ML070180741

Issuance of Amendment No. 200 to Facility Operating License Re ORIGEN-ARP Methodology for Calculating Fuel Pool Decay Heat Load

ML070180741
Person / Time
Site:	Columbia
Issue date:	02/08/2007
From:	Lyon C NRC/NRR/ADRO/DORL/LPLIV
To:	Parrish J Energy Northwest
Lyon C Fred, NRR/DORL/LPL4, 301-415-2296
Shared Package
ML070180448	List: ML070180741
References
TAC MD1227
Download: ML070180741 (11)

Text

February 8, 2007 Mr. J. V. Parrish Chief Executive Officer Energy Northwest P.O. Box 968 (Mail Drop 1023)

Richland, WA 99352-0968

SUBJECT:

COLUMBIA GENERATING STATION - ISSUANCE OF AMENDMENT RE: ADOPTION OF ORIGEN-ARP METHODOLOGY FOR CALCULATING FUEL POOL DECAY HEAT LOAD (TAC NO. MD1227)

Dear Mr. Parrish:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 200 to Facility Operating License No. NPF-21 for the Columbia Generating Station. The amendment consists of changes to the Final Safety Analysis Report in response to your application dated April 17, 2006.

This amendment changes the method for calculating fuel pool decay heat load from the original licensing basis methodology of ORIGEN to ORIGEN-ARP.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely,

/RA/

Carl F. Lyon, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-397

Enclosures:

1. Amendment No. 200 to NPF-21

2. Safety Evaluation cc w/encls: See next page

ML070180741 *Previously concurred OFFICE NRR/LPL4/PM NRR/LPL4/LA NRR/SBPB OGC-NLO NRR/LPL4/BC NAME FLyon LFeizollahi* JSegala AHodgdon DTerao DATE 1/24/07 1/23/07 1/29/07 2/2/07 2/6/07 Columbia Generating Station cc:

Mr. W. Scott Oxenford (Mail Drop PE04) Mr. Dale K. Atkinson (Mail Drop PE08)

Vice President, Technical Services Vice President, Nuclear Generation Energy Northwest Energy Northwest P.O. Box 968 P.O. Box 968 Richland, WA 99352-0968 Richland, WA 99352-0968 Mr. Albert E. Mouncer (Mail Drop PE01) Mr. William A. Horin, Esq.

Vice President, Corporate Services/ Winston & Strawn General Counsel/CFO 1700 K Street, N.W.

Energy Northwest Washington, DC 20006-3817 P.O. Box 968 Richland, WA 99352-0968 Mr. Matt Steuerwalt Executive Policy Division Chairman Office of the Governor Energy Facility Site Evaluation Council P.O. Box 43113 P.O. Box 43172 Olympia, WA 98504-3113 Olympia, WA 98504-3172 Ms. Lynn Albin Mr. Douglas W. Coleman (Mail Drop PE20) Washington State Department of Health Manager, Regulatory Programs P.O. Box 7827 Energy Northwest Olympia, WA 98504-7827 P.O. Box 968 Richland, WA 99352-0968 Technical Services Branch Chief FEMA Region X Mr. Gregory V. Cullen (Mail Drop PE20) 130 228th Street, S.W.

Supervisor, Licensing Bothell, WA 98201-9796 Energy Northwest P.O. Box 968 Ms. Cheryl M. Whitcomb (Mail Drop PE03)

Richland, WA 99352-0968 Vice President, Organizational Performance & Staffing/CKO Regional Administrator, Region IV Energy Northwest U.S. Nuclear Regulatory Commission P.O. Box 968 611 Ryan Plaza Drive, Suite 400 Richland, WA 99352-0968 Arlington, TX 76011-4005 Assistant Director Chairman Nuclear Safety and Energy Siting Division Benton County Board of Commissioners Oregon Department of Energy P.O. Box 190 625 Marion Street, NE Prosser, WA 99350-0190 Salem, OR 97301-3742 Senior Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 69 Richland, WA 99352-0069 August 2006

ENERGY NORTHWEST DOCKET NO. 50-397 COLUMBIA GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 200 License No. NPF-21

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Energy Northwest (licensee), dated April 17, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended to authorize changes to the Final Safety Analysis Report to reflect the method for calculating fuel pool decay heat load from the original licensing basis methodology of ORIGEN to ORIGEN-ARP, as described in the licensees application dated April 17, 2006.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of the date of issuance. In addition, the licensee shall include the revised information in the next Final Safety Analysis Report update submitted to the NRC in accordance with 10 CFR 50.71(e), as described in the licensees application dated April 17, 2006, and evaluated in the staffs safety evaluation enclosed with this amendment.

