cd HTC-P3-C10 0.99653 0.00021 0.00322 0.1288 1.6 Cd HTC-P3-C11 0.99619 0.00019 0.00322 0.1364 1.6 Cd HTC-P3-C12 0.99965 0.00024 0.00254 0.1120 1.6 HTC-P3-C13 0.99956 0.00028 0.00254 0.1110 1.6 HTC-P3-C14 0.99990 0.00023 0.00254 0.1111 1.6 HTC-P3-C15 0.99938 0.00017 0.00254 0.1103 1.6 HTC-P3-C16 0.99949 0.00027 0.00254 0.1098 1.6 HTC-P3-C17 0.99991 0.00023 0.00254 0.1079 1.6 HTC-P3-C18 0.99955 0.00023 0.00254 0.1060 1.6 HTC-P3-C19 1.00008 0.00022 0.00254 0.1036 1.6 HTC-P3-C20 0.99967 0.00023 0.00254 0.1016 1.6 HTC-P3-C21 1.00014 0.00022 0.00254 0.1041 1.6 HTC-P3-C22 1.00056 0.00023 0.00254 0.1065 1.6 HTC-P3-C23 0.99996 0.00023 0.00254 0.1141 1.6 HTC-P3-C24 0.99981 0.00025 0.00254 0.1497 1.6 HTC-P3-C25 0.99952 0.00023 0.00254 0.1261 1.6 HTC-P3-C26 0.99922 0.00026 0.00254 0.1148 1.6 HTC-P4-ST-C1 1.00334 0.00018 0.00499 0.1519 1.6 Borated SS HTC-P4-ST-C2 0.99844 0.00018 0.00305 0.1496 1.6 Borated SS HTC-P4-ST-C3 0.99817 0.00019 0.00305 0.1457 1.6 Borated SS HTC-P4-ST-C4 0.99799 0.00020 0.00305 0.1416 1.6 Borated SS HTC-P4-ST-C5 0.99806 0.00019 0.00318 0.1386 1.6 Borated SS HTC-P4-ST-C6 0.99732 0.00018 0.00318 0.1373 1.6 Borated SS HTC-P4-ST-C7 0.99677 0.00018 0.00318 0.1360 1.6 Borated SS HTC-P4-ST-C8 0.99674 0.00019 0.00232 0.1351 1.6 Borated SS HTC-P4-ST-C9 0.99634 0.00021 0.00232 0.1341 1.6 Borated SS HTC-P4-ST-C10 0.99813 0.00020 0.00232 0.1336 1.6 Borated SS

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 269 of 300 HTC-P4-ST-C11 0.99846 0.00018 0.00232 0.1315 1.6 Borated SS HTC-P4-ST-C12 1.00095 0.00018 0.00511 0.1347 1.6 Boral HTC-P4-ST-C 13 0.99674 0.00020 0.00511 0.1330 1.6 Boral HTC-P4-ST-C 14 1.00411 0.00017 0.00616 0.1486 1.6 Cd HTC-P4-ST-C 15 0.99846 0.00020 0.00406 0.1423 1.6 Cd HTC-P4-ST-C16 0.99788 0.00018 0.00190 0.1358 1.6 Cd HTC-P4-ST-C 17 0.99585 0.00018 0.00190 0.1349 1.6 Cd HTC-P4-ST-C18 0.99561 0.00019 0.00190 0.1335 1.6 Cd HTC-P4-ST-C19 0.99517 0.00018 0.00190 0.1324 1.6 Cd HTC-P4-ST-C20 0.99438 0.00019 0.00190 0.1316 1.6 Cd HTC-P4-ST-C21 0.99887 0.00018 0.00237 0.1330 1.6 Cd HTC-P4-ST-C22 1.00070 0.00023 0.00432 0.1724 1.6 HTC-P4-ST-C23 1.00096 0.00024 0.00432 0.1650 1.6 HTC-P4-ST-C24 0.99964 0.00024 0.00470 0.1573 1.6 HTC-P4-ST-C25 0.99960 0.00021 0.00470 0.1557 1.6 HTC-P4-ST-C26 0.99967 0.00017 0.00470 0.1543 1.6 HTC-P4-ST-C27 0.99917 0.00024 0.00470 0.1533 1.6 HTC-P4-ST-C28 0.99909 0.00025 0.00470 0.1523 1.6 HTC-P4-ST-C29 0.99914 0.00023 0.00470 0.1431 1.6 HTC-P4-ST-C30 0.99993 0.00023 0.00090 0.1335 1.6 HTC-P4-ST-C31 0.99925 0.00021 0.00090 0.1278 1.6 HTC-P4-ST-C32 0.99990 0.00028 0.00090 0.1244 1.6 HTC-P4-ST-C33 0.99961 0.00022 0.00090 0.1224 1.6

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 270 of 300 Table A.4.3: Results of Low Enriched MOX Critical Experiments Calculated with SCALE

...  ::* . ., ./ . i >.*

. . Reference i : keff :.,,:*:** :. sigma

• EAl.F Pu>* \Pu 240 ... i *. *.

Case ID

• .:* .... .. .,........  : ,.. :*:** .. :***  : ***:::':**/** . ., *. (eVf *. *****wto/J .*wt%. !mi4t/U~38

.:* :::::::. ;:; :t* "'

093arrav OECD-7 1.00201 0.00024 0.189 2 16 6.816E-05 105al OECD-7 0.99513 0.00025 0.136 2 16 7.554E-05 105arrav OECD-7 0.99680 0.00026 0.137 2 16 7.554E-05 105b1 OECD-7 0.99202 0.00027 0.137 2 16 7.554E-05 105b2 OECD-7 0.99304 0.00023 0.137 2 16 7.554E-05 105b3 OECD-7 0.99359 0.00024 0.137 2 16 7.554E-05 105b4 OECD-7 0.99450 0.00024 0.136 2 16 7.554E-05 1143arra OECD-7 0.99785 0.00024 0.116 2 16 8.134E-05 132array OECD-7 0.99689 0.00022 0.095 2 16 8.134E-05 1386arra OECD-7 0.99511 0.00023 0.090 2 16 6.965E-05 epri70b OEC0-2 0.99893 0.00025 0.712 2 7.8 7.287E-05 epri70un OECD-2 0.99719 0.00026 0.536 2 7.8 7.287E-05 epri87b OECD-2 1.00160 0.00022 0.269 2 7.8 7.287E-05 epri87un OECD-2 0.99862 0.00025 0.184 2 7.8 7.287E-05 epri99b OECD-2 1.00083 0.00021 0.176 2 7.8 7.287E-05 epri99un OECD-2 1.00140 0.00028 0.133 2 7.8 7.287E-05 k1mct009 OECD-9 0.99917 0.00026 0.508 1.5 8 1.058E-05 k2mct009 OECD-9 0.99440 0.00024 0.290 1.5 8 9.774E-06 k3mct009 OECD-9 0.99460 0.00024 0.151 1.5 8 8.960E-06 k4mct009 OECD-9 0.99240 0.00024 0.114 1.5 8 8.960E-06 k5mct009 OECD-9 0.99287 0.00022 0.094 1.5 8 8.960E-06 k6mct009 OECD-9 0.99375 0.00024 0.090 1.5 8 9.774E-06 omct61 OECD-6 0.99669 0.00024 0.373 2 8 2.609E-05 omct62 OECD-6 0.99726 0.00024 0.190 2 8 2.271 E-05 omct63 OECD-6 0.99684 0.00024 0.137 2 8 2.512E-05 omct64 OECD-6 0.99727 0.00024 0.116 2 8 2.222E-05 omct65 OECD-6 0.99868 0.00022 0.095 2 8 2.271 E-05 omct66 OECD-6 0.99708 0.00021 0.090 2 8 2.367E-05 mct8c1 OECD-8 1.00038 0.00024 0.137 2 24 7.931 E-05 mct8c2 OECD-8 0.99496 0.00027 0.138 2 24 7.270E-05 mct8c3 OECD-8 0.99457 0.00022 0.137 2 24 8.586E-05 mct8c4 OECD-8 0.99527 0.00024 0.137 2 24 9.882E-05 mct8c5 OECD-8 0.99458 0.00023 0.137 2 24 9.560E-05 mct8c6 OECD-8 0.99833 0.00024 0.372 2 24 7.270E-05 mct8cal OECD-8 0.99770 0.00024 0.247 2 24 8.586E-05 mct8cb1 OECD-8 1.00459 0.00026 0.171 2 24 8.586E-05 mct8cb3 OECD-8 1.00137 0.00023 0.142 2 24 8.586E-05 mctcb2 OECD-8 1.00384 0.00026 0.106 2 24 8.586E-05 mctcb4 OECD-8 1.00345 0.00023 0.092 2 24 8.586E-05 mixo251k OECD-5 1.00686 0.00024 0.087 4 18 1.587E-04 mixo252k OECD-5 0.99585 0.00025 0.354 4 18 1.587E-04 mixo253k OECD-5 0.99984 0.00026 0.187 4 18 1.587E-04

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 271 of 300 Wt'1il mixo254k OECD-5 0.99546 0.00026 0.137 4 18 1.587E-04 mixo255k OECD-5 1.00068 0.00024 0.116 4 18 1.587E-04 mixo256k OECD-5 1.00039 0.00022 0.095 4 18 1.587E-04 mixo257k OECD-5 0.99846 0.00024 0.090 4 18 1.587E-04 saxtn104o OECD-3 1.00023 0.00026 0.099 6.6 8.6 0.0 saxtn56bo OECD-3 0.99971 0.00028 0.612 6.6 8.6 0.0 saxtn735o OECD-3 0.99919 0.00028 0.182 6.6 8.6 0.0 saxtn792o OECD-3 1.00015 0.00028 0.150 6.6 8.6 0.0 saxton52o OECD-3 1.00000 0.00029 0.848 6.6 8.6 0.0 saxton56o OECD-3 1.00067 0.00029 0.517 6.6 8.6 0.0 tca1 OECD-4 0.99609 0.00026 0.141 3.01 22.02 1.045E-04 tca10 OECD-4 0.99949 0.00024 0.079 3.01 22.02 9.307E-05 tca11 OECD-4 0.99960 0.00025 0.079 3.01 22.02 2.058E-04 tca2 OECD-4 0.99693 0.00023 0.140 3.01 22.02 1.993E-04 tca3 OECD-4 0.99716 0.00025 0.140 3.01 22.02 2.962E-04 tca4 OECD-4 0.99695 0.00027 0.117 3.01 22.02 9.878E-05 tca5 OECD-4 0.99768 0.00025 0.116 3.01 22.02 2.018E-04 tca6 OECD-4 0.99797 0.00026 0.115 3.01 22.02 3.903E-04 tea? OECD-4 0.99826 0.00024 0.091 3.01 22.02 8.878E-05 tca8 OECD-4 0.99817 0.00023 0.091 3.01 22.02 2.034E-04 tca9 OECD-4 0.99906 0.00023 0.091 3.01 22.02 3.019E-04 A.5. Statistical Analysis of the Data The statistical treatment used follows the guidance provided in NUREG/CR-6698 [A.2]. The NUREG approach weights the calculated.k's by the experimental uncertainty. This approach means the higher quality experiments (i.e.: lower uncertainty) affect the results more than the low quality experiments. The uncertainty weighting is used for the analysis of the set of experiments as a whole, as well as for the analysis for trends.

