CP-201300257, License Amendment Request (LAR) 13-01, Revision to Technical Specifications 3.7.16, Fuel Storage Pool Boron Concentration, 3.7.17, Spent Fuel Assembly Storage, 4.3, Fuel Storage, and 5.5 Programs & Manuals

License Amendment Request (LAR) 13-01, Revision to Technical Specifications 3.7.16, Fuel Storage Pool Boron Concentration, 3.7.17, Spent Fuel Assembly Storage, 4.3, Fuel Storage, and 5.5 Programs & Manuals

ML13095A023
Person / Time
Site:	Comanche Peak
Issue date:	03/28/2013
From:	Madden F Luminant Power, Luminant Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CP-201300257, TXX-13045, CAW-13-3663
Download: ML13095A023 (110)

Text

Rafael Flores Luminant Power Senior Vice President P.O. Box 1002

& Chief Nuclear Officer 6322 North FM 56 rafael.flores@luminant.com Glen Rose, TX 76043 Luminant T 2548975590 c 817 5590403 F 2548976652 CP-201300257 Ref:

10CFR50.90 TXX-13045 10CFR2.390 March 28, 2013 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

COMANCHE PEAK NUCLEAR POWER PLANT (CPNPP)

DOCKET NOS. 50-445 AND 50-446, LICENSE AMENDMENT REQUEST (LAR) 13-01, REVISION TO TECHNICAL SPECIFICATIONS 3.7.16, "FUEL STORAGE POOL BORON CONCENTRATION,"

3.7.17, "SPENT FUEL ASSEMBLY STORAGE," 4.3, "FUEL STORAGE,"

AND 5.5 "PROGRAMS AND MANUALS"

REFERENCES:

1. Letter logged TXX-13001, dated January 15, 2013 regarding License Amendment Request 12-06 for Spent Fuel Storage from Rafael Flores (Luminant Power) to the NRC (ML13032A240)

2. Letter logged TXX-12148, dated October 9, 2012 regarding Spent Fuel Pool Criticality Analysis from Rafael Flores (Luminant Power) to the NRC (ML12292A193)

Dear Sir or Madam:

Pursuant to 10CFR50.90, Luminant Generation Company LLC (Luminant Power) hereby requests an amendment to the Comanche Peak Nuclear Power Plant (CPNPP) Unit 1 Operating License (NPF-87) and CPNPP Unit 2 Operating License (NPF-89) by incorporating the attached change into the CPNPP Unit 1 and 2 Technical Specifications. This change request applies to both Units.

Per Reference 1, Luminant Power provided a License Amendment Request (LAR) to the NRC with regard to Technical Specifications for Spent Fuel Pool Storage. The LAR in Reference 1 only provides for a short term solution for spent fuel storage at CPNPP, allowing time to complete a modern criticality analysis and submit that analysis to the NRC. This letter provides the LAR based on a modem criticality analysis methodology and proposes changes to TS 3.7.16 entitled "Fuel Storage Pool Boron Concentration," TS 3.7.17 entitled "Spent Fuel Assembly Storage", TS 4.3 entitled "Fuel Storage," and TS 5.5 entitled "Programs and Manuals." TS 3.7.16 describes the proposed minimum concentration of dissolved boron in the fuel storage pools. TS 3.7.17 describes proposed storage configurations allowed in Region II high density storage racks based on minimum bumup limitations generated from a spent fuel pool (SFP) criticality analysis. The proposed TS 4.3 describes the fuel storage design requirements in the Fuel Building. TS 5.5 provides a proposed Neutron Absorber Monitoring Program. provides a detailed description of the proposed changes, a technical analysis of the proposed changes, Luminant Power's determination that the proposed changes do not involve a significant hazard consideration, a regulatory analysis of the proposed changes and an environmental evaluation. Attachment 2 provides the affected Technical Specification (TS) pages marked-up to reflect A member of the STARS Alliance

/L Callaway Comanche Peak. Diablo Canyon Palo Verde San Onofre - South Texas Project. Wolf Creek

U. S. Nucdear ReguLatory Commission TXX-13045 Page 2 of 3 03/28/2013 the proposed changes. Attachment 3 provides proposed changes to the Technical Specification Bases for information only. These changes will be processed per CPNPP site procedures. Attachment 4 provides retyped Technical Specification pages which incorporate the requested changes. Attachment 5 provides retyped Technical Specification Bases pages which incorporate the proposed changes. Attachment 6 provides a markup of the FSAR (for information only). provides a description of the administrative controls proposed in conjunction with this license amendment request to support spent fuel storage management and ensure that the conditions evaluated in the criticality safety analysis discussed below remain bounding.

The following two enclosures, were provided by Westinghouse to support the changes:

One (1) copy of WCAP-17728-P "Comanche Peak Nuclear Power Plant Units I & 2 Spent Fuel Pool Criticality Safety Analysis" (Proprietary) (Enclosure 2)

One (1) copy of WCAP-17728-NP "Comanche Peak Nuclear Power Plant Units 1 & 2 Spent Fuel Pool Criticality Safety Analysis" (Non-Proprietary) (Enclosure 3)

Also enclosed is the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-13-3663, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice (Enclosure 4).

As Enclosure 2 contains information proprietary to Westinghouse Electric Company LLC, 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 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 CFR 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-13-3663 and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

Luminant Power requests approval of the proposed License Amendment by January 24, 2014, to be implemented within 90 days of the issuance of the license amendment. The requested schedule for this amendment would support MODE 6 operations in the Unit 2 Spring refueling outage.

In accordance with 10CFR50.91(b), Luminant Power is providing the State of Texas with a copy of this proposed amendment.

This communication contains no new licensing basis commitments and completed the following commitment identified in Reference 2.

Commitment No.

Commitment Description Status 4486411 Luminant will request a License Amendment to revise TS 3.7.17, Complete "Spent Fuel Assembly Storage." The schedule for submittal of this license amendment is March 31, 2013.

U. S. Nuclear Regulatory Commission TXX-13045 Page 3 of 3 03/28/2013 Should you have any questions, please contact Mr. J. D. Seawright at (254) 897-0140.

I state under penalty of perjury that the foregoing is true and correct.

Executed on March 28, 2013.

Sincerely, Luminant Generation Company, LLC Rafael Flores By:

A!z 2777 Frerd W. Madden Director, Oversight and Regulatory Affairs Attachments 1.

2.

3.

4.

5.

6.

Description and Assessment Proposed Technical Specifications Changes (Mark-up)

Proposed Technical Specifications Bases Changes (Mark-up For Information Only)

Retyped Technical Specification Pages Retyped Technical Specification Bases Pages (For Information Only)

Proposed Final Safety Analysis Report Changes (Mark-up For Information Only)

Enclosures

1. Comanche Peak Nuclear Power Plant Spent Fuel Pool (SFP) Configuration Controls

2.

WCAP-17728-P "Comanche Peak Nuclear Power Plant Units I & 2 Spent Fuel Pool Criticality Safety Analysis" (Proprietary)

3.

WCAP-17728-NP "Comanche Peak Nuclear Power Plant Units I & 2 Spent Fuel Pool Criticality Safety Analysis" (Non-proprietary)

4.

Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-13-3663, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice c -

• E. E. Collins, Region IV

• A. G. Howe, Region IV B. K. Singal, NRR

• Resident Inspectors, CPNPP Alice Hamilton Rogers, P.E.

Inspection Unit Manager Texas Department of State Health Services Mail Code 1986 P. 0. Box 149347 Austin TX 78714-9347

• w/o Enclosure 2

• Westinghouse U.S. Nuclear Regulatory Commission Document Control Desk 11555 Rockville Pike Rockville, MD 20852 Westinghouse Electric Company Nuclear Services 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA Direct tel: (412) 374-4643 Direct fax: (724) 720-0754 e-mail: greshaja@westinghouse.com Proj letter: NF-TB-13-38 CAW-13-3663 March 18,2013 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

WCAP-1 7728-P, Revision 0, "Comanche Peak Nuclear Power Plant Units I and 2 Spent Fuel Pool Criticality Safety Analysis" (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-13-3663 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. 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 Luminant Generation Company LLC.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference CAW-13-3663 and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

Very truly yours, L/James A. Gresham, Manager Regulatory Compliance Enclosures

CAW-13-3663 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

Before me, the undersigned authority, personally appeared James A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

James A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this 18th day of March 2013 Notary Public r

COMMONWEALTH OF PENNSYLVANIA I Arne M. Stegman, Notary Public Ndotrlal Seal Unity Twp., Westmoreland County IMy Conmission Expires Aug. 7, 2016 MENGE&

PENNSYLVAIA ASSOCIAION Of NOTARIES

2 CAW-13-3663 (1)

I am Manager, Regulatory Compliance, in Nuclear Services, 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) 1 am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3)

I have personal knowledge 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 constitutes Westinghouse policy and provides 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

3 CAW-13-3663 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.

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.

4 CAW-13-3663 (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.

(iii)

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

(iv)

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.

(v)

The proprietary information sought to be withheld in this submittal is that which is appropriately marked in WCAP-17728-P, Revision 0, "Comanche Peak Nuclear Power Plant Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis" (Proprietary), dated March 2013, for submittal to the Commission, being transmitted by Luminant Generation Company LLC. letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with Westinghouse's request for NRC approval of WCAP-17728, and may be used only for that purpose.

5 CAW-13-3663 This information is part of that which will enable Westinghouse to:

(a)

Obtain NRC approval of WCAP-17728, "Comanche Peak Nuclear Power Plant Units I and 2 Spent Fuel Pool Criticality Safety Analysis".

(b)

Demonstrate the sub-criticality of the Comanche Peak spent fuel pools.

Further this information has substantial commercial value as follows:

(a)

Westinghouse plans to sell the use of similar information to its customers for the purpose of demonstrating the sub-criticality of the spent fuel pool.

(b)

Westinghouse can sell support and defense of spent fuel pool criticality analysis.

(c)

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/or non-proprietary versions of documents 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 CFR 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 CFR 2.390(b)(1).

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 CFR 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.

ATTACHMENT 1 to TXX-13045 DESCRIPTION AND ASSESSMENT to TXX-13045 Page 1 of 17 LICENSEE'S EVALUATION

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

7.0 REFERENCES

to TXX-13045 Page 2 of 17

1.0 DESCRIPTION

Luminant Power requests to amend Comanche Peak Nuclear Power Plant (CPNPP) Unit 1 Operating License (NPF-87) and CPNPP Unit 2 Operating License (NPF-89) by incorporating the attached change for fuel storage requirements into the CPNPP Unit 1 and 2 Technical Specifications. The license amendment request (LAR 13-01) proposes changes to the spent fuel storage and affects Technical Specifications (TS) 3.7.16, "Fuel Storage Pool Boron Concentration,"

TS 3.7.17, "Spent Fuel Assembly Storage," TS 4.3 "Fuel Storage," and TS 5.5 "Programs and Manuals" for CPNPP Units 1 and 2. The analysis to support the proposed changes takes credit for empty storage cells and soluble boron in the Region ]I spent fuel storage racks. Credit is taken for soluble boron and neutron absorbing panels in Region I. In accordance with 10 CFR 50.68, Criticality accident requirements, the effective neutron multiplication factor (keff) limits for Region I and Region H storage racks are based on analyses to maintain kef less than 1.0 when flooded with unborated water, and less than or equal to 0.95 when crediting soluble boron during normal operations. The proposed change is evaluated for both normal operation and accident conditions.

2.0 PROPOSED CHANGE

The proposed change would revise three Technical Specifications and add a new Technical Specifications Program. Changes to the Technical Specifications are described below and evaluated in Section 4.0 of this attachment.

LAR 12-06 (Reference 7.3) submitted in January 2013 also proposed changes to TS 3.7.16, TS 3.7.17, and included a "Spent Fuel Assembly Dispersion Program" as a TS program as Section 5.5.22. The proposed changes described below (and the mark-up changes of Attachments 2 and

3) represent changes from the current CPNPP Technical Specifications, rather than assuming the changes proposed in LAR 12-06 are implemented.

TS 3.7.16, "Fuel Storage Pool Boron Concentration" Revise LCO 3.7.16 from 2000 ppm boron concentration to 2400 ppm boron concentration.

If the changes proposed in LAR 12-06 are implemented, TS 3.7.16 will not change, since LAR 12-06 also proposed the same increase in required boron concentration to 2400 ppm.

TS 3.7.17, "Spent Fuel Assembly Storage" Replace the entire TS 3.7.17 to update the LCO and Surveillance requirements and include allowed configurations and restrictions for storage of fuel in Region II high density storage racks based on minimum burnup limits generated from a modem spent fuel pool (SFP) criticality analysis.

If the changes proposed in LAR 12-06 are implemented, TS 3.7.17 contents will still be replaced in their entirety by the changes included in this LAR.

TS 4.3. "Fuel Storage" Updated to reflect the changes to the Criticality Safety Analysis and TS 3.7.17.

to TXX-1 3045 Page 3 of 17 TS 5.5, "Programs and Manuals" Add new TS program, 5.5.22, to include a Neutron Absorber monitoring program for the Region I spent fuel storage racks.

If the changes proposed in LAR 12-06 are implemented, the contents of 5.5.22 "Spent Fuel Assembly Dispersion Program" will be removed and replaced with the Neutron Absorber Monitoring Program.

Mark-ups of the proposed Technical Specification changes are provided in Attachment 2. A copy of the proposed mark-up of the Technical Specification Bases is provided (for information only) in Attachment

3. Revised (clean) Technical Specification and Technical Specification Bases pages are provided in and 5, respectively. Attachment 6 provides a markup of the CPNPP Final Safety Analysis Report (for information only).

3.0 BACKGROUND

CPNPP has two pools, Spent Fuel Pool 1 (SFP1) and Spent Fuel Pool 2 (SFP2), containing spent fuel racks for storage of spent nuclear fuel. The spent fuel racks are designed to accommodate a safe shutdown earthquake, handling loads, and the dead load of the spent fuel assemblies.

The spent fuel assemblies in SFP1 and SFP2 are stored in both Region I and Region II racks. The total number of usable storage locations for SFP1 is 1,684 cells, and is 1,689 for SFP2. This provides a total of 3,373 storage locations. The actual storage capacity is limited by Technical Specification 3.7.17, and is dependent upon the fuel characteristics of the CPNPP fuel inventory.

The Region I and Region II racks are composed of vertical cells fastened together in a checkerboard arrangement to produce a matrix structure. The cells are welded to a baseplate and to one another to form an integral structure without the use of a supporting grid structure. The center to center spacing between cells within a Region I rack is a nominal 10.65 inches by a nominal 11.05 inches. The Region I racks use a flux trap design and have neutron absorbing BORAL (BORAL) panels between adjacent storage cells to provide neutron attenuation. The center to center spacing between cells within a Region II rack is a nominal 9.0 inches. The Region II racks do not use a flux trap design and have no special neutron absorbing material. The Region II racks were originally designed to use Boraflex panels, but the Boraflex material was removed from the CPNPP storage racks prior to installation.

SFP1 and SFP2 each contain two (2) 10 x 8 Region I rack modules, one (1) 9 x 8 Region I rack module, six (6) 12 x 14 Region II rack modules, and three (3) 11 x 14 Region II rack modules (twelve racks total). Some of the storage cells have been modified to allow for fuel inspection.

Operation of the spent fuel pool includes periodic chemical analyses and operational surveillance for determining concentrations of chloride, fluoride and boron. The current chemical limits used in monitoring the spent fuel pools are, as follows:

Chlorides 0.15 ppm (maximum)

Fluorides 0.15 ppm (maximum)

Boron Concentration 2400 ppm (minimum) (Note the current TS 3.7.16 minimum is 2000 ppm)

Additional descriptions may be found in Section 9.1 of the FSAR.

to TXX-13045 Page 4 of 17 This request for a License Amendment (LAR) is to revise Technical Specification (TS) 3.7.16, "Fuel Storage Pool Boron Concentration," TS 3.7.17, "Spent Fuel Assembly Storage," TS 4.3 "Fuel Storage," and TS 5.5, "Programs and Manuals," considering fuel discharged from reactor operation at 3612 MWt. Luminant Power has prepared this LAR addressing the most recent guidance from NEI and recommendations of the most recent NRC guidance.

4.0 TECHNICAL ANALYSIS

The Criticality Safety Analysis which supports this License Amendment Request (LAR) is included as Enclosure 2 (WCAP-17728-P, Proprietary Version) and Enclosure 3 (WCAP-17728-NP, Non-proprietary Version). A description of the CPNPP SFP Configuration Controls which will be utilized to ensure the conditions evaluated in the Criticality Safety Analysis remain bounding is included as Enclosure 1.

The LAR was modeled based on the guidance of NEI 12-16, Rev. 0, "Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants", March 2013 (Reference 7.1). CPNPP utilized the NEI guidance to ensure that the proper considerations were made in the analysis and controls. The supporting Criticality Safety Analysis also reflects the guidance of DSS-ISG-2010-1, "Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools".

This section includes a brief statement related to each applicable topic discussed in NEI 12-16, and summarizes the analysis or proposed controls applicable to each area. Also included is a discussion of the proposed Technical Specification changes. Supporting details are found in either Enclosure 1 or Enclosure 2, as referenced below.

Analytical Techniques to Calculate klff NEI 12-16, Section 2.1 describes the Acceptance Limits for Criticality Analysis, which reflects the requirements of 10CFR50.68.

The Criticality Safety Analysis was performed to satisfy the requirements of 10CFR50.68(b)(4),

which states in part:

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

Section 2.1 of the Criticality Safety Analysis (Enclosure 2) reflects the requirements of 10CFR50.68(b)(4) as the Acceptance Criteria for the analysis.

NEI 12-16, Section 2 also describes the basic equation for calculating the maximum ker, for comparison to the acceptance limit.

m n

knta-x = keff +

Bias1 +

Uncertainty2 i=O

- j=O to TXX-13045 Page 5 of 17 The Criticality Safety Analysis demonstrates that the kuff, including all applicable biases and uncertainties which account for the statistical 95/95 confidence levels, satisfy the Acceptance Criteria. Uncertainties are statistically combined (the total Uncertainty is calculated as the square root of the sum of the squares of each individual Uncertainty), while the Biases are summed, as described in Section 5.3.2 of the Criticality Safety Analysis (Enclosure 2). All applicable biases and uncertainties are utilized in both the borated and the non-borated analysis.

Computer Codes NEI 12-16, Section 3 describes the different types of computer codes that may be used in a criticality analysis. This section also discusses the validation of the computer codes used in the criticality analysis.

The analysis utilized in this submittal employs the following computer codes and cross-section libraries:

(1)

The two dimensional (2-D) transport lattice code PARAGON and its cross-section library based on Evaluated Nuclear Data File Version VI.3 (ENDF/B-VI.3), and (2)

SCALE Version 5.1 with the 44-group cross-section library based on ENDF/B-V.

The Computer Codes used in this application are discussed in Section 2.3 of the Criticality Safety Analysis (Enclosure 2). The application specific validation of SCALE 5.1 is provided in Appendix A of the Criticality Safety Analysis.

Rack and Fresh Fuel Modeling NEI 12-16, Section 4 describes generally acceptable methods of modeling fresh fuel assemblies and fuel storage racks, including considerations for Rack Neutron Absorbers.