FOR THE NUCLEAR REGULATORY COMMISSION:

/RA/

David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Page 3 of the Facility Operating License Date of Issuance: February 8, 2007

ATTACHMENT TO LICENSE AMENDMENT NO. 200 FACILITY OPERATING LICENSE NO. NPF-21 DOCKET NO. 50-397 Replace the following page of the Facility Operating License No. NPF-21 with the attached revised page. The revised page is identified by amendment number and contains a vertical line indicating the area of change.

Facility Operating License REMOVE INSERT (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

(6) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to store byproduct, source and special nuclear materials not intended for use at Columbia Generating Station. The materials shall be no more than 9 sealed neutron radiation sources designed for insertion into pressurized water reactors and no more than 40 sealed beta radiation sources designed for use in area radiation monitors. The total inventory shall not exceed 24 microcuries of strontium-90, 20 microcuries of uranium-235, 30 curies of plutonium-238, and 3 curies of americium-241.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3486 megawatts thermal). Items in Attachment 1 shall be completed as specified. Attachment 1 is hereby incorporated into this license.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 200 and the Environmental Protection Plan contained in l Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

a. For Surveillance Requirements (SRs) not previously performed by existing SRs or other plant tests, the requirement will be considered met on the implementation date and the next required test will be at the interval specified in the Technical Specifications as revised in Amendment No. 149.

Amendment No. 199, 200

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 200 TO FACILITY OPERATING LICENSE NO. NPF-21 TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397

1.0 INTRODUCTION

By letter dated April 17, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML061110153), Energy Northwest (the licensee) requested U.S.

Nuclear Regulatory Commission (NRC) approval to allow the use of the ORIGEN-ARP code, instead of the ORIGEN code, when calculating fuel pool decay heat load. The proposed amendment to the licensing basis will resolve a nonconformance associated with this methodology that currently exists in the Final Safety Analysis Report (FSAR) Section 9.1.3, Spent Fuel Pool Cooling and Cleanup System, for Columbia Generating Station (Columbia).

Currently, Columbias fuel pool heat load values are calculated using ORIGEN, and are based on estimated refueling data. As described in the FSAR, refueling specific analyses are done to validate that the maximum fuel pool temperatures are within the licensing basis acceptance criteria. In its April 17, 2006, submittal, the licensee stated that the original ORIGEN code is no longer supported by the industry and has been replaced by ORIGEN-ARP, and requested that ORIGEN be replaced by ORIGEN-ARP for calculating spent fuel pool (SFP) decay heat loads at Columbia. The licensee stated that the use of ORIGEN-ARP will facilitate future compliance with the current FSAR SFP temperature limits by improving the accuracy and ease of performing future outage specific calculations.

2.0 REGULATORY EVALUATION

Part 50, Appendix A of Title 10 of the Code of Federal Regulations (10 CFR) General Design Criterion (GDC) 61 for fuel storage and handling and radioactivity control, specifies, in part, that fuel storage systems shall be designed with residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat removal, and with the capability to prevent significant reduction in fuel storage coolant inventory under accident conditions.

NUREG-0800, Standard Review Plan (SRP), Section 9.1.3, provides criteria for design and performance of the SFP cooling and cleanup (SFPCC) system. It provides the acceptance criteria for the SFPCC system and make-up water system. An acceptable design as stated in

SRP Section 9.1.3 is one in which the integrated design is in accordance with the specified requirements in GDC 2, 4, 5, 44, 45, 61, 63, and 10 CFR Part 20. to Matrix 5 of Section 2.1 of NRC Review Standard RS-001, Revision 0, provides the NRC staff review guidance used to determine the adequacy of SFP cooling capability. It supercedes the guidance of paragraphs III.1.d. and III.1.h. of SRP (NUREG-0800) Section 9.1.3.

Section 3.1 of RS-001 states that the licensee demonstrates adequate SFP cooling capacity by either performing a bounding evaluation or committing to a method of performing outage-specific evaluations. The analysis conditions to be assumed for bounding and cycle-specific analysis are given in Section 3.1.1 of RS-001. Section 3.2 of RS-001 provides guidance regarding requirements for adequate makeup supply.