Spent fuel goes from having little plutonium to having about 1.5 wt% plutonium at discharge burnups. Since the bias is not the same for plutonium critical experiments as it is for uranium critical experiments, the bias would be expected to be a function of burnup. Rather than attempt to make the bias a function of burnup, analysis of the U02 and MOX critical experiments are separated and the most limiting bias and uncertainty from the two sets will be used in the analysis of the spent fuel pool. The fresh fuel storage uses only the U02 critical experiments.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 272 of 300 The set of MOX experiments is more limited in geometric variation. Because of this, the only trending parameter used for the analysis of the MOX fuel is the spectral index Energy of the Average Lethargy of the neutrons causing Fission, EALF.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 273 of 300 A.5.1 Statistical Analysis of the U02 Critical Experiments This section follows closely NUREG/CR-6698 so in order to help matching with the NU REG the equation numbers from the NUREG are given in parentheses.

The first step of the analysis is force all the experiments to be critical so the analysis is consistent over the entire set. This is done by converting supercritical experiments to critical experiments using the following equation (9):

knorm = kcaic / kexp NUREG/CR-6698 recommends weighting the data by its uncertainty. The combined error for each experimented is calculated (3):

The weighted mean keff (6):

1 L (Y'!- kett,t keff = i 1 L"""z" (Yi The bias is calculated as follows (8):

Bias = ketf - 1; the bias is set to zero if calculated to be greater than zero.

The variance about the mean (4):

The average total uncertainty (5):

n jj2 - - -

- 1 Lo-~l The square root of the pooled variance (7):

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 274 of 300 The uncertainty is calculated by multiplying the square root of the pooled variance by the one-sided lower tolerance factor. Since all the analysis has a sample size greater than 50, 2.065 is used as the single-sided lower tolerance factor.

The weighted mean is 0.9974 and the weighted standard deviation is 0.0024. The average uncertainty of the experiments (interpreted as one sigma) is 0.0022. Since the total one sigma standard deviation is only 0.0024, this suggests that the experimental uncertainty dominates the uncertainty and there is little to be gained with improved methods. Unless stated otherwise, all the results presented will come from the weighted analysis. The bias of the set as a whole is 1 - 0.9974 = 0.0026. The uncertainty is the standard deviation multiplied by the single-sided lower tolerance factor (taken as 2.065 from Table 2.1 of Reference 2), so it is 0.0050.

As recommended by NUREG/CR-6698, the results of the validation are checked for normality.

Normality of the U02 results were tested using the Shapiro-Wilks test via the J. P. Royston algorithm [A.9]. The Shapiro-Wilks test results indicates that the null hypothesis for the U02 set being normally distributed should be rejected at the a=0.05 significance level (i.e. 95%

confidence). The p-value for the SW Test statistic is 0.02, which indicate that the SW Test would pass with 98% confidence. While we reject the null hypothesis for the U02 set at the 95%

confidence level, a qualitative visual examination of the histogram (Figure A.5.1) and the Q-Q plot (Figure A.5.2) does not show any obvious indication of non-normal behavior, trend, or skewness. Given no obvious non-normal behavior with the histogram and Q-Q plot, a normal distribution assumption may be acceptable. Notice that the calculated k's are a little closer to the mean and a little higher (conservative) than expected in a normal distribution. This means assuming a normal distribution may be conservative for this data.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 275 of 300 35 ,, ._ __

30

• Ill

~ 25 ....

....l3

~ 20 ...................................................................................................................

