ML050750534

Issuance of License Amendment 233 Revising TS 5.9.5 to Be Consistent with Specification 5.6.5 of NUREG-1432

ML050750534
Person / Time
Site:	Fort Calhoun
Issue date:	03/11/2005
From:	Wang A NRC/NRR/DLPM/LPD4
To:	Ridenoure R Omaha Public Power District
Wang A, NRR/DLPM, 415-1445
Shared Package
ML050750567	List: ML050750059 ML050750534
References
NUREG-1432, TAC MC4304
Download: ML050750534 (10)

Text

March 11, 2005 Mr. R. T. Ridenoure Division Manager - Nuclear Operations Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.

Post Office Box 550 Fort Calhoun, NE 68023-0550

SUBJECT:

FORT CALHOUN STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT (TAC NO. MC4304)

Dear Mr. Ridenoure:

The Commission has issued the enclosed Amendment No. 233 to Renewed Facility Operating License No. DPR-40 for the Fort Calhoun Station, Unit No. 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated September 7, 2004.

The amendment revises TS 5.9.5, "Core Operating Limits Report," to be consistent with Specification 5.6.5 of NUREG-1432, "Standard Technical Specifications Combustion Engineering Plants." In addition, the list of core reload analysis methodologies contained in TS 5.9.5b used to determine the core operating limits has been updated. Many of these references were moved to the Omaha Public Power District core reload analysis methodology documents OPPD-NA-8301, 8302, and 8303, which are also listed in TS 5.9.5b. However, OPPD-NA-8302 has been revised to incorporate use of the code CASMO-4 in lieu of the previously approved CASMO-3 code.

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/

Alan B. Wang, Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosures:

1. Amendment No. 233 to DPR-40

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

March 11, 2005 Mr. R. T. Ridenoure Division Manager - Nuclear Operations Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.

Post Office Box 550 Fort Calhoun, NE 68023-0550

SUBJECT:

FORT CALHOUN STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT (TAC NO. MC4304)

Dear Mr. Ridenoure:

The Commission has issued the enclosed Amendment No. 233 to Renewed Facility Operating License No. DPR-40 for the Fort Calhoun Station, Unit No. 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated September 7, 2004.

The amendment revises TS 5.9.5, "Core Operating Limits Report," to be consistent with Specification 5.6.5 of NUREG-1432, "Standard Technical Specifications Combustion Engineering Plants." In addition, the list of core reload analysis methodologies contained in TS 5.9.5b used to determine the core operating limits has been updated. Many of these references were moved to the Omaha Public Power District core reload analysis methodology documents OPPD-NA-8301, 8302, and 8303, which are also listed in TS 5.9.5b. However, OPPD-NA-8302 has been revised to incorporate use of the code CASMO-4 in lieu of the previously approved CASMO-3 code.

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/

Alan B. Wang, Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-285 DISTRIBUTION:

PUBLIC GHill (2)

Enclosures:

1. Amendment No. 233 to DPR-40 PDIV-2 Reading

2. Safety Evaluation RidsNrrDlpmPdiv (HBerkow)

RidsNrrPMAWang cc w/encls: See next page RidsNrrLALFeizollahi RidsOgcRp RidsAcrsAcnwMailCenter RidsRegion4MailCenter (K. Kennedy)

TBoyce RidsNrrDlpmPdiv2 (RGramm)

TS: ML050750059 NRR-100 PKG.: ML050750567 ACCESSION NO.: ML050750534 NRR-058 OFFICE PDIV-2/PM PDIV-2/LA IROB/SCA OGC PDIV-2/SC NAME AWang LFeizollahi TBoyce RWeisman RGramm DATE 2/4/05 2/4/05 2/9/05 2/ /05 2/ /05 DOCUMENT NAME: E:\Filenet\ML050750534.wpd OFFICIAL RECORD COPY

OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 233 License No. DPR-40

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

A. The application for amendment by the Omaha Public Power District (the licensee) dated September 7, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and 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 license 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, Renewed Facility Operating License No. DPR-40 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No. DPR-40 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 233, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. The license amendment is effective as of its date of issuance, and shall be implemented within 60 days from date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Robert A. Gramm, Chief, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 11, 2005

ATTACHMENT TO LICENSE AMENDMENT NO. 233 RENEWED FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

REMOVE INSERT 5.0 - Page 8 5.0 - Page 8 5.0 - Page 9 5.0 - Page 9 5.0 - Page 10 5.0 - Page 10

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 233 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET NO. 50-285