Fuel Assemblies are divided into two Fuel Groups in the Criticality Analysis. Fuel Group F1 includes fuel with a nominal fuel rod outer diameter of 0.374 inches, a fuel design which is no longer used at CPNPP in new core designs. Fuel Group F2 includes the current fuel design, which has a nominal outer diameter of 0.360 inches. Controls which ensure future fuel designs satisfy the assumptions of the analysis are discussed in the SFP Configuration Controls (Enclosure 1).

The fuel manufacturing tolerances are included in the criticality analysis through either analysis, or use of bounding values. Details of the fuel designs used are provided in Section 3.1 of the Criticality Safety Analysis (Enclosure 2). Details of the fuel manufacturing tolerance are provided in Section 5.3.2 of the Criticality Safety Analysis.

The analysis includes updated criticality analysis for the Region I and Region II storage racks.

The Region I racks are based on a "flux trap" design which uses two BORAL panels between each assembly to reduce the neutronic interaction between adjacent assemblies. The analysis was performed using conservative assumptions for the neutron absorber B-10 areal density (Section 5.1.1.2 of the Criticality Safety Analysis (Enclosure 2)). Additionally, analysis was performed which demonstrates that the potential reactivity affects of blisters forming on the BORAL panels is acceptable, as described in Section 5.1.2.4 of the Criticality Safety Analysis. Blistering is a phenomenon which has occurred at other sites using this material.

to TXX-1 3045 Page 6 of 17 The Region II racks are a high density rack design which was originally designed to use one Boraflex panel between each assembly to reduce the neutronic interaction of adjacent assemblies.

Due to industry operating experience issues with Boraflex degradation which were discovered prior to the installation of the racks, the Boraflex material was removed from the CPNPP Region II racks. The racks in SFP2 maintained the thin stainless steel wrapper which was designed to hold the Boraflex material, while the racks in SFP1 do not contain this wrapper. The wrappers were conservatively ignored in the analysis for both pools, and this assumption is justified in Section 3.2.2 of the Criticality Safety Analysis.

Details of the storage rack parameters are provided in Section 3.2 of the Criticality Safety Analysis. Details of the storage rack tolerance are provided in Section 5.3.2 of the Criticality Safety Analysis.

Configuration Modeling and Soluble Boron Credit NEI 12-16, Section 5 describes considerations for configuration modeling, including a description of Normal Conditions, Interface considerations, and abnormal / accident conditions which should be considered. NEI 12-16, Section 6 describes considerations for soluble boron credit under normal and accident conditions, and considerations for a boron dilution accident.

The analysis demonstrates acceptable results for kef for both Normal Conditions (Section 5.5 of Criticality Safety Analysis, Enclosure 2) and Accident Conditions (Section 5.7 of the Criticality Safety Analysis. Normal Conditions include normal storage, fuel movement, and other procedurally controlled activities in the SFPs. Region I and II storage arrays which define allowable storage are defined in Section 5.2 of the Criticality Safety Analysis, which have been incorporated into the proposed Technical Specification changes. Controls which ensure that the proposed Technical Specification limitations for storage are maintained are discussed in the Licensee Controls section.

Accident Conditions include single and multiple misload scenarios, temperature excursions, a dropped or misplaced fuel assembly, and seismic events.

For the limiting Normal Condition, 400 ppm of soluble boron is credited to ensure the maximum kef satisfies the acceptance criteria of kff< 0.95.

The limiting analyzed Accident Condition was an event which involves misloading multiple fuel assemblies due to a common cause.

In the multiple misload accident cases, the maximum keff satisfies the acceptance criteria of kef <

0.95 when crediting the proposed Technical Specification limit of 2400 ppm (Table 5-21 of the Criticality Safety Analysis). Enclosure 1 Section 2 demonstrates that the analyzed multiple misload cases are not credible based on the SFP Configuration Controls in place to prevent such an accident. This is based on the Double Contingency Principle as described in NEI 12-16, Section 1.4:

The double contingency principle, as described in ANSI/ANS 8.1, Section 4.2.2 [9], states that "process designs should incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible". In other words, the nuclear criticality analysis is required to demonstrate that criticality cannot occur without at least two unlikely, independent and concurrent incidents and abnormal occurrences. This will ensure that no single occurrence can lead to a criticality. The to TXX-13045 Page 7 of 17 double contingency principle means that a realistic condition may be assumed for the criticality analysis in calculating the effects of incidents or abnonnal occurrences.

In the multiple misload accident cases, two independent and unlikely errors would be required to achieve the analyzed scenarios. First, a significant error during the fuel movement planning and reviews would be required to generate a fuel movement plan which would result in the severe TS 3.7.17 non-compliances which were analyzed. Secondly, fuel handling personnel would have to violate fuel handling procedural requirements to either create a "4 out of 4" array in an area of the SFP which is not adjacent to a wall, or to place fresh fuel assemblies adjacent to other fuel assemblies.

Although it is demonstrated that the analyzed multiple misload accident cases are not credible, the acceptability demonstrated for these scenarios bounds other multiple misload accidents which are considered credible, such as multiple misplaced once-burned fuel assemblies in a non-compliant configuration.

Other analyzed accident conditions, such as a single misload event, temperature excursions, and seismic events, satisfied the acceptance criteria when analyzed at 1867 ppm.

The CPNPP SFP boron dilution analysis was not revised based on the results of the revised criticality analysis. The CPNPP dilution analysis was submitted to support LAR-00-05, which was approved by the NRC in 2001 (Reference 7.4). The SFP boron dilution analysis concluded that an event which would result in the dilution of the SFP boron concentration from 1,900 ppm (the Technical Specification minimum value at that time was 2000 ppm, and a value of 100 ppm below this value was used in the dilution analysis for conservatism) to 800 ppm (the maximum boron value credited in the 2001 analysis for normal conditions) was not a credible event.

From the Technical Analysis section of the LAR-00-05:

"Each SFP has a water inventory of 300,000 gallons. Assuming a well-mixed SFP, the volume required to dilute the SFP from 1900 ppm to 800 ppm is approximately 259,500 gallons of non-borated Water. The various events that were considered included dilution from the reactor makeup water system, demineralized water system, component cooling water system, fire protection system, and chemical and volume control system letdown. Other events that may affect the boron concentration of the SFP such as pipe cracks and loss of offsite power were also evaluated. All pipes in the vicinity of the SFP are seismically qualified and supported. As such, a random pipe break was not considered in the analyses. The guidance of Reference 4, Branch Technical Position, Mechanical Engineering Branch 3-1 (MEB 3-1), for selection of the potential pipe cracks, was followed.

The analysis concludes that an unplanned or inadvertent event that would dilute the SFP boron concentration from 1,900 ppm to 800 ppm would be readily detectable by plant personnel via alarms, flooding in the fiel and auxiliary buildings, or by normal operator rounds through the SFP area. The combination of the large volume of water required for a dilution event, TS-controlled SFP concentration and seven-day sampling requirement, and plant personnel rounds would adequately detect a dilution event prior to keff reaching 0.95 (800 ppm). Therefore, the analysis and TS controls are acceptable for the boron dilution aspects of this submittal."

The proposed Technical Specification Limit for minimum SFP soluble boron has been increased (from 2,000 ppm to 2,400 ppm), and the soluble boron value credited in the analysis for normal conditions has decreased (from 800 ppm to 400 ppm). No changes to the CPNPP Spent Fuel Pool or cooling systems have been made which would invalidate this analysis. A much larger volume to TXX-13045 Page 8 of 17 of water would be required to dilute the SFP from 2,400 to 400 ppm, compared to the volume to dilute from 1,900 ppm to 800 ppm, which was determined non-credible in the 2001 analysis.

Therefore, the dilution analysis approved in 2001 remains bounding to support this Criticality Safety Analysis.

Reactivity Effects of Depletion NEI 12-16, Section 7 describes appropriate considerations for calculating reactivity effects of fuel depletion. Key parameters which could affect the reactivity of discharged fuel include:

Relative power during depletion (which impacts the moderator and fuel temperatures during depletion)

Soluble boron during depletion Presence of burnable absorbers Rodded operation Axial burnup shapes The depletion analysis is described in Section 4 of the Criticality Safety Analysis (Enclosure 2),

and was performed using assumptions which bound each of these parameters (unless specifically addressed). Controls which ensure future fuel designs satisfy the assumptions of the analysis are discussed in the Licensee Controls section. Depletion Uncertainty is discussed in Section 5.3.2.1.3 of the Criticality Safety Analysis.

Other Credits NEI 12-16, Section 8 describes additional factors which reduce reactivity in fuel assemblies, and therefore may be credited in a Criticality Safety Analysis with proper justification.

Section 4.2.1 of the Criticality Safety Analysis (Enclosure 2) describes that credit is taken for the isotopic decay of 241Pu to 241Am. This credit is reflected in the decay time limitations of TS 3.7.17.

The maximum decay time calculated in the analysis is 20 years, and interpolation beyond this value is not permitted. Interpolation of decay time values is justified in section 6.1.3 of the Criticality Safety Analysis.

Other credits discussed in the NEI guidance document, including credit for fresh burnable absorbers, control rods, or other types of movable Absorber Inserts, were not utilized in the Criticality Safety Analysis.

Licensee Controls NEI 12-16, Section 9 describes controls intended to ensure that the conditions evaluated in the nuclear criticality safety analysis are and remain bounding to the current plant operating parameters. It discusses considerations for controls such as administrative controls for fuel move planning and execution, changes to fuel designs, fuel characterization, and neutron absorber monitoring programs.

CPNPP will establish controls which ensure the SFP configuration and other applicable conditions evaluated in the Criticality Safety Analysis remain bounding when compared to current fuel design and plant operating parameters. Enclosure 1 provides a description of the controls which will be implemented, to ensure:

to TXX-13045 Page 9 of 17

1.

TS 3.7.17 compliance is maintained at all times. Enclosure 1 describes the controls which will be established to ensure that all fuel movement plans into Spent Fuel Pool Region H are prepared in a manner which ensures continual compliance with the limitations of proposed TS 3.7.17, including all intermediate steps during fuel movement.

2.

A misloading event beyond the analyzed accident conditions is not credible. Enclosure 1 describes the controls which will be established to ensure that an error in the fuel move planning does not have the potential to result in a misloading accident which is not bound by the Safety Analysis.

3.

Assumptions related to Fuel Characterization and reactor operation remain valid. describes the controls which will be established to ensure that conditions evaluated in the Criticality Safety Analysis will remain bounding, for both future fuel design changes (pre-irradiation fuel characterization) and future operating conditions (post-irradiation fuel characterization).

4.

Neutron Absorber assumptions remain valid and bounding. Enclosure I describes the controls which will be established to ensure that Safety Analysis assumptions for Neutron Absorber panels utilized in Region I storage cells remain conservative when considering aging affects or abnormal degradation. This program is captured in the proposed Technical Specification section 5.5.22.

Proposed Technical Specification Changes TS 3.7.16 - Fuel Storage Pool Boron Concentration The proposed change to TS 3.7.16 increases the minimum SFP Boron Concentration from 2000 ppm to 2400 ppm. Note that this change was previously submitted to the NRC for review in January of 2013 per CPNPP LAR 12-06. This increase is appropriate and justified since the revised criticality analysis was performed considering Multiple Misload events as a potential accident scenario, which was not considered in the previous analysis.

TS 3.7.17 - Spent Fuel Assembly Storage The proposed change to TS 3.7.17 completely revises the section based on the new criticality analysis. A change to TS 3.7.17 was previously submitted to the NRC for review in January of 2013 per CPNPP LAR 12-06. Regardless of the implementation of LAR-12-06, the entire contents of TS 3.7.17 will be replaced with the changes included in this LAR.

The proposed changes to TS 3.7.17 include:

The proposed TS 3.7.17 surveillance requirement now requires verification of the fuel move plans and the final configuration. This is a significant improvement compared to the current surveillance requirement, which only requires verification of fuel parameters versus the burnup limits, and is not a verification of the fuel movement plan. This ensures that all steps in the fuel movement plan will be reviewed for acceptability. A sequence which creates and resolves a TS 3.7.17 violation within the same fuel movement plan will therefore not satisfy the Surveillance Requirement (SR).

The proposed SR Frequency was changed to "Prior to moving a fuel assembly into any Region II storage location", rather than "storing" fuel in Region II.

to "XX-13045 Page 10 of 17 The current TS Figures 3.7.17-1 through 3.7.17-4, which contained "burnup vs.

enrichment" plots for each allowable configuration, and a graphical description of the four configurations, have been removed. These figures will be replaced by a proposed new Figure 3.7.17-1, and three data tables related to Fitting Coefficients.

The proposed Figure 3.7.17-1 contains graphical and verbal descriptions of the five allowed Region II storage arrays, and includes notes required for proper interpretation of the limits.

The proposed Figure 3.7.17-1 explains the required terms for proper interpretation of the Region II limits, including "Array", "Fuel Category", and "Fuel Group".

The notes of proposed Figure 3.7.17-1 include references to the appropriate Table to determine Fitting Coefficients, and the equation used to calculate the Minimum Burnup for each Fuel Category.

The limitations of the proposed TS 3.7.17 are more complex than the previous limits.

o The allowable configurations, or Array Definitions, have increased from 4 to 5.

o Unlike previous limits, one of the Array Definitions is based on a non-uniform pattern (contains fuel assemblies with 2 separate limits in the array). One of the Arrays is restricted to the outer two rows adjacent to the SFP wall.

o The CPNPP fuel inventory is now divided into two Fuel Groups and utilizes Decay Time, which requires tracking of additional parameters.

o Improved Administrative Controls, as discussed in Enclosure 1, will be implemented to ensure that the increased complexity will not result in an increased risk of a TS 3.7.17 violation due to an error made during fuel movement planning.

The proposed Figure 3.7.17-1 Note 2 implements a limitation which requires that all Fresh Fuel Assemblies be categorized as Fuel Category 1 (the category with the highest reactivity), even for low values of Initial Enrichment. This limitation is being implemented based on the CPNPP SFP Configuration Controls, and is conservative relative to the requirements of the Criticality Safety Analysis.

o For example, assume a 3.5% enriched fresh fuel assembly is being placed into Region II. Per the Criticality Safety Analysis, this assembly could be stored in a checkerboard pattern (Array II-D). However, since all fresh fuel must be considered Fuel Category 1 fuel assemblies per the proposed limits of TS 3.7.17, this assembly will be limited to a pattern surrounded by empty water locations in Region II (Array H-E).

Figure 3.7.17-1 Note 2 is included in the proposed changes for consistency with the CPNPP SFP Configuration Controls (see Enclosure 1, Section 2) to ensure that Multiple Misload events beyond analyzed cases are not credible. This section states "Limitations will be implemented into Fuel Handling field procedures to ensure that any Fresh Fuel Assemblies (which are visually distinct from discharged fuel assemblies) which are placed into Region II are ONLY placed into a cell which contains empty water cells in all adjacent surrounding Region Il cells, including diagonally adjacent locations."

TS 4.3 - Fuel Storage The proposed changes to Technical Specification Section 4.3 reflect the analysis and proposed changes to TS 3.7.17. The changes to the nominal Region I center-to-center spacing in 4.3.1.1.e are administrative, and were made in order to display the same number of significant figures as was used in the analysis (the actual values did not change).

to TXX-13045 Page 11 of 17 Specifications 4.3.1.1.g through 4.3.1.1.i were consolidated into a single Specification with reference to the proposed TS 3.7.17.

Note that the Criticality Analysis will be incorporated by reference into FSAR Section 4.3; therefore the references to 4.3.1.1.b and 4.3.1.1.c for uncertainties are appropriate.

Proposed updates to the Storage Capacity discussion of 4.3.3 reflects that CPNPP cannot utilize all 3373 storage locations previously licensed, since the only Array which allows a solid storage pattern is limited to the outer two rows of the SFP.

TS 5.5 - Programs and Manuals As discussed in Enclosure 1 Section 4, CPNPP is proposing to implement a Neutron Absorber monitoring program to ensure the criticality analysis assumptions related to BORAL remain bounding, considering potential aging affects and degradation. This program is being captured in the Technical Specifications as a new proposed Program, 5.5.22, "Spent Fuel Storage Rack Neutron Absorber Monitoring Program". The acceptance criteria for the monitoring program are established conservatively relative to the assumptions utilized in the criticality analysis.

Note that LAR-12-06, which was submitted for NRC review in January 2013, included a proposed "Spent Fuel Assembly Dispersion Program" as TS 5.5.22. This program was implemented as a short term solution to increase the reactivity margin in the SFP. The analysis included in this LAR eliminates the need for a Spent Fuel Assembly Dispersion Program. Therefore, the Neutron Absorber monitoring program will be implemented as 5.5.22, either as a new section (if changes submitted in LAR-12-06 are not implemented), or as a replacement for 5.5.22, Spent Fuel Assembly Dispersion Program (if changes submitted in LAR-12-06 are implemented).

Conclusion In conclusion, the proposed changes to TS 3.7.16, TS 3.7.17, TS 4.3, and TS 5.5 allow safe storage of new and used fuel at CPNPP.

to TXX-13045 Page 12 of 17

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Lummnant Power has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10CFR50.92, "Issuance of amendment," as discussed below:

1.

Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response

No All proposed Technical Specification changes are related to changes for storage limitations in the spent fuel pool storage racks. The Region I analysis was updated to use modern methods, but other than inclusion of a neutron absorber monitoring program, the limitations for storage were unchanged.

Applicable accidents previously evaluated include the boron dilution accident, and fuel misload accident.

The probability for a boron dilution accident is not affected by the proposed Technical Specification change to storage limits. The consequences of a dilution accident are not increased, since the proposed change to the minimum spent fuel pool boron concentration increases the limit from 2000 ppm to 2400 ppm, and the boron required to ensure k-effective less than or equal to 0.95 has been decreased from 800 ppm to 400 ppm. Therefore, a much larger volume of water would be required to approach the acceptance criteria than assumed in the previous analysis.

The limitations associated with the proposed Technical Specification 3.7.17 utilize the same basic concepts as the current limits, in that fuel parameters are utilized to determine the minimum burnup requirements for multiple allowed storage configurations. The acceptability of a fuel configuration is then verified administratively prior to moving fuel in the Region II racks. However, the proposed limits are more complex than the current limits in several aspects.

The fuel inventory is divided into two Fuel Groups (under the current Technical Specifications, the same limits apply to all fuel designs).

Region II has 5 allowable storage arrays (up from 4).

One of the allowable arrays is limited to rows adjacent to the spent fuel pool wall.

One of the allowable arrays contains assemblies of two fuel categorizations in a specific pattern, and is limited to a single Fuel Group.

Calculation of the minimum burnup for each fuel category will require Decay Time, which is not an input parameter in the current limits.

With the proper administrative controls, the proposed increase in complexity would not result in an increase in the probability of a fuel misload accident due to a fuel move planning error. For example, determining the acceptability of a to TXX-13045 Page 13 of 17 loading pattern based on the current Technical Specification limits can be performed visually by reviewing a color coded spent fuel pool map. Violations of the storage pattern are easily recognizable and easily identified. Although color coded spent fuel pool maps will still be useful, the increased complexity of the proposed Technical Specification 3.7.17 limits results in increased difficulty for identifying non-compliant configurations by simple visual methods.