41

.Q

§ 15 -+---------

z

~~~~~~~~~~~#~###~~~~~~~

$$$$$$$$$$$$$$$~~~~~~~~

k-effective by 0.00050 bins Figure A.5.1: Distribution of the Calculated k's Around the Mean for All the U02 Benchmarks 4

3 2 +----------------------

1 ii e

~ 0

s VI -1

-2

..4 0.9920 0 .9940 0.9960 0.9980 1.0000 1.()020 1 .0040 keff Figure A.5.2: Q-Q Normality Plot for the U02 Benchmarks

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 276 of 300 Numerous sources [A.10, A.11, A.12] suggest that for the large sample size used here, normality testing is not important. For example, in the textbook, Statistics for Social Science by R. Mark Sirkin [A.12], it states:

"Law of large numbers. A Jaw that states that if the size of the sample, n, is sufficiently large (no less than 30; preferably no Jess than 50), then the central limit theory will apply even if the population is not normally distributed along variable x ...

If: Then:

n>= 100 It is always safe to relax the normality assumption 50<=n<100 It is almost always safe 30<=n<50 It is probably safe."

The analysis in this validation assumes that the techniques used here are sufficiently robust for the limited normality data. In conclusion, although we reject the null hypothesis that the data is normally distributed at the 95% confidence level, there is no indication that the data is not normally distributed, thus it will be treated as normally distributed.

The next step in the analysis is to look for trends in the data. In the past it was assumed that unless there is a high confidence level (95%} that the slope was non-zero, the analysis would assume a zero slope (no trend) on the given parameter. The analysis will include consideration of the data as non-trended, however it is more conservative to assume there is a trend.

Inverting the statistical test to requiring a high confidence that the slope is zero will result in all cases having a trend. At this time, although a test on the confidence of the trend is performed, the analysis assumes all calculated trends are real.

Before presenting the results of the analysis the following provides the equations used for the analysis. For these equations parameter y is for the dependent variable (kett), and parameter x is for the independent variables (e.g., enrichment, EALF).

First, the linear equation for the fit(1 O):

Y(x) =a+ bx The coefficients are calculated using the next three equations (11, 12, 13):

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 277 of 300 The weighted mean for the independent parameter (15):

1 I-x-(J'f i x-- i 1

I (J'f i

The bias is calculated as follows (23):

Bias = kfit(x) - 1; the bias is set to zero if calculated to be greater than zero.

Finally, the uncertainty is computed from (23):

SP fit { 2Fa (2,n-2) [1"ii_+ 2 (x - i) ]

I(xi - x)2 + Zzp-1 (n - 2)}

Xi-y,n-2 where p = desired confidence (0.95) and the remaining parameters are computed as follows (25, 30, 28):

1-p y=--

2 2

_ ~ I {-tr-[kett,i - ktitCxa] }

2 sfit - 1 1 nI (J'"!-

i The width of the tolerance band is a function of the trending parameter. When the value for the independent variable is known, it is used in the calculation of the uncertainty. For simplicity sometimes the maximum width of the tolerance band over the range of data is taken as the uncertainty.

In the final analysis, the calculated k of the system must be less than the minimum of k(x) minus the uncertainty minus the administrative safety margin. The uncertainty in k from other independent uncertainties, such as the manufacturing tolerances, burnup, and depletion uncertainties can be statistically combined with the uncertainty in the criticality validation. Now

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 278 of 300 this section will evaluate the trends in k as a function of trending parameters using the methods described above.

Neutron spectrum Trends in the calculated k of the benchmarks were sought as a function of the neutron spectrum. Since a large number of things can affect the spectrum, a single index calculated by -

SCALE is used. This index is the Energy (eV) of the Average Lethargy causing Fission (EALF).

Figure A.5.3 shows the distribution of k's around the mean k, which is shown as the red line.

Visual inspection of the graph and the statistical analysis of the results of the statistical analysis suggest that there is a statistically significant trend on neutron spectrum. Using NUREG/CR-6698 [A.2] equations from above and the data from Table A.4.1, the predicted mean k as a function of EALF is:

k(EALF) - = 0.99853 - 0.00491

• EALF The units for EALF are eV. The uncertainty (in terms of k) about the trend is a function of" EALF and is shown on Table A.5-1.

Criticality Safety Eval_uation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 279 of 300 Table A.5.1: Bias and Uncertainty Based on the EALF Trend from U02 Critical Expenments Fresh Fuel Maximum Bias Uncertainty EALF (ev) 0.35 0.0032 0.0047 0.40 0.0034 0.0047 0.45 0.0037 0.0047 0.50 0.0039 0.0048 0.55 0.0042 0.0049 0.60 0.0044 0.0050 0.65 0.0047 0.0051 0.70 0.0049 0.0052 0.75 0.0052 0.0053 0.80 0.0054 0.0054 0.85 0.0056 0.0055 0.90 0.0059 0.0057 0.95 0.0061 0.0058 1.00 0.0064 0.0059 1.05 0.0066 0.0060 1.10 0.0069 0.0061 1.15 0.0071 0.0062 1.20 0.0074 0.0063

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 280 of 300 1.00400 * .............. **************************

1.CXl200 ., ............................................................................................... ,,............................................................................................................................. ..................................................... .

~'fy

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~ 0.99600 *,!*********-*-*---*,.:***-*-**--

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0.99400 J____. . . , ~-----~--------* ***-**-*-- + - -?- - -

0.992()0 .; .................

O.OOE+OO 1.00E-01 2.00E-01 3.00E-01 4.00£-01 5.00E-01 6.00E-01 7.00E-01 8.00E-01 9.00E-01 Energy of Average Lethargy Causing Fission (eV)

Figure A.5.3: Calculated k for the U02 Critical Benchmarks as a Function of EALF Geometry Tests Two trend tests were performed to determine if lattice/geometric parameters are adequately treated by SCALE 6.0. The first parameter is the fuel pin diameter. A small, statistically significant trend was found when the critical experiment analysis results were correlated to the fuel pin diameter. The second lattice parameter tested is the lattice pitch. A statistically significant trend on lattice pitch was found. The trend on pitch or pin diameter could be caused by the spectral trend found in the previous subsection.

Using NUREG/CR-6698 [A.2] equations from above and the data from Table A'.4.1, the predicted mean k as a function of pin diameter is:

k(Pin Diameter) = 0.99421 + ( 2.82E-03)*Pin Diameter where the pin diameter is in cm. The predicted mean k as a function of pitch is:

k(Pitch) = 0.99433 + ( 1.84E-03)*Pitch where lattice pitch is in cm.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 281 of 300 The tolerance band widths, using the second term of NUREG/CR-6698 [2] equation 23, for the MPS3 pin diameter (0.95 cm) and the pin pitch (1.26 cm) are 0.0048 and 0.0047 respectively.

Figures A.5.4 and A.5.5 graphically present kett as a function of the pin diameter and the lattice pitch.

1.00400

• ----**----*---**-----*------------~*-**-

1.00200 ...._ ..............._ .............. - ......__,.... - ........ _ .......... - ......_____..._. __,, _ _ _ _ _ _ _ ,, ___ ._...... _ ........____................_ ......_ ..__ ..... __ ,, .... _.............

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(.) 0.99600 .,............................................................... , ........................................................................................................................... w ..........................,.................................... ................................... .. ...........................

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~.....................................................................................................................¢¥.,........................................................................................................................................

99200 + ............................................ ,.............................................,............................................ ,.............................................. ,.............................................Y'..........................................., ............................................... ,

0.800 0.900 1.000 1.100 1.200 1.300 1.400 1.500 Pin Diameter (cm)

Figure A.5.4: Calculated k for the U02 Critical Benchmarks as a Function of Pin Diameter

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 282 of 300

~---*------**-*---***--**-***-**-*------- - - -

• 1.00400 ., ......................................................................................................................................................................................................................................................................................................................................... .

1.00200 .,................................................ _ ........................................................................_ ........................ _ ..........................., ......................................................................................... _ ..................................................

0.99400 .; ................................................................................................................................"'.................................................................................................................................................................................................. .

0.99200 . ,........................................ ,.......................................,.........................................

1.000 1.200 1.400 1.600 1.800 2.000 2.200 2.400 2.600 Pitch (cm)

Figure A.5.5: Calculated k for the U02 Critical Benchmarks as a Function of Fuel Pin Pitch Enrichment 235 235 The fuel to be stored in the racks may range in enrichment from 1.7 wt% U to 5 wt% U. It was determined that there is not a statistically significant trend on enrichment. However, to be conservative, both the zero slope and the calculated fit are used for determining the limiting k as a function of enrichment. Using NUREG/CR-6698 [A.2] equations from above and the data from Table A.4.1, the trend in the mean k is:

k(Enrichment) = 0.99699 + ( 0.01134)*Enrichment 235 where Enrichment is wt% U.

The maximum tolerance band width is 0.0051. Figure A.5.6 graphically presents the results.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 283 of 300

---

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1.00400 **, ....................................... .

1.00200 .!........ ......................... + .......

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0.99200 .! ...................................... ~.............. ......................... . ..... ......... .

1.00% 1.50% 2.00% 2.50% 3.00% 3.50% 4.00% 4.50% 5.00%

wt% U-235 Figure A.5.6: Calculated k for the U02 Critical Benchmarks as a Function of Enrichment Soluble Boron Content A fit of the calculated k's as a function of the soluble boron ppm was performed using the data from Table A.4.1 and Table A.5.2. The trend on soluble boron concentration is not statistically significance test compared to a zero slope. However, to be conservative, both the zero slope and the calculated fit are used for determining the limiting k as a function of soluble boron content.

The following equation is the best fit of the data fork versus soluble boron. Figure A.5.7 shows the results of the analysis. The uncertainty around the mean value given in the following equation is 0.0053 at 2550 ppm (maximum uncertainty in the range of interest).

k(ppm soluble boron) = 0.99739 + ( 6.32E-08)*ppm

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 284 of 300 Table A.5.2: U02 Critical Experiment With Soluble Boron Results with SCALE 6.0 and ENDF/8-VII i @enchmark I Case No; S9lµble .* MeaSUl'~ment . . kJtf L <

10 .. 1 Boron ppm Uncertainty

......... i < ......... , :*::::.,,*,: ._.\;,: *******.: .......

.. . . Ccfeltak)\ . .*. . *. . <J . . i i .. T >

LCT-8 1 1511 0.0012 0.9965 2 1334 0.0012 0.9969 3 1337 0.0012 0.9974 4 1183 0.0012 0.9969 5 1181 0.0012 0.9962 6 1034 0.0012 0.9966 7 1031 0.0012 0.9965 8 794 0.0012 0.9960 9 779 0.0012 0.9964 10 1245 0.0012 0.9966 11 1384 0.0012 0.9970 12 1348 0.0012 0.9970 13 1348 0.0012 0.9969 14 1363 0.0012 0.9967 15 1362 0.0012 0.9962 16 1158 0.0012 0.9973 17 921 0.0012 0.9963 LCT-11 2 1037 0.0032 0.9963 3 769 0.0032 0.9967 4 764 0.0032 0.9974 5 762 0.0032 0.9967 6 753 0.0032 0.9968 7 739 0.0032 0.9975 8 721 0.0032 0.9968 9 702 0.0032 0.9969 LCT-35 1 70 0.0018 0.9981 2 147.7 0.0019 0.9971 LCT-50 3 822 0.0010 0.9970 4 822 0.0010 0.9967 5 5030 0.0010 0.9983 6 5030 0.0010 0.9986 7 5030 0.0010 0.9988 LCT-51 1 C10 143 0.0020 0.9960 2 c11a 510 0.0024 0.9979 3 c11b 514 0.0024 0.9973 4 c11c 501 0.0024 0.9970 5 c11d 493 0.0024 0.9974 6 c11e 474 0.0024 0.9967 7 c11f 462 0.0024 0.9972 8 c11q 432 0.0024 0.9970 9 c12 217 0.0019 0.9968

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 285 of 300 1.0{)400 - , - - - - - - - - - - - - - - - - - * - - - - - - - - - - - - - -

1.00200 t ...................................................................................................................................................................................................................................................................................................

a, 1.00000

~

i "C 0.99800

!A V

~ ..,.

A V

¢ 3(.) ¢, ~ ¢, #

i;

~ $ ~~ ¢ 0  ;  ! ++ ,$- +

0.99600 ,i--<>------*w------*--------*------**-*--------*-----**-******--*------*

0.99400 .....................................................................................

0.99200 ., .........................................................-;-**

0 1000 2000 30{)0 40{)0 5000 Boron Concentration (ppm)

Figure A.5. 7: Calculated k for the U02 Critical Benchmarks as a Function of Soluble Boron Establishing the Bias and the Uncertainty To make the incorporation of the bias and bias uncertainty in the criticality analysis conservative, the most limiting bias and bias uncertainty from the trends in the range of interest is used. The lattice pitch for Westinghouse fuel 17x17 fuel is 1.26 cm. The bias from the pitch trend is 0.0034. The pin diameter is 0.95 cm where the bias as a function of pin diameter is 0.0031. The maximum bias from the enrichment trend is only 0.0027. The maximum bias as a function of soluble boron content is only 0.0026. There is a fairly strong trend with EALF. The bias increases with harder spectrum. The conservative bias to be used is 0.0034 or the EALF bias, whichever is greater. For all EALF less than 0.4 eV the bias from the pin pitch trend is most limiting. Table A.5.3 is a combination of the 0.0034 and the EALF values from Table A.5.1

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 286 of 300 The maximum uncertainty is also used but for this analysis if the bias is less than 0.0034, it is appropriate to subtract the differences in biases from the calculated uncertainties. The uncertainty from the pin pitch is 0.0047. Although the uncertainty from the pin diameter is slightly higher (0.0048) the bias for the pin diameter is less by 0.0003 so the pitch based uncertainty of 0.0047 should be used. Similarly, the uncertainty from the enrichment and soluble boron based trends are 0.0004 and 0.0006 higher than 0.0047 but the maximum bias from either of those two trends is less by 0.0007. Thus for all the trends other than the trend on EALF the conservative uncertainty is 0.0047. For all EALF under 0.4 eV the uncertainty based on the pin pitch is most limiting. Table A.5.3. provides the most limiting uncertainty using the uncertainties from Table A.5.1 (EALF) and 0.0047 from the other trends.

Table A.5.3: Bias and Uncertainty Based on the EALF Trend from U02 Critical Expenments Fresh Fuel Maximum Bias Uncertainty EALF (ev) 0.40 0.0034 0.0047 0.45 0.0037 0.0047 0.50 0.0039 0.0048 0.55 0.0042 0.0049 0.60 0.0044 0.0050 0.65 0.0047 0.0051 0.70 0.0049 0.0052 0.75 0.0052 0.0053 0.80 0.0054 0.0054 0.85 0.0056 0.0055 0.90 0.0059 0.0057 0.95 0.0061 0.0058 1.00 0.0064 0.0059 1.05 0.0066 0.0060 1.10 0.0069 0.0061 1.15 0.0071 0.0062 1.20 0.0074 0.0063 Impact of Cd and boron cases on the bias The mean unweighted k of the 232 cases used for the bias and trends is 0.99775. The mean unweighted k of the 17 cases including Cd is 0.99821. The delta k between these means is

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 287 of 300 0.0005. This is less than one third of the standard deviation of the set as a whole. Including the Cd set of experiments has no significant effect on the set of experiments used in the analysis.

Likewise the mean unweighted k of the 69 cases that contain boron is 0.99745. This is well within the uncertainty of the set as a whole so including boron cases has no significant effect on the set of experiments used in the analysis.

A.5.2 Statistical Analysis of MOX Critical Experiments Tables A.4.2 and A.4.3 provides *the raw results of the analysis of the MOX critical experiments.

From the calculated k's provided in Tables A.4.1, A.4.2 and A.4.3 there is a trend on the plutonium content. Figure A.5.8 shows this trend.

Since there is a strong trend on plutonium content, the critical experiments with plutonium content out of the range of spent fuel have been excluded. Only the critical experiments with 2 wt% plutonium or less are included in the trending analysis for MOX critical experiments. This set of MOX experiments consists of 136 critical experiments. Similar to the U02 benchmark set, for the Shapiro-Wilks test, we reject the null hypothesis for the MOX benchmark set at the 95%

confidence level. The SW Test statistic for the MOX benchmark set is 0.01, indicating that it would pass the SW test at the 99% confidence level. Figure A.5.9a shows the histogram for the 136 calculated k's, and A.5.9b shows the Q-Q plot. The visual inspection of the Q-Q plot does not show any clear indication of non-normal behavior, such as non-linear trend, uneven data point groupings, or skewness. Therefore even though the null hypothesis for the SW test is rejected at the 95% level, it is reasonable to treat the data as normally distributed.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 288 of 300 1.008 ~ - - - - - - - - - - - -- -- -- -----------

1.006 ...............................................

1.004 i' **-**************-******t*********-******i *-******** . . . . . . . . . . . .. . . . . . . . . . . . . . . .

1.002 1

i:

41

..:i::

0 .998 0.996 ************************************************ * ** * ...................

0 .994 t

0 .992 *~****** ............... **************---******-********-*--*-******************-*************************---********************--*******************-**********************

0 .99 0 .0 1.0 2 .0 3.0 4 .0 5.0 6 .0 7 .0 8 .0 Puwt%

Figure A.5.8: Calculated k for the Critical Benchmarks as a Function of Plutonium Content

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 289 of 300

1. 0 15 ********************************************************************************************************************************************************************************************* 1111**************************************************************************************************************

10 , ......................................................................................................................... .

0