1.0 INTRODUCTION

By application dated September 7, 2004, Omaha Public Power District (OPPD) requested changes to the Technical Specifications (Appendix A to Renewed Facility Operating License No. DPR-40) for the Fort Calhoun Station, Unit No. 1 (FCS). The proposed amendment would revise Technical Specification (TS) 5.9.5, "Core Operating Limits Report," to be consistent with Specification 5.6.5 of NUREG-1432, "Standard Technical Specifications Combustion Engineering Plants." In addition, the list of core reload analysis methodologies, contained in TS 5.9.5b and used to determine the core operating limits, would be updated. Many of these references were previously moved to the OPPD core reload analysis methodology documents OPPD-NA-8301, 8302, and 8303, which are also listed in TS 5.9.5b. In addition, OPPD-NA-8302 would be revised to incorporate use of the code CASMO-4 in lieu of the previously approved CASMO-3 code.

2.0 REGULATORY EVALUATION

The proposed technical specification changes are primarily intended to achieve consistency with Standard Technical Specifications or to consolidate the list of Nuclear Regulatory Commission (NRC) approved analytical methods of TS 5.9.5b into OPPD core reload analysis documents. TS 5.9.5b requires that the analytical methods used to determine core operating limits shall have been approved by the NRC. As such, the OPPD core reload methodology documents, which are several of the analytical methods referenced in TS 5.9.5b were submitted for NRC approval. The applicant is proposing to revise the OPPD core reload methodology documents to incorporate references removed from TS 5.9.5b, delete characters designating the documents as proprietary/approved, and incorporate a description of the CASMO-4 computer code for use for Cycle 23 core reload analysis.

OPPD originally intended to incorporate the CASMO-4 methodology into OPPD-NA-8302 under the provisions of 10 CFR 50.59, as set forth in section 4.3.8 of NEI 96-07, Revision 1, "Guidelines for 10 CFR 50.59 Implementation," dated November 2000. NEI 96-07, Revision 1 was endorsed by the NRC in Regulatory Guide 1.187, "Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments," dated November 2000. However, subsequent

discussions with NRC staff concluded that a license amendment request was needed to incorporate the CASMO-4 methodology as the staff concluded that going from CASMO-3 to CASMO-4 was an upgrade to the CASMO code rather than a revision.

The licensee states that the adoption of the CASMO-4 computer code complies with FCS Design Criterion 6, Reactor Core Design, which is similar to 10 CFR Part 50, Appendix A, General Design Criteria (GDC) 10 for Nuclear Power Plants, "Reactor Design." FCS Design Criterion 6 states that the reactor core shall be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits, which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all off-site power.

Similarly, the licensee states that the adoption of the CASMO-4 computer code complies with FCS Design Criterion 7, "Suppression of Power Oscillations," which is similar to 10 CFR Part 50, Appendix A, GDC 12, "Suppression of Reactor Power Oscillations." FCS Design Criterion 7 states the core design, together with reliable controls, shall ensure that power oscillations which could cause damage in excess of acceptable fuel damage limits are not possible or can be readily suppressed.

Finally, the licensee states that the adoption of the CASMO-4 computer code complies with FCS Design Criterion 8, "Overall Power Coefficient," which is similar to 10 CFR Part 50, Appendix A, GDC 11, "Reactor Inherent Protection." FCS Design Criterion 8 states the reactor shall be designed so that the overall power coefficient in the power operating range shall not be positive.

3.0 TECHNICAL EVALUATION

By letter dated September 7, 2004, the OPPD requested an amendment to the TSs. The proposed amendment request four changes, of which the first three changes are administrative in nature. The first proposed change would revise TS 5.9.5, "Core Operating Limits Report," to be consistent with Specification 5.6.5 of NUREG-1432, "Standard Technical Specifications Combustion Engineering Plants" (Reference 1). The staff has reviewed these changes and agree that the proposed revisions are editorial and do not change substantive requirements, and therefore are acceptable. The changes are consistent with NUREG-1432.

The second proposed change would revise pertinent references in TS 5.9.5b and relocate them to OPPD core reload analysis methodology documents (References 2, 3, 4). These changes are purely administrative in nature in that they incorporate several non-OPPD core reload analysis methodologies by reference into OPPD-controlled documents that are set forth in the COLR methodologies, and delete the methodologies no longer applicable to FCS. The staff has reviewed these changes and agree that the revised list of core reload methodologies are consistent with the previous TS in that the references have been either relocated to the OPPD core reload methodology documents or were deleted because they were no longer applicable.

Therefore, and since these OPPD documents are controlled by 10 CFR 50.59, the proposed changes are acceptable.

The third proposed change revises three OPPD core reload analysis methodologies by removing the "P" and the "A" designations that designate proprietary and approved. OPPD has reviewed these documents and determined that they no longer contain proprietary information and that the approved designation is no longer necessary. The staff agrees that the removal of these designations is administrative in nature and does not change substantive requirements, and therefore, are acceptable.