Improvements in the administrative controls are necessary to ensure that this type of simple visual verification is not relied upon, and to ensure that the increased complexity will not result in an increased risk of a Technical Specification 3.7.17 non-compliance.

The proposed Technical Specification 3.7.17 Surveillance Requirement now requires verification of the fuel move plans and the final configuration. This is a significant improvement compared to the current surveillance requirement, which only requires verification of fuel parameters versus the burnup limits, and is not a verification of the fuel movement plan. Fitting coefficients, rather than burnup-vs-enrichment curves, allow increased accuracy and reduce the chance of software errors.

Fuel handling procedure changes implement a second layer of defense to ensure that a multiple misload event, beyond the analyzed accident condition, is not credible. The proposed Technical Specification 3.7.17 includes a requirement to consider all fresh fuel assemblies Category 1 (the highest reactivity fuel category) regardless of initial enrichment. This limit, which is conservative relative to the criticality safety analysis, can be easily verified during fuel handling, since fresh fuel is visually distinct from irradiated fuel, due to lack of oxidation discoloration. All Category 1 fuel assemblies stored in Region II must be surrounded by empty storage cells, which can also be visually verified during fuel movement. Additionally, the proposed Technical Specification 3.7.17 limits only have a small area of the spent fuel pool where a solid fuel configuration is allowed, in the two rows adjacent to the spent fuel pool walls. Compliance with this limitation can also be verified during fuel handling. Since fuel handling procedures will direct the fuel handler to stop fuel movement if the above situations are encountered, regardless of the approved fuel move plans, these additional verifications reduce the probability of many fuel misload events.

The probability of other evaluated accidents, such as a seismic event, a dropped fuel assembly, or a temperature excursion, is not affected by the revised limitations. Analysis has been completed which demonstrates the consequences of these accidents are not significantly increased.

Therefore the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Attachment I to TXX-13045 Page 14 of 17

2.

Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response

No The proposed changes which ensure the maintenance of the fuel storage pool boron concentration and storage configuration do not represent new concepts.

The actual boron concentration in the fuel storage pool has been previously maintained at or above 2,400 ppm for spent fuel pool 1 and spent fuel pool 2. The criticality analysis determined that a boron concentration of 400 ppm (non-accident) and 2,400 ppm (accident) results in a k-effective less than 0.95.

The possibility of a fuel storage pool dilution is not affected by the proposed change to the Technical specifications. Therefore, increasing the Technical Specification controls for the soluble boron will not create the possibility of a new or different kind of accidental pool dilution.

The potential for criticality in the spent fuel pool is not a new or different type of accident. All storage configurations allowed by Technical Specification 3.7.17 have been analyzed to demonstrate that the pool remains subcritical.

The criticality safety analysis includes analysis of a multiple misload accident scenario; only single misload events were previously analyzed. This analysis was performed in light of recent industry operating experience which demonstrates that misload events beyond a single misload event are credible. The inclusion of this analysis does not imply the creation of the possibility of a new accident, but simply expands the boundaries of the analyzed accident conditions to ensure that all potential accidents are properly considered.

There is no significant change in plant configuration, equipment design, or usage of plant equipment. The safety analysis for boron dilution remains bounding.

The criticality analyses assure that the pool will remain subcritical with no credit for soluble boron.

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

to TXX-13045 Page 15 of 17

3.

Do the proposed changes involve a significant reduction in a margin of safety?

Response

No Proposed Technical Specifications 3.7.16, 3.7.17, and 4.3 and the associated fuel storage requirements will provide adequate margin to assure that the fuel storage array in Region H will remain subcritical by the margins required in 10CR50.68.

The criticality analysis for both Region I and Region II utilized credit for soluble boron, and the storage configurations have been defined using k-effective calculations to ensure that the spent fuel rack k-effective will be less than 1.0 with no soluble boron. Soluble boron credit is used to offset off-normal conditions (such as a misplaced assembly) and to provide subcritical margin such that the fuel storage pool k-effective is maintained less than or equal to 0.95. The loss of substantial amount of soluble boron from the spent fuel pools which could lead to exceeding a k-effective of 0.95 has been evaluated and shown not to be credible. These evaluations show that the dilution of the spent fuel pools boron concentration from 1,900 ppm to 800 ppm is not credible and that the fuel stored in Region II racks will remain less than 1.0 k-effective when flooded with unborated water.

The thermal-hydraulic conditions of spent fuel pool cooling, when considering the stretch power uprate, were considered in License Amendment 146, and found to be acceptable by the NRC. The spent fuel pool cooling system continues to maintain the temperature of the bulk spent fuel pool water within the limits of the existing licensing basis. Thus, the existing licensing basis remains valid, and there is no significant reduction in the margin of safety for the thermal-hydraulic design or spent fuel cooling.

The main safety function of the spent fuel pool racks is to maintain the spent fuel assemblies in a safe configuration through normal and abnormal operating conditions. The structural considerations of the spent fuel pool storage racks continue to maintain margin of safety against tilting and deflection or movement, such that the racks do not impact each other or the pool walls, damage spent fuel assemblies, or cause criticality concerns. Thus, the margin of safety with respect to mechanical, material or structural considerations is not changed by this proposed License Amendment Request.

The addition of a Spent Fuel Pool Rack Neutron Absorber Monitoring program (proposed Technical Specification section 5.5.22) provides a method to identify potential degradation in the neutron absorber material prior to challenging the assumptions of the Criticality Safety Analysis related to the material.

Additionally, the revised analysis utilized more conservative assumptions relative to the current Analysis of Record. Therefore, the addition of this monitoring program does not reduce the margin of safety, but ensures the margin of safety is maintained for the planned life of the spent fuel storage racks.

Therefore the proposed change does not involve a reduction in a margin of safety.

to rXX-1 3045 Page 16 of 17 Based on the above evaluations, Luminant Power concludes that the proposed amendment(s) present no significant hazards under the standards set forth in 10CFR50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria Criterion 2 of 10 CFR 50.36(c)(2)(ii), "A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:.

A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier."

GDC 61 - Fuel Storage and Handling and Radioactivity Control, "The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions.

These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventory under accident conditions."

GDC 62 - Prevention of Criticality in Fuel Storage and Handling, "Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations."

NUREG-0800, Standard Review Plan 9.1.2, "Spent Fuel Storage": "Nuclear reactor plants include storage facilities for the wet storage of spent fuel assemblies. The safety function of the spent fuel pool and storage racks is to maintain the spent fuel assemblies in a safe and subcritical array during all credible storage conditions and to provide a safe means of loading the assemblies into shipping casks."

10 CFR 50.68 (b)(4):"If no credit for soluble boron is taken, 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 unborated water. 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 proposed changes consider the types of fuel used for past, current and planned future operating conditions for CPNPP Units 1 and 2. Storage of this fuel in the spent fuel storage pools does not change the compliance with the above general design criteria and are also consistent with the above Standard Review Plan.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the to T)(X-1 3045 Page 17 of 17 issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

Luminant Power has determined that the proposed amendment would change requirements with respect to the installation or use of a facility component located within the restricted area, as defined in 10CFR20, or would change an inspection or surveillance requirement. Luminant Power has evaluated the proposed changes and has determined that the changes do not involve (1) a significant hazards consideration, (2) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (3) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10CFR51.22(c)(9).

Therefore, pursuant to 10CFR51.22(b), an environmental assessment of the proposed change is not required.

7.0 REFERENCES

7.1 NEI 12-16, Rev. 0, "Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants", March 2013 7.2 Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 146 to Facility Operating License Nos. NPF-87 and NPF-89 Texas Utilities Electric Company Comanche Peak Steam Electric Station, Units 1 and 2, Docket Nos. 50-445 and 50-446 (Accession #: ML081510173) dated June 27, 2008.

7.3 Comanche Peak Nuclear Power Plant (CPNPP) Docket Nos. 50-445 AND 50-446, License Amendment Request (LAR) 12-06, Revision to Technical Specifications 3.7.16, "FUEL STORAGE POOL BORON CONCENTRATION," 3.7.17, "SPENT FUEL ASSEMBLY STORAGE" AND 5.5 "PROGRAMS AND MANUALS", (Accession #: ML13032A240) dated January 15, 2013.

7.4 Letter from C. L. Terry, TXU Electric, to U.S. NRC,

Subject:

"Comanche Peak Steam Electric Station (CPSES) - Docket Nos. 50-445 and 50-446 - License Amendment Request (LAR) 00 Revision to Technical Specification Spent Fuel Assembly Storage Racks and Fuel Storage Capacity," (Accession #: ML003760128) October 4, 2000.

Westinghouse Non-Proprietary Class 3 Attachment to NF-TB-13-38 March 18, 2013 Comanche Peak Nuclear Power Plant Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis Non-Proprietary (145 Pages Attached)

©2013 Westinghouse Electric Company LLC. All rights reserved.

ATTACHMENT 2 to TXX-13045 PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

Pages:

3.7-36 3.7-37 3.7-38 3.7-39 3.7-40 3.7-41 3.7-42 4.0-2 5.5-17 Insert for 3.7.37 Insert for 3.7.38 Insert for 3.7.39 (5 Pages)

Insert for 5.5.17 to TXX-1 3045 Page 1 of 17 Fuel Storage Pool Boron Concentration 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Fuel Storage Pool Boron Concentration LCO 3.7.16 APPLICABILITY:

The fuel storage pool boron concentration shall be Ž_ 2000 ppm.

When fuel assemblies are stored in the fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool boron

-NOTE--------..

concentration not within LCO 3.0.3 is not applicable.

limit.

A.1 Suspend movement of fuel Immediately assemblies in the fuel storage pool AND A.2 Initiate action to restore fuel storage Immediately pool boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify the fuel storage pool boron concentration is within In accordance with

limit, the Surveillance Frequency Control Program.

COMANCHE PEAK - UNITS 1 AND 2 3.7-36 Amendment No. 4fQ-, 156 to TXX-1304,4 Page 2 of 17 Delete LCO, APPLICABILITY, and ACTIONS and insert revised LCO, APPLICABILITY, and ACTIONS (Insert 1).

Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage LCO 3.7.17 The combination of initial enrichment, burnup and decay time of each spent fuel assembly stored in Region II racks shall be within either (1) the "acceptable" domain of Figure 3.7.17-1 in a 4 out of 4 configuration, (2) the "acceptable" domain of Figure 3.7.17-2 in a 3 out of 4 configuration, (3) the "a'ceptable" domain of Figure 3.7.17-3 in a 2 out of 4 configuration, or (4) all be stored in a 1 out of 4 configuration. The acceptable storage config rations are shown in Figure 3.7.17-4.

APPLICABILITY:

Whenever any el assembly is stored in Region II racks of the spent fuel storage po ACTIONS CONDITION REQUI ED ACTION COMPLETION TIME A. Requirements of the LCO A.1 ------------------ NOT--------------

not met.

LCO 3.0.3 is not appli ble.

Initiate action to move the Immediately noncomplying fuel assembly t n

acceptable storage location.

COMANCHE PEAK - UNITS 1 AND 2 3.7-37 Amendment No. 46Gý 156 to TXX-13045 Delete SUF Page T3 179REQUIRE*

revised SU REQUIREI*

SURVEILLANCE RE IREMENTS WVEILLANCE AENTS and insert RVEILLANCE AENTS (Insert 2).

Spent Fuel Assembly Storage 3.7.17 URVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administra i means the initial enrichment, Prior to storing the sI a en burnup and decay time the fuel assembly is in fuel assembly in e

the a ie b

accordance with either (1)

"acceptable" domain of Region 11 racks ey e ti ichm is in ur iý out 0 le domain of i

ue'acceptab Figure 3.7.17-1 in a 4 out of 4 figuration, (2) the im acceptable" domain of Figure 3.7.

-2 in a 3 out of 4 configuration, (3) the "acceptable" do of Figure 3.7.17-3 in a 2 out of 4 configurati or (4) a 1 out of 4 configuration. The acceptable storage c igurations Ic 1 )

m are shown in Figure 3.7.17-4.

I t

COMANCHE PEAK - UNITS 1 AND 2 3.7-38 Amendment No. 460-,156 to TXX-13045 Page 4 of 17 Spent Fuel Assembly Storage 3.7.17 Delete Figure 3.7.17-1 and insert revised Figure 3.7.17-1 and Tables 3.7.17-1, 3.7.17-2, and 3.7.17-3 (Insert 3).

Initial U-235 Enrichment (w/o)

Figure 3.7.17-1 (page 1 of 1)

Fuel Assembly Burnup vs. U-235 Enrichments vs. Decay Time Lii For a 4 out of 4 Storage Configuration in Region II Racks COMANCHE PEAK - UNITS 1 AND 2 3.7-39 Amendment No. 460-,156 to TXX-13045 Page 5 of 17 Spent Fuel Assembly Storage 3.7.17 Delete Figure 3.7.17-2 1 I.

E2 LL Initial U-235 Enrichment (w/o)

Figure 3.7.17-2 (page 1 of 1)

Minimum Burnup vs. Initial U-235 Enrichment vs. Decay Time For a 3 out of 4 Storage Configuration in Region II Racks COMANCHE PEAK - UNITS 1 AND 2 3.7-40 Amendment No. 46Gý 156 to TXX-13045 Page 6 of 17 20000 15000 F-10000 3

E 0

LL 5000 0

Spent Fuel Assembly Storage 3.7.17 IDelete Figure 3.7.17-3 1

--V 17 2.8 3.0 3.2 3.4 3.6 3.8 4.0 4.2 4.4 4.6

4.

5.0 Initial U-235 Enrichment (w/o)

Figure 3.7.17-3 (page 1 of 1)

Minimum Burnup vs. Initial U-235 Enrichment For a 2 out of 4 Storage Configuration in Region II Racks COMANCHE PEAK - UNITS 1 AND 2 3.7-41 Amendment No. 46Gý 156 to TXX-13045 Page 7 of 17 Spent Fuel Assembly Storage A

B C

D Note:

Delete Figure 3.7.17-4 I3.7.17 A

A A

A A

A A

A A

A A

A "AAA A

A A

AýA IAIA A

A A 1A A

A A

A 1A NA A

A !A C

C

• C C

\\

C

• C C

C C\\N, B

B B

B B

B B

B B

B B

B B

B B

B B

B B

B B

B B

B B

B B

D D

D D

D D

C C

I ID D

VD Region 11 (4/4), new or partially spent fu assemblies in the "acceptable" domain of Figure 3.7.17-1.

Region II (3/4), new or partially spent fuel a emblies in the "acceptable" domain of Figure 3.7.17-2.

Region 11 (2/4), new or partially spent fuel asse lies in the "acceptable" domain of Figure 3.7.17-3.

Region 11 (1/4), new or partially spent fuel assemblie which are stored in an expanded checkerboard (1 out of 4).

- empty All possible 2 by 2 matrices containing Region II rack cells shall comply with at least one of the following: (1) within the "acceptable" domain of Figure

.7.17-1 in a 4 out of 4 configuration, (2) within the "acceptable" domain of Figure 3.7. 7-2 in a 3 out of 4 configuration, (3) within the "acceptable" domain of Figure 3.7.173 in a 2 out of 4 configuration, or (4) a 1 out of 4 configuration.

Region I and Region II interface restrictions: The Region II 1 out of onfiguration shall be oriented such that the single fuel assembly resides in the internal r with the empty cells facing Region I. There are no interface restrictions between the Re *on II (2/4, 3/4, 4/4) and Region I configurations.

Figure 3.7.17-4 (page 1 of 1)

Storage Configurations (4/4, 3/4, 2/4, 1/4) in Region II Racks COMANCHE PEAK - UNITS 1 AND 2 3.7-42 Amendment No. 460-,156 to TXX-13045 Page 8 of 17 Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b.

keff < 1.0 when fully flooded with unborated water which includes an allowance for uncertainties as described in Section 4.3 of the FSAR;

c.

keff < 0.95 if fully flooded with water borated to 80N ppm, which includes an allowance for uncertainties as described in Section 4.3 of the FSAR;

d.

A nominal 9 inch center to center distance between fuel storage locations in Region II fuel storage racks; S10.65I

e.

A nominal 4-1@ inch by nominal -- nc center to center distance between fuel assemblies placed in Region I fuel storage racks; Storage of new or spent fuel assemblies in Region II storage racks must comply with 3.7.17 Spent Fuel Assembly Storage.

New or partially spent fuel assemblies may be allowed restricted storage in a 1 out of 4 configuration in Region 11 fuel storage racks (as shown in Figure 3

),or unrestricted storage in Region I fuel storage racks;

-3.17AraII

9.

Now or partially cpont fuel assomblies with a dicc6harge lbFRnUP in tho "accoeptablo" domfafin Of Figuro 3.7.17 2 may be allowod URoctrictod sterago in a 3 out of 4 configuration in Region 11 fuol ctorage Fackc as chown-, in Figurc 3.7.17 4,; ad

14. Now or partially sepot fuolacoblo with a dischaF0o buIRnup-in the "accosptablo" domain of Figuro 3.7-.17 3 may be al!owod roctrictod setoago in a 2 out of 4 configuration in Rogion 11 fuol tetr~ago rackc ac chown in Figuro 3.7.17 4;.~

nt~g Fak IIseninFq~

COMANCHE PEAK - UNITS 1 AND 2 4.0-2 Amendment No. 150 to TXX-13045 Page 9 of 17 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.21 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a.

The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b.

Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI-04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c.

The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to ilnsert 41___the Frequencies established in the Surveillance Frequency Control Program.

COMANCHE PEAK - UNITS 1 AND 2 5.5-17 Amendment No. 156 to TXX-13045 Page 10 of 17 Insert I LCO 3.7.17 APPLICABILITY:

New or spent fuel assemblies will be stored in compliance with Figure 3.7.17-1.

Whenever any fuel assembly is stored in Region II of the spent fuel storage pool.

ACTIONS COMPLETION CONDITION REQUIRED ACTIONS TIME TIME A. Requirements of the LCO not A.1 NOTE --.....-------

met.

LCO 3.0.3 is not applicable Initiate action to move fuel as Immediately necessary to restore compliance.

to TXX-13045 Page 11 of 17 Insert 2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the acceptability of Prior to moving a fuel fuel movement plans and the resulting storage assembly into any Region configuration in accordance with Figure 3.7.17-1.

II storage location.

to TXX-13045 Page 12 of 17 Insert 3 Figure 3.7.17-1 (Page 1 of 3)

Spent Fuel Pool Loading Restrictions All 2x2 Region II storage cell arrays shall comply with one of the Arrays Definitions below. Each storage location is a corner location for up to 4 separate 2x2 arrays.

A. Arrays Il-A through II-E designate the pattern of fuel which may be stored in any 2x2 Array, and are dependent upon Fuel Category.

B. Fuel Categories 1-6 are determined based on Fuel Burnup, Initial Enrichment, Decay Time, and Fuel Group.

C. Fuel Group F1 assemblies have a nominal rod outer diameter of 0.374 inches. Fuel Group F2 assemblies have a nominal rod outer diameter of 0.360 inches.