Auxiliary Systems Branch Technical Position (BTP) 9-2, Residual Decay Energy for Light-Water Reactors for Long-Term Cooling, describes acceptable assumptions and formulations that may be used to calculate the residual decay energy release rate for light-water reactors (LWRs) for long-term cooling of the reactor facility.

The methodology in BTP 9-2 was developed based on experimental data published from 1958 to 1973 relating to energy release from the decay of fission products and is consistent with the American Nuclear Society standard 5.1 (ANS 5.1 standard), American National Standard for Decay Heat Power in Light Water Reactors. The actual equation resulting from the curve-fit in BTP 9-2 is more complex than that in the ANS standard. However, in general, the results of the curve-fit equations agree with each other within +/-5 percent.

The ORIGEN computer code modeled fissile material behavior during periods of irradiation and decay by computing time-dependent concentrations and source terms of a large number of isotopes that are simultaneously generated or depleted through neutronic transmutation, fission, radioactive decay, and physical or chemical removal rates. The primary advantage of ORIGEN over earlier burnup codes was its capability to treat the full isotopic transition matrix rather than a limited number of transmutation chains. The original ORIGEN program utilized a standard cross-section library designed for the analysis of standard LWR fuel. The required data input for ORIGEN consisted of data relating to the specific problem to be analyzed, including input for fissile isotope concentrations (i.e., bundle enrichments and uranium weight), bundle power during irradiation, irradiation length, and length of decay period. Taking these factors into consideration, the staff determined, as described in NRC Information Notice 93-39, that ORIGEN provided a rigorous approach for performing calculations of all decay heat inputs.

3.0 TECHNICAL EVALUATION

3.1 NRC Staff Evaluation The licensee stated that essentially all features of the original ORIGEN were retained, expanded, or supplemented within new computations in the development of ORIGEN-ARP. The original ORIGEN code was intended principally for the generic investigation of fuel cycle operations over a wide range of fuel design parameters. ORIGEN2 was released as an update to ORIGEN and the associated database. Similar to the original ORIGEN code, ORIGEN2 was designed to operate as a stand-alone tool with fixed cross-section data libraries provided for several reactor models. However, ORIGEN2 is no longer supported by the industry. The use

of ORIGEN2 in SFP applications was approved by the NRC staff in license amendments to Duane Arnold Energy Center (dated September 21, 2001, ADAMS Accession No. ML012500246) and Virgil C. Summer Nuclear Station (dated August 30, 2002, ADAMS Accession No. ML022330203).

A comparison between ORIGEN2 and ORIGEN-ARP is documented in Westinghouse Report A-GEN-BWR-90. The bias for decay heat from each methodology (when compared to the measured data) is within 1.2 percent of each other with a standard deviation of 0.55 percent. The total error at the 95/95 confidence level is 15.83 percent for ORIGEN2 and 15.76 percent for ORIGEN-ARP.

One difference between the assumptions and inputs evaluated in the two referenced license amendments above for ORIGEN2 and the intended application at Columbia is the treatment of power measurement uncertainty. In the referenced ORIGEN2 applications, a 2-percent factor for power measurement uncertainty was included. This is normally applied by multiplying the specific power by 1.02. The licensee stated that ORIGEN-ARP bias and uncertainty terms were derived by comparing code predictions to actual measured data. Therefore, the uncertainty in power measurement is inherently included in the error terms. To ensure that the error term adequately encompasses the required power uncertainty factor, the licensee compared (1) the results assuming only the 2-percent power uncertainty factor and (2) the results assuming no power uncertainty, but applying the error associated with the best estimate analysis. The decay heat with purely the bias term applied (i.e., the best estimate) was greater than the decay heat calculated with the 1.02 multiplier on specific power for all cases. Since the best estimate approach represents the minimum error that would be applied to the decay heat results, the requirement to include a 2-percent factor for power measurement uncertainty is satisfied by using the methodology bias procedure defined in A-GEN-BWR-90 for any confidence level.

3.2 NRC Staff Conclusion

Based on the results of the benchmarking analysis described in Westinghouse Report A-GEN-BWR-90, which is referenced by the licensee in its amendment request, and the procedure used to determine the methodology bias and uncertainties as compared to measured data, the NRC staff concludes that ORIGEN-ARP provides an acceptable means to calculate decay heat loads in the SFP and approves its use at the Columbia Generating Station.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Washington State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The

Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (71 FR 29674; published on May 23, 2006). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (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 issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: J. Hernandez Date: February 8, 2007