In the forth proposed change, OPPD requests NRC approval to utilize the CASMO-4 computer code for core reload analysis. The currently approved OPPD core reload analysis methodologies (References 2,3,4) use the Studsvik Scandpower codes, MICBURN-3 and CASMO-3 for cross-section generation, and SIMULATE-3 for core simulation. The licensee has stated that the implementation of the CASMO-4 computer code in the OPPD core reload analysis methodology is consistent with the requirements set forth in 10 CFR Part 50, Appendix A, GDC 10, GDC 11, and GDC 12.

OPPD has extensively evaluated the CASMO-4 computer code utilizing the guidance of Generic Letter 83-11, Supplement 1 (Reference 5). To this end, OPPD performed cold critical benchmarking with CASMO-4 (Reference 6), as well as benchmarking against plant-specific data calculated with CASMO-3 (Reference 7). These calculations demonstrate that the CASMO-4/SIMULATE-3 model does not provide any margin gains. In addition, the uncertainties associated with the peaking factors do not change, and therefore, CASMO-4 does not reduce any previously approved margins.

The approval of the use of CASMO-4 for core reload analysis is based on the FCS reactor core analyses over the range of operating parameters that span the plant specific FCS data to verify the neutronic model. Accordingly, this safety evaluation is limited to the operating conditions at FCS, and is not applicable to any other nuclear power plant. The licensee submitted extensive benchmarking analyses comparing the CASMO-4 code and the CASMO-3 code results. This benchmarking included comparisons of various core parameters at the beginning of cycle, hot zero power, hot full power and end of cycle. The results demonstrate that the CASMO-4/SIMULATE-3 code model does not provide any margin gains, or reduce any previously approved margins. Therefore, the staff finds that the requested changes comply with the conditions set forth in FCS Design Criteria 6, 7, and 8, which are similar to GDC 10, 11, and 12 of 10 CFR Part 50, Appendix A, and therefore are acceptable. In addition, the application of the CASMO-4 code in core reload analysis has been approved by the NRC for cross-section generation at the Palo Verde power station of the Arizona Public Service Company, the Prairie Island power station of the Nuclear Management Company and the North Anna and Surry power stations of Virginia Electric Power Company.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

This amendment relates to changes in record keeping, administrative procedures or requirements. 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 (69 FR 60683; dated October 12, 2004). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). 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.

7.0 REFERENCES

1. U.S. Nuclear Regulatory Commission, "Standard Technical Specifications Combustion Engineering Plants," NUREG-1432, Revision 3, March 2004.

2. Omaha Public Power District, "Reload Core Analysis Methodology Overview," OPPD-NA-8301.

3. Omaha Public Power District, "Reload Core Analysis Methodology, Neutronics Design Methods and Verification," OPPD-NA-8302.

4. Omaha Public Power District, "Reload Core Analysis Methodology, Transient and Accident Methods and Verification," OPPD-NA-8303.

5. U.S. Nuclear Regulatory Commission, "Licensee Qualification for Performing Safety Analyses," NRC Generic Letter 83-11, Supplement 1, June 24, 1999.

6. Omaha Public Power District, "B&W Cold Critical Experiments with CASMO-4," OPPD Engineering Analysis EA-FC-02-027, Revision 0.

7. Omaha Public Power District, "CASMO-4 Benchmarking," OPPD Engineering Analysis EA-FC-03-146, Revision 0.

Principal Contributor: Y. Orechwa Date: March 11, 2005

Ft. Calhoun Station, Unit 1 cc:

Winston & Strawn Mr. Daniel K. McGhee ATTN: James R. Curtiss, Esq. Bureau of Radiological Health 1400 L Street, N.W. Iowa Department of Public Health Washington, DC 20005-3502 401 SW 7th Street, Suite D Des Moines, IA 50309 Chairman Washington County Board of Supervisors P.O. Box 466 Blair, NE 68008 Mr. John Hanna, Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 310 Fort Calhoun, NE 68023 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Ms. Sue Semerera, Section Administrator Nebraska Health and Human Services Systems Division of Public Health Assurance Consumer Services Section 301 Centential Mall, South P.O. Box 95007 Lincoln, NE 68509-5007 Mr. David J. Bannister, Manager Fort Calhoun Station Omaha Public Power District Fort Calhoun Station FC-1-1 Plant P.O. Box 550 Fort Calhoun, NE 68023-0550 Mr. John B. Herman Manager - Nuclear Licensing Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.

P.O. Box 550 Fort Calhoun, NE 68023-0550