Array Definition Illustration Array II-A Category 6 assembly in every cell. Only valid for two rows adjacent 6

6 to the SFP wall. The two rows adjacent to Array II-A must be Array Array r-II-B I-13, and the empty cell in Array Il-B must be adjacent to Array II-A.

6 6

Array Il-B Category 4 fuel assembly in 3 out of 4 cells, with empty cell in the forth cell.

4 4

X 4

Array II-C Pattern which contains fuel in 3 out of 4 cells, including two diagonally-opposed 5

Category 5 assemblies, one Category 3 assembly, and one empty location. Only Fuel Group F2 assemblies may be stored in Array II-C.

X 5

Array II-D Checkerboard pattern of two diagonally-opposed Category 2 assemblies with two 2

X diagonally-opposed empty cells.

X 2

Array II-E 1 out of 4 storage array, with 3 empty cells.

X X

x 1

to TXX-13045 Page 13 of 17 Insert 3 (Continued)

Figure 3.7.17-1 (Page 2 of 3)

Notes:

1. Fuel Categories are ranked in order of relative reactivity, from Category 1 to 6. Fuel Category 1 assemblies have the highest reactivity, and Fuel Category 6 assemblies have the lowest.

2. All Fresh Fuel Assemblies (assemblies with a burnup value of 0.0 MWD/MTU) should be considered Category 1 fuel, independent of Fuel Group or Enrichment.

a. In Fuel Group F1, Fuel Category 1 is fresh fuel up to 3.5 weight percent U-235 Initial Enrichment.

b. In Fuel Group F2, Fuel Category 1 is fresh fuel up to 5.0 weight percent U-235 Initial Enrichment.

3. Fuel Category 2 is any Non-Fresh fuel assembly up to 3.5 weight percent U-235 Initial Enrichment (Burnup Requirement is > 0 MWD/MTU).

4. For all other fuel, Fuel Categories are determined as follows:

a. For Initial Enrichment values below the Minimum Applicable Initial Enrichment values of Table 3.7.17-1, the Fitting Coefficients of Tables 3.7.17-2 and 3.7.17-3 are not applicable. The Minimum Burnup Requirement for the corresponding Category is >

0 MWD/MTU.

b. For Fuel Group F1 assemblies, determine the Fitting Coefficients A1 - A4 using Table 3.7.17-2. Note that Table 3.7.17-2 is only applicable to fuel with > 10 years of decay time, and an Initial Enrichment of -< 3.5 weight percent.

c. For Fuel Group F2 assemblies, determine the Fitting Coefficients A1 - A4 using Table 3.7.17-3.

d. The required Minimum Burnup value (in MWD/MTU) for each Fuel Category is calculated based on Initial Enrichment (En) and the appropriate fitting coefficients, using the equation below. If the fuel assembly burnup is greater than or equal to the calculated Minimum Bumup value, then the fuel may be classified into this Fuel Category.

Minimum Burnup = 1,000 x [Al x En 3 + A2 x En 2 + A3 x En + A4]

e. All relevant uncertainties are explicitly included in the criticality analysis. No additional allowance for burnup uncertainty or initial enrichment uncertainty is required.

. Attachment 2 to TXX-13045 Page 14 of 17 Insert 3 (Continued)

Figure 3.7.17-1 (Page 3 of 3)

Notes (continued):

4. (continued)

f. Conservatively low values of Decay Time may be used to calculate the Minimum Burnup value, or interpolation may be used. If interpolation is used, Minimum Burnup values for tabulated Decay Time values above and below the actual value should first be determined. Next, linear interpolation between these values may be used to determine the Minimum Burnup value. No extrapolation beyond 20 years is permitted.

g. Initial Enrichment (En) is the nominal U-235 weight percent enrichment of the central zone region of fuel, excluding axial blankets, prior to fuel depletion.

h. If the computed Minimum Burnup value < 0 MWD/MTU, the Minimum Burnup Requirement is > 0 MWD/MTU.

5. In all Arrays, an assembly with a higher Fuel Category number can be utilized in place of any fuel assembly with a lower Fuel Category Number, with the following exception.

a. Fuel Group F1 assemblies are not allowed to be stored in Array II-C, regardless of Fuel Category.

6. An empty (water-filled) cell can be substituted for any fuel-containing cell in all storage arrays.

7. Any storage array location designated for a fuel assembly can be replaced with non-fissile hardware. Items other than Fuel Assemblies which contain fissile material are restricted to storage in Region I.

8. Fuel assembly inserts approved for use in the reactor core can be inserted in a stored assembly (in the Spent Fuel Pool) without affecting the fuel category.

to TXX-13045 Page 15 of 17 Insert 3 (Continued)

Table 3.7.17-1 Minimum Applicable Initial Enrichment for Table 3.7.17-2 and Table 3.7.17-3 Fitting Coefficients Fuel Category Fuel Group F1 Fuel Group F2 6

1.25 1.20 5

N/A 1.30 4

1.35 1.45 3

N/A 1.45 2

N/A 3.55 Table 3.7.17-2 Fuel Group F1 Nominal Fuel Rod Outer Diameter of 0.374" Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Decay Time and Initial Enrichment (En)

Fuel Decay Fitting Coefficients Category (yrs)

A1 A 2 A 3 A4 6

10 1.4351

-17.3247 73.3805

-67.4585 6

15 1.7078

-18.7916 74.6322

-67.2637 6

20 0.5289

-9.9969 1 53.7741

-52.6302 4

10

-0.0444

-1.3474 22.7039

-28.0852 4

15 0.2015

-2.6257 24.1016

-28.2473 4

20 0.4646

-4.1432 26.3891

-29.2170 Note: Fuel must have at least 10 years of decay time and less than or equal to 3.5 weight percent Initial Enrichment to utilize Table 3.7.17-2.

to TXX-13045 Page 16 of 17 Insert 3 (Continued)

Table 3.7.17-3 Fuel Group F2 Nominal Fuel Rod Outer Diameter of 0.360" Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Decay Time and Initial Enrichment (En)

Fuel Decay Fitting Coefficients Category (yrs)

A1 A2 A3 A4 6

0 0.5789

-7.4498 42.4056

-41.1591 6

5 0.5247

-6.8992 39.7676

-38.6927 6

10 0.2701

-4.4306 31.9841

-32.4674 6

15 0.3105

-4.5582 31.1825

-31.3916 6

20 0.2374

-3.8754 28.8900

-29.4975 5

0 0.9373

-11.2553 54.7226

-54.1769 5

5 0.6169

-8.1494 44.7801

-45.7968 5

10 0.5380

-7.1852 40.7044

-41.9545 5

15 0.5385

-7.0180 39.2299

-40.3213 5

20 0.5200

-6.7906 38.0244

-39.0979 4

0 0.2553

-3.9826 30.6152

-36.7967 4

5 0.2366

-3.6430 28.2160

-33.9749 4

10 0.4387

-5.6018 33.3609

-37.9327 4

15 0.5450

-6.6302 36.0760

-40.0315 4

20 0.6327

-7.4663 38.2724

-41.7257 3

0 0.5317

-6.1006 32.7118

-36.2263 3

5 0.5228

-5.9434 31.2846

-34.4602 3

10 0.5432

-6.1075 31.1578

-33.9933 3

15 0.5206

-5.8897 30.1768

-32.9600 3

20 0.5158

-5.7796 29.4050

-32.0577 2

0 0.0000 1.6738

-8.5396 9.2206 to TXX-1 3045 Page 17 of 17 INSERT 4 5.5.22 Spent Fuel Storage Rack Neutron Absorber Monitoring Program The Region I storage cells in the CPNPP Spent Fuel Pool utilize the neutron absorbing material BORAL, which is credited in the Safety Analysis to ensure the limitations of Technical Specification 4.3.1.1 are maintained.

In order to ensure the reliability of the Neutron Poison material, a monitoring program is required to routinely confirm that the assumptions utilized in the criticality analysis remain valid and bounding. The Neutron Absorber Monitoring Program is established to monitor the integrity of neutron absorber test coupons periodically as described below.

A test coupon "tree" shall be maintained in each SFP. Each coupon tree originally contained 8 neutron absorber surveillance coupons. Detailed measurements were taken on each of these 16 coupons prior to installation, including weight, length, width, thickness at several measurement locations, and B-10 content (g/cm2). These coupons shall be maintained in the SFP to ensure they are exposed to the same environmental conditions as the neutron absorbers installed in the Region I storage cells, until they are removed for analysis.

One test coupon from each SFP shall be periodically removed and analyzed for potential degradation, per the following schedule. The schedule is established to ensure adequate coupons are available for the planned life of the storage racks.

Year Coupon Number Year Coupon Number 2013 1

2028 5

2015 2

2033 6

2018 3

2043 7

2023 4

2053 8

Further evaluation of the absorber materials, including an investigation into the degradation and potential impacts on the Criticality Safety Analysis, is required if:

• A decrease of more than 5% in B-10 content is from the initial value observed in any test coupon as determined by neutron attenuation.

" An increase in thickness at any point is greater than 25% of the initial thickness at that point.

ATTACHMENT 3 to TXX-13045 PROPOSED TECHNICAL SPECIFICATIONS BASES CHANGES (Markup For Information Only)

Pages:

B 3.7-72 B 3.7-73 B 3.7-74 B 3.7-75 B 3.7-76 B 3.7-77 B 3.7-78 B 3.7-79 Insert 1 for TS Bases Insert 2 for TS Bases (3.7.16 - 6 pages)

(3.7.17 - 4 pages) to TXX-13045 Page 1 of 18 Fuel Storage Pool Boron Concentration B 3.7.16 Insert 1, Revised Bases for 3.7.16 B 3.7 PLANT SYSTEMS B 3.7.16 Fuel Storage Pool Boron Concentration BASES BACKGROUND A common Fuel Building houses facilities for storage and transfer of new and spent fuel. Two pools are provided for CPSES spent fuel storage. Each ol may be used to store fuel from either or both of the CPSES units.

In t Region II rack (References 1 and 2) design, the spent fuel storage pool mbers 1 and 2 (SFP1 and SFP2) permit four different configurations (as sh n in Figure 3.7.17-4) which, for the purpose of criticality consider tions, are considered as separate pools. Region II racks, with 1462 and 470 storage positions in SFP1 and SFP2 respectively (2932 total), are d igned to accommodate fuel of various initial enrichments which have accumu ted minimum burnups and decay times within either (1) the "acceptable" d ain of Figure 3.7.17-1 in a 4 out of 4 configuration, (2) the "acceptable" do in of Figure 3.7.17-2 in a 3 out of 4 configuration, (3) the "acceptable" dom

• of Figure 3.7.17-3 in a 2 out of 4 configuration, or (4) a 1 out of 4 configura n as shown in Figure 3.7.17-4.

Region I racks (Refere es 1 and 2) with 222 and 219 storage positions located in SFP1 and SF respectively (441 total), constitute a fifth configuration within the po Is. These Region I racks are designed to accommodate new fuel with maximum enrichment of 5.0 w/t % U-235 or spent fuel regardless of the di charge fuel burnup or decay time. Soluble boron is not credited for the sto ge of spent fuel assemblies within the Region I racks, and there are no torage pattern restrictions associated with the Region I racks. The neutron a sorber material Boral is credited for the storage of spent fuel assemblies wit in the Region I racks to maintain keff less than or equal to 0.95.

Soluble boron is not credited for the stora e of fuel assemblies within the Region II racks in the 1 out of 4 and 2 out 4 configurations. Criticality analyses have been performed (Reference which demonstrate that the multiplication factor, keff, of the fuel and spent el storage racks is less than or equal to 0.95.

In order to maintain keff less than or equal to 0.95, e presence of fuel pool soluble boron is credited for the storage of fuel assem lies within the Region II racks in the 3 out of 4 and 4 out of 4 configurations.

description of how credit for fuel storage pool soluble boron is used under n rmal storage configuration conditions is found in References 2, 3, and The storage configuration is defined using calculations to ensure that keff ill be less than 1.0 with no soluble boron under normal storage conditions in ding continued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.7-72 Revision 67 to TXX-13045 Fuel Storage Pool Boron Concentration Page 2 of 18 B 3.7.16 BASES BACKGROUND (cn'ued) lrances and uncertainties. Soluble boron credit is then used to maintain lss than or equal to 0.95. Criticality analyses have been performed (R erence 3) which demonstrate that the pools require 800 ppm of soluble boro to maintain keff less than or equal to 0.95 for all allowed combinations of sto ge configurations, enrichments, burnups, and decay time limits. The effect o B-1 0 depletion on the boron concentration for maintaining keff less than or ual to 0.95 is negligible.

Criticality an lyses considering accident conditions have also been performed (R erences 2 and 3). These analyses establish the amount of soluble boron cessary to ensure that keff will be maintained less than or equal to 0.95 sh Id pool temperatures fall outside the assumed range or a fuel assembly mis ad occur. The total amount of soluble boron required to mitigate these even is 1900 ppm.

For an occurrence of t above postulated accident condition, the double contingency principle of NSI/ANS 8.1-1983 (Reference 6) can be applied.

This states that one is no equired to assume two unlikely, independent, concurrent events to ensur protection against a criticality accident. Thus, for these postulated acciden onditions, the presence of additional soluble boron in the storage pool wate (above the concentration required for normal conditions and reactivity equiva ncing) can be assumed as a realistic initial condition since not assuming its esence would be a second unlikely event.

A boron concentration equal to or g ater than 2000 ppm assures that a dilution event which will result in a keff reater than 0.95 is not credible. This is demonstrated by a boron dilution ana sis performed for the CPSES Spent Fuel pools. This conclusion is based on e following: (1) a substantial amount of water is needed in order to dilu the SFP to the design keff of 0.95, (2) since such a large water volume tu over is required, a SFP dilution event would be readily detected by plant pers nnel via alarms, flooding in the fuel and auxiliary buildings or by normal op rator rounds through the SFP area, and (3) evaluations indicate that, bas on the flow rates of non-borated water normally available to the SFP, take in conjunction with significant operator errors, and equipment failures, fficient time is available to detect and respond to a dilution event. In addition, there is significant conservatism built into this evaluation; for example, th cooling of the spent fuel pools can be performed by one train supplying com on water to both pools. This cooling configuration would allow credit of th volume of both pools and substantially increase the dilution time estimates resented.

However, because the flexibility exists for the cooling syste to be totally dedicated to one pool, only one pool volume is considered in is evaluation.

continued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.7-73 Revision 67 to TXX-13045 Page 3 of 18 Fuel Storage Pool Boron Concentration B 3.7.16 BASES BACKGROUND (cntinued)

It should be noted that this boron dilution evaluation considered the boron ilution volumes required to dilute the SFP from 1900 ppm to 800 ppm. The 0 ppm end point was utilized to ensure that keff for the spent fuel racks w

Id remain less than or equal to 0.95. However, as discussed above, calc lations for Region II 3 out of 4 and 4 out of 4 configurations have been perfo* ed on a 95/95 basis to show that the spent fuel rack keff remains less than 1. with non-borated water in the pool. Thus, even if the SFP were diluted t concentrations approaching zero ppm, the fuel in the Region II racks wou remain subcritical and the health and safety of the public would be protecte The storage o uel with initial enrichments up to and including 5.0 weight percent U-235 i the Comanche Peak fuel storage pools has been evaluated. For th Region II storage racks, the resulting enrichment, burnup, and decay *me limits for the pool are shown in Figures 3.7.17-1 through 3.7.174.

APPLICABLE SAFETY ANALYSES Most fuel storage pool ac 'dent conditions will not result in a significant increase in keff. Examples such accidents are the drop of a fuel assembly on top of a rack, and the dro of a fuel assembly outside but adjacent to the rack modules.

A dropped assembly accident oc rs when a fuel assembly is dropped onto the storage racks. The rack struct e is not excessively deformed. An assembly, in its most reactive conditi n, is considered in the criticality evaluation. Accident analyses have b n performed which demonstrate that the dropped assembly which comes to st horizontally on top of the rack has sufficient water separating it from the ctive fuel height of stored assemblies to preclude neutronic interactio. This is true even with unborated water. For the borated water con ition, the potential for interaction is even less since the water contai boron which is an additional thermal neutron absorber.

However, three accidents can be postulated for ea h storage configuration that could increase reactivity beyond the analyzed c ndition. The first postulated accident would be a change in pool tempe ture to outside the range of normal operating temperatures assumed in th criticality analyses (50°F to 1500F). The second accident would be droppin a fuel assembly into an already loaded cell. The third would be the misloa *ng of a fuel assembly within the racks into a cell for which the restrictio on location, enrichment, burnup, or decay time are not satisfied or adjace to but outside the racks.

COMANCHE PEAK - UNITS 1 AND 2 B 3.7-74 Revision 67 to TXX-13045 Fuel Storage Pool Boron Concentration Page 4 of 18 B 3.7.16 BASES APPLICABLE SAFE NALYSES (continued)

V iations in the temperature of the water passing through the stored fuel ass mblies outside the normal operating range were considered in the critic ity analysis. The reactivity effects of a temperature range from 320F to 212°F ere evaluated. The increase in reactivity due to the change in tempera re is bounded by the misloading accident.

For the acc ent of dropping a fuel assembly into an already loaded cell, the upward axial akage of that cell will be reduced; however, the overall effect on the rack re tivity will be insignificant. This is because minimizing the upward-only lea ge of just a single cell will not cause any significant increase in reacti

y. Furthermore, the neutronic coupling between the dropped assembly d the already loaded assembly will be low due to several inches of ass mbly nozzle structure which would separate the active fuel regions. Therefor this accident would clearly be bounded by the misloading accident.

The fuel assembly misloadi accident involves placement of a fuel assembly in a location for wh h it does not meet the requirements for enrichment, burnup, or decay ti e including the placement of an assembly in a location that is required to be I empty. The result of the misloading is to add positive reactivity, increasing ff toward 0.95. The maximum required boron to compensate for this event 1900 ppm, which is below the LCO limit of 2000 ppm.

The concentration of dissolved boron in e fuel storage pool satisfies Criterion 2 of the 10CFR50.36(c)(2)(ii).

LCO The fuel storage pool boron concentration is req ired to be _> 2000 ppm. The specified concentration of dissolved boron in the el storage pool preserves the assumptions used in the analyses of the poten I criticality accident scenarios as described in Reference 5. The amoun of soluble boron required to offset each of the above postulated accide ts was evaluated for all of the proposed storage configurations. The specifie minimum boron concentration of 2000 ppm assures that the concentratiocwill remain above these values.\\

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel storage pool.

(con -ued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.7-75 Revision 67 to TXX-13045 Page 5 of 18 Fuel Storage Pool Boron Concentration B 3.7.16 BASES (continued)

N ACTIONS A. and A.2 Wh the concentration of boron in the fuel storage pool is less than requi d, immediate action must be taken to preclude the occurrence of an accide t or to mitigate the consequences of an accident in progress. This action is ost efficiently achieved by immediately suspending the movement of fuel as emblies. The concentration of boron is restored simultaneously with suspe ding movement of fuel assemblies. Prior to resuming movement of fuel asse blies, the concentration of boron must be restored. This requirement es not preclude movement of a fuel assembly to a safe position.