~~~~~~~~~~~~~~/~~~#~#F~

<:,* <:, * <:,* <:,~ <:, * <:, * <:, * <:, * <:, * <:, * <:,* Q* Q' Q* Q* ",* ",* V V V '> V V k-<offectlve by O.OOOSOblns Figure A.5.9a: Distribution of Calculated ks for the MOX Critical Benchmarks

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 290 of 300 3 - . - - - - - - - - - - - - - - - - - - - - -- - - -

iv E

0 - - - - - - ~ -- ~ - - ~ - - ~

z0 1J U)

- - - - - - - - ~ - - - ~ - - - - - - - - ~ - - - - - - - -

0.9900 0.9920 0.9940 0.9960 0.9980 1.0000 1.0020 1.0040 1.0060 Data Figure A.5.9b: Q-Q Normality Plot for the MOX Benchmarks As with the U02 set, the MOX set has a tr~nd with EALF. That trend is:

k(EALF) = 0.999247 - 0.00264

• EALF This is graphically presented as Figure A.5 .10. The bias and uncertainty as a function of EALF is found on Table A.5.4. For accident conditions where the full Technical Specification soluble boron is used, the EALF may reach 0.75 eV. If partial voiding of the spent fuel pool is assumed, then the EALF approaches 1.0 eV but these cases are never close to limiting allowing a large amount of margin to cover the extrapolation above the highest EALF in the critical experiments of 0.71 eV. Spent fuel is not allowed in the new fuel vault, so the optimum moderation hard spectrum is not a concern for the burned fuel.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 291 of 300 1.005 * *

• 1.003 ! *

. i:

. +

Cl)  :

=-~~,~:.~. ~:!..:!.:*~. .

t; I

~

*r1 =

.. =****=****..:

• • • • • =.**=***=***=****=***=*

• .=**.=***=**.=***.=*.=**.~** ~====~**~***:*~** *~*=~*

• * ~.*;.~*.. ...

'iij 0

0.995

• 0.993 ; ......

............................ T ....... .

0.991

• 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 Ii 0 Energy of Average Lethargy Causing Fission (EALF) (eV)

L.*-*-*

Figure A.5.10: Calculated MOX Critical k as a Function of EALF

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 292 of 300 Table A.5.4: Bias and Uncertainty Based on the EALF Trend from MOX Critical Expenmen ts Fresh Fuel Maximum Bias Uncertainty EALF (ev) 0.30 0.0017 0.0079 0.35 0.0017 0.0083 0.40 0.0018 0.0088 0.45 0.0019 0.0094 0.50 0.0021 0.0099 0.55 0.0022 0.0104 0.60 0.0023 0.0110 0.65 0.0025 0.0115 0.70 0.0026 0.0121 0.75 0.0027 0.0126 0.80 0.0029 0.0132 0.85 0.0030 0.0137 0.90 0.0031 0.0143 0.95 0.0033 0.0148 1.00 0.0034 0.0154 1.05 0.0035 0.0159 1.10 0.0037 0.0165 1.15 0.0038 0.0170 1.20 0.0039 0.0176 The bias for the MOX is smaller than the U02 set, but the uncertainty for the MOX data is much higher. Until the combined total of the uncertainties in the spent fuel storage area is considered, it is not clear which is more limiting. Therefore both the U02 and MOX bias and uncertainty are used in determining the most limiting condition.

A.5.3 Subcritical Margin In the USA, the NRC has established subcritical margins for rack analysis. The subcritical margin for borated spent fuel pools, casks, and fully flooded dry storage racks is O when the analysis is performed with unborated water. This is actually saying the subcritical margin is contained in the uncredited soluble boron. To make sure there is sufficient soluble boron, analysis is also performed with soluble boron and a subcritical margin of 5% in k is required.

For dry storage racks analyzed with optimum moderation, the subcritical margin is 2% and 5%

with full moderation. In the analysis of 232 critical experiments, which generously cover the

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 293 of 300 range of expected conditions, the lowest calculated k was 0.9928. The additional 136 MOX experiments also support this subcritical margin since the lowest calculated k is 0.9920. The subcritical margin is more than sufficient.

A.6. Area of Applicability {Benchmark Applicability)

The critical benchmarks selected cover both the new fuel storage area and the spent fuel pool of MPS3. To summarize the range of the benchmark applicability (or area of applicability),

Tabl_e A.6.1 is provided below.

Benchmark A Comments '** .

Fissionable Material/Physical The fuel material is the same as in the Form benchmark ex eriments Enrichment {wt% U-235) 2.35 to 4.74 The analysis covers possible low enrichments in the future that would require extrapolation. Assuming the enrichment is zero (y intercept of the trend) the selected bias generously covers the extrapolated bias.

Therefore extrapolation of the bias to lower enrichments is justified.

An extrapolation from 4.74 to 5 wt%

will also be needed but in this direction the bias is decreasing so the data is ade uate for this extra olation.

Spectrum 0.0605 to Expected range in spent fuel pool

- EALF (eV) 0.8485 applications withoutvoiding:

0.1 to 0.8 eV Voided cases have significant margin to allow for the extrapolation.

For optimum moderation in new fuel storage area the EALF may be as high as 1.1 eV for this some extra olation of the data is done.

Lattice Characteristics Hex lattices have been excluded.

Type Square W 17x17 pin pitch is 1.26.

Pin Pitch cm 1.075 to 2.54

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 294 of 300 Parameter Rarige \( ..... ** Comments ****. ,., .

Assembly Spacing in Racks Oto 15.4 This covers all spacing. Neutron Distance between Assemblies transport through larger than 15.4 cm (cm) has a small effect on k. Note that the spacing is assumed to be filled with full density water. If the water density is less, this separation effectively decreases. Therefore, optimum moderation cases of wide spaced racks are covered.

Absorbers Oto 5030 All designs are within this range.

Soluble Boron Concentration ppm Absorbers Cd Absorber Although Cd is in panels, the inclusion Cd (for Ag-In-Cd rods) panels or exclusion had no significant effect on the bias and uncertainty so credit for control rods is acceptable.

Reflector Reflectors Most racks are reflected by water, Experiments included water adequately steel, and concrete which was and steel covered covered in the set of experiments.

Temperature Room Most criticality calculations are Temperature performed with the fuel at low (Reference 9 temperatures. A separate set of provides a experiments are used for a bias for up to temperature bias covered in 85° C) Reference 9.

Moderating material Water The moderator in all benchmark experiments is water, therefore water as a moderating material is covered A.7. Summary and Recommendations 232 U02 and 136 MOX critical experiments were analyzed with SCALE 6.0 and the 238 group ENFD/B-VII cross section set. The calculated k's were analyzed for trends using the statistical approach recommended in NUREG/CR-6698. Table A.7.1 provides the maximum bias and uncertainty for each trend.

For the spent fuel pool, the bias and uncertainty depends on the burnup since at low burnup the dominant fissile material is U-235 and at high burnup the dominant fissile material is Pu-239. In order to avoid trying to properly weight the critical experiments for the amount of U-235 and Pu-239, it is recommended to use two sets of bias and uncertainty, one from the fresh U02 critical experiments and one from the MOX critical experiments. The final bias and uncertainty to be used will be that which produces the highest 95/95 k. The U02 critical experiments have a

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 295 of 300 higher bias but lower uncertainty than the MOX experiments. Since the uncertainty is statistically combined with other uncertainties it is not possible to determine which set is more limiting until the other uncertainties due to factors such as manufacturing tolerances are determined.