The Required Acions are modified by a Note indicating that LCO 3.0.3 does not apply. If the L 0 is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO.0.3 would not be applicable. If moving irradiated fuel assemblies while in ODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Th efore, inability to suspend movement of fuel assemblies is not suffici nt reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.16.1_

REQUIREMENTS This SR verifies that the conce tration of boron in the fuel storage pool is within the required limit. As Ion as this SR is met, the analyzed accidents are fully addressed. The Surveill ce Frequency is controlled under the Surveillance Frequency Control Pr ram.

REFERENCES

1.

FSAR, Section 9.1.

2.

License Amendment Requests 94-2, 98-08, and 00-05, Spent Fuel Storage Capacity Increase, Docket OS 50-445 and 50-446, CPSES.

3.

Comanche Peak High Density Spent Fu I Rack Criticality Analysis using Soluble Boron Credit and No Oute Wrapper Plate, dated July, 2001 (Enclosure 2 to TXX-01118).

4.

WCAP-14416 NP-A, Rev. 1, "Westinghouse ent Fuel Rack Critical-ity Analysis Methodology," November 1996.

5.

FSAR, Section 15.7.4.

6.

American Nuclear Society, "American National Stan ard for Nuclear Criticality Safety in Operations with Fissionable Mate *als Outside Reactors," ANSI/ANS-8.1-1983, October 7, 1983.

I COMANCHE PEAK - UNITS 1 AND 2 B 3.7-76 Revision 67 to TXX-13045 Spent Fuel Assembly Storage Page 6 of 18 B 3.7.17 Insert 2, Revised Bases for 3.7.17 B 3.7 PLANT SYSTEMS B 3.7.17 Spent Fuel Assembly Storage BASES BACKGROUND A common Fuel Building houses facilities for storage and transfer of new and spent fuel. Two pools are provided for CPSES spent fuel storage. Each aol may be used to store fuel from either or both of the CPSES units.

In t Region II rack (References 1 and 2) design, the spent fuel storage pool mbers 1 and 2 (SFP1 and SFP2) permit four different configurations (as sh n in Figure 3.7.17-4) which, for the purpose of criticality consider ions, are considered as separate pools. Region II racks, with 1462 and 70 storage positions in SFP1 and SFP2 respectively (2932 total), are d igned to accommodate fuel of various initial enrichments which have accumu I

ted minimum burnups and decay times within either (1) the "acceptable" do ain of Figure 3.7.17-1 in a 4 out of 4 configuration, (2) the "acceptable" domr in of Figure 3.7.17-2 in a 3 out of 4 configuration, (3) the "acceptable" doma of Figure 3.7.17-3 in a 2 out of 4 configuration, or (4) a 1 out of 4 configurati as shown in Figure 3.7.17.4.

Region I racks (Referen s 1 and 2) with 222 and 219 storage positions located in SFP1 and SFP respectively (441 total) constitute a fifth configuration within the poo

• These Region I racks are designed to accommodate new fuel with aximum enrichment of 5.0 w/t % U-235 or spent fuel regardless of the dis arge fuel burnup. Soluble boron is not credited for the storage of spent el assemblies within the Region I racks, and there are no storage pattern r trictions associated with the Region I racks. The neutron absorber materid Boral is credited for the storage of spent fuel assemblies within the Regi I racks to maintain keff less than or equal to 0.95.

A discussion of how soluble boron is credit for the storage of spent fuel assemblies is contained in the BACKGROU for B 3.7.16.

Within the SFP1 Region II racks, there exist two ersized (2x2) cells.

Within the SFP2 Region I racks, there exists one o rsized (2x2) cell. These oversized cells are not approved for storage of eithe fresh or spent fuel.

However, they can be used as a place in the pool for assembly to be lowered and raised while being inspected. Prior to use the inspection cells certain prerequisites must be met. Criticality analyses (R ference 3) have been performed which demonstrate that there is no increas in reactivity relative to the approved Region II storage configurations (the urrent licensing basis requirements for the spent fuel pool are still me provided that administrative prerequisites are maintained for the oversize cells in (c

tinued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.7-77 Revision 67 to TXX-13045 Page 7 of 18 Spent Fuel Assembly Storage B 3.7.17 BASES BACKGROUND ontinued)

SFP1 Region II racks. The prerequisite for the use of the oversized cells in Region II racks is that all the Region II cells in the first row surrounding the versized cell remain empty. This results in a total of 8 empty Region II cells a acent to the oversized cell in the SFP I Region II rack adjacent to the Re ion I rack and a total of 5 empty Region II cells adjacent to the oversized cell 1 the SFP1 Region II racks adjacent to the spent fuel pool walls. There are norerequisites for the use of the oversized cell in SFP2 Region I racks since th criticality analyses (Reference 3) demonstrate there is no increase in reactivi relative to the approved Region I storage configuration.

APPLICABLE SAFETY ANALYSES Adiscussion o he criticality analysis for the storage of spent fuel assemblies is co tained in the APPLICABLE SAFETY ANALYSES for B 3.7.16.

Most fuel storage poo accident conditions will not result in a significant increase in keff. Examp s of such accidents are the drop of a fuel assembly on top of a rack, and the op of a fuel assembly outside but adjacent to the rack modules. However, a idents can be postulated for each rack storage configuration which could inc ase reactivity beyond the analyzed condition.

A discussion of these acciden is contained in B 3.7.16.

By closely controlling the movem t of each assembly and by checking the location of each assembly after mo ment, the time period for potential accidents may be limited to a small fr ction of the total operating time.

The configuration of fuel assemblies in t fuel storage pool satisfies Criterion 2 of 10CFR50.36(c)(2)(ii).

LCO The restrictions on the placement of fuel assem lies within the spent fuel pool, in accordance with Figures 3.7.17-1 through

.7.17-4, in the accompanying LCO, ensures the keff of the spent f I storage pool will always remain <0.95, assuming the pool to be floode with borated water.

NOTE: The oversized inspection cells within the racks a not approved storage locations and are not covered by the LCO. Admin trative controls which govern the use of the inspections cells are described the BACKGROUND.

nue

)

( tined COMANCHE PEAK - UNITS 1 AND 2 B 3.7-78 Revision 67 to TXX-13045 Page 8 of 18 Spent Fuel Assembly Storage B 3.7.17 BASES (continue\\)

APPLICABILITY This LCO applies whenever any fuel assembly is stored in Region II racks of he fuel storage pool.

ACTIONS A.1 When th configuration of fuel assemblies stored in Region II racks of the spent fuel orage pool is not in accordance with Figures 3.7.17-1 through 3.7.17-4, the immediate action is to initiate action to make the necessary fuel assembly mo ment(s) to bring the configuration into compliance with Figures 3.7.17-through12 3.7.17-4.

Required Action A. *s modified by a Note indicating that LCO 3.0.3 does not apply. If unable to m ye irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not b applicable. If unable to move irradiated fuel assemblies while in MO 1, 2, 3, or 4, the action is independent of reactor operation. Therefore, inab 'ty to move fuel assemblies is not sufficient reason to require a reactor s utdown.

SURVEILLANCE SR 3.7.17.1 REQU IREM ENTS This SR verifies, by administrative me s, that the initial enrichment, burnup and decay time of the fuel assembly is i accordance with Figures 3.7.17-1 through 3.7.17-4 in the accompanying LC REFERENCES

1.

FSAR Section 9.1.

2.

License Amendment Request 94-22, 98-08, nd 00-05 Spent Fuel Storage Capacity Increase, Docket Nos. 50-4 and 50-446, CPSES.

3.

Criticality Safety Analysis of Holtec Spent Fuel Ra ks, dated January, 2003 (Holtec Report HI-2002436, Revision 9).

COMANCHE PEAK - UNITS 1 AND 2 B 3.7-79 Revision 67 to TXX-13045 Page 9 of 18 Insert 1, Revised Bases for TS 3.7.16 B 3.7 PLANT SYSTEMS B 3.7.16 Fuel Storage Pool Boron Concentration BASES BACKGROUND A common Fuel Building houses facilities for storage and transfer of new and spent fuel. Two pools are provided for CPNPP spent fuel storage. Each pool may be used to store fuel from either or both of the CPNPP units.

In the Region II rack (References I and 2) design, the spent fuel storage pool numbers 1 and 2 (SFP1 and SFP2) permit five different configurations (as shown in Figure 3.7.17-1). Region II racks, with 1462 and 1470 storage positions in SFP1 and SFP2 respectively (2932 total), are designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups and decay times as required by Figure 3.7.17-1.

Region I racks (References I and 2) with 222 and 219 storage positions located in SFP1 and SFP2 respectively (441 total), constitute a sixth configuration within the pools. These Region I racks are designed to accommodate new fuel with a maximum enrichment of 5.0 w/t % U-235 or spent fuel regardless of the discharge fuel burnup or decay time. There are no storage pattern restrictions associated with the Region I racks. The neutron absorber material Boral is credited for the storage of spent fuel assemblies within the Region I racks to maintain keff less than or equal to 1.0 at 0 ppm soluble boron concentration.

In order to maintain kefr less than or equal to 0.95, the presence of fuel pool soluble boron is credited for the storage of fuel assemblies within the Region I and Region II racks. A description of how credit for fuel storage pool soluble boron is used under normal storage configuration to TXX-13045 Page 10 of 18 conditions is found in Reference 3. The storage configurations are defined using calculations to ensure that keff will be less than 1.0 with no soluble boron under normal storage conditions including biases and uncertainties. Soluble boron credit is then used to maintain keff less than or equal to 0.95. Criticality analyses have been performed (Reference 3) which demonstrate that the pools require 400 ppm of soluble boron to maintain keff less than or equal to 0.95 for all allowed combinations of storage configurations, enrichments, burnups, and decay time limits. The effect of B-10 depletion on the boron concentration for maintaining keff less than or equal to 0.95 is accounted for in Reference 3.

Criticality analyses considering accident conditions have also been performed (Reference 3). These analyses establish the amount of soluble boron necessary to ensure that keff will be maintained less than or equal to 0.95 should pool temperatures fall outside the assumed range or multiple fuel assembly misload events occur. The total amount of soluble boron required to mitigate these events is 2400 ppm.

For an occurrence of the above postulated accident condition, the double contingency principle of ANSI/ANS 8.1-1983 (Reference 6) can be applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for these postulated accident conditions, the presence of additional soluble boron in the storage pool water (above the concentration required for normal conditions, and up to the minimum value required by Technical Specifications) can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely and independent event.

A boron concentration equal to or greater than 2400 ppm assures that a dilution event which will result in a kef greater than 0.95 is not credible. This is demonstrated by a boron dilution analysis performed for the CPNPP Spent Fuel pools. This conclusion is based on the following: (1) a substantial amount of water is needed in order to dilute the SFP to the design keff of 0.95, (2) since such a large water volume turnover is required, a SFP dilution event would be readily detected by plant personnel via alarms, flooding in the fuel and auxiliary buildings or by normal operator rounds through the SFP area, and (3) to TXX-13045 Page 11 of 18 evaluations indicate that, based on the flow rates of nonborated water normally available to the SFP, taken in conjunction with significant operator errors, and equipment failures, sufficient time is available to detect and respond to a dilution event. In addition, there is significant conservatism built into this evaluation; for example, the cooling of the spent fuel pools can be performed by one train supplying common water to both pools. This cooling configuration would allow credit of the volume of both pools and substantially increase the dilution time estimates presented. However, because the flexibility exists for the cooling system to be totally dedicated to one pool, only one pool volume is considered in this evaluation.

It should be noted that this boron dilution evaluation considered the boron dilution volumes required to dilute the SFP from 1900 ppm to 800 ppm. The 800 ppm end point was utilized to ensure that keff for the spent fuel racks would remain less than or equal to 0.95. More recent analysis (Reference 3) demonstrates keff is maintained less than 0.95 with only 400 ppm of SFP boron, and therefore the dilution evaluation remains bounded. However, as discussed above, calculations for Region I and Region II configurations have been performed on a 95/95 basis to show that the spent fuel rack keff remains less than 1.0 with non-borated water in the pool. Thus, even if the SFP were diluted to concentrations approaching zero ppm, the fuel storage racks would remain subcritical and the health and safety of the public would be protected.

The storage of fuel with initial enrichments up to and including 5.0 weight percent U-235 in the Comanche Peak fuel storage pools has been evaluated. For the Region II storage racks, the resulting enrichment, burnup, and decay time limits for the pool are shown in Figure 3.7.17-1.

APPLICABLE SAFETY ANALYSIS Most fuel storage pool accident conditions will not result in a significant increase in keff. Examples of such accidents are the drop of a fuel assembly on top of a rack, and the drop of a fuel assembly outside but adjacent to the rack modules.

to TXX-13045 Page 12 of 18 A dropped assembly accident occurs when a fuel assembly is dropped onto the storage racks. The rack structure is not excessively deformed. An assembly, in its most reactive condition, is considered in the criticality evaluation. The dropped assembly, which comes to rest on top of the rack, has sufficient water separating it from the active fuel height of stored assemblies to preclude neutronic interaction. This is true even with unborated water. For the borated water condition, the potential for interaction is even less since the water contains boron which is an additional thermal neutron absorber.

However, five accidents can be postulated for each storage configuration that could increase reactivity beyond the analyzed condition. The first postulated accident would be a change in pool temperature to outside the range of normal operating temperatures assumed in the criticality analyses (50°F to 1501F). The second accident would be dropping a fuel assembly into an already loaded cell. The third would be the misloading of a fuel assembly within the racks into a cell for which the restrictions on location, enrichment, burnup, or decay time are not satisfied. A forth would be the misload of a fuel assembly adjacent to but outside the racks.

The fifth accident would be misloading of multiple fuel assemblies, in series, into unacceptable storage locations.

Variations in the temperature of the water passing through the stored fuel assemblies outside the normal operating range were considered in the criticality analysis. The reactivity effects of a temperature range from 320F to 212°F were evaluated. The increase in reactivity due to the change in temperature is bounded by the misloading accident.

For the accident of dropping a fuel assembly into an already loaded cell, the upward axial leakage of that cell will be reduced; however, the overall effect on the rack reactivity will be insignificant. This is because minimizing the upward-only leakage of just a single cell will not cause any significant increase in reactivity. Furthermore, the neutronic coupling between the dropped assembly and the already loaded assembly will be low due to several inches of assembly nozzle structure which would separate the active fuel regions. Therefore, this accident would clearly be bounded by the misloading accident.

The fuel assembly misloading accident involves placement of a single fuel assembly in a location for which it does not meet the requirements for enrichment, burnup, or decay time including the placement of an assembly in a location that is required to be left empty. The result of the misloading is to add positive reactivity, increasing keff toward 0.95. The maximum required boron to compensate for this event is 1867 ppm, which is below the LCO limit of 2400 ppm.

to TXX-1 3045 Page 13 of 18 The multiple fuel assembly misloading accident involves the misplacement of assemblies, in series, due to a common cause. The result of this misloading is the addition of positive reactivity, increasing the kea toward 0.95. The maximum required boron concentration to compensate for this event is 2400 ppm, which is the LCO limit.

The concentration of dissolved boron in the fuel storage pool satisfies Criterion 2 of the 10CFR50.36(c)(2)(ii).

LCO The fuel storage pool boron concentration is required to be > 2400 ppm.

The specified concentration of dissolved boron in the fuel storage pool preserves the assumptions used in the analyses of the potential criticality accident scenarios as described in Reference 3. The amount of soluble boron required to offset each of the above postulated accidents was evaluated for all of the proposed storage configurations. The specified minimum boron concentration of 2400 ppm assures that the concentration will remain above these values.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel storage pool.

ACTIONS A.1 and A.2 When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This action is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies. Prior to resuming movement of fuel assemblies, the concentration of boron must be restored. This requirement does not preclude movement of a fuel assembly to a safe position.

The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply. If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend to TXX-1 3045 Page 14 of 18 movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE REQUIREMENTS SR 3.7.16.1 This SR verifies that the concentration of boron in the fuel storage pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES

1.

FSAR, Section 9.1.

2.

License Amendment Request 13-01.

3.

WCAP-1 7728-P, "Comanche Peak Nuclear Power Plant Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis", March 2013

4.

Deleted

5.

FSAR, Section 15.7.4.

6.

American Nuclear Society, "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," ANSI/ANS-8.1-1983, October 7, 1983.

to TXX-13045 Page 15 of 18 Insert 2, Revised Bases for TS 3.7.17 B 3.7 PLANT SYSTEMS B 3.7.17 Spent Fuel Assembly Storage BASES BACKGROUND A common Fuel Building houses facilities for storage and transfer of new and spent fuel. Two pools are provided for CPNPP spent fuel storage.

Each pool may be used to store fuel from either or both of the CPNPP units.

In the Region II rack (References 1 and 2) design, the spent fuel storage pool numbers 1 and 2 (SFP1 and SFP2) permit five different configurations (as shown in Figure 3.7.17-1). Region II racks, with 1462 and 1470 storage positions in SFP1 and SFP2 respectively (2932 total),

are designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups and decay times as shown in Figure 3.7.17-1.

Region I racks (References I and 2) with 222 and 219 storage positions located in SFP1 and SFP2 respectively (441 total) constitute a sixth configuration within the pools. These Region I racks are designed to accommodate new fuel with a maximum enrichment of 5.0 w/t % U-235 or spent fuel regardless of the discharge fuel bumup. There are no storage pattern restrictions associated with the Region I racks. The neutron absorber material Boral is credited for the storage of spent fuel assemblies within the Region I racks to maintain kef less than or equal to 1.0 at 0 ppm boron concentration.

A discussion of how soluble boron is credited for the storage of spent fuel assemblies is contained in the BACKGROUND for B 3.7.16.

Within the SFP1 Region II racks, there exist two oversized (2x2) cells.

Within the SFP2 Region I racks, there exists one oversized (2x2) cell.

These oversized cells are not approved for storage of either fresh or spent fuel. However, they can be used as a place in the pool for a single assembly to be lowered and raised while being inspected. Prior to use of to TXX-13045 Page 16 of 18 the inspection cells certain prerequisites must be met. Criticality analyses (Reference 3) have been performed which demonstrate that there is no increase in reactivity relative to the approved Region II storage configurations (the current licensing basis requirements for the spent fuel pool are still met) provided that administrative prerequisites are maintained for the oversized cells in SFP1 Region II racks. The prerequisite for the use of the oversized cells in Region II racks is that all the Region II cells in the first row surrounding the oversized cell remain free of fuel assemblies. This prerequisite applies to a total of 8 Region II cells adjacent to the oversized cell in the SFP I Region II rack adjacent to the Region I rack and a total of 5 Region II cells adjacent to the oversized cell in the SFP1 Region II racks adjacent to the spent fuel pool walls.

There are no prerequisites for the use of the oversized cell in SFP2 Region I racks since the criticality analyses (Reference 3) demonstrate there is no increase in reactivity relative to the approved Region I storage configuration.

APPLICABLE SAFETY ANALYSIS A discussion of the criticality analysis for the storage of spent fuel assemblies is contained in the APPLICABLE SAFETY ANALYSES for B 3.7.16.

Most fuel storage pool accident conditions will not result in a significant increase in keff. Examples of such accidents are the drop of a fuel assembly on top of a rack, and the drop of a fuel assembly outside but adjacent to the rack modules. However, accidents can be postulated for each rack storage configuration which could increase reactivity beyond the analyzed condition. A discussion of these accidents is contained in B 3.7.16.

By closely controlling the movement of each assembly and by checking the location of each assembly after movement, the time period for potential accidents may be limited to a small fraction of the total operating time.

The configuration of fuel assemblies in the fuel storage pool satisfies Criterion 2 of 10CFR50.36(c)(2)(ii).

to TXX-13045 Page 17 of 18 LCO The restrictions on the placement of fuel assemblies within the spent fuel pool, in accordance with Figure 3.7.17-1, in the accompanying LCO, ensures the kef of the spent fuel storage pool will always remain < 0.95, assuming the pool to be flooded with borated water.

NOTE: The oversized inspection cells within the racks are not approved storage locations and are not covered by the LCO. Administrative controls which govern the use of the inspections cells are described in the BACKGROUND.

APPLICABILITY This LCO applies whenever any fuel assembly is stored in Region II racks of the fuel storage pool.

ACTIONS A.1 When the configuration of fuel assemblies stored in Region II racks of the spent fuel storage pool is not in accordance with Figure 3.7.17-1, the immediate action is to initiate action to make the necessary fuel assembly movement(s) to bring the configuration into compliance with Figure 3.7.17-1.

Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. If unable to move irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the action is independent of reactor operation. Therefore, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE REQUIREMENTS SR 3.7.17.1 This SR applies to fuel movement plans as well as the final storage configuration, since it is possible to create and resolve a non-compliant storage configuration in a single fuel movement sequence. The SR verifies, by administrative means, compliance with Figure 3.7.17-1 in the accompanying LCO.

to TXX-13045 Page 18 of 18 The surveillance frequency requirement, to perform prior to moving a fuel assembly into any Region II storage location, ensures that all fuel movement which could impact the acceptability of TS 3.7.17 is properly reviewed.

REFERENCES

1.

FSAR Section 9.1.

2.

License Amendment Request 13-01.

3.

WCAP-1 7728-P, "Comanche Peak Nuclear Power Plant Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis", March 2013

ATTACHMENT 4 to TXX-13045 RETYPED TECHNICAL SPECIFICATION PAGES Pages 3.7-36 3.7-37 3.7-38 3.7-39 3.7-40 3.7-41 3.7-42 3.7-43 3.7-44 3.7-45 3.7-46 3.7-47 3.7-48 3.7-49 4.0-2 4.0-3 5.5-17 5.5-18 to TXX-13045 page 1 of 18 Fuel Storage Pool Boron Concentration 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Fuel Storage Pool Boron Concentration LCO 3.7.16 APPLICABILITY:

The fuel storage pool boron concentration shall be _> 2400 ppm.

When fuel assemblies are stored in the fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool boron NOTE -----------------------

concentration not within LCO 3.0.3 is not applicable.

lim it.

A.1 Suspend movement of fuel Immediately assemblies in the fuel storage pool AND A.2 Initiate action to restore fuel storage Immediately pool boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify the fuel storage pool boron concentration is within In accordance with

limit, the Surveillance Frequency Control Program.

COMANCHE PEAK - UNITS 1 AND 2 3.7-36 Amendment No. 450-, 4We-,

to TXX-13045 page 2 of 18 Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage LCO 3.7.17 APPLICABILITY:

New or spent fuel assemblies will be stored in compliance with Figure 3.7.17-1.

Whenever any fuel assembly is stored in Region II racks of the spent fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1

---

NOTE not met.

LCO 3.0.3 is not applicable.

Initiate action to move fuel as Immediately necessary to restore compliance.

COMANCHE PEAK - UNITS 1 AND 2 3.7-37 Amendment No. 4EQ-, 466-,

to TXX-13045 page 3 of 18 Spent Fuel Assembly Storage 3.7.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the acceptability of fuel Prior to moving a movement plans and the resulting storage configuration in fuel assembly into accordance with Figure 3.7.17-1.

any Region II storage location.

COMANCHE PEAK - UNITS 1 AND 2 3.7-38 Amendment No. 4W-, 4W6-,

to TXX-1 3045 page 4 of 18 Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-1 (page 1 of 3)

Spent Fuel Pool Loading Restrictions All 2x2 Region II storage cell arrays shall comply with one of the Arrays Definitions below. Each storage location is a corner location for up to 4 separate 2x2 arrays.

A.

Arrays Il-A through II-E designate the pattern of fuel which may be stored in any 2x2 Array, and are dependent upon Fuel Category.

B.

Fuel Categories 1-6 are determined based on Fuel Burnup, Initial Enrichment, Decay Time, and Fuel Group.

C.

Fuel Group F1 assemblies have a nominal rod outer diameter of 0.374 inches. Fuel Group F2 assemblies have a nominal rod outer diameter of 0.360 inches.

Array Definition Illustration Array Il-A W

6 6

Array Category 6 assembly in every cell. Only valid for two rows adjacent to the A

Il-B SFP wall. The two rows adjacent to Array II-A must be Array Il-B, and the L

empty cell in Array Il-B must be adjacent to Array Il-A.

L 6

6 Array 1I-B 4

4 Cateory 4 fuel assembly in 3 out of 4 cells, with empty cell in the fourth cell.

X 4

Array Il-C 5

3 Pattern which contains fuel in 3 out of 4 cells, including two diagonally-opposed Category 5 assemblies, one Category 3 assembly, and one empty location. Only Fuel Group F2 X

5 assemblies may be stored in Array Il-C.

Array 11-D 2

x Checkerboard pattern of two diagonally-opposed Category 2 assemblies with two diagonally-opposed empty cells.

X 2

COMANCHE PEAK - UNITS 1 AND 2 3.7-39 Amendment No. 464-, 46&,

to TXX-13045 page 5 of 18 Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-1 (page 2 of 3)

Notes:

1.

Fuel Categories are ranked in order of relative reactivity, from Category 1 to 6. Fuel Category 1 assemblies have the highest reactivity, and Fuel Category 6 assemblies have the lowest.

2.

All Fresh Fuel Assemblies (assemblies with a burnup value of 0.0 MWD/MTU) should be considered Category 1 fuel, independent of Fuel Group or Enrichment.

a.

In Fuel Group F1, Fuel Category 1 is fresh fuel up to 3.5 weight percent U-235 Initial Enrichment.

b.

In Fuel Group F2, Fuel Category 1 is fresh fuel up to 5.0 weight percent U-235 Initial Enrichment.

3.

Fuel Category 2 is any Non-Fresh fuel assembly up to 3.5 weight percent U-235 Initial Enrichment (Burnup Requirement is > 0 MWD/MTU).

4.

For all other fuel, Fuel Categories are determined as follows:

a.

For Initial Enrichment values below the Minimum Applicable Initial Enrichment values of Table 3.7.17-1, the Fitting Coefficients of Tables 3.7.17-2 and 3.7.17-3 are not applicable.

The Minimum Burnup Requirement for the corresponding Category is > 0 MWD/MTU.

b.

For Fuel Group F1 assemblies, determine the Fitting Coefficients A1 - A4 using Table 3.7.17-2. Note that Table 3.7.17-2 is only applicable to fuel with > 10 years of decay time, and an Initial Enrichment of < 3.5 weight percent.

c.

For Fuel Group F2 assemblies, determine the Fitting Coefficients A1 - A4 using Table 3.7.17-3.

d.

The required Minimum Burnup value (in MWD/MTU) for each Fuel Category is calculated based on Initial Enrichment (En) and the appropriate fitting coefficients, using the equation below. If the fuel assembly burnup is greater than or equal to the calculated Minimum Burnup value, then the fuel may be classified into this Fuel Category.

Minimum Burnup = 1,000 x [A1 x En 3 + A2 x En 2 + A3 x En + A4]

e.

All relevant uncertainties are explicitly included in the criticality analysis. No additional allowance for burnup uncertainty or initial enrichment uncertainty is required.

COMANCHE PEAK - UNITS 1 AND 2 3.7-40 Amendment No. 460-, 46&,

to TXX-1 3045 page 6 of 18 Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-1 (page 3 of 3)

Notes (continued):

f.

Conservatively low values of Decay Time may be used to calculate the Minimum Burnup value, or interpolation may be used. If interpolation is used, Minimum Burnup values for tabulated Decay Time values above and below the actual value should first be determined. Next, linear interpolation between these values may be used to determine the Minimum Burnup value. No extrapolation beyond 20 years is permitted.

g.

Initial Enrichment (En) is the nominal U-235 weight percent enrichment of the central zone region of fuel, excluding axial blankets, prior to fuel depletion.

h.

If the computed Minimum Burnup value < 0 MWD/MTU, the Minimum Burnup Requirement is > 0 MWD/MTU.

5.

In all Arrays, an assembly with a higher Fuel Category number can be utilized in place of any fuel assembly with a lower Fuel Category Number, with the following exception.

a.

Fuel Group F1 assemblies are not allowed to be stored in Array II-C, regardless of Fuel Category.

6.

An empty (water-filled) cell can be substituted for any fuel-containing cell in all storage arrays.

7.

Any storage array location designated for a fuel assembly can be replaced with non-fissile hardware. Items other than Fuel Assemblies which contain fissile material are restricted to storage in Region I.

8.

Fuel assembly inserts approved for use in the reactor core can be inserted in a stored assembly (in the Spent Fuel Pool) without affecting the fuel category.

COMANCHE PEAK - UNITS 1 AND 2 3.7-41 Amendment No. 46Q, 4W to TXX-13045 page 7 of 18 Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-1 Minimum Applicable Initial Enrichment for Table 3.7.17-2 and Table 3.7.17-3 Fitting Coefficients Fuel Category Fuel Group F1 Fuel Group F2 6

1.25 1.20 5

N/A 1.30 4

1.35 1.45 3

N/A 1.45 2

N/A 3.55 Table 3.7.17-2 Fuel Group F1 Nominal Fuel Rod Outer Diameter of 0.374" Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Decay Time and Initial Enrichment (En)

Fuel Decay Category (yrs)

Fitting Coefficients A1 A2 A3 A4 6

10 1.4351

-17.3247 73.3805

-67.4585 6

15 1.7078

-18.7916 74.6322

-67.2637 6

20 0.5289

-9.9969 53.7741

-52.6302 4

10

-0.0444

-1.3474 22.7039

-28.0852 4

15 0.2015

-2.6257 24.1016

-28.2473 4

20 0.4646

-4.1432 26.3891

-29.2170 Note: Fuel must have at least 10 years of decay time and less than or equal to 3.5 weight percent Initial Enrichment to utilize Table 3.7.17-2 COMANCHE PEAK - UNITS 1 AND 2 3.7-42 Amendment No. 46ý9-, 4We-,

to TXX-13045 page 8 of 18 Spent Fuel Assembly Storage 3.7.17 Table 3.7.17-3 Fuel Group F2 Nominal Fuel Rod Outer Diameter of 0.360" Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Decay Time and Initial Enrichment (En)

Fuel Decay Category (yrs)

Fitting Coefficients A1 A2 A3 A4 6

0 0.5789

-7.4498 42.4056

-41.1591 6

5 0.5247

-6.8992 39.7676

-38.6927 6

10 0.2701

-4.4306 31.9841

-32.4674 6

15 0.3105

-4.5582 31.1825

-31.3916 6

20 0.2374

-3.8754 28.8900

-29.4975 5

0 0.9373

-11.2553 54.7226

-54.1769 5

5 0.6169

-8.1494 44.7801

-45.7968 5

10 0.5380

-7.1852 40.7044

-41.9545 5

15 0.5385

-7.0180 39.2299

-40.3213 5

20 0.5200

-6.7906 38.0244

-39.0979 4

0 0.2553

-3.9826 30.6152

-36.7967 4

5 0.2366

-3.6430 28.2160

-33.9749 4

10 0.4387

-5.6018 33.3609

-37.9327 4

15 0.5450

-6.6302 36.0760

-40.0315 4

20 0.6327

-7.4663 38.2724

-41.7257 3

0 0.5317

-6.1006 32.7118

-36.2263 3

5 0.5228

-5.9434 31.2846

-34.4602 3

10 0.5432

-6.1075 31.1578

-33.9933 3

15 0.5206

-5.8897 30.1768

-32.9600 3

20 0.5158

-5.7796 29.4050

-32.0577 2

0 0.0000 1.6738

-8.5396 9.2206 COMANCHE PEAK-UNITS 1 AND 2 3.7-43 Amendment No. 460-, 456-,

to TXX-13045 page 9 of 18 Secondary Specific Activity 3.7.18 3.7 PLANT SYSTEMS 3.7.18 Secondary Specific Activity LCO 3.7.18 APPLICABILITY:

The specific activity of the secondary coolant shall be <! 0.10 pCi/gm DOSE EQUIVALENT 1-131 MODES 1, 2, 3, and 4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Specific activity not within A.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> limit.

AND A.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.18.1 Verify the specific activity of the secondary coolant is In accordance with

• 0.10 pCi/gm DOSE EQUIVALENT 1-131.

the Surveillance Frequency Control Program.

COMANCHE PEAK - UNITS 1 AND 2 3.7-44 Amendment No. 460-, 46&,

to TXX-13045 page 10 of 18 Safety Chilled Water 3.7.19 3.7 PLANT SYSTEMS 3.7.19 Safety Chilled Water LCO 3.7.19 APPLICABILITY:

Two safety chilled water trains shall be OPERABLE MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One safety chilled water A.1 Restore safety chilled water train to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> train inoperable.

OPERABLE status.

B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> COMANCHE PEAK - UNITS 1 AND 2 3.7-45 Amendment No. 45Gý 4WG-,

to TXX-13045 page 11 of 18 Safety Chilled Water 3.7.19 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i

SR 3.7.19.1 NOTE Isolation of safety chilled water flow to individual components does not render the safety chilled water system inoperable.

Verify each safety chilled water manual, power operated, and automatic valve servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.

In accordance with the Surveillance Frequency Control Program.

SR 3.7.19.2 Verify each safety chilled water pump and chiller starts on In accordance with an actual or simulated actuation signal.

the Surveillance Frequency Control Program.

COMANCHE PEAK - UNITS 1 AND 2 3.7-46 Amendment No. 46G,- 4-5&,

to TXX-1 3045 page 12 of 18 UPS HVAC System 3.7.20 3.7 PLANT SYSTEMS 3.7.20 UPS HVAC System LCO 3.7.20 APPLICABILITY:

Two UPS HVAC System Trains shall be OPERABLE MODES 1, 2, 3, and 4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One UPS HVAC System A.1 Verify the affected UPS & Distribution Immediately train inoperable.

Room is supported by an OPERABLE UPS A/C Train.

AND A.2 Restore the inoperable UPS HVAC 30 days train to OPERABLE status.

B. Two UPS HVAC System B.1 Verify air circulation is maintained by Immediately trains inoperable, at least one UPS A/C Train.

OR AND Required Action A.1 and B.2 Verify the air temperature in the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion affected UPS & Distribution Room(s)

Time not met.

does not exceed the maximum AND temperature limit for the room(s).

Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter AND B.3 Restore UPS HVAC System train to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status.

COMANCHE PEAK - UNITS 1 AND 2 3.7-47 Amendment No. 4EQ-, 4We-,

to TXX-13045 page 13 of 18 UPS HVAC System 3.7.20 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action B.1 and C.1 Restore the required support.

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> associated Completion Time not met.

D. Required Action and D.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Required AND Action A.2, B.2, B.3 or C.1 not met.

D.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> COMANCHE PEAK - UNITS 1 AND 2 3.7-48 Amendment No. 4f)@-, 4WG-,

to TXX-13045 page 14 of 18 UPS HVAC System 3.7.20 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.20.1 Verify each required UPS & Distribution Room Fan Coil In accordance with Unit operates > 1 continuous hour.

the Surveillance Frequency Control Program.

SR 3.7.20.2 Verify each required UPS A/C train operates for > 1 In accordance with continuous hour.

the Surveillance Frequency Control Program.

SR 3.7.20.3 Verify each required UPS A/C train actuates on an actual or In accordance with simulated actuation signal.

the Surveillance Frequency Control Program.

COMANCHE PEAK - UNITS 1 AND 2 3.7-49 Amendment No. 4f)@-, 4-5&,

to TXX-13045 Design Features page 15 of 18 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b.

keff < 1.0 when fully flooded with unborated water which includes an allowance for uncertainties as described in Section 4.3 of the FSAR;

c.

keff < 0.95 if fully flooded with water borated to 400 ppm, which includes an allowance for uncertainties as described in Section 4.3 of the FSAR;

d.

A nominal 9 inch center to center distance between fuel storage locations in Region II fuel storage racks;

e.

A nominal 10.65 inch by nominal 11.05 inch center to center distance between fuel assemblies placed in Region I fuel storage racks;

f.

New or partially spent fuel assemblies may be allowed restricted storage in a I out of 4 configuration in Region II fuel storage racks (as shown in Figure 3.7.17, Array II-E) or unrestricted storage in Region I fuel storage racks;

g.

Storage of new or spent fuel assemblies in Region II storage racks must comply with 3.7.17 Spent Fuel Assembly Storage.

4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b.

keff < 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 4.3 of the FSAR;

c.

keff < 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in Section 4.3 of the FSAR; and COMANCHE PEAK - UNITS 1 AND 2 4-0-2 Amendment No. 469-,

to TXX-13045 page 16 of 18 Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued)

d.

A nominal 21 inch center to center distance between fuel assemblies placed in the storage racks.

4.3.2 Drainage The spent fuel storage pools are designed and shall be maintained to prevent inadvertent draining of the pool below elevation 854 ft.

4.3.3 Capacity The spent fuel storage pools are designed and shall be maintained with a storage capacity limited to no more than 3373 fuel assemblies.

COMANCHE PEAK - UNITS 1 AND 2 4.0-3 Amendment No. 46Q, to TXX-13045 Programs and Manuals page 17 of 18 5.5 5.5 Programs and Manuals 5.5.21 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a.

The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b.

Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI-04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c.

The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

5.5.22 Spent Fuel Storage Rack Neutron Absorber Monitoring Program The Region I storage cells in the CPNPP Spent Fuel Pool utilize the neutron absorbing material BORAL, which is credited in the Safety Analysis to ensure the limitations of Technical Specification 4.3.1.1 are maintained.

In order to ensure the reliability of the Neutron Poison material, a monitoring program is required to routinely confirm that the assumptions utilized in the criticality analysis remain valid and bounding. The Neutron Absorber Monitoring Program is established to monitor the integrity of neutron absorber test coupons periodically as described below.

A test coupon "tree" shall be maintained in each SFP. Each coupon tree originally contained 8 neutron absorber surveillance coupons. Detailed measurements were taken on each of these 16 coupons prior to installation, including weight, length, width, thickness at several measurement locations, and B-10 content (g/cm2). These coupons shall be maintained in the SFP to ensure they are exposed to the same environmental conditions as the neutron absorbers installed in the Region I storage cells, until they are removed for analysis.

One test coupon from each SFP shall be periodically removed and analyzed for potential degradation, per the following schedule. The schedule is established to ensure adequate coupons are available for the planned life of the storage racks.