Table A.7.2 provides the most limiting bias and uncertainty from both the U02 and MOX sets as a function of EALF.

Table A.7.1: sis Maximum Bias 002 Critical Experiments No trend n/a 0.0026 0.0050 EALF 0.99853-0.00491 *EALF (ev) Bias as a function Uncertainty as a of EALF is used function of EALF is used Fuel Pin Diameter 0.99421 +2.82E-03*Pin Dia. 0.0031 0.0048 (cm)

Lattice Pitch 0.99433+1.84E-03*Pitch (cm) 0.0034 0.0047 Enrichment 0.99699+ O.Ol 134*U235wt%t 0.0027 0.0051 Soluble Boron 0.99739+6.32E-08*ppm 0.0026 0.0053 MOX Critical Experiments No trend n/a 0.0011 0.0065 EALF 0.999247-0.00264*EALF (ev) Used as a Used as a function of EALF function of EALF

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 296 of 300 Table A.7.2: Final Bias and Uncertainty (For burned fuel calculate with both U02 and MOX bias and uncertainty and use the set that provides the highest 95/95 k)

U02 MOX Maximum Maximum Bias Uncertainty Bias Uncertainty EALF(evl EALF (evl 0.35 0.0034 0.0048 0.35 0.0017 0.0083 0.40 0.0034 0.0048 0.40 0.0018 0.0088 0.45 0.0037 0.0048 0.45 0.0019 0.0094 0.50 0.0039 0.0048 0.50 0.0021 0.0099 0.55 0.0042 0.0049 0.55 0.0022 0.0104 0.60 0.0044 0.0050 0.60 0.0023 0.0110 0.65 0.0047 0.0051 0.65 0.0025 0.0115 0.70 0.0049 0.0052 0.70 0.0026 0.0121 0.75 0.0052 0.0053 0.75 0.0027 0.0126 0.80 0.0054 0.0054 0.80 0.0029 0.0132 0.85 0.0056 0.0055 0.85 0.0030 0.0137 0.90 0.0059 0.0057 0.90 0.0031 0.0143 0.95 0.0061 0.0058 0.95 0.0033 0.0148 1.00 0.0064 0.0059 1.00 0.0034 0.0154 1.05 0.0066 0.0060 1.05 0.0035 0.0159 1.10 0.0069 0.0061 1.10 0.0037 0.0165 1.15 0.0071 0.0062 1.15 0.0038 0.0170 1.20 0.0074 0.0063 1.20 0.0039 0.0176 For unburned fuel in the spent fuel pool and new fuel vault use only the U02 set from the above.

A.8. Temperature Bias All of the critical experiments used thus far have been at room temperature. There could be a bias in k in the temperature range of interest to spent fuel pools and dry storage racks (0 to 100 C). There is one critical benchmark evaluation in the OECD/NEA handbook [A.3] that performed measurements with elevated temperatures in this range, LEU-COMP-THERM-046 (shortened to LCT-046). LCT-046 consists of 22 experiments but the last 5 experiments contain copper rods. Since copper is not in North Anna's spent fuel pool only the first 17 experiments are analyzed.

The 17 LCT-046 experiments have been analyzed using SCALE 6.0 [A.2] and the 238 ENDF/B-VII cross section library. Section 3 of LCT-046 provides the details for analysis of the critical

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 297 of 300 benchmark. The SCALE models used follows that specification. All the expansion factors from Table 29 of LCT-046 were applied to all the x-y dimensions. That means that the same stainless steel component expansion factor was applied to pitch and the inner and outer diameter of the clad. This is consistent with the MCNP samples given in the Appendix of LCT-046. For the axial expansion only the fuel was expanded. As with the MCNP sample input, the same expansion factor was used for the radius and the axial direction. The fuel column is 54.84 cm long (unexpanded). Due to the control rod bottom plug which hangs into the fuel region, the fuel is modeled as a 53.44 cm long (unexpanded) zone followed by two shorter zones. For this effort only the 53.44 cm long segment was expanded axially by the expansion factor. This approach assures that the axial position of the control rod and bottom plug is not changed.

Table A.8.1 shows the corrected SCALE 6.0 ENDF/8-VII results for the 17 critical experiments.

Corrected results in this case means they were divided by the k of the benchmark which was not quite 1.0.

Table A.8.1: LCT-046 with Full Thermal Expansion Calculated with SCALE 6.0 and ENDF/B-VII Case Temperature <Kl corrected**scAtE:k*.** SCAl..Esigma 1 297.05 0.998901 0.00007 2 310.41 0.998867 0.00007 3 315.43 0.998710 0.00007 4 319.96 0.998915 0.00007 5 324.93 0.998558 0.00007 6 332.53 0.998697 0.00007 7 287.22 0.999163 0.00007 8 315.91 0.998854 0.00006 9 330.27 0.998669 0.00007 10 337.44 0.998566 0.00007 11 351.99 0.998625 0.00007 12 303.60 0.998632 0.00007 13 312.95 0.998616 0.00007 14 321.16 0.998511 0.00007 15 328.24 0.998258 0.00007 16 338.26 0.998147 0.00007 17 358.31 0.998057 0.00007 Figure A.8.1 plots the results of the analysis as a function of case. As can be seen from this plot there does seem to be a small trend with temperature. Figure A.8.2 is the data plotted against temperature with the least squares linear fit. The nominal slope of the fit is -1.14E-05

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 298 of 300

~k/~C. Using the EXCEL regression function the most limiting slope with 95% certainty is -1.?E-05 ~k/~C.

k versus Case (three sets with increasing temperature in each set) 0.999400 ---*---*-*------*---******** ~----*--**-------**-* -**----*--***-------------

0.999200 *+************************************************************************************************************************** ........................................................................................... **************************************************************************************************************************

X 0.999000 - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

.:.i: >< X* X X

] 0.998800 ----**-*--*--*---**--*----<t------

... X j o.998600 . -------------------------- - -*-----x______)(__ ------* ---- ........'.':______x ___.x._ __x ---X------:::---- *-**------------------*-

0.998400 -t------------------------------

X 0 .9 9 8 2 0 0 .. **********-**************-*-***************-****-**-**-**---*-****-******-*------********* *-**-****************-*****--*-**********-**-**-********************- **********************-*-*-***-******-****-******-*****************-*-************************-********-***

X Rectangular Set Rounded Set 4Gd Rods Set X 0.998000 *,-----, -**-*-*******-*************.-*-*-**************************.,-**--* ~*******---......--***-************1***********-*-- --*,.......---*-*******r**************** ...............*----*--*****1 0 2 4 6 8 10 12 14 16 18 Case Number Figure A.8.1: LCT-046 Corrected Calculated k per Case

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 299 of 300 0.99980 0.99960 0.99940 0.99920

~

"Cl Ill "a 0.99900

~

u 0.99880 0.99860 0.99840 0.99820 0 10 20 30 40 50 60 70 80 90 Temperature (0 C)

Figure A.8.2: LCT-046 k versus Temperature The analysis of the only set of thermal critical experiments in the International Handbook that uses elevated temperatures in the range of O to 100 C has shown a small increase in the bias with temperature. This increase can be conservatively handled by a bias from room temperature of 1.?E-05 ~k/°C. This bias is the lower (most negative slope) 95% confidence slope of the fit line.

Criticality Safety Evaluation Report Serial No.18-039 Docket No. 50-423 Attachment 6, Page 300 of 300 A.9. Appendix References

[A.1] SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, ORNUTM-2005/39, Version 6, Volumes 1-3, January 2009.

[A.2] J.C. Dean and R.W. Tayloe, Jr., Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, Nuclear Regulatory Commission, Washington, DC January 2001

[A.3] International Handbook of Evaluated Criticality Safety Benchmark Experiments,

• NEA/NSC/DOC(95)3, Volumes IV and VI, Nuclear Energy Agency, OECD, Paris, September, 2010.

[A.4] D. E. Mueller, K. R. Elam, and P. B. Fox, Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data, NUREG/CR-6979 (ORNL/TM-2007/083),

prepared for the US Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., September 2008.

[A.5] F. Femex, "Programme HTC-Phase 1 : Reseaux de crayons dans l'eau pure (Water-moderated and reflected simple arrays) Reevaluation des experiences," DSU/SEC/T/2005-33/D.R., Institut de Radioprotection et de Surete Nucleaire, 2008.

[A.6] F. Femex, ProgrammeHTC - Phase 2 : Reseaux simples en eau empoisonnee (bore et gadolinium) (Reflected simple arrays moderated by poisoned water with gadolinium or boron) Reevaluation des experiences, DSU/SEC/T/2005-38/D.R., Institut de Radioprotection et de Surete Nucleaire, 2008.

[A.7] F. Femex, Programme HTC-Phase 3: Configurations "stockage en piscine" (Pool storage)

Reevaluation des experiences, DSU/SEC/T/2005-37/D.R., Institut de Radioprotection et de Surete Nucleaire, 2008.

[A.8] F. Femex, Programme HTC- Phase 4: Configurations "chateaux de transport" Reevaluation des experiences, DSU/SEC/T/2005-36/D.R., Institut de Radioprotection et de Surete Nucleaire, 2008.

[A.9] J. P. Royston, "Approximating the Shapiro-Wilk W-test for non-normality", Statistics and Computing, Vol. 2,, September 1992.

[A.1 O] Kleinbaum, Kupper, and Muller, Applied Regression Analysis and Other Multivariable Methods, Second Edition, page 48, PWS-KENT Publishing Company, Boston, MA 1988.

[A.11] "PROPHET StatGuide: Examining normality test results,"

http://www.basic.northwestern.edu/statguidefiles/n-dist exam res.html, located on 6/8/09.

[A.12] R. Mark Sirkin, Statistics for the Social Sciences, Third Edition, 2005, page 245, Sage Publications, Thousand Oaks, CA.

Serial No.18-039 Docket No. 50-423 ATTACHMENT 7 SPENT FUEL POOL BORON DILUTION ANALYSIS DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 2 of 21 Table of Contents

1. lntroduction ...................................................................................................................3

2. Spent Fuel Pool and Related System Features ........................................................... .4 2.1. Spent Fuel Pool .............................................................................,.............................. 4 2.2. Spent Fuel Pool Storage Racks .................................................................................... 4 2.3. Spent Fuel Pool Cooling ............................................................................................... 5 2.4. Spent Fuel Pool lnstrumentation ................................................................................... 5

. 2.5. Spent Fuel Pool Administrative Procedure .................................................................... 6 2.6. Boration Sources ..........................................................................................................6

3. Spent Fuel Pool Dilution Event ..................................................................................... 7 3.1. Calculation of SFP Volume ........................................................................................... 7 3.2. Plant Equipment Operator Response ........................................................................... 7 3.3. Calculation of SFP Boron Dilution Water Flow Rate ..................................................... 8

4. Dilution Source Path Evaluation .................................... '. ............................................ 10 4.1. Unintentional Overfilling from Primary Grade Water (PGS) System ............................ 1O 4.2. Unintentional overfilling from normally isolated Service Water System ....................... 1O 4.3. SFP Cooling System Heat Exchanger Tube Rupture .................................................. 1O 4.4. Excess Demineralized Water Addition from Rinsing Transfer Cask and Equipment ... 11 4.5. Unintentional overfilling using RWST connection to SFP ............................................ 11

5. Pipe Breaks and Leaks ............................................................................................... 12 5.1. Component Cooling, Fire Protection, and Hot Water Pre-Heating Water Line Breaks 13 5.2. Hot Water Heating (HVH) Line Break ......................................................................... 13 5.3. Roof Drain Line Break ................................................................................................ 15 5.4. Fuel Pool Cooling Lines .............................................................................................. 16 5.5. Fire Water Hose Station Pipe Leakage ....................................................................... 16 5.6. Demineralized Water Pipe and Valve Leakage ........................................................... 16 5.7. Hot Water Heating System leakage ............................................................................ 16 5.8. Domestic Water Pipe Leakage ................................................................................... 16 5.9. Roof Drain pipe leakage .: ........................................................................................... 17

6. Securing the Dilution Source ...................................................................................... 18

7. Conclusions ................................................................................................................ 19

8. References .................................................................................................................21

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 3 of 21

1. Introduction A Spent Fuel Pool (SFP) criticality reanalysis has been completed for crediting soluble boron in the Millstone Unit 3 (MPS3} SFP under both normal and accident conditions. As a result of the soluble boron credit, a boron dilution analysis is required and is presented herein. This analysis includes the following plant specific features and potential events:

• Level indicator instrumentation

• Administrative procedures

• Boration sources

• Dilution sources

• Dilution flow rates

• Boron dilution initiating events

• Boron dilution times and volumes

• Boron dilution mitigation actions The boron dilution analysis, along with the criticality reanalysis, ensures that sufficient time is available to detect and mitigate the dilution before the design basis limit on the effective multiplication factor (kett = 0.95) is reached.

The new, proposed Technical Specifications will change the minimum SFP boron concentration to 2600 ppm. This dilution analysis will assume the dilution event starts at a boron concentration of 2300 ppm and the lower boron concentration limit is 700 ppm. The starting boron concentration of 2300 ppm was conservatively chosen prior to the completion of the criticality accident analysis that established 2600 ppm as the Technical Specifications minimum boron value.

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 4 of 21

2. Spent Fuel Pool and Related System Features This section provides background information on the SFP and related systems.

2. 1. Spent Fuel Pool The spent fuel pool is an L-shaped structure located in the southwestern quadrant of the fuel building. Two adjacent areas, which are accessible from the spent fuel pool by means of sealable gates, are the transfer canal and the spent fuel shipping cask pit.

The spent fuel pool is designed to accommodate fuel racks that store new and spent fuel assemblies. The storage racks are located under water in the spent fuel pool. (UFSAR Section 9.1.2.2, Ref. 1).

2.2. Spent Fuel Pool Storage Racks There are three different fuel storage regions in the spent fuel pool. The spent fuel storage pool contains 350 Region 1 storage locations, 673 Region 2 storage locations and 756 Region 3 storage locations, for a total of 1779 total available fuel storage locations. An additional Region 2 rack with 81 storage locations has been licensed by the NRC, but is not physically installed in the spent fuel pool. If this additional rack is installed, the Region 2 storage capacity is 754 storage locations. The total storage capacity of the spent fuel pool is limited to no more than 1860 fuel assemblies.

The Region 1 fuel storage racks are made up of 5 rack modules. Each rack module is free standing and is made up of a 7 by 1O array of storage cells. The Region 1 racks have a nominal 10.0 inch (North/South) and a nominal 10.455 inch (East/West) center-to-center spacing between adjacent fuel storage locations. The Region 1 storage racks have a neutron flux trap design, which uses BORAL as the active neutron absorber. BORAL panels are included on all peripheral rack locations.

The Region 2 fuel storage racks are made up of 9 rack modules. Each rack module is free standing with several different storage array sizes. The Region 2 racks have a nominal 9.017 inch center-to-center spacing between adjacent fuel storage locations. Like the Region 1 racks, the Region 2 storage racks use BORAL as the active neutron absorber. The Region 2 storage

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 5 of 21 racks have a single BORAL panel between adjacent fuel assemblies. BORAL panels .are included on all peripheral rack locations.

The Region 3 fuel storage racks are made up of 21 rack modules. Each rack module is free standing and is made up of a 6 by 6 array of storage cells. Each rack consists of cells welded to a grid base and welded together at the top through an upper grid to form an integral structure.

The Region 3 racks have a nominal 10.35 inch center-to-center spacing between adjacent fuel s~orage locations. The Region 3 storage racks have a neutron flux trap design, which uses Boraflex as the active neutron absorber. However, Boraflex is no longer credited in the criticality analysis of these racks. Each rack module is provided with adjustable leveling pads at the center of the four corner cells within the module. (UFSAR Section 9.1.2.2, Ref. 1).

2.3. Spent Fuel Pool Cooling Cooling for the spent fuel pool consists of two cooling mechanisms. The first is the active cooling provided by the fuel pool heat exchangers. The spent fuel pool water flows from the fuel pool discharge through either of the two fuel pool cooling pumps and through the tube side of a fuel pool cooler, and then returns to the fuel pool. One fuel pool cooling pump and cooler are normally in service. Cooling for the fuel pool coolers is provided by the reactor plant component cooling water system. The second mechanism is the passive cooling provided by evaporative cooling from the surface of the pool. The spent fuel pool cooling system has been analyzed to remove the decay heat load of up to 1960 fuel assemblies and maintain a bulk pool temperature at or below 150°F using a single train of spent fuel pool cooling. (UFSAR Section 9.1.3.2, Ref.

1).

2.4. Spent Fuel Pool Instrumentation The fuel pool has redundant safety grade low level alarm lights and temperature indicators provided in the main control room. Non-safety grade level indication is provided locally and high and low level alarms are provided both locally and in the main control room. Continuous non-safety augmented quality wide range level indication is provided remotely in the Auxiliary Building. (UFSAR Section 9.1.3.5, Ref. 1).

If the pool were to overflow, the water would eventually collect into and auto-start the fuel building sump pump (FSAR Section 9.3.3.3, Ref. 1). The control room would get an indication

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 6 of 21 of the pump's auto-start. This analysis assumes that these automatic safety systems failed, and the only way operations can discover a dilution event is during operator rounds ..

2.5. Spent Fuel Pool Administrative Procedure Currently, Technical Specifications (TS) requires the soluble boron concentration in the SFP to always be greater than or equal to 800 ppm. The SFP soluble boron concentration must also be greater than or equal to 2600 ppm during refueling operations when fuel is being transferred between the fuel pool and refueling cavity to preclude uncontrolled dilution of the filled portion of the RCS (UFSAR Section 9.1.3.2, Ref. 1). Chemistry practice maintains the SFP greater than or equal to 2600 ppm since that is the boron concentration assumed in the "Fuel Pool Cooling and Purification" operating procedure. TS requirements are proposed to be changed to require the soluble boron concentration to be maintained .:: 2600 ppm. The spent fuel pool boron concentration must be measured every 7 days in accordance with plant surveillance procedure.

Plant surveillance procedure states the shift readings must be checked and completed in the first 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of each shift. These shifts start at 0700 and 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br />.

2. 6. Boration Sources Normal makeup to the fuel pool, necessitated by losses due to evaporation, is primary grade water from the primary grade water system. Borated water from the RWST can be used to fill the fuel pool at a concentration matching that used in the refueling cavity during refueling operations (UFSAR Section 9.1.3.2, Ref. 1). The RWST is maintained between 2700 ppm and 2900 ppm (LCO 3.5.4, Ref. 2).

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 7 of 21

3. Spent Fuel Pool Dilution Event

3. 1. Calculation of SFP Volume Excluding the Spent Fuel Shipping Cask Loading Area, the gate areas, Spent Fuel Shipping Cask Storage Area, and transfer tube, the SFP gross volume is 59,712.08 ft3 up to the low level alarm. This is conservative since the minimum gross volume results in the minimum volume required for dilution. The low level alarm is used for conservatism as this will have the least amount of boric acid to be diluted. The borated water volume is calculated by subtracting fuel .

assembly (FA) volume, RCCA volume, and Fuel Rack volume from the gross SFP volume. This calculation is as follows:

SFPborated = SFPlowlevelalarm -SFPFA -SFPRCCA -SFPFuelRack (3.1)

SFPborated = 59,712.08 ft 3 - 5,040.6 ft 3 - 582.9 ft 3 - 1,511 ft 3 = 52,577.6 ft 3 Where:

• SF Plow level alarm= 59,712.08 ft3

• SFPFA = 5040.6 ft3 (=1860 FAs

• 2.71 ft3/FA)

• SFPRCCA = 582.9 ft3 (=1860

• 0.313373 ft3/RCCA)

• SFPFuelRack = 1511 ft3 Note: This calculation models the SFP with the same number of RCCAs as the number of FAs; this is conservative in that it reduces the volume of the SFP water.

This calculation results in a borated water volume of 52,577.6 ft3 or 393,308 gallons.

3.2. Plant Equipment Operator Response As discussed in Section 2.4, the automatic safety systems that could detect a dilution event are assumed to have failed leaving operator rounds as the only method to discover the event. Plant surveillance procedure states the shift readings must be checked and completed in the first 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of each shift. The ranges of times that the shift readings must be checked and completed are 0700-1000 and 1900-2200. The maximum amount of time that can pass between checking these procedural requirements would be 0700-2200 and 1900-1000 on a 2400 hour0.0278 days <br />0.667 hours <br />0.00397 weeks <br />9.132e-4 months <br /> scale

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 8 of 21 (beginning of earlier or later shift to end of the next shift). The time transpired between these two times is 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and is used as the conservative estimate of PEO response time.

3.3. Calculation of SFP Boron Dilution Water Flow Rate The total amount of unborated water that can be added to the SFP to reduce the boron concentration from 2300 ppm to 700 ppm is determined using continuous dilution methodology (Ref. 3) consistent with the approach employed by Millstone Unit 2 SFP dilution calculation (Ref.

4 & 5). The continuous dilution method models unborated water being added at a constant rate with an equally constant rate of removal of a homogenous mixture of borated water. For conservatism, the SFP volume used excludes areas outside the SFP liner envelope (see Section 3.1 for further detail).

The equation used to calculate the volume required to dilute the SFP boron concentration to 700 ppm is derived in Air Contaminants and Industrial Hygiene Ventilation (Ref. 3).

Vin = VsF~

• ln (~;) (3.2) 393,308 gal* ln ( 2300ppm) 700ppm

= 467,873 gal Where:

• Vin= Required unborated water to dilute to C1 (gallons)

• VsFP= Volume of SFP to the low level alarm calculated in Section 3.1 (393,308 gallons)

• C0 = Initial boron concentration in SFP (2300 ppm)

• c1 = Final boron concentration in SFP (700 ppm)

The maximum allowable dilution flow rate is determined by dividing the required unborated water volume to dilute the SFP (Vin) by the maximum response time by a Plant Equipment Operator (PEO) which is 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />.

Vin 1hr Qdilution = tpEo

• 60 min (3.3) 467,873 gal * --2.!!!:_ = SZO m 15 hrs 60 min gp

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 9 of 21 Where:

• tPEo = Maximum response time by the PEO (15 hrs)

• Qdilution = Minimum flow rate to dilute the SFP below 700 ppm (467,873 gpm)

Therefore, the minimum flow rate that could dilute the SFP from 2300 ppm to 700 ppm in 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> is 520 gpm. Note that the SFP would need to fill to overflowing prior to applying the feed and bleed methodology, however, the outlined approach is conservative since it results in boron exiting the SFP immediately.

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 10 of 21

4. Dilution Source Path Evaluation This section evaluates the potential for dilution of the SFP from the SFP Cooling System and from external sources within the SFP building.

4. 1. Unintentional Overfilling from Primary Grade Water (PGS) System The rated flow from the PGS pumps is 450 gpm combined which is less than the maximum allowable dilution flow rate of 520 gpm (Section 3.3). Since a flow rate of 520 gpm dilutes the SFP to 700 ppm in 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, a flow rate of 450 gpm would take 17.33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> to dilute the pool the same amount.

15 hrs

• 520 gpm 450 gpm = 17.33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> 4.2. Unintentional overfilling from normally isolated Service Water System Pipe 3-SFC-004-41-3(2-), which provides access to the SFP from the Service Water System, is only opened during emergency operation. If standard operating procedures fail, this method is deemed acceptable for filling the SFP. Maximum operating conditions for the Service Water System are listed as 66 psig and 95/80°F per FSAR Tables 3.6-4 and 3.6-3 (Ref. 1). Using the methodology described in Section 5, the Service Water flow rate to the SFP is about 4000 gpm based on an open ended pipe with a driving pressure of 66 psig and a loss coefficient of 1 (pipe exit, Ref. 6). However, this method will only be used in emergency situations. Additionally, the water level of the SFP is monitored during this process of filling the pool and it is unreasonable for the PEO to be away from the pool for an extended period of time in an emergency situation.

Therefore, it is not credible for the Service Water System to dilute the SFP beyond the critical lower boron limit of 700 ppm.

4.3. SFP Cooling System Heat Exchanger Tube Rupture The SFP cooling heat exchangers have component cooling water (CCP) on the shell side and SFP Fuel Pool Cooling water on the tube side. Conservatively, the CCP (non-borated) water is considered to flow directly from high CCP pressure to atmosphere, giving the maximum possible flow through a potential tube break. FSAR Table 3.6-4 (Ref. 1) indicates that the CCP system is moderate energy and has maximum operating conditions of 186 psig and 137°F in the fuel building. However, for conservatism, the maximum CCP design pressure and temperature

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 11 of21 (250 psig and 150°F) are used for this calculation. This state point is used with Equations 5.1 and 5.2 with a tube I.D. of 5/8". This produces a flow of 152 gpm per ruptured tube. Thus, 4 tubes would have to break and leak out directly to atmospheric conditions in the SFP to exceed the maximum allowable dilution flow rate of 520 gpm (Section 3.3). Assuming one ruptured tube, a flow rate of 152 gpm would take 51 hours5.902778e-4 days <br />0.0142 hours <br />8.43254e-5 weeks <br />1.94055e-5 months <br /> to dilute the SFP to 700 ppm.

15 hrs

• 520 gpm 1 5 2gpm

= 51 hours5.902778e-4 days <br />0.0142 hours <br />8.43254e-5 weeks <br />1.94055e-5 months <br /> 4.4. Excess Demineralized Water Addition from Rinsing Transfer Cask and Equipment Per procedure, purification demineralizer flow shall not exceed 60 gpm which is less than the maximum allowable dilution flow rate of 520 gpm (Section 3.3). A flow rate of 60 gpm would take 130 hours0.0015 days <br />0.0361 hours <br />2.149471e-4 weeks <br />4.9465e-5 months <br /> to dilute the SFP to 700 ppm.

15 hrs

• 520 gpm 60 gpm

= 130 hours0.0015 days <br />0.0361 hours <br />2.149471e-4 weeks <br />4.9465e-5 months <br /> 4.5. Unintentional overfilling using RWST connection to SFP The minimum RWST borated concentration is 2700 ppm per Table 6.3-9 of the FSAR (Ref. 1) and would not dilute the SFP, as the RWST system minimum borated concentration of 2700 ppm is higher than the initial SFP boron concentration of 2300 ppm and the proposed T.S. limit of 2600 ppm.

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 12 of 21

5. Pipe Breaks and Leaks This section evaluates potential pipe breaks that are in close proximity to the SFP.

The lines in piping systems near the SFP are considered moderate energy lines since their design pressure and temperature (or maximum normal operating pressure and temperature for the Hot Water Pre-Heating system) are both less than 200°F and 275 psig (high energy designation requires excess of 275 psig or 200°F; UFSAR Section 3.6.1.1.2, Ref. 1). In a moderate energy piping system, through-wall leakage cracks are modeled for these systems instead of double ended guillotine breaks per FSAR Sections 3.6.2.1.2 and 3.6.2.1.3 (Ref. 1).

The area of a crack is considered equal to that of a rectangle one-half pipe diameter in length and one-half pipe wall thickness in width. For conservatism, the outside diameter of the pipe is used in this calculation as the diameter is in the numerator, and therefore a larger value is more conservative. The flow rate through this crack area can be calculated using Equation 3-14 of Crane Technical Paper 41 O (Ref. 6).

Q= (5.1)

Where:

• Q = Flow rate through the opening (gpm)

• b.P = Pressure differential across the opening (psi)

• d = equivalent hydraulic diameter of round break area (in). Defined in Eq. 5.2.

• K = Loss coefficient of the pipe exit, in this evaluation a value of 1.5 (-) is used (summation of exit and entrance losses from Ref. 6)

• p = Density of water before the pipe exit (lb/ft3)

(5.2)

Where:

• L = Length of rectangular crack (in)

• W = Width of rectangular crack (in)

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 13 of 21

5. 1. Component Cooling, Fire Protection, and Hot Water Pre-Heating Water Line Breaks The CCP, FPW, and HVG pipes are classified as moderate energy lines. Using Equations 5.1 and 5.2, the maximum flow value from these lines is 23.3 gpm which results from a postulated crack in line 3-FPW-006-252-4. A flow rate of 23.3 gpm would take approximately 335 hours0.00388 days <br />0.0931 hours <br />5.539021e-4 weeks <br />1.274675e-4 months <br /> to dilute the SFP to 700 ppm.

15 hrs

• 520 gpm 23 .3 gpm

= 335 hours0.00388 days <br />0.0931 hours <br />5.539021e-4 weeks <br />1.274675e-4 months <br /> 5.2. Hot Water Heating (HVH) Line Break The HVH system has a fixed initial inventory in the lines as well as a hot water expansion tank since it is a closed loop system. A break in this system would first release inventory and then fill the system with make-up pump flow.

To maintain 700 ppm of boron in the SFP, the maximum permissible amount of unborated water

_in the 15 hour1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> period between PEO shift checks is 467,873 gal (Section 3.3). The hot water heating make-up pump has an approximate runout flow of 200 gpm. If the pump runs continuously for 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, the total theoretical flow passed through is 180,000 gal (= 200 gpm

• 15 hr* 60 min/hr). The remaining unborated water necessary to dilute the SFP, providing the make-up flow goes directly to the SFP, would be 287,873 gal (= 467,873 gal -180,000 gal) or 38,483 ft3 (using 7.48052 gal/ft3).

The volume of the HVH tank is calculated using the three following equations from Ref. 7:

/Dz Vcylinder = Tr 4

• H (5.3)

Vcylinder - Tr (4,5 ft)2 4

• 11 f t -- 17 5 f t 3 Where:

• ID= Internal Diameter (4'-6")

• H = Length of Cylinder (11 ')

VDished Head,Ellipsoidal -

rrR z (2a) 3 (5.4)

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 14 of 21

_ ft _

Vvished Head,Ellipsoidal - 1f * ( 2.25 ft )2

• 2*1.125 3

- 12 ft 3

Where:

• R = Tank Radius (2'-3")

• a = Dished head height (1 '-1.5" - radius to height ratio is 2: 1)

Vrank = Vcylinder + 2

• VDished Head,Ellipsoidal (5.5)

Vrank = 175 ft 3 + 2

• 12 ft 3 = 199 ft 3 The largest pipe diameter associated with the HVH system is 1O" with internal transverse area of 0.5475 ft2 (at Standard Schedule) (Ref. 6). With this area, there would need to be 69,925

(=(38,483ft3-199ft3)/0.5475ft2) linear feet of 1O" piping to release its inventory into the SFP to dilute to below 700 ppm.

Since 70,000 linear feet (more than 13 miles) of 10" piping does not exist in this system, it is reasonable to conclude that HVH line breaks would not result in an unacceptable bulk dilution of the SFP.

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 15 of 21

5. 3. Roof Drain Line Break There are roof drain lines that travel above the SFP to disperse environmental precipitation.

These lines are shown on Figure 5-1. Using this drawing in conjunction with other drawings, roof areas to consider for rain precipitation can be determined. The roof areas considered are highlighted in Figure 5.1.

Figure 5-1: Roof Drain Surface Area f &.JH},t~~1N&

1 ,*,1,$,;'t.

v'"'" l ccmr.: ~.4$

'"---v,x:i<t.~ -.'<><:>%1

} nim

,.. ~-.****** ,********.:,---~~,----"~--~ v*' V'f <<;,W 11 I a:J1~,eA~J*!{;a u~ i~tt:td

• 1 The roof areas outside the outlined sections A and B illustrated in Figure 5.1 are not considered to flow to the drain lines that flow within the SFP area because they either have a lower roof elevation, or are segregated due to a parapet. The surface areas of Section A and B are 2 2 4156.25 ft and 2770.7 ft , respectively. The total roof drain (DNR) drainage surface area is

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 16 of 21 6927 ft2. The rainfall required to collect a critical dilution volume is calculated by the following equation:

H _ Vin (gal)

Rain - SAvNR*7.48052 gal/ft 3 (5.6)

The inches of rain required to dilute the SFP to 700 ppm is calculated using Equation 5.6 and is as follows:

H . = 467,873 gal = 9 ft = 108 in Ram 6927 ft 2 *7.48052 gal/ft 3 This amount of rain is far more than what would conservatively occur in 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />. Therefore, excessive dilution due to rain is not credible.

5.4. Fuel Pool Cooling Lines Fuel Pool cooling lines are borated to the same value as the SFP as this system is fed from and exhausts to the pool itself. Breaks in these lines will not dilute the SFP or borate the system.

5.5. Fire Water Hose Station Pipe Leakage A walk-down did not identify any fire water hose stations in the vicinity of the SFP. Further, fire protection pipe leakage would be bounded by the crack in the fire protection lines considered in Section 5.1.

5. 6. Demineralized Water Pipe and Valve Leakage A walk-down did not identify any demineralized water piping in the vicinity of the SFP. Further, per FSAR Section 9.2.8.2 (Ref. 1), the PGS consists of demineralized and deaerated *water. As procedure fails to differentiate between PGS and Demineralized water pipes and valves, these lines are considered to be PGS lines and the leakage would be ultimately bounded by the pump flow discussed in Section 4.1.

5.7. Hot Water Heating System leakage These lines are addressed in Section 5.2.

5.8. Domestic Water Pipe Leakage

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 17 of 21 A walk-down did not identify any domestic water pipes in the vicinity of the SFP. Further, per FSAR Section 9.2.2.1.2 (Ref. 1), the CCP system supplies make up water to reactor plant component cooling water piping and sill cock plumbing in various buildings. Thus the CCP piping has a domestic water function (plumbing). As procedure fails to differentiate between domestic water and CCP lines and valves, the domestic water is considered in the modeling of CCP lines justified in Section 5.1.

5. 9. Roof Drain pipe leakage These lines are addressed in Section 5.4.

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 18 of 21

6. Securing the Dilution Source Two activities are needed to cease a dilution event: the dilution needs to be discovered, and then the dilution source needs to be secured. Sections 4 and 5 address the possible dilution sources and their associated flow rates. Table 6-1 shows the amount of time needed to dilute the pool to 700 ppm using the credible source flow rates discussed in the previous sections.

Table 6-1: Times to 700 ppm for Various Dilution Sources Flow Time to Section Dilution Initialization Scenario Rate 700 ppm Calculated (Qpm) (hrs)

Unintentional Overfilling from Primary Grade Water System 450 17.33 4.1 SFP Heat Exchanger Tube Rupture 152 51 4.3 Excess demin water from rinsing transfer cask and equipment 60 130 4.4 CCP, FPW, and HVG Line Breaks 23.3 335 5.1 Table 6-1 shows that unintentionally overfilling the SFP from the PG water system is the bounding scenario. Using the assumptions from Sections 2.4 and 3.2, 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> are needed to discover the dilution event. Therefore, operations must be able to secure the PG water system in 2.33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br />.

When a fuel pool high level alarm is received locally or in the control room, station procedures direct operations to close the PG valve, PG bypass valve and to stop makeup to the spent fuel pool. While there is no formal time requirement on these procedural actions, discussions with Operations estimated that the process should take no more than 15 minutes from the time of discovery. Note that an operator will already be at the SFP since the event was assumed to be discovered via operator rounds. Since the estimated mitigation time has over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or 89%

margin to the required mitigation time limit, it can be reasonably assumed that a dilution event from the PG water system can be discovered and secured before the SFP reaches 700 ppm.

The other dilution scenarios have much more time to determine and secure the source of the dilution; therefore they are bounded by the dilution scenario from the PG Water System.

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 19 of 21

7. Conclusions This engineering analysis of potential scenarios which could dilute the boron concentration in the SFP demonstrates that sufficient time is available to detect and mitigate a boron dilution from 2300 ppm prior to reaching 700 ppm.

The systems which could dilute the spent fuel pool, either by direct connection to the spent fuel pool or by a potential pipe crack/break, have been analyzed via a bleed and feed methodology.

However, in reality, the addition of unborated water to the SFP will lead to increased SFP water level and, if not controlled, an overflow of the SFP.

The ability to prevent the SFP soluble boron concentration from being diluted from 2300 ppm to a value of 700 ppm has been demonstrated by showing that each potential dilution source meets one of the following two criteria:

• Any dilution source not capable of supplying 467,873 gallons of unborated water will not be capable of diluting the SFP soluble boron concentration from 2300 ppm to 700 ppm.

• If the dilution flow rate of unborated water is < 520 gpm, then at least 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> will be needed for the SFP soluble boron concentration to be reduced from 2300 ppm to 700 ppm.

All dilution scenarios evaluated here will eventually cause SFP high water level alarms either detected directly by control room alarm, by the PEO detecting high SFP water levels or SFP overflow. The largest, credible dilution flow rate was 450 gpm (Unintentionally Overfilling from the Primary Grade Water System). The time to dilute the SFP from 2300 ppm to 700 ppm with a 450 gpm dilution source is 17.33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br />. Since the conservatively longest time between PEO rounds is 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, operators need to be able to mitigate the dilution event within 2.33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br />.

Discussions with operations estimate that securing the dilution source the dilution path would take no more than 15 minutes once discovered.

The postulated dilution break sources in the vicinity of the SFP have been conservatively evaluated. None of the sources result in a final SFP concentration less than 700 ppm of boron from an initial concentration of 2300 ppm. The major conservatisms of this evaluation are as follows:

• The proposed TS boron concentration limit is 2600 ppm rather than the 2300 ppm assumed initial condition.

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 20 of 21

• The SFP is assumed to contain the least amount of borated water possible when the dilution event initiates o Water level at the low level alarm set point o 1860 assemblies with 1860 RCCAs are stored in the SFP o The Spent Fuel Shipping Cask Loading Area, gate areas, Spent Fuel Shipping Cask Storage Area, and transfer tube are isolated from the SFP.

• Water level alarms do not detect the dilution event. It is assumed that this occurs if either the level indicators fail or the Bleed and Feed methodology keeps the water level constant.

• Operators are not notified of the dilution event via the Fuel Building Sump Pump auto start up.

• Operator rounds are the only way to detect a dilution event. It is assumed that no other workers or operators can discover a 467,873 gallon dilution event.

• The time interval between operator rounds is 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> (the maximum interval procedurally allowable).

Serial No.18-039 Docket No. 50-423 Attachment 7, Page 21 of 21 B. References External

References:

1) MP3 UFSAR, Rev. 30.1, "Millstone Power Station Unit 3 Final Safety Analysis Report,"

09/28/17.

2) MP3 T.S., Change No. 293, "Facility Operating License/Safety Technical Specifications,"

03/09/17.

3) Air Contaminants and Industrial Hygiene Ventilation -A Handbook of Practical Calculations, Problems, and Solutions, Roger L. Wabeke, 1998 by CRC Press LLC.

4) Letter from J. Alan Price (DNC) to USN RC, "Millstone Power Station, Unit No. 2, Technical Specifications Change Request (TSCR) 2-10-01, Fuel Pool Requirements," November 6, 2001. (ADAMS Accession No. ML013510295).

5) Letter from Richard B. Ennis (NRG) to J. A. Price (DNC), "MILLSTONE POWER STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: SPENT FUEL POOL REQUIREMENTS (TAC NO. MB3386)," April 1, 2003. (ADAMS Accession No. ML030910485).

6) Crane Technical Paper No. 410, Flow of Fluids through Valves, Fittings, and Pipe, 1988.

7) Rules of Thumb for Chemical Engineers - fifth edition 2012, Stephen Hall.