COMANCHE PEAK - UNITS 1 AND 2 5.5-17 Amendment No. 46&,

to TXX-13045 page 18 of 18 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.22 Spent Fuel Storage Rack Neutron Absorber Monitoring Program (continued)

Year Coupon Number Year Coupon Number 2013 1

2028 5

2015 2

2033 6

2018 3

2043 7

2023 4

2053 8

Further evaluation of the absorber materials, including an investigation into the degradation and potential impacts on the Criticality Safety Analysis, is required if:

A decrease of more than 5% in B-10 content from the initial value is observed in any test coupon as determined by neutron attenuation.

An increase in thickness at any point is greater than 25% of the initial thickness at that point.

COMANCHE PEAK - UNITS 1 AND 2 5.5-18 Amendment No.

ATTACHMENT 5 to TXX-13045 RETYPED TECHNICAL SPECIFICATION BASES PAGES (For Information Only)

Pages B 3.7-72 B 3.7-73 B 3.7-74 B 3.7-75 B 3.7-76 B 3.7-77 B 3.7-78 B 3.7-79 to TXX-1 3045 Fuel Storage Pool Boron Concentration page 1 of 8 B 3.7.16 B 3.7 PLANT SYSTEMS B 3.7.16 Fuel Storage Pool Boron Concentration BASES BACKGROUND A common Fuel Building houses facilities for storage and transfer of new and spent fuel. Two pools are provided for CPNPP spent fuel storage. Each pool may be used to store fuel from either or both of the CPNPP units.

In the Region II rack (References 1 and 2) design, the spent fuel storage pool numbers 1 and 2 (SFP1 and SFP2) permit five different configurations (as shown in Figure 3.7.17-1). Region II racks, with 1462 and 1470 storage positions in SFP1 and SFP2 respectively (2932 total), are designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups and decay times as required by Figure 3.7.17-1.

Region I racks (References 1 and 2) with 222 and 219 storage positions located in SFP1 and SFP2 respectively (441 total), constitute a sixth configuration within the pools. These Region I racks are designed to accommodate new fuel with a maximum enrichment of 5.0 w/t % U-235 or spent fuel regardless of the discharge fuel burnup or decay time. There are no storage pattern restrictions associated with the Region I racks. The neutron absorber material Boral is credited for the storage of spent fuel assemblies within the Region I racks to maintain keff less than or equal to 1.0 at 0 ppm soluble boron concentration.

In order to maintain keff less than or equal to 0.95, the presence of fuel pool soluble boron is credited for the storage of fuel assemblies within the Region I and Region II racks. A description of how credit for fuel storage pool soluble boron is used under normal storage configuration conditions is found in Reference 3. The storage configurations are defined using calculations to ensure that keff will be less than 1.0 with no soluble boron under normal storage conditions including biases and uncertainties. Soluble boron credit is then used to maintain keff less than or equal to 0.95. Criticality analyses have been performed (Reference 3) which demonstrate that the pools require 400 ppm of soluble boron to maintain keff less than or equal to 0.95 for all allowed combinations of storage configurations, enrichments, burnups, and decay time limits. The effect of B-1 0 depletion on the boron concentration for maintaining keff less than or equal to 0.95 is accounted for in Reference 3.

Criticality analyses considering accident conditions have also been performed (Reference 3). These analyses establish the amount of soluble boron necessary to ensure that keff will be maintained less than or equal to (continued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.7-72 Revision to TXX-13045 Fuel Storage Pool Boron Concentration page 2 of 8 B 3.7.16 BASES BACKGROUND (continued) 0.95 should pool temperatures fall outside the assumed range or multiple fuel assembly misload events occur. The total amount of soluble boron required to mitigate these events is 2400 ppm.

For an occurrence of the above postulated accident condition, the double contingency principle of ANSI/ANS 8.1-1983 (Reference 6) can be applied.

This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for these postulated accident conditions, the presence of additional soluble boron in the storage pool water (above the concentration required for normal conditions, and up to the minimum value required by Technical Specifications) can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely and independent event.

A boron concentration equal to or greater than 2400 ppm assures that a dilution event which will result in a keff greater than 0.95 is not credible. This is demonstrated by a boron dilution analysis performed for the CPNPP Spent Fuel pools. This conclusion is based on the following: (1) a substantial amount of water is needed in order to dilute the SFP to the design keff of 0.95, (2) since such a large water volume turnover is required, a SFP dilution event would be readily detected by plant personnel via alarms, flooding in the fuel and auxiliary buildings or by normal operator rounds through the SFP area, and (3) evaluations indicate that, based on the flow rates of non-borated water normally available to the SFP, taken in conjunction with significant operator errors, and equipment failures, sufficient time is available to detect and respond to a dilution event. In addition, there is significant conservatism built into this evaluation; for example, the cooling of the spent fuel pools can be performed by one train supplying common water to both pools. This cooling configuration would allow credit of the volume of both pools and substantially increase the dilution time estimates presented.

However, because the flexibility exists for the cooling system to be totally dedicated to one pool, only one pool volume is considered in this evaluation.

It should be noted that this boron dilution evaluation considered the boron dilution volumes required to dilute the SFP from 1900 ppm to 800 ppm. The 800 ppm end point was utilized to ensure that keff for the spent fuel racks would remain less than or equal to 0.95. More recent analysis (Reference 3) demonstrates keff is maintained less than 0.95 with only 400 ppm of SFP boron, and therefore the dilution evaluation remains bounded. However, as discussed above, calculations for Region I and Region II configurations have been performed on a 95/95 basis to show that the spent fuel rack keff remains less than 1.0 with non-borated water in the pool. Thus, even if the SFP were diluted to concentrations approaching zero ppm, the fuel storage COMANCHE PEAK - UNITS 1 AND 2 B 3.7-73 Revision to TXX-13045 Fuel Storage Pool Boron Concentration page 3 of 8 B 3.7.16 BASES (continued)

BACKGROUND (continued) racks would remain subcritical and the health and safety of the public would be protected.

The storage of fuel with initial enrichments up to and including 5.0 weight percent U-235 in the Comanche Peak fuel storage pools has been evaluated. For the Region II storage racks, the resulting enrichment, burnup, and decay time limits for the pool are shown in Figure 3.7.17-1.

APPLICABLE Most fuel storage pool accident conditions will not result in a significant SAFETY ANALYSES increase in keff. Examples of such accidents are the drop of a fuel assembly on top of a rack, and the drop of a fuel assembly outside but adjacent to the rack modules.

A dropped assembly accident occurs when a fuel assembly is dropped onto the storage racks. The rack structure is not excessively deformed. An assembly, in its most reactive condition, is considered in the criticality evaluation. The dropped assembly, which comes to rest on top of the rack, has sufficient water separating it from the active fuel height of stored assemblies to preclude neutronic interaction. This is true even with unborated water. For the borated water condition, the potential for interaction is even less since the water contains boron which is an additional thermal neutron absorber.

However, five accidents can be postulated for each storage configuration that could increase reactivity beyond the analyzed condition. The first postulated accident would be a change in pool temperature to outside the range of normal operating temperatures assumed in the criticality analyses (500F to 1500F). The second accident would be dropping a fuel assembly into an already loaded cell. The third would be the misloading of a fuel assembly within the racks into a cell for which the restrictions on location, enrichment, burnup, or decay time are not satisfied. A forth would be the misload of a fuel assembly adjacent to but outside the racks. The fifth accident would be misloading of multiple fuel assemblies, in series, into unacceptable storage locations.

Variations in the temperature of the water passing through the stored fuel assemblies outside the normal operating range were considered in the criticality analysis. The reactivity effects of a temperature range from 320F to 212°F were evaluated. The increase in reactivity due to the change in temperature is bounded by the misloading accident.

(continued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.7-74 Revision to TXX-13045 Fuel Storage Pool Boron Concentration page 4 of 8 B 3.7.16 BASES APPLICABLE SAFETY ANALYSES (continued)

For the accident of dropping a fuel assembly into an already loaded cell, the upward axial leakage of that cell will be reduced; however, the overall effect on the rack reactivity will be insignificant. This is because minimizing the upward-only leakage of just a single cell will not cause any significant increase in reactivity. Furthermore, the neutronic coupling between the dropped assembly and the already loaded assembly will be low due to several inches of assembly nozzle structure which would separate the active fuel regions. Therefore, this accident would clearly be bounded by the misloading accident.

The fuel assembly misloading accident involves placement of a single fuel assembly in a location for which it does not meet the requirements for enrichment, burnup, or decay time including the placement of an assembly in a location that is required to be left empty. The result of the misloading is to add positive reactivity, increasing keff toward 0.95. The maximum required boron to compensate for this event is 1867 ppm, which is below the LCO limit of 2400 ppm.

The multiple fuel assembly misloading accident involves the misplacement of assemblies, in series, due to a common cause. The result of this misloading is the addition of positive reactivity, increasing the keff toward 0.95. The maximum required boron concentration to compensate for this event is 2400 ppm, which is the LCO limit.

The concentration of dissolved boron in the fuel storage pool satisfies Criterion 2 of the 10CFR50.36(c)(2)(ii).

LCO The fuel storage pool boron concentration is required to be > 2400 ppm. The specified concentration of dissolved boron in the fuel storage pool preserves the assumptions used in the analyses of the potential criticality accident scenarios as described in Reference 3. The amount of soluble boron required to offset each of the above postulated accidents was evaluated for all of the proposed storage configurations. The specified minimum boron concentration of 2400 ppm assures that the concentration will remain above these values.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel storage pool.

(continued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.7-75 Revision to TXX-13045 page 5 of 8 Fuel Storage Pool Boron Concentration B 3.7.16 BASES (continued)

ACTIONS A.1 and A.2 When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This action is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies. Prior to resuming movement of fuel assemblies, the concentration of boron must be restored. This requirement does not preclude movement of a fuel assembly to a safe position.

The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply. If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.16.1 REQUIREMENTS This SR verifies that the concentration of boron in the fuel storage pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES

1.

FSAR, Section 9.1.

2.

License Amendment Request 13-01.

3.

WCAP-17728-P, "Comanche Peak Nuclear Power Plant Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis", March 2013.

4.

Deleted.

5.

FSAR, Section 15.7.4.

6.

American Nuclear Society, "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," ANSI/ANS-8.1-1983, October 7, 1983.

COMANCHE PEAK - UNITS 1 AND 2 B 3.7-76 Revision to TXX-1 3045 Spent Fuel Assembly Storage page 6 of 8 B 3.7.17 B 3.7 PLANT SYSTEMS B 3.7.17 Spent Fuel Assembly Storage BASES BACKGROUND A common Fuel Building houses facilities for storage and transfer of new and spent fuel. Two pools are provided for CPNPP spent fuel storage. Each pool may be used to store fuel from either or both of the CPNPP units.

In the Region II rack (References 1 and 2) design, the spent fuel storage pool numbers 1 and 2 (SFP1 and SFP2) permit five different configurations (as shown in Figure 3.7.17-1). Region II racks, with 1462 and 1470 storage positions in SFP1 and SFP2 respectively (2932 total), are designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups and decay times as shown in Figure 3.7.17-1.

Region I racks (References 1 and 2) with 222 and 219 storage positions located in SFP1 and SFP2 respectively (441 total) constitute a sixth configuration within the pools. These Region I racks are designed to accommodate new fuel with a maximum enrichment of 5.0 w/t % U-235 or spent fuel regardless of the discharge fuel burnup. There are no storage pattern restrictions associated with the Region I racks. The neutron absorber material Boral is credited for the storage of spent fuel assemblies within the Region I racks to maintain keff less than or equal to 1.0 at 0 ppm boron concentration.

A discussion of how soluble boron is credited for the storage of spent fuel assemblies is contained in the BACKGROUND for B 3.7.16.

Within the SFP1 Region II racks, there exist two oversized (2x2) cells.

Within the SFP2 Region I racks, there exists one oversized (2x2) cell. These oversized cells are not approved for storage of either fresh or spent fuel.

However, they can be used as a place in the pool for a single assembly to be lowered and raised while being inspected. Prior to use of the inspection cells certain prerequisites must be met. Criticality analyses (Reference 3) have been performed which demonstrate that there is no increase in reactivity relative to the approved Region II storage configurations (the current licensing basis requirements for the spent fuel pool are still met) provided that administrative prerequisites are maintained for the oversized cells in SFP1 Region II racks. The prerequisite for the use of the oversized cells in Region II racks is that all the Region II cells in the first row surrounding the oversized cell remain free of fuel assemblies. This prerequisite applies to a total of 8 Region II cells adjacent to the oversized cell in the SFP I Region II rack adjacent to the Region I rack and a total of 5 Region II cells adjacent to the oversized cell in the SFP1 Region II racks adjacent to the spent fuel pool (continued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.7-77 Revision to TXX-13045 page 7 of 8 Spent Fuel Assembly Storage B 3.7.17 BASES BACKGROUND (continued) walls. There are no prerequisites for the use of the oversized cell in SFP2 Region I racks since the criticality analyses (Reference 3) demonstrate there is no increase in reactivity relative to the approved Region I storage configuration.

APPLICABLE SAFETY ANALYSES A discussion of the criticality analysis for the storage of spent fuel assemblies is contained in the APPLICABLE SAFETY ANALYSES for B 3.7.16.

Most fuel storage pool accident conditions will not result in a significant increase in keff. Examples of such accidents are the drop of a fuel assembly on top of a rack, and the drop of a fuel assembly outside but adjacent to the rack modules. However, accidents can be postulated for each rack storage configuration which could increase reactivity beyond the analyzed condition.

A discussion of these accidents is contained in B 3.7.16.

By closely controlling the movement of each assembly and by checking the location of each assembly after movement, the time period for potential accidents may be limited to a small fraction of the total operating time.

The configuration of fuel assemblies in the fuel storage pool satisfies Criterion 2 of 10CFR50.36(c)(2)(ii).

LCO The restrictions on the placement of fuel assemblies within the spent fuel pool, in accordance with Figure 3.7.17-1, in the accompanying LCO, ensures the keff of the spent fuel storage pool will always remain < 0.95, assuming the pool to be flooded with borated water.

NOTE: The oversized inspection cells within the racks are not approved storage locations and are not covered by the LCO. Administrative controls which govern the use of the inspections cells are described in the BACKGROUND.

APPLICABILITY This LCO applies whenever any fuel assembly is stored in Region II racks of the fuel storage pool.

(continued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.7-78 Revision to TXX-13045 page 8 of 8 Spent Fuel Assembly Storage B 3.7.17 BASES (continued)

ACTIONS A.1 When the configuration of fuel assemblies stored in Region II racks of the spent fuel storage pool is not in accordance with Figure 3.7.17-1, the immediate action is to initiate action to make the necessary fuel assembly movement(s) to bring the configuration into compliance with Figure 3.7.17-1.

Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. If unable to move irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the action is independent of reactor operation. Therefore, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE REQUIREMENTS SR 3.7.17.1 This SR applies to fuel movement plans as well as the final storage configuration, since it is possible to create and resolve a non-compliant storage configuration in a single fuel movement sequence. The SR verifies, by administrative means, compliance with Figure 3.7.17-1 in the accompanying LCO.

The surveillance frequency requirement, to perform prior to moving a fuel assembly into any Region II storage location, ensures that all fuel movement which could impact the acceptability of TS 3.7.17 is properly reviewed.

REFERENCES

1.

FSAR Section 9.1.

2.

License Amendment Request 13-01.

3.

WCAP-17728-P, "Comanche Peak Nuclear Power Plant Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis", March 2013.

COMANCHE PEAK - UNITS 1 AND 2 B 3.7-79 Revision

ENCLOSURE 1 to TXX-13045

Enclosure I to TXX-13045 Page 1 of 15 Comanche Peak Nuclear Power Plant Spent Fuel Pool (SFP) Configuration Controls

Purpose:

Ensure the SFP configuration and other applicable conditions evaluated in the nuclear criticality safety analysis (WCAP-17728-P, March 2013) remain bounding when compared to current fuel design and plant operating parameters. This document provides a description of the controls supporting the proposed Technical Specification 3.7.17, to ensure:

1. TS 3.7.17 compliance is maintained at all times. All fuel movement plans into Spent Fuel Pool Region II are prepared in a manner which ensures continual compliance with the proposed limitations of TS 3.7.17, including all intermediate steps during fuel movement.

2. A misloading event beyond the analyzed accident conditions is not credible. An error in the fuel move planning does not have the potential to result in a misloading accident which is not bounded by the Criticality Safety Analysis.

3. Assumptions related to Fuel Characterization and reactor operation remain valid. Conditions evaluated in the nuclear criticality safety analysis remain bounding, for both future fuel design changes (pre-irradiation fuel characterization) and future operating conditions (post-irradiation fuel characterization).

4. Neutron Absorber assumptions remain valid and bounding. Safety Analysis assumptions for Neutron Absorber panels utilized in Region I storage cells remain conservative when considering aging affects or abnormal degradation. The proposed TS program 5.5.22, "Spent Fuel Storage Rack Neutron Absorber Monitoring Program" incorporates these controls into the Technical Specifications programs.

Enclosure i to TXX-13045 Page 2 of 15 Section 1 - Ensure TS 3.7.17 compliance is maintained at all times License Amendment Request 13-01 includes a proposed revision to the limitations associated with TS 3.7.17. The proposed Region II configurations are more complex than the previous limits associated with SFP storage. Some examples of the increased complexity include:

Fuel Groups: Storage limitations are dependent upon two Fuel Groups (F1 and F2), which are differentiated based on fuel parameters (primarily rod outer diameter).

Fuel Category: Six Fuel Categories (1-6) are utilized, ranked from highest reactivity (category

1) to lowest reactivity (category 6).

o The Minimum Burnup for each Fuel Category is calculated using Fitting Coefficients, based on Fuel Group, Initial Enrichment, and Decay Time.

o The proposed TS 3.7.17 limits use tables of Fitting Coefficients in place of graphically displayed limit curves.

Storage Array: Six storage array definitions (five in Region II, plus the all-cell configuration in Region I) define the allowable storage pattern for a 2x2 array. The array definitions are more complex than the previous limits at Comanche Peak Nuclear Power Plant (CPNPP).

o Array I-A applies only to Region I, and is an all cell configuration. This is not different than the current limits, and this Array is not discussed in TS 3.7.17, which is applicable only to Region II.

o Arrays Il-A through II-E apply in Region II.

o Array Il-A is a "4 of 4" configuration which is limited to "wall adjacent" locations, and has specific interface requirements for the adjacent array.

o Array Il-C (one of the two "3 of 4" configurations) utilizes a pattern containing 2 Fuel Categories, with a higher reactivity assembly being surrounded by lower reactivity assemblies.

o Array II-C is limited to Fuel Group F2, which creates a requirement to ensure no F1 assemblies are stored within this Array.

In response to the increase in complexity, improved administrative controls ensure the proposed TS 3.7.17 limits remain satisfied at all times. The improvements in the administrative controls include both the implementation of a new Configuration Confirmation Software program, as well as procedural and process changes. These improvements ensure that the increased complexity does not result in an increased risk of a TS 3.7.17 non-compliance due to an error made during fuel movement planning.

Enclosure i to TXX-13045 Page 3 of 15 Configuration Confirmation Software Functionality A Quality Assurance controlled software program is proposed, which provides the functionality described below. The purpose of this software is to ensure that the proposed limitations of TS 3.7.17 are satisfied for any fuel movement plan.

This software interfaces with two other QA (Quality Assurance) software programs at Comanche Peak:

1.

ShuffleWorks-This commercial software is used at many nuclear power plants for configuration control of the Spent Fuel Pools, and is used to plan movement of fuel and fuel components. At CPNPP, ShuffleWorks is used for fuel movement planning and SFP configuration control, but does NOT perform verification of TS 3.7.17 limits.

A ShuffleWorks "Configuration File" represents a snapshot in time for the entire CPNPP fuel inventory (either actual or planned), only excluding fuel stored at the Dry Cask storage pad. This file contains the location for each fuel assembly and fuel insert at a single point in time, but does not contain detailed fuel information or movement history.

A ShuffleWorks "Sequence File" represents a planned fuel or fuel insert movement sequence. These files include step-by-step changes in the configuration in a specified order. As an example, a core offload Sequence File would contain 193 data lines, each one moving an assembly (with the insert it contains) from a specific core location to a specific pool location.

2.

TARPIT -This software, developed internally at CPNPP, is primarily used to verify the acceptability of Dry Cask Canister loading patterns. The software contains a QA controlled database with all of the necessary fuel information needed to support validation of the dry cask storage limitations. Since much of this information is also utilized to perform fuel categorization, this database is proposed to be used as the source of necessary fuel parameters. Examples include fuel data such as burnup, initial enrichment, and discharge date. An expansion of the database is proposed to include the Fuel Group association of each fuel assembly.

to TXX-13045 Page 4 of 1S Software Features:

The key features of the proposed Configuration Confirmation Software are as follows.

1. Categorize each assembly by Fuel Category (1 to 6).

a.

The software determines the Fuel Category for each fuel assembly in the ShuffleWorks database. This determination is performed by comparing data obtained directly from the TARPIT database for enrichment, discharge date, burnup, and Fuel Group to the burnup limits calculated per TS Figure 3.7.17-1.

b.

No input files need to be created to perform this determination, since the input data will be directly obtained from the TARPIT database. This eliminates the potential for errors which could be introduced when transferring the fuel parameter data or formatting input files.

c.

The software generates graphical comparisons of fuel parameters (burnup vs.

enrichment graphs), including the ability to compare the data to the applicable TS 3.7.17 limits or compare results to previous surveillance reports. This is intended to aid in the review process and help identify potential errors or anomalies in the fuel parameter database.

2.

Determine acceptability of a given SFP Configuration:

a.

A current or planned SFP Configuration (obtained directly from a ShuffleWorks Configuration File) will be evaluated by the software for compliance with the proposed TS 3.7.17 limits. This evaluation is proposed to include a comparison of each categorized fuel assembly and all surrounding assemblies to the limits of TS 3.7.17 for fuel category and interface requirements.

b. Color-coded Spent Fuel Pool maps are proposed to aid in the review of the TS Surveillance Report, or for fuel movement planning purposes.

3.

Review ShuffleWorks Sequence Files:

a.

Using the same data as used to validate a Configuration File, validate a ShuffleWorks Sequence File. This is proposed to be performed on a move-by-move basis, to ensure that potential violations are identified (even when the potential violation would be resolved prior to completion of the fuel moves).

b.

Generate a report demonstrating either acceptability, or noting unacceptable fuel movement steps and reasons for failures.

Enclosure I to TXX-1304S Page S of 15

4. Situations which will result in a FAILURE notification include, but are not limited to:

a.

A fuel assembly being stored in a configuration not allowed by TS 3.7.17, either temporarily (in a Sequence file) or in a Configuration file.

To determine acceptability of storage, each storage location is evaluated as a member of 4 separate 2x2 arrays, as shown in Figure 1. If any one of the 4 2x2 Arrays which contains the fuel is not an approved Storage Array per TS 3.7.17, the proposed software generates a failure alert.

Figure 1 - Example of Array Verification Scope for each assembly 1

2 3

4

b.

Interface requirements not being satisfied. Examples of interface requirement failures would include:

i. An Array II-A configuration being located in any location other than 2 outer rows, adjacent to the SFP Wall (this configuration is not allowed in the central area of Region II, or along the Region II/Region I boundary).

ii. A configuration adjacent to the Array II-A configuration which does not meet the interface requirements (any configuration other than Array Il-B is not allowed in the 2x2 array adjacent to the Array II-A, and the row adjacent to Array Il-A must contain an empty cell).

c.

Any necessary Fuel Information missing from the database (such as missing burnup/enrichment values, or not being assigned a Fuel Group, see Section 3 "Assumptions related to Fuel Characterization and reactor operation remain valid").

d.

Any input parameter being outside of the analyzed range (for example enrichments above 5% for Fuel Group F2, above 3.5% for Fuel Group F1, or Decay Time < 10 years for Fuel Group Fl).

e.

Multiple Items stored in the same location in a Configuration File, or a Sequence file which temporarily places more than one item into a location.

f.

Items in invalid locations for fuel storage (such as non-existent storage locations, into oversized inspection cells, etc).

to TXX-13045 Page 6 of 15

g.

A fuel assembly is stored or is moved into Region II which has been flagged as

'restricted' in the fuel parameter database (see Section 3, "Assumptions related to Fuel Characterization and reactor operation remain valid").

h.

Non-fuel items (such as trash baskets, etc) stored in cells which were assumed to be only filled with water in the analysis. Note that all non-fuel items stored in a fuel storage location are already required to be tracked in the ShuffleWorks program.

5.

The Configuration Confirmation Software generates proposed reports to support Surveillance Requirement (SR) 3.7.17.1.

a.

Surveillance Requirement SR 3.7.17.1 is required at a frequency of "Prior to moving a fuel assembly into any Region II storage location." Therefore, the supporting report will contain a validation of each fuel movement plan, including the acceptability of the starting Configuration File, Sequence File (or files), and the final Configuration File.

b. The proposed report will provide information to document compliance with the TS limits, and include file information for all applicable database and ShuffleWorks data files for QA control and traceability purposes.

6.

The proposed Configuration Confirmation Software and associated data files will be controlled per the CPNPP software quality assurance program, which addresses design documentation, cyber security, configuration control, and media/access controls. The CPNPP software quality assurance program requires:

a.

Software features independently tested to ensure accuracy and completeness, reliability, functionality, and ease of use prior to approval.

b. Station procedures control all input data files, and require independent review for any changes, including routine updates.

c.

Independent testing of any changes to the software prior to approval of the software revision.

to TXX-13045 Page 7 of 15 Proposed Procedural Controls and Limitations

1.

Applicable procedures for preparing, reviewing, and performing fuel moves will be revised.

a.

The revised Technical Specification limits and interface requirements will be fully incorporated into the procedure which performs the TS 3.7.17 Surveillance.

b.

Fuel movement planning procedures will be revised to reflect the Technical Specification changes and interface requirements. Requirements will be incorporated into the procedures for validating both fuel movements and final pool configurations, using the Configuration Confirmation Software (performance of SR 3.7.17.1 using methods other than QA-controlled software will not be permitted).

c.

Changes to the fuel move plan within Region II will require storing the assembly somewhere other than the Region II storage racks, or the change will require the generation a new fuel movement plan (including generation of a new SR 3.7.17.1 report and Surveillance). This includes fuel location changes as well as fuel move sequence changes.

d.

Fuel movement field procedures will incorporate necessary precautions and limitations (see Section 2).

e.

Following each fuel movement campaign, a visual verification of the as-loaded SFP storage configuration will be performed. This verification ensure that no errors were made during fuel movement which placed a fuel assembly in an unapproved location.

i. Note that this verification will not include verification of assembly IDs using an underwater camera, but will verify the pattern of fuel and water locations in the SFP.

ii. This is adequate to ensure that any misplaced fuel assembly is discovered in a timely fashion. The only potential misloading event which would be undetected by this method is a series of two independent misloading errors, in which the final locations were "swapped" without the error being discovered.

Enclosure I to TXX-13045 Page 8 of 15 iii.

Two independent fuel handling mistakes in which the final locations are swapped without detection is not a credible occurrence. Fuel handling procedures require that two individuals independently verify both the location specified by the fuel move plan and the actual physical location for placement of fuel, prior to lowering the assembly into the storage cell.

f.

See Figure 2 for a flowchart which summarizes the proposed fuel move planning process.

2.

Current CPNPP procedures require all fuel movement plans be prepared and independently reviewed by Qualified Core Performance Engineers (CPE), and approved by a Fuel Handling Qualified Senior Reactor Operator (SRO). Any changes to an in-progress fuel movement plan must be prepared by a Qualified CPE and approved by a Fuel Handling qualified SRO (note that these types of field changes will no longer be allowed in Region II as described above, but will continue to be allowed in Region I or other Item Control Areas).

These requirements will not change, but the requirements for obtaining these qualifications will be updated to reflect the revised Technical Specifications and associated procedure and software changes. Training will be provided to all currently qualified personnel in these positions related to the changes.

Enclosure I to TXX-13045 Page 9 of 15 Figure 2 - Fuel Movement Planning Process Chart Initiation of fuel movement planning which affects Region 11 (other than only removing* fuel from Region 11, such as for cask loading)J Generate color coded planning map for the Starting Configuration using software tools Plan fuel movements using ShuffleWorks, and save Sequence File(s) and planned final Configuration File

+-ý iDetermine and correct cause 1~~

I Yes Generate Draft Surveillance Report for SR 3.7.17.1.

This report will validate Sequence and Final Configuration, including all interface requirements, and identify any TS 3.7.17 failures which would result if the fuel movement plan were executed.

1 I

I No I

Finalize and Approve Surveillance Report for SR 3.7.17.1, which includes file information for both Configuration Files and Sequence Files.

I Lr Complete Review and Approval process for fuel movement sequences. This will involve ensuring the ShuffleWorks files from the TS Surveillance Report exactly match the Fuel Movement Plans.

j Perform fuel Movement per the approved Movement Plan.

If Alterations are required, they will NOT be permitted to move fuel to Region II.

Sequence Order changes will NOT be permitted.

I

[

Perform visual verification of the pattern of fuel in the affected Spent Fuel Storage Racks.

This will identify any potential errors made during fuel movement which involve misplacing an assembly.

I

Enclosure I to TMX-13045 Page 10 of 15 Section 2 - Ensure a misloading event beyond the analyzed accident conditions is not credible The criticality safety analysis which supports LAR-13-01 (WCAP-17728-P) analyzed several misloading cases in the accident analysis, including multiple misload events, and concluded that keff would remain less than 0.95, including biases and uncertainties, with SFP Boron concentration at the minimum Technical Specification Boron value. The Multiple Assembly Misload analysis is described in WCAP-17728-P Section 5.7.

Other accidents, including single assembly misload accidents, dropped assemblies, seismic events, and SFP Temperature excursions, were also evaluated and determined to be acceptable when evaluated at a SFP Boron value of 1867 ppm.

The following Administrative Controls, in addition to the fuel movement planning controls described in Section 1, ensure that an error introduced during fuel move planning could not result in a multiple misloading accident which is not bounded by the accident analysis:

1.

Limitations will be implemented into Fuel Handling field procedures to ensure that any Fresh Fuel Assemblies (which are visually distinct from irradiated fuel assemblies) which are placed into Region II are ONLY placed into a cell which contains empty water cells in all adjacent surrounding Region II cells, including diagonally adjacent locations. This limitation is reflected in TS Figure 3.7.17-1, note 2, and is conservative relative to the supporting Criticality Analysis.

2.

Limitations will be implemented into Fuel Handling field procedures to ensure that if a fuel move will result in the creation of a 2x2 array which contains 4 fuel assemblies, this 2x2 array must be located adjacent to a SFP wall. This is also reflected in TS Figure 3.7.17-1, in the definition of Array II-A.

3.

If these visually verifiable limitations are not satisfied, the Fuel Handling personnel will be procedurally required to stop fuel movement prior to placing the fuel assembly into the planned storage location, regardless of the instructions provided on the fuel move plan.

The analyzed multiple misload cases, and other potential misload events which were not analyzed, are not considered credible accidents when considering the Administrative Controls in place to prevent such accidents. Based on the Double Contingency Principle, a condition may be considered "not credible" if two unlikely, independent, and concurrent incidences would need to occur to achieve this condition. In this case, both a significant error during planning, and multiple violations of fuel handling procedures during fuel movement would be required to achieve a condition matching the limiting analyzed case, or an unanalyzed condition. Both of these incidences are unlikely and are independent.

Further evidence supporting the determination that both the analyzed multiple misload cases and other unanalyzed multiple misload cases are not credible accidents, is provided by examining the CPNPP fuel inventory. Based on the current CPNPP inventory (see Figure 3) and CPNPP operational strategy, it is not credible to assume that Region II would be used to store a significant number of

Encosure I to TXX-13045 Page 11 of 15 assemblies of equivalent reactivity to that assumed in the limiting accident case. Based on the fact that the average fuel assembly burnup following the first cycle of operation (for modern cycles) is over 20,000 MWD/MTU, it is highly unlikely that the CPNPP fuel inventory would ever include a significant number of non-fresh fuel with such a high level of reactivity, or with a level of reactivity close to the reactivity of a fresh fuel assembly.

Figure 3 - Spent Fuel Pool Inventory CPNPP SFP Inventory of Fuel Assemblies March 2013 60000

. Fuel Group F1 50000

• Fuel Group F2 S

~40000 U

30000

___W E

20000.

U i

10000 U

0 0

1 2

3 4

Initial Enrichment (%)

5 6

Although it is demonstrated that the analyzed multiple misload accident cases are not credible, the acceptability demonstrated for these scenarios bounds multiple misload accidents which are considered credible, such as multiple once-burned fuel assemblies misplaced into a non-compliant configuration.

The Multiple Misload accident cases were calculated assuming 2400 ppm soluble boron in the SFP, the minimum value required by the proposed Technical Specifications. Other accidents, including Single Misload events and Temperature Excursions, were performed at 1867 ppm. A significant SFP dilution accident which would reduce the SFP boron value below the Technical Specification requirement is also unlikely, and the cause would be independent from fuel move planning or handling errors.

Enclosure I to TXX-13045 Page 12 of 15 To summarize, any potential errors made during the fuel movement planning process, or independent errors made during fuel move planning, will not result in a misloading accident which is not bounded by the Criticality Safety Analysis. These accidents are prevented by independent visual verifications required during fuel movement. Two unlikely, independent, and concurrent incidences would need to occur to achieve a condition beyond the analyzed cases (first a significant error during planning, followed by a violation of fuel handling limitations or a significant SFP dilution event). Fuel handling errors which result in a misload event will be detected in a timely fashion due to required post-fuel move reviews of the storage configuration pattern. Therefore the resulting misloaded configuration is not considered a credible accident per the Double Contingency Principle.

to TXX-13045 Page 13 of 15 Section 3 - Assumptions related to Fuel Characterization and reactor operation remain valid WCAP-17728-P Section 6.2 details the Analysis Area of Applicability, which summarizes the data which needs to be confirmed for each cycle of operation to assure that the results of the Criticality Safety Analysis remain valid. Each parameter in this section will be captured as a limitation in applicable CPNPP procedures. Certain parameters will be verified during the core design process, such as fuel design parameters. Other parameters which are dependent upon actual operational history will be verified following fuel depletion.

If a fuel parameter falls outside of the Analysis Area of Applicability, its storage will be restricted. If the fuel parameter is only utilized in the fuel depletion calculations (such as burnable poisons, RCS boron, etc), this restriction will only apply to Region II, and the assembly can be stored in Region I.

Other parameters, such as fuel lattice, rod OD, and fuel material, will result in a Region I storage restriction as well. An evaluation will be required prior to removing this restriction, including a 10 CFR 50.59 review to determine whether NRC review and approval is required.

The procedure for adding new fuel assemblies to the TARPIT databases will be updated to incorporate definitions for classifying fuel assemblies by the appropriate Fuel Group, and will include all key assumptions related fuel design, such as pellet density, fuel diameter, and axial blanket characterization, etc. Fuel which does not satisfy all of these requirements will not be classified into either of the Fuel Groups utilized to perform the Technical Specification surveillances, therefore adding an additional software barrier to ensure these assemblies are not stored in Region II storage racks.

Following each cycle, the operating history will be reviewed to ensure assumptions related to depletion history were met. For fuel which has operated outside the bounds of the analysis, restrictions will be added to the fuel classification database, adding a software barrier to ensure these assemblies are not stored in Region II storage racks.

Enclosure I to TXX-13045 Page 14 of 15 Section 4 - Neutron Absorber assumptions remain valid and bounding Region I storage cells at CPNPP contain the neutron absorbing material BORAL (BORAL), and the criticality analysis takes credit for these absorbers to ensure the regulatory limits are maintained.

Although the Region II storage cells were originally designed to use the BORAFLEX material, all of the Region II neutron absorber material removed prior to installation of the racks.

Results of test programs and industry experience have demonstrated that BORAL neutron absorber panels used in fuel storage racks similar to CPNPP Region I racks have performed satisfactory for reactivity control, and it is expected that the panels will fulfill their design function over the lifetime of the racks. However, due to the potential for an unknown degradation method, it is appropriate to establish a monitoring program to routinely confirm that the assumptions utilized in the criticality analysis remain valid and bounding.

CPNPP will establish a BORAL coupon surveillance program to monitor the integrity and performance of neutron absorbers on a continuing basis. The requirements for this program are captured in the proposed changes to Technical Specifications as Section 5.5.22. The goal of this program will be to assure that slow, long-term processes do not result in significant absorber degradation. The surveillance program will be capable of detecting the onset of any significant degradation with ample time to take corrective action as necessary.

At the time the Region I racks were installed in the SFP, two test coupon "trees" were also placed in the SFP, one tree in each SFP. Each tree contains 8 neutron absorber surveillance coupons. Detailed measurements were taken on each of these 16 coupons prior to installation, including weight, length, width, thickness at several measurement locations, and B-10 content (g/cm 2). These coupons have endured the same SFP environment conditions as the actual installed neutron absorbers for chemistry and temperature since installation of the Region I racks.

CPNPP will establish a surveillance program which will involve periodically removing one of the test coupons from each SFP and having it analyzed for any potential degradation. The surveillance program will be established to ensure adequate coupons are available for testing for 40 years. This schedule allows the ability to perform testing for the duration of the current CPNPP license, and to support a potential 20 year plant life extension.

The Criticality Analysis (WCAP-17728-P) assumed very conservative and bounding conditions for the BORAL panels. The Neutron Absorber Loading was assumed to be below the minimum value of the as-built material. The analysis also performed a study which demonstrated that blistering or growth on the neutron absorbers would not challenge the acceptance criteria.

The acceptance criteria for the surveillance program will be established conservatively to ensure not only that the analytical assumptions will not be challenged, but to ensure that no unexpected degradation is occurring which would warrant further investigation and study.

Enclosure I to TXX-13045 Page 15 of 15 Further evaluation of the absorber materials, including an investigation into the degradation and potential impacts on the Criticality Analysis, will be required if:

A decrease of more than 5% in B-10 content is observed in any test coupon as determined by neutron attenuation.

An increase in thickness at any point is greater than 25% of the initial thickness at